WorldWideScience

Sample records for tungsten plasma-facing components

  1. Tungsten thick coatings for plasma facing components

    International Nuclear Information System (INIS)

    Riccardi, B.; Pizzuto, A.; Orsini, A.; Libera, S.; Visca, E.; Bertamini, L.; Casadei, F.; Severini, E.; Montanari, R.; Litunovsky, N.

    1998-01-01

    The aim of the R and D activity was to realize thick W coatings on CuCrZr hollow bars and to test the mock ups with respect to thermal fatigue. Eight mock ups provided of 4 mm thick W coating were finally manufactured. The bonding integrity between coating and substrate was checked by means of an Ultrasonic apparatus. Characterisation of coatings was performed in order to assess microstructure, impurity content, density, tensile strength, adhesion strength, thermal conductivity and thermal expansion coefficient. Macroscopic residual strain measurements were performed by means of 'hole drilling' technique. The activities performed demonstrated the feasibility of thick Tungsten coatings on geometries with more complex residual strain distribution. These coatings are reliable armour of medium heat flux plasma facing component. (author)

  2. Tungsten-microdiamond composites for plasma facing components

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  3. Tungsten fibre-reinforced composites for advanced plasma facing components

    OpenAIRE

    Neu, R.; Riesch, J.; Müller, A.v.; Balden, M.; Coenen, J.W.; Gietl, H.; Höschen, T.; Li, M.; Wurster, S.; You, J.-H.

    2016-01-01

    The European Fusion Roadmap foresees water cooled plasma facing components in a first DEMO design in order to provide enough margin for the cooling capacity and to only moderately extrapolate the technology which was developed and tested for ITER. In order to make best use of the water cooling concept copper (Cu) and copper-chromium-zirconium alloy (CuCrZr) are envisaged as heat sink whereas as armour tungsten (W) based materials will be used. Combining both materials in a high heat flux comp...

  4. Tungsten fibre-reinforced composites for advanced plasma facing components

    Directory of Open Access Journals (Sweden)

    R. Neu

    2017-08-01

    Full Text Available The European Fusion Roadmap foresees water cooled plasma facing components in a first DEMO design in order to provide enough margin for the cooling capacity and to only moderately extrapolate the technology which was developed and tested for ITER. In order to make best use of the water cooling concept copper (Cu and copper-chromium-zirconium alloy (CuCrZr are envisaged as heat sink whereas as armour tungsten (W based materials will be used. Combining both materials in a high heat flux component asks for an increase of their operational range towards higher temperature in case of Cu/CuCrZr and lower temperatures for W. A remedy for both issues- brittleness of W and degrading strength of CuCrZr- could be the use of W fibres (Wf in W and Cu based composites. Fibre preforms could be manufactured with industrially viable textile techniques. Flat textiles with a combination of 150/70 µm W wires have been chosen for layered deposition of tungsten-fibre reinforced tungsten (Wf/W samples and tubular multi-layered braidings with W wire thickness of 50 µm were produced as a preform for tungsten-fibre reinforced copper (Wf /Cu tubes. Cu melt infiltration was performed together with an industrial partner resulting in sample tubes without any blowholes. Property estimation by mean field homogenisation predicts strongly enhanced strength of the Wf/CuCrZr composite compared to its pure CuCrZr counterpart. Wf /W composites show very high toughness and damage tolerance even at room temperature. Cyclic load tests reveal that the extrinsic toughening mechanisms counteracting the crack growth are active and stable. FEM simulations of the Wf/W composite suggest that the influence of fibre debonding, which is an integral part of the toughening mechanisms, and reduced thermal conductivity of the fibre due to the necessary interlayers do not strongly influence the thermal properties of future components.

  5. 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.)

  6. Silicon Carbide as a tritium permeation barrier in tungsten plasma-facing components

    Science.gov (United States)

    Wright, G. M.; Durrett, M. G.; Hoover, K. W.; Kesler, L. A.; Whyte, D. G.

    2015-03-01

    The control of tritium inventory is of great importance in future fusion reactors, not only from a safety standpoint but also to maximize a reactor's efficiency. Due to the high mobility of hydrogenic species in tungsten (W) one concern is the loss of tritium from the system via permeation through the tungsten plasma-facing components (PFC). This can lead to loss of tritium through the cooling channels of the wall thereby mandating tritium monitoring and recovery methods for the cooling system of the first wall. The permeated tritium is then out of the fuel cycle and cannot contribute to energy production until it is recovered and recycled into the system.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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

  12. 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)

  13. 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)

  14. 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

  15. TOWARD TUNGSTEN PLASMA-FACING COMPONENTS IN KSTAR: RESEARCH ON PLASMA-METAL WALL INTERACTION

    Czech Academy of Sciences Publication Activity Database

    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.; 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, Michael; van den Berg, M.; De Temmerman, G.; Pitts, R.

    2015-01-01

    Roč. 68, č. 1 (2015), s. 36-43 ISSN 1536-1055. [International Conference on Open Magnetic Systems for Plasma Confinement (OS 2014)/10./. Daejeon, 26.08.2014-29.08.2014] Institutional support: RVO:61389021 Keywords : Plasma-metal wall interaction * Tungsten technology Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.799, year: 2015 http://dx.doi.org/10.13182/FST14-897

  16. 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.

  17. 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)

  18. 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)

  19. 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.

  20. 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.

  1. 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

  2. Improvement of non destructive infrared test bed SATIR for examination of actively cooled tungsten armour Plasma Facing Components

    Energy Technology Data Exchange (ETDEWEB)

    Vignal, N., E-mail: nicolas.vignal@cea.fr; Desgranges, C.; Cantone, V.; Richou, M.; Courtois, X.; Missirlian, M.; Magaud, Ph.

    2013-10-15

    Highlights: • Non destructive infrared techniques for control ITER like PFCs. • Reflective surface such as W induce a measurement temperature error. • Numerical data processing by evaluation of the local emissivity. • SATIR test bed can control metallic surface with low and variable emissivity. -- Abstract: For steady state (magnetic) thermonuclear fusion devices which need large power exhaust capability and have to withstand heat fluxes in the range 10–20 MW m{sup −2}, advanced Plasma Facing Components (PFCs) have been developed. The importance of PFCs for operating tokamaks requests to verify their manufacturing quality before mounting. SATIR is an IR test bed validated and recognized as a reliable and suitable tool to detect cooling defaults on PFCs with CFC armour material. Current tokamak developments implement metallic armour materials for first wall and divertor; their low emissivity causes several difficulties for infrared thermography control. We present SATIR infrared thermography test bed improvements for W monoblocks components without defect and with calibrated defects. These results are compared to ultrasonic inspection. This study demonstrates that SATIR method is fully usable for PFCs with low emissivity armour material.

  3. Improvement of non destructive infrared test bed SATIR for examination of actively cooled tungsten armour Plasma Facing Components

    International Nuclear Information System (INIS)

    Vignal, N.; Desgranges, C.; Cantone, V.; Richou, M.; Courtois, X.; Missirlian, M.; Magaud, Ph.

    2013-01-01

    Highlights: • Non destructive infrared techniques for control ITER like PFCs. • Reflective surface such as W induce a measurement temperature error. • Numerical data processing by evaluation of the local emissivity. • SATIR test bed can control metallic surface with low and variable emissivity. -- Abstract: For steady state (magnetic) thermonuclear fusion devices which need large power exhaust capability and have to withstand heat fluxes in the range 10–20 MW m −2 , advanced Plasma Facing Components (PFCs) have been developed. The importance of PFCs for operating tokamaks requests to verify their manufacturing quality before mounting. SATIR is an IR test bed validated and recognized as a reliable and suitable tool to detect cooling defaults on PFCs with CFC armour material. Current tokamak developments implement metallic armour materials for first wall and divertor; their low emissivity causes several difficulties for infrared thermography control. We present SATIR infrared thermography test bed improvements for W monoblocks components without defect and with calibrated defects. These results are compared to ultrasonic inspection. This study demonstrates that SATIR method is fully usable for PFCs with low emissivity armour material

  4. Effect of microstructural anisotropy on the mechanical properties of K-doped tungsten rods for plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Nogami, Shuhei, E-mail: shuhei.nogami@qse.tohoku.ac.jp; Guan, Wenhai, E-mail: wenhai.guan@jupiter.qse.tohoku.ac.jp; Fukuda, Makoto, E-mail: fukuda@jupiter.qse.tohoku.ac.jp; Hasegawa, Akira, E-mail: akira.hasegawa@qse.tohoku.ac.jp

    2016-11-01

    Highlights: • K-doping led to improve the tensile strength regardless of the test direction and temperature. • K-doping did not alter the elongation regardless of the test direction in the ductile fracture temperature range. • The ductility at lower temperature range was improved by the K-doping. • The lowest temperature of ductile fracture along both axial and radial directions decreased because of K-doping. • K-doping could suppress the influence of microstructural anisotropy on tensile properties, especially ductility, in large diameter W rods. - Abstract: The effect of microstructural anisotropy in pure tungsten (W) and potassium (K) doped W rods (20 mm in diameter) on their mechanical properties was investigated by tensile tests along the axial and radial directions at temperatures from 473 K to 1573 K and fracture analysis. K-doping led to improved tensile strength regardless of the test direction and temperature. K-doping did not alter the elongation regardless of the test direction in the temperature range showing ductile fracture. The ductility at lower temperature range was improved by the K-doping, especially in tensile tests along the radial direction. The lowest temperature of ductile fracture along both axial and radial directions decreased from 1373 K to 973 K because of K-doping. Thus, K-doping could suppress the influence of microstructural anisotropy on tensile properties, especially ductility, in large diameter W rods.

  5. 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.

  6. 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

  7. Recent progress in R and D on tungsten alloys for divertor structural and plasma facing materials

    Energy Technology Data Exchange (ETDEWEB)

    Wurster, S., E-mail: stefan.wurster@oeaw.ac.at [Erich Schmid Institute of Materials Science, Austria and Association EURATOM-ÖAW, Jahnstrasse 12, A-8700 Leoben (Austria); Baluc, N.; Battabyal, M. [Ecole Polytechnique Fédérale de Lausanne (EPFL), Villigen PSI (Switzerland); Crosby, T. [University of California, Mechanical and Aerospace Engineering Department, Los Angeles, CA (United States); Du, J. [Max-Planck-Institut für Plasmaphysik, Garching (Germany); García-Rosales, C. [Centro de Estudios e Investigaciones Técnicas de Gipuzkoa (CEIT), San Sebastián (Spain); Hasegawa, A. [Department of Quantum Science and Energy Engineering, Faculty of Engineering, Tohoku University (Japan); Hoffmann, A. [Plansee Metall GmbH, Reutte (Austria); Kimura, A. [Institute of Advanced Energy, Kyoto University (Japan); Kurishita, H. [International Research Center for Nuclear Material Science, Institute for Materials Research, Tohoku University (Japan); Kurtz, R.J. [Pacific Northwest National Laboratory, Richland, WA (United States); Li, H. [Erich Schmid Institute of Materials Science, Austria and Association EURATOM-ÖAW, Jahnstrasse 12, A-8700 Leoben (Austria); Chair of Atomistic Modelling and Design of Materials, University of Leoben, Leoben (Austria); Noh, S.; Reiser, J. [Karlsruhe Institute of Technology, Karlsruhe (Germany); Riesch, J. [Max-Planck-Institut für Plasmaphysik, Garching (Germany); Rieth, M. [Karlsruhe Institute of Technology, Karlsruhe (Germany); Setyawan, W. [Pacific Northwest National Laboratory, Richland, WA (United States); Walter, M. [Karlsruhe Institute of Technology, Karlsruhe (Germany); You, J.-H. [Max-Planck-Institut für Plasmaphysik, Garching (Germany); and others

    2013-11-15

    Tungsten materials are candidates for plasma-facing components for the International Thermonuclear Experimental Reactor and the DEMOnstration power plant because of their superior thermophysical properties. Because these materials are not common structural materials like steels, knowledge and strategies to improve the properties are still under development. These strategies discussed here, include new alloying approaches and microstructural stabilization by oxide dispersion strengthened as well as TiC stabilized tungsten based materials. The fracture behavior is improved by using tungsten laminated and tungsten wire reinforced materials. Material development is accompanied by neutron irradiation campaigns. Self-passivation, which is essential in case of loss-of-coolant accidents for plasma facing materials, can be achieved by certain amounts of chromium and titanium. Furthermore, modeling and computer simulation on the influence of alloying elements and heat loading and helium bombardment will be presented.

  8. 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

  9. Tritium Removal from Carbon Plasma Facing Components

    International Nuclear Information System (INIS)

    Skinner, C.H.; Coad, J.P.; Federici, G.

    2003-01-01

    Tritium removal is a major unsolved development task for next-step devices with carbon plasma-facing components. The 2-3 order of magnitude increase in duty cycle and associated tritium accumulation rate in a next-step tokamak will place unprecedented demands on tritium removal technology. The associated technical risk can be mitigated only if suitable removal techniques are demonstrated on tokamaks before the construction of a next-step device. This article reviews the history of codeposition, the tritium experience of TFTR (Tokamak Fusion Test Reactor) and JET (Joint European Torus) and the tritium removal rate required to support ITER's planned operational schedule. The merits and shortcomings of various tritium removal techniques are discussed with particular emphasis on oxidation and laser surface heating

  10. ITER plasma facing components, design and development

    International Nuclear Information System (INIS)

    Vieider, G.; Cardella, A.; Akiba, M.; Matera, R.; Watson, R.

    1991-01-01

    The paper summarizes the collaborative effort of the ITER Conceptual Design Activity (CDA) on Plasma Facing Components (PFC) which focused on the following main tasks: (a) The definition of basic design concepts for the First Wall (FW) and Divertor Plates (DP), (b) the analysis of the performance and likely lifetime of these PFC designs including the identification of major critical issues, (c) the start of R and D work giving already first results, and the definition of the required further R and D program to support the contemplated ITER Engineering Design Activity (EDA). From the ITER CDA effort on PFC it is mainly concluded that: (a) The expected PFC operating conditions lead to design solutions at the limit of present technology in particular for the divertor, which may constrain the overall machine performance, (b) the development of convincing PFC designs requires an intensified R and D effort both on PFC technology and plasma physics. (orig.)

  11. Beryllium application in ITER plasma facing components

    International Nuclear Information System (INIS)

    Raffray, A.R.; Federici, G.; Barabash, V.; Cardella, A.; Jakeman, R.; Ioki, K.; Janeschitz, G.; Parker, R.; Tivey, R.; Pacher, H.D.; Wu, C.H.; Bartels, H.W.

    1997-01-01

    Beryllium is a candidate armour material for the in-vessel components of the International Thermonuclear Experimental Reactor (ITER), namely the primary first wall, the limiter, the baffle and the divertor. However, a number of issues arising from the performance requirements of the ITER plasma facing components (PFCs) must be addressed to better assess the attractiveness of Be as armour for these different components. These issues include heat loading limits arising from temperature and stress constraints under steady state conditions, armour lifetime including the effects of sputtering erosion as well as vaporisation and loss of melt during disruption events, tritium retention and permeation, and chemical hazards, in particular with respect to potential Be/steam reaction. Other issues such as fabrication and the possibility of in-situ repair are not performance-dependent but have an important impact on the overall assessment of Be as PFC armour. This paper describes the present view on Be application for ITER PFCs. The key issues are discussed including an assessment of the current level of understanding based on analysis and experimental data; and on-going activities as part of the ITER EDA R and D program are highlighted. (orig.)

  12. 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.

  13. 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)

  14. 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

  15. 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

  16. 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.

  17. 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.)

  18. 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

  19. European development of carbon armoured plasma facing components for ITER

    International Nuclear Information System (INIS)

    Merola, M.; Vieider, G.; Wu, C.; Schedler, B.; Chappuis, P.; Escourbiac, F.; Schlosser, J.; Duwe, R.; Roedig, M.; Febvre, M.; Grattarola, M.; Tahtinen, S.; Vesprini, R.

    2001-01-01

    After a brief description of the rationale of the material and geometry selection for each carbon armoured plasma facing components, this paper describes the European development of the two basic geometries, namely the monoblock and the flat tile. An overview of the non-destructive inspection techniques specifically developed for these components is also presented. (orig.)

  20. Retention and release of hydrogen isotopes in tungsten plasma-facing components: the role of grain boundaries and the native oxide layer from a joint experiment-simulation integrated approach

    Science.gov (United States)

    Hodille, E. A.; Ghiorghiu, F.; Addab, Y.; Založnik, A.; Minissale, M.; Piazza, Z.; Martin, C.; Angot, T.; Gallais, L.; Barthe, M.-F.; Becquart, C. S.; Markelj, S.; Mougenot, J.; Grisolia, C.; Bisson, R.

    2017-07-01

    Fusion fuel retention (trapping) and release (desorption) from plasma-facing components are critical issues for ITER and for any future industrial demonstration reactors such as DEMO. Therefore, understanding the fundamental mechanisms behind the retention of hydrogen isotopes in first wall and divertor materials is necessary. We developed an approach that couples dedicated experimental studies with modelling at all relevant scales, from microscopic elementary steps to macroscopic observables, in order to build a reliable and predictive fusion reactor wall model. This integrated approach is applied to the ITER divertor material (tungsten), and advances in the development of the wall model are presented. An experimental dataset, including focused ion beam scanning electron microscopy, isothermal desorption, temperature programmed desorption, nuclear reaction analysis and Auger electron spectroscopy, is exploited to initialize a macroscopic rate equation wall model. This model includes all elementary steps of modelled experiments: implantation of fusion fuel, fuel diffusion in the bulk or towards the surface, fuel trapping on defects and release of trapped fuel during a thermal excursion of materials. We were able to show that a single-trap-type single-detrapping-energy model is not able to reproduce an extended parameter space study of a polycrystalline sample exhibiting a single desorption peak. It is therefore justified to use density functional theory to guide the initialization of a more complex model. This new model still contains a single type of trap, but includes the density functional theory findings that the detrapping energy varies as a function of the number of hydrogen isotopes bound to the trap. A better agreement of the model with experimental results is obtained when grain boundary defects are included, as is consistent with the polycrystalline nature of the studied sample. Refinement of this grain boundary model is discussed as well as the inclusion

  1. Numerical simulation of runaway electron effect on Plasma Facing Components

    International Nuclear Information System (INIS)

    Ezato, Koichiro; Suzuki, Satoshi; Akiba, Masato; Kunugi, Tomoaki

    1998-07-01

    The runaway electron effects on Plasma Facing Components (PFCs) are studied by the numerical analyses. The present study is the first investigation of time-dependent thermal response of PFCs caused by runaway electron impact. For this purpose, we developed a new integrated numerical code, which consists of the Monte Carlo code for the coupled electrons and photons transport analysis and the finite element code for the thermo-mechanical analysis. In this code, we apply the practical incident parameters and distribution of runaway electrons recently proposed by S. Putvinski, which can express the time-dependent behavior of runaway electrons impact. The incident parameters of electrons in this study are the energy density ranging from 10 to 75 MJ/m 2 , the average electrons' energy of 12.5 MeV, the incident angle of 0.01deg and the characteristic time constant for decay of runaway electrons event of 0.15sec. The numerical results showed that the divertor with CFC (Carbon-Fiber-Composite) armor did not suffer serious damage. On the other hand, maximum temperatures at the surface of the divertor with tungsten armor and the first wall with beryllium armor exceed the melting point in case of the incident energy density of 20 and 50 MJ/m 2 . Within the range of the incident condition of runaway electrons, the cooling pipe of each PFCs can be prevented from the melting or burn-out caused by runaway electrons impact, which is one of the possible consequences of runaway electrons event so far. (author)

  2. Manufacturing technology development for vacuum vessel and plasma facing components

    International Nuclear Information System (INIS)

    Laitinen, Arttu; Liimatainen, Jari; Hallila, Pentti

    2005-01-01

    Vacuum vessel and plasma facing components of the ITER construction including shield modules and primary first wall panels have great impact on the production costs and reliability of the installation. From the manufacturing technology point of view, accuracy of shape, properties of the various austenitic stainless steel/austenitic stainless steel interfaces or CuCrZr/austenitic stainless steel interfaces as well as those of the base materials are crucial for technical reliability of the construction. The current approach in plasma facing components has been utilisation of solid-HIP technology and solid-powder-HIP technology. Due to the large size of especially shield modules shape, control of the internal cavities and cooling channels is extremely demanding. This requires strict control of the raw materials and manufacturing parameters

  3. 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

  4. Development of plasma facing components with functionally gradient layers

    Energy Technology Data Exchange (ETDEWEB)

    Morimoto, M.; Kudough, F. [Mitsubishi Atomic Power Industries, Inc., Yokohama (Japan); Onozuka, M.; Tsunoda, H.; Toyoda, M. [Mitsubishi Heavy Industries, Ltd., Yokohama (Japan)

    1994-11-01

    The use of functionally graded layers (FGLs) for plasma facing components (PFCs), owing to moderate or piecewise transition in material properties from low-Z surface materials to metal substrates, can provide reduction in thermal stresses, and also provide high thermal load resistance to PFCs. This article deals with the comparison of high heat flux testing and thermal stress analysis results on PFCs. Thermal stress analyses confirmed the thermal loading test results.

  5. 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.

  6. 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)

  7. Hydrogen in tungsten as plasma-facing material

    Science.gov (United States)

    Roth, Joachim; Schmid, Klaus

    2011-12-01

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

  8. 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.).

  9. 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.

  10. 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

  11. Heat Loads On Tore Supra ICRF Launchers Plasma Facing Components

    International Nuclear Information System (INIS)

    Bremond, S.; Colas, L.; Chantant, M.; Beaumont, B.; Ekedahl, A.; Goniche, M.; Moreau, P.; Mitteau, R.

    2005-01-01

    Understanding the heat loads on Ion Cyclotron Range of Frequency launchers plasma facing components is a crucial task both for operating present tokamaks and for designing ITER ICRF launchers as these loads may limit the RF power coupling capability. Tore Supra facility is particularly well suited to take this issue. Parametric studies have been performed which enables to get an overall detailed picture of the different heat loads on several areas, pointing to different mechanisms at the origin of the heat power fluxes. Lessons are drawned both with regards to Tore Supra possible operational limits and to ITER ICRF launcher design

  12. Interaction of stochastic boundary layer with plasma facing components

    International Nuclear Information System (INIS)

    Nguyen, F.; Ghendrih, P.; Grosman, A.

    1997-01-01

    To alleviate the plasma-wall interaction problems in magnetic confinement devices, a stochastic layer is used at the edge of the Tore Supra tokamak (ergodic divertor). A very important point is to determine the power deposition on the plasma facing components. Two different kinds of transport can be identified in such a configuration: Stochastic transport surrounding the confined plasma, with a random walk process, and scrape-off layer (SOL) like transport, a laminar transport, near the plasma facing components. The laminar regime is investigated in terms of a simple criterion, namely that the power deposition is proportional to the radial penetration of the laminar zone flux tubes over a finite parallel length. The magnetic connection properties of the first wall components are then determined. The connection lengths are quantified with two characteristic scales. The larger corresponds to one poloidal turn and appears to be the characteristic parallel length for laminar transport. A field line tracing code MASTOC (magnetic stochastic configuration) is used to computer the complex topology and the statistics of the connection in the real tokamak geometry. The numerical simulations are then compared with the experimental heat deposition on the modules and neutralizer plates of the Tore Supra ergodic divertor. Good agreement is found. Further evidence of laminar transport is also provided by the tangential view of such structures revealed from H α structures in detached plasma experiments. (author). 27 refs, 14 figs

  13. 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

  14. 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

  15. Advanced qualification methodology for actively cooled plasma facing components

    Science.gov (United States)

    Durocher, A.; Escourbiac, F.; Grosman, A.; Boscary, J.; Merola, M.; Cismondi, F.; Courtois, X.; Farjon, J. L.; Missirlian, M.; Schlosser, J.; Tivey, R.

    2007-12-01

    The use of high heat flux plasma facing components (PFCs) in steady state fusion devices requires high reliability. These components have to withstand heat fluxes in the range 10-20 MW m-2 involving a number of severe engineering constraints. Feedback from the experience of various industrial manufacturings showed that the bonding of the refractory armour material onto the metallic heat sink causes generic difficulties strongly depending on material qualities and specific design. As the heat exhaust capability and lifetime of PFCs during plasma operation are directly linked to the manufacturing quality, a set of qualification activities such as active infrared thermography, lock-in and acoustic measurements were performed during the component development phases following a qualification route. This paper describes the major improvements stemming from better measurement accuracy and refined data processing and analyses recent developments aimed at investigating the capability to qualify the component in situ during its lifetime.

  16. Advanced qualification methodology for actively cooled plasma facing components

    International Nuclear Information System (INIS)

    Durocher, A.; Escourbiac, F.; Grosman, A.; Boscary, J.; Merola, M.; Cismondi, F.; Courtois, X.; Farjon, J.L.; Missirlian, M.; Schlosser, J.; Tivey, R.

    2007-01-01

    The use of high heat flux plasma facing components (PFCs) in steady state fusion devices requires high reliability. These components have to withstand heat fluxes in the range 10-20 MW m -2 involving a number of severe engineering constraints. Feedback from the experience of various industrial manufacturings showed that the bonding of the refractory armour material onto the metallic heat sink causes generic difficulties strongly depending on material qualities and specific design. As the heat exhaust capability and lifetime of PFCs during plasma operation are directly linked to the manufacturing quality, a set of qualification activities such as active infrared thermography, lock-in and acoustic measurements were performed during the component development phases following a qualification route. This paper describes the major improvements stemming from better measurement accuracy and refined data processing and analyses recent developments aimed at investigating the capability to qualify the component in situ during its lifetime

  17. Heat loads on Tore Supra ICRF Launchers Plasma Facing Components

    International Nuclear Information System (INIS)

    Bremond, S.; Colas, L.; Beaumont, B.; Chantant, M.; Goniche, M.; Mitteau, R.

    2005-01-01

    Understanding the heat loads on Ion Cyclotron Range of Frequency (ICRF) launchers plasma-facing components is a crucial task both for operating present tokamaks and for designing ITER ICRF launchers as these loads may limit the RF power coupling capability. Tore Supra facility is particularly well suited to take this issue. Parametric studies have been performed which enables to get an overall detailed picture of the different heat loads on several areas, pointing to different mechanisms at the origin of the heat power fluxes. It is found that the most critical items for Tore-Supra operation are localized heat loads on the Faraday screen top left corner and vertical edges. Warming up close to maximum temperature limit originally set for protection of the plasma-facing components is found of high power pulses, but no erosion was observed after detailed inspection of the launcher in Tore-Supra vessel. Yet, the associated heat loads could be limiting for Tore-Supra operation in the future, and some dedicated work is under progress to improve the understanding of these power fluxes, pointing out the importance of getting a better knowledge of particle flows in the scrape of layer

  18. Design of plasma facing components for the SST-1 tokamak

    International Nuclear Information System (INIS)

    Jacob, S.; Chenna Reddy, D.; Choudhury, P.; Khirwadkar, S.; Pragash, R.; Santra, P.; Saxena, Y.C.; Sinha, P.

    2000-01-01

    Steady state Superconducting Tokamak, SST-1, is a medium sized tokamak with major and minor radii of 1.10 m and 0.20 m respectively. Elongated plasma operation with double null poloidal divertor is planned with a maximum input power of 1 MW. The Plasma Facing Components (PFC) like Divertors and Baffles, Poloidal limiters and Passive stabilizers form the first material boundary around the plasma and hence receive high heat and particle fluxes. The PFC design should ensure efficient heat and particle removal during steady state tokamak operation. A closed divertor geometry is adopted to ensure high neutral pressure in the divertor region (and hence high recycling) and less impurity influx into the core plasma. A set of poloidal limiters are provided to assist break down, current ramp-up and current ramp down phases and for the protection of the in-vessel components. Two pairs of Passive stabilizers, one on the inboard and the other on the outboard side of the plasma, are provided to slow down the vertical instability growth rates of the shaped plasma column. All PFCs are actively cooled to keep the plasma facing surface temperature within the design limits. The PFCs have been shaped/profiled so that maximum steady state heat flux on the surface is less than 1 MW/m 2 . (author)

  19. Heat loads on plasma facing components during disruptions on JET

    International Nuclear Information System (INIS)

    Arnoux, G.; Riccardo, V.; Fundamenski, W.; Loarte, A.; Huber, A.

    2009-01-01

    For the first time, fast measurements of heat loads on the main chamber plasma facing components (about 1 ms time resolution) during disruptions are taken on JET. The timescale of energy deposition during the thermal quench is estimated and compared with the timescale of the core plasma collapse measured with soft x-ray diagnostic. The energy deposition time is 3-8 times longer than the plasma energy collapse during density limit disruptions or radiative limit disruptions. This factor is rather in the range 1.5-4 for vertical displacement events. The heat load profiles measured during the thermal quench show substantial broadening of the power footprint on the upper dump plate. The scrape-off layer power width is increased by a factor of 3 for the density limit disruptions. The far scrape-off layer is characterized by a steeper gradient which could be explained by shadowing of the dump plate by other main chamber plasma facing components such as the outer limiter.

  20. Net erosion measurements on plasma facing components of Tore Supra

    International Nuclear Information System (INIS)

    Tsitrone, E.; Chappuis, P.; Corre, Y.; Gauthier, E.; Grosman, A.; Pascal, J.Y.

    2001-01-01

    Erosion of the plasma facing components is a crucial point of investigation in long pulse operation of future fusion devices. Therefore erosion measurements have been undertaken in the Tore Supra tokamak. After each experimental campaign, different plasma facing components have been monitored in situ by non-destructive means, in order to evaluate their net erosion following a long plasma exposure. This paper presents the results obtained over three experimental campaigns on the Tore Supra ergodic divertor B 4 C-coated neutralisers and CFC Langmuir probes. The erosion on the Langmuir probes after one year of plasma exposure can reach 100 μm, leading to an effective erosion coefficient of around 5x10 -3 to 10 -2 , in reasonable agreement with values found on other tokamaks. The erosion of the ergodic divertor neutraliser plates is lower (10 μm). This is coherent with the attenuated particle flux due to a lower incidence angle, and might also be due to some surface temperature effect, since the neutralisers are actively cooled while the Langmuir probes are not. Moreover, the profile along the neutraliser shows net erosion in zones wetted by the plasma and net redeposition in shadowed zones

  1. Design of the ITER Plasma-Facing Components

    Energy Technology Data Exchange (ETDEWEB)

    Merola, M.

    2009-07-01

    The ITER plasma-facing components cover an area of about 850 m{sup 2} and consist of the Divertor, the Blanket and the Test Blanket Modules (TBMs) with their corresponding frames. The Divertor is located at the bottom of the plasma chamber and is aimed at exhausting the major part of the plasma thermal power (including alpha power) and at minimizing the helium and impurity content in the plasma. It consists of 54 cassette assemblies. Each assembly has 3 plasma-facing components (PFCs), namely the inner and outer target and the dome, which are mounted onto a steel support structure, the cassette body. The targets directly intercept the magnetic field lines and are designed to withstand heat fluxes as high as 20 MW/m{sup 2}. CFC is the reference design solution for the armour of the lower part of the targets. However, the resultant high erosion rate could potentially limit machine operation in the DT phase (due to co-deposition with T). Therefore, prior to the DT phase, the divertor PFCs will be replaced with a new set entirely covered with W armour. The Divertor is a RH Class 1 component, which is planned to be replaced 3 times during the 20 years of the ITER operation. The construction phase of the ITER Divertor is being launched. The Blanket covers the largest fraction of the plasma-facing surface. Each of the 440 Blanket modules consists of a first wall (FW) panel, which is mechanically attached onto a Shield Module (SM). The design heat flux is set up to 1 or 5 MW/m{sup 2}. The FW panels are covered by Be tiles, which are joined onto a copper alloy (CuCrZr) heat sink, which is in turn intimately joined onto a 316L(N) stainless steel part. The SM is a block of 316L(N)-IG steel, where an array of cooling channels are obtained by machining and welding. The TBMs are mock-ups of DEMO breeding blankets. There are three ITER equatorial ports devoted to TBM testing, each of them allocating two TBMs, inserted in a thick steel frame. The frame is a water-cooled 316L

  2. 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

  3. Investigation of plasma facing components in JT-60U operation

    International Nuclear Information System (INIS)

    Masaki, K.; Ando, T.; Kodama, K.; Arai, T.; Neyatani, Y.; Yoshino, R.; Tsuji, S.; Yagyu, J.; Kaminaga, A.; Sasajima, T.; Ouchi, Y.; Koike, T.; Shimizu, M.

    1995-01-01

    The mechanical fracture of three carbon fiber composite (CFC) first wall tiles was observed. This damage was probably caused by the electromagnetic force due to halo current during disruption. The required current to break the CFC tile is estimated to be 25 kA. The broken tile was rotated poloidally around the plasma with a speed of about 10 m/s during the following discharge. A possible driving force of this rotation might be the electromagnetic force due to the scrape-off layer (SOL) current. The required current to rotate the piece of the broken tile is 1 kA. These results indicate that electromagnetic interaction between SOL plasma and the plasma facing components is important in the research on the plasma wall interactions in fusion devices. ((orig.))

  4. High quality actively cooled plasma facing components for fusion

    International Nuclear Information System (INIS)

    Nygren, R.

    1993-01-01

    This paper interweaves some suggestions for developing actively-cooled PFCs (plasma facing components) for future fusion devices with supporting examples taken from the design, fabrication and operation of Tore Supra's Phase III Outboard Pump Limiter (OPL). This actively-cooled midplane limiter, designed for heat and particle removal during long pulse operation, has been operated in essentially thermally steady state conditions. From experience with testing to identify braze flaws in the OPL, recommendations are made to analyze the impact of joining flaws on thermal-hydraulic performance of PFCs and to validate a method of inspection for such flaws early in the design development. Capability for extensive in-service monitoring of future PFCs is also recommended and the extensive calorimetry and IR thermography used to confirm and update safe operating limits for power handling of the OPL are reviewed

  5. Carbon fiber composites application in ITER plasma facing components

    Science.gov (United States)

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

    1998-10-01

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

  6. 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

  7. Carbon fiber composites application in ITER plasma facing components

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  8. Towards intelligent video understanding applied to plasma facing component monitoring

    International Nuclear Information System (INIS)

    Martin, V.; Travere, J.M.; Moncada, V.; Bremond, F.

    2011-01-01

    In this paper, we promote intelligent plasma facing component video monitoring for both real-time purposes (machine protection issues) and post event analysis purposes (plasma-wall interaction understanding). We propose a vision-based system able to automatically detect and classify into different pre-defined categories thermal phenomena such as localized hot spots or transient thermal events (e.g. electrical arcing) from infrared imaging data of PFCs. This original computer vision system is made intelligent by endowing it with high level reasoning (i.e. integration of a priori knowledge of thermal event spatio-temporal properties to guide the recognition), self-adaptability to varying conditions (e.g. different thermal scenes and plasma scenarios), and learning capabilities (e.g. statistical modelling of event behaviour based on training samples). (authors)

  9. 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.

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

    Science.gov (United States)

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

    2016-02-01

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

  11. 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)

  12. 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

  13. Technological challenges at ITER plasma facing components production in Russia

    Energy Technology Data Exchange (ETDEWEB)

    Mazul, I.V., E-mail: mazuliv@niiefa.spb.su [Efremov Institute, 196641 St. Petersburg (Russian Federation); Belyakov, V.A.; Gervash, A.A.; Giniyatulin, R.N.; Guryeva, T.M.; Kuznetsov, V.E.; Makhankov, A.N.; Okunev, A.A. [Efremov Institute, 196641 St. Petersburg (Russian Federation); Sevryukov, O.N. [MEPhI, 115409 Moscow (Russian Federation)

    2016-11-01

    Highlights: • Technological aspects of ITER PFC manufacturing in Russia are presented. • Range of technologies to be used during manufacturing of ITER PFC at Efremov Institute has been, in general, defined and their complexity, originality and difficulty are described. • Some features and challenges of welding, brazing and various tests are discussed. - Abstract: Major part of ITER plasma facing components will be manufactured in the Russian Federation (RF). Operational conditions and other requirements to these components, as well as the scale of production, are quite unique. These unique features and related technological solutions found in the frame of the project are discussed. Procedure breakdown and results of qualification for the proposed technologies and potential producers are presented, based on mockups production and testing. Design of qualification mockups and prototypes, testing programs and results are described. Basic quantitative and qualitative parameters of manufactured components and methods of quality control are presented. Critical manufacturing issues and prospects for unique production for future fusion needs are discussed.

  14. Fatigue life of the plasma-facing components in PULSAR

    International Nuclear Information System (INIS)

    Crowell, J.A.; Blanchard, J.P.

    1994-01-01

    The PULSAR project is a multi-institutional effort to determine the advantages that can be gained by building a tokamak without current drive. This machine would reduce the capital and operating costs of the machine by avoiding the need for complex current drive hardware but it must compensate for this with an energy storage scheme and with increased structural requirements due to cyclic fatigue. This paper presents the results of the fatigue analysis for the plasma-facing components of PULSAR. The structural analysis is carried out using two-dimensional finite element models and a variety of boundary conditions to account for the third dimension. In some cases the temperature distribution is modified to simulate behaviors which cannot normally be modeled with two-dimensional finite element models. PULSAR features two major engineering designs: a liquid metal-cooled vanadium design and a helium-cooled SiC/SiC design. Results are given for each. It is shown that the superior thermal and strength properties of the vanadium alloy simplify the component design process significantly. The SiC composite properties cause significantly more difficulty for the designer and, in particular, no credible design is found for a divertor fabricated solely from the SiC composite. This conclusion is based on current data for the thermophysical properties and fatigue strength of SiC fiber composites, so developments in these areas could allow the fabrication of a SiC/SiC divertor for a pulsed tokamak

  15. High quality actively cooled plasma-facing components for fusion

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1995-01-01

    This paper interweaves some suggestions for developing actively cooled plasma-facing components (PFCs) for future fusion devices, with supporting examples taken from the design, fabrication and operation of Tore Supra's Phase III outboard pump limiter (OPL). This actively cooled midplane limiter, designed for heat and particle removal during long-pulse operation, has been operated under essentially thermally steady state conditions. Testing to identify braze flaws, analysis of the impact of joining flaws on the thermal-hydraulic performance of the OPL, and the extensive calorimetry and IR thermography used to confirm and update safe operating limits for power handling of the OPL are reviewed. This experience suggests that, for PFCs in future fusion devices, flaw-tolerant designs are possible; analyses of the impacts of flaws on performance can provide criteria for quality assurance; and validating appropriate methods of inspection for such flaws early in the design development of PFCs is prudent. The need for in-service monitoring is also discussed. (orig.)

  16. Surface temperature measurement of plasma facing components in tokamaks

    International Nuclear Information System (INIS)

    Amiel, Stephane

    2014-01-01

    During this PhD, the challenges on the non-intrusive surface temperature measurements of metallic plasma facing components in tokamaks are reported. Indeed, a precise material emissivity value is needed for classical infrared methods and the environment contribution has to be known particularly for low emissivities materials. Although methods have been developed to overcome these issues, they have been implemented solely for dedicated experiments. In any case, none of these methods are suitable for surface temperature measurement in tokamaks.The active pyrometry introduced in this study allows surface temperature measurements independently of reflected flux and emissivities using pulsed and modulated photothermal effect. This method has been validated in laboratory on metallic materials with reflected fluxes for pulsed and modulated modes. This experimental validation is coupled with a surface temperature variation induced by photothermal effect and temporal signal evolvement modelling in order to optimize both the heating source characteristics and the data acquisition and treatment. The experimental results have been used to determine the application range in temperature and detection wavelengths. In this context, the design of an active pyrometry system on tokamak has been completed, based on a bicolor camera for a thermography application in metallic (or low emissivity) environment.The active pyrometry method introduced in this study is a complementary technique of classical infrared methods used for thermography in tokamak environment which allows performing local and 2D surface temperature measurements independently of reflected fluxes and emissivities. (author) [fr

  17. 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.

  18. 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

  19. Towards intelligent video understanding applied to plasma facing component monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Martin, V.; Bremond, F. [INRIA, Pulsa team-project, Sophia Antipolis (France); Travere, J.M. [CEA IRFM, Saint Paul-lez-Durance (France); Moncada, V.; Dunand, G. [Sophia Conseil Company, Sophia Antipolis (France)

    2011-07-01

    Infrared thermography has become a routine diagnostic in many magnetic fusion devices to monitor the heat loads on the plasma facing components (PFCs) for both physics studies and machine protection. The good results of the developed systems obtained so far motivate the use of imaging diagnostics for control, especially during long pulse tokamak operation (e.g. lasting several minutes). In this paper, we promote intelligent monitoring for both real-time purposes (machine protection issues) and post event analysis purposes (PWI understanding). We propose a vision-based system able to automatically detect and classify into different pre-defined categories phenomena as localized hot spots, transient thermal events (e.g. electrical arcing), and unidentified flying objects (UFOs) as dusts from infrared imaging data of PFCs. This original vision system is made intelligent by endowing it with high-level reasoning (i.e. integration of a priori knowledge of thermal event spatial and temporal properties to guide the recognition), self-adaptability to varying conditions (e.g. different plasma scenarios), and learning capabilities (e.g. statistical modelling of thermal event behaviour based on training samples). This approach has been already successfully applied to the recognition of one critical thermal event at Tore Supra. We present here latest results of its extension for the recognition of others thermal events (e.g., B{sub 4}C flakes, impact of fast particles, UFOs) and show how extracted information can be used during plasma operation at Tore Supra to improve the real time control system, and for further analysis of PFC aging. This document is composed of an abstract followed by the slides of the presentation. (authors)

  20. 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.

  1. 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.

  2. 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.

  3. ITER vacuum vessel, in vessel components and plasma facing materials

    International Nuclear Information System (INIS)

    Ioki, Kimihiro; Enoeda, M.; Federici, G.

    2007-01-01

    in a protected position flush with the FW. There are no sliding supports inside the vacuum, to keep the reliability of the system. Driving mechanisms are located outside the vacuum boundary. The divertor activities have progressed with the aim of launching the procurement according to the ITER project schedule. They include: (a) the consolidation of the design and manufacturing technologies for the plasma facing components (PFCs); (b) the prequalification programme by the parties prior to entering into the procurement phase, (c) the diagnostics integration into the divertor design, (d) the development of suitable acceptance criteria for the divertor PFCs including the required fabrication control methods; (e) the development of remote handling procedures for the first installation and for the following replacements of the divertor cassettes. (orig.)

  4. Actively cooled plasma facing components qualification, commissioning and health monitoring

    International Nuclear Information System (INIS)

    Escourbiac, F.; Durocher, A.; Grosman, A.; Courtois, X.; Farjon, J.-L.; Schlosser, J.; Merola, M.; Tivey, R.

    2006-01-01

    In modern steady state magnetic fusion devices, actively cooled plasma facing components (PFC) have to handle heat fluxes in the range of 10-20 MW/m 2 . This generates a number of engineering constraints: the armour materials must be refractory and compatible with plasma wall interaction requirements (low sputtering and/or low atomic number); the heat sink must offer high thermal conductivity, high mechanical resistance and sufficient ductility; the component cooling system -which is generally based on the circulation of pressurized water in the PFC's heat sink - must offer high thermal heat transfer efficiency. Furthermore, the assembling of the refractory armour material onto the metallic heat sink causes generic difficulties strongly depending on thermo-mechanical properties of materials and design requirements. Life time of the PFC during plasma operation are linked to their manufacturing quality, in particular they are reduced by the possible presence of flaw assembling. The fabrication of PFC in an industrial frame including their qualification and their commissioning - which consists in checking the manufacturing quality during and at the end of manufacture - is a real challenge. From experience gained at Tore Supra on carbon fibre composite flat tiles technology components, it was assessed that a set of qualifications activities must be operated during R(and)D and manufacturing phases. Dedicated Non Destructive Technique (NDT) based on advanced active infrared thermography was developed for this purpose, afterwards, correlations between NDT, high heat flux testing and thermomechanical modelling were performed to analyse damage detection and propagation, and define an acceptance criteria valuable for industrial application. Health monitoring using lock-in technique was also recently operated in-situ of the Tore Supra tokamak for detection of possible defect propagation during operations, presence of acoustic precursor for critical heat flux detection induced

  5. On “bubbly” structures in plasma facing components

    International Nuclear Information System (INIS)

    Krasheninnikov, S.I.; Smirnov, R.D.

    2013-01-01

    The theoretical model of “fuzz” growth describing the main features observed in experiments is discussed. This model is based on the assumption of enhancement of plasticity of tungsten containing significant fraction of helium atoms and clusters. The results of molecular dynamics (MD) simulations support this idea and demonstrate strong reduction of the yield strength for all temperature range. The MD simulations also show that the “flow” of tungsten strongly facilitates coagulation of helium clusters, which otherwise practically immobile, and the formation of nano-bubbles

  6. Progress of ITER full tungsten divertor technology qualification in Japan: Manufacturing full-scale plasma-facing unit prototypes

    International Nuclear Information System (INIS)

    Ezato, Koichiro; Suzuki, Satoshi; Seki, Yohji; Yamada, Hirokazu; Hirayama, Tomoyuki; Yokoyama, Kenji; Escourbiac, Frederic; Hirai, Takeshi

    2016-01-01

    Highlights: • JADA has demonstrated the feasibility of manufacturing the full-W plasma-facing units (W-PFU). • The surface profiles of the W monoblocks of the W-PFU prototypes on the test frame to mimic the support structure of the ITER OVT were examined by using an optical three-dimensional measurement system. The results show the most W monoblock surface in the target part locates within + 0.25 mm from the CAD data. • The strict profile control with the profile tolerance of ±0.3 mm is imposed on the OVT to prevent the leading edges of the W monoblocks from over-heating. • The present full-scale prototyping demonstrates to satisfy this requirement on the surface profile. • It can be concluded that the technical maturities of JADA and its suppliers are as high as to start series manufacturing the ITER divertor components. - Abstract: Japan Atomic Energy Agency (JAEA) is in progress for technology demonstration toward Full-tungsten (W) ITER divertor outer vertical target (OVT), especially, W monoblock technology that needs to withstand the repetitive heat load as high as 20 MW/m 2 for 10 s. Under the framework of the W divertor qualification program developed ITER organization, JAEA as Japanese Domestic Agency (JADA) manufactured seven full-scale plasma-facing unit (PFU) prototypes with the Japanese industries. Four prototypes that have 146 W monoblock joint with casted copper (Cu) interlayer passed successfully the ultrasonic testing. In the other three prototypes that have the different W/Cu interlayer joint, joint defects were found. The dimension measurements reveal the requirements of the gap between W monoblocks and the surface profile of PFU are feasible.

  7. Progress of ITER full tungsten divertor technology qualification in Japan: Manufacturing full-scale plasma-facing unit prototypes

    Energy Technology Data Exchange (ETDEWEB)

    Ezato, Koichiro, E-mail: ezato.koichiro@jaea.go.jp [Department of ITER Project, Naka Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency (Japan); Suzuki, Satoshi; Seki, Yohji; Yamada, Hirokazu; Hirayama, Tomoyuki; Yokoyama, Kenji [Department of ITER Project, Naka Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency (Japan); Escourbiac, Frederic; Hirai, Takeshi [ITER Organization, route de vinon sur Verdon, 13067 St Paul lez Durance (France)

    2016-11-01

    Highlights: • JADA has demonstrated the feasibility of manufacturing the full-W plasma-facing units (W-PFU). • The surface profiles of the W monoblocks of the W-PFU prototypes on the test frame to mimic the support structure of the ITER OVT were examined by using an optical three-dimensional measurement system. The results show the most W monoblock surface in the target part locates within + 0.25 mm from the CAD data. • The strict profile control with the profile tolerance of ±0.3 mm is imposed on the OVT to prevent the leading edges of the W monoblocks from over-heating. • The present full-scale prototyping demonstrates to satisfy this requirement on the surface profile. • It can be concluded that the technical maturities of JADA and its suppliers are as high as to start series manufacturing the ITER divertor components. - Abstract: Japan Atomic Energy Agency (JAEA) is in progress for technology demonstration toward Full-tungsten (W) ITER divertor outer vertical target (OVT), especially, W monoblock technology that needs to withstand the repetitive heat load as high as 20 MW/m{sup 2} for 10 s. Under the framework of the W divertor qualification program developed ITER organization, JAEA as Japanese Domestic Agency (JADA) manufactured seven full-scale plasma-facing unit (PFU) prototypes with the Japanese industries. Four prototypes that have 146 W monoblock joint with casted copper (Cu) interlayer passed successfully the ultrasonic testing. In the other three prototypes that have the different W/Cu interlayer joint, joint defects were found. The dimension measurements reveal the requirements of the gap between W monoblocks and the surface profile of PFU are feasible.

  8. Plasma Facing Components Generic Facilities Review Panel (PFC-GFRP): Final report

    International Nuclear Information System (INIS)

    McGrath, R.; Allen, S.; Hill, D.; Brooks, J.; Mattas, R.; Davis, J.; Lipschultz, B.; Ulrickson, M.

    1993-10-01

    The Plasma Facing Components (PFC) Facilities Review Panel was chartered by the US Department of Energy, Office of Fusion Energy, ITER (International Thermonuclear Experimental Reactor) and Technology Division, to outline the program plan and identify the supporting test facilities that lead to reliable, long-lived plasma facing components for ITER. This report summarizes the panel's findings and identifies the necessary and sufficient set of test facilities required for ITER PFC development

  9. Joining technologies for the plasma facing components of ITER

    International Nuclear Information System (INIS)

    Barabash, V.; Kalinin, G.; Matera, R.

    1998-01-01

    An extensive R and D program on the development of the joining technologies between armour (beryllium, tungsten and carbon fibre composites)/copper alloys heat sink and copper alloys/ stainless steel has been carried out by ITER Home Teams. A brief review of this R and D program is presented in this paper. Based on the results, reference technologies for use in ITER have been selected and recommended for further development. (author)

  10. 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.

  11. 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)

  12. 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.

  13. 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.

  14. 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

  15. The design of actively cooled plasma-facing components

    International Nuclear Information System (INIS)

    Scheerer, M.; Smid, I.; Bolt, H.; Gervash, A.; Linke, J.

    2001-01-01

    In future fusion devices, like in the stellarator Wendelstein 7-X, the target plates of the divertor will be exposed to heat loads up to power densities of 10 MW/m 2 for 1000 s. For this purpose actively cooled target elements with an internal coolant flow return, made of 2-D CFC armor tiles brazed onto a two tube cooling structure were developed and manufactured at the Forschungszentrum Juelich. Individual bent- and coolant flow reversal elements were used to achieve a high flexibility in the shape of the target elements. A special brazing technology, using a thin layer of plasma-arc deposited titanium was used for the bonding of the cooling structure to the plasma facing armor (PFA). FEM-simulations of the thermal and mechanical behavior show that a detachment of about 25% of the bonded area between the copper tubes and the PFA can be tolerated, without exceeding the critical heat flux at 15 MW/m 2 or a surface temperature of 1400 C at 10 MW/m 2 by using twisted tape inserts with a twist ratio of 2 at a cooling water velocity of 10 m/s. Thermal cycling tests in an electron beam facility up to a power density level 10.5 MW/m 2 show a very good behavior of parts of the target elements, which confirms the performance under fusion relevant conditions. Even defected parts in the bonding interface of the target elements, known from ultrasonic inspections before, show no change in the thermal performance under cycling, which confirms also the structural integrity of partly defected regions. (orig.)

  16. 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.)

  17. 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.

  18. 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.

  19. Beryllium assessment and recommendation for application in ITER plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Barabash, V.; Tanaka, S.; Matera, R. [ITER Joint Central Team, Muenchen (Germany)

    1998-01-01

    The design status of the ITER Plasma Facing Components (PFC) is presented. The operational conditions of the armour material for the different components are summarized. Beryllium is the reference armour material for the Primary Wall, Baffle and Limiter and the back-up material for the Divertor Dome. The activities on the selection of the Be grades and the joining technologies are reviewed. (author)

  20. Edge loading of plasma facing components in fusion devices

    International Nuclear Information System (INIS)

    Mohanti, R.; Deksnis, E.; Lomas, P.; Pick, M.

    1993-03-01

    The new poloidal and the inner wall guard limiter tiles of the Joint European Torus Experiment (JET) have been shaped to maximise power handling capability. The existing design of the divertor tiles of JET have been modified to reduce edge exposure. All of these components consist of discrete tiles with finite gaps. Under the assumption that the particle power flow is along field lines, the leading edges of the tiles are exposed due to field line penetration between gaps. The peak loading of these tiles to be at the edges. The report presents a generalised solution to the edge problem which indicates the steps required to shape the tiles for maximum power handling capability. (Author)

  1. Development and application of W/Cu flat-type plasma facing components at ASIPP

    International Nuclear Information System (INIS)

    Li, Q; Sun, Z X; Xu, Y; Li, B; Wei, R; Wang, W J; Xie, C Y; Wang, J C; Wang, X L; Yang, Z S; Luo, G-N; Zhao, S X; Qin, S G; Shi, Y L; Liu, G H; Missirlian, M; Guilhem, D

    2017-01-01

    W/Cu flat-type plasma facing components (PFCs) were widely used in divertor of fusion device because of its advantages, such as low cost, light in weight and good machinability. However, it is very difficult to manufacture them due to the large mismatch between the thermo-mechanical properties of W and Cu. Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) has successfully developed W/Cu flat-type PFCs for EAST W/Cu divertor project by hot isostatic pressing (HIP) technology. This paper presents the development and application of W/Cu flat-type PFCs at ASIPP. The optimized manufacturing process is to cast pure copper onto the rear side of W tiles at temperature of 1200 °C firstly, and then to HIP the W/Cu tiles onto CuCrZr heat sink at temperature of 600 °C, pressure of 150 MPa and duration of 3 h. W/Cu flat-type testing mock-up for EAST survived 1000 cycles at heat load of 5 MW m −2 in high heat flux tests. And then ASIPP prepared two mock-ups for CEA’s tungsten environment in steady-state tokamak (WEST) project. One mock-up withstood successfully 302 cycles of 20 MW m −2 , which are far beyond the design requirement. Since 2014, W/Cu flat-type PFCs were wildly used in EAST upper divertor as baffle and dome components which showed excellent performance in 2015 and 2016 campaigns. Given the success in EAST upper divertor, W/Cu flat-type concept is as well applied in the design of actively cooled Langmuir probes which will be mounted onto EAST divertor targets soon. (paper)

  2. Development and application of W/Cu flat-type plasma facing components at ASIPP

    Science.gov (United States)

    Li, Q.; Zhao, S. X.; Sun, Z. X.; Xu, Y.; Li, B.; Wei, R.; Wang, W. J.; Qin, S. G.; Shi, Y. L.; Xie, C. Y.; Wang, J. C.; Wang, X. L.; Missirlian, M.; Guilhem, D.; Liu, G. H.; Yang, Z. S.; Luo, G.-N.

    2017-12-01

    W/Cu flat-type plasma facing components (PFCs) were widely used in divertor of fusion device because of its advantages, such as low cost, light in weight and good machinability. However, it is very difficult to manufacture them due to the large mismatch between the thermo-mechanical properties of W and Cu. Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) has successfully developed W/Cu flat-type PFCs for EAST W/Cu divertor project by hot isostatic pressing (HIP) technology. This paper presents the development and application of W/Cu flat-type PFCs at ASIPP. The optimized manufacturing process is to cast pure copper onto the rear side of W tiles at temperature of 1200 °C firstly, and then to HIP the W/Cu tiles onto CuCrZr heat sink at temperature of 600 °C, pressure of 150 MPa and duration of 3 h. W/Cu flat-type testing mock-up for EAST survived 1000 cycles at heat load of 5 MW m-2 in high heat flux tests. And then ASIPP prepared two mock-ups for CEA’s tungsten environment in steady-state tokamak (WEST) project. One mock-up withstood successfully 302 cycles of 20 MW m-2, which are far beyond the design requirement. Since 2014, W/Cu flat-type PFCs were wildly used in EAST upper divertor as baffle and dome components which showed excellent performance in 2015 and 2016 campaigns. Given the success in EAST upper divertor, W/Cu flat-type concept is as well applied in the design of actively cooled Langmuir probes which will be mounted onto EAST divertor targets soon.

  3. IAEA consultants' meeting on thermal response of plasma facing materials and components

    International Nuclear Information System (INIS)

    Janev, R.K.

    1990-07-01

    The present Summary Report contains brief proceedings and the main conclusions and recommendations of the IAEA Consultants' Meeting on ''Thermal Response of Plasma Facing Materials and Components'', which was organized by the IAEA Atomic and Molecular Data Unit and held on June 11-13, 1990, in Vienna, Austria. The Report also includes a categorization and assessment of currently studied plasma facing materials, a classification scheme of material properties data, required in fusion reactor design, and a survey of the urgently needed material properties data. (author)

  4. 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.

  5. 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)

  6. Qualification, commissioning and in situ monitoring of high heat flux plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Escourbiac, F. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint Paul Lez Durance (France)], E-mail: frederic.escourbiac@cea.fr; Durocher, A.; Grosman, A.; Cismondi, F.; Courtois, X.; Farjon, J.L.; Schlosser, J. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint Paul Lez Durance (France); Merola, M.; Tivey, R. [ITER Team, CEA/Cadarache, F-13108 Saint Paul Lez Durance (France)

    2007-10-15

    Up-to-date development of actively cooled high heat flux (HHF) plasma facing components (PFC) prototypes only allows reduced margins with regards to the ITER thermal requirements. Additionally, perfect quality cannot be ensured along series manufacturing: the presence of flaws which impair the heat transfer capability of the component, in particular at the interface armour/heat sink appears to be statistically unavoidable. In order to ensure a successful series production, a qualification methodology of actively cooled high heat flux plasma facing components is proposed. Secondly, advanced non-destructive techniques developed for HHF PFC commissioning are detailed with definition of acceptance criteria. Finally, innovative diagnostics for in situ monitoring during plasma operations or tokamak shutdowns are investigated in order to prevent immediate damage (safety monitoring); or evaluate component degradation (health monitoring). This work takes into account the relevance to Tore Supra, and is applied to W7X and ITER Divertor HHF PFC.

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

    Science.gov (United States)

    Bolt, H.; Calén, H.

    1991-03-01

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

  8. Qualification, commissioning and in situ monitoring of high heat flux plasma facing components

    International Nuclear Information System (INIS)

    Escourbiac, F.; Durocher, A.; Grosman, A.; Cismondi, F.; Courtois, X.; Farjon, J.L.; Schlosser, J.; Merola, M.; Tivey, R.

    2007-01-01

    Up-to-date development of actively cooled high heat flux (HHF) plasma facing components (PFC) prototypes only allows reduced margins with regards to the ITER thermal requirements. Additionally, perfect quality cannot be ensured along series manufacturing: the presence of flaws which impair the heat transfer capability of the component, in particular at the interface armour/heat sink appears to be statistically unavoidable. In order to ensure a successful series production, a qualification methodology of actively cooled high heat flux plasma facing components is proposed. Secondly, advanced non-destructive techniques developed for HHF PFC commissioning are detailed with definition of acceptance criteria. Finally, innovative diagnostics for in situ monitoring during plasma operations or tokamak shutdowns are investigated in order to prevent immediate damage (safety monitoring); or evaluate component degradation (health monitoring). This work takes into account the relevance to Tore Supra, and is applied to W7X and ITER Divertor HHF PFC

  9. 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.

  10. 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.

  11. Application of lock-in thermography non destructive technique to CFC armoured plasma facing components

    International Nuclear Information System (INIS)

    Escourbiac, F.; Constans, S.; Courtois, X.; Durocher, A.

    2007-01-01

    A non destructive testing technique - so called modulated photothermal thermography or lock-in thermography - has been set-up for plasma facing components examination. Reliable measurements of phase contrast were obtained on 8 mm carbon fiber composite (CFC) armoured W7-X divertor component with calibrated flaws. A 3D finite element analysis allowed the correlation of the measured phase contrast and showed that a 4 mm strip flaw can be detected at the CFC/copper interface

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    Science.gov (United States)

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

    2009-12-01

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

  14. Brazing and machining of carbon based materials for plasma facing components

    International Nuclear Information System (INIS)

    Brossa, M.; Guerreschi, U.; Rossi, M.

    1994-01-01

    Carbon based materials in the recent years have often been considered and used as armour material in plasma facing components for several fusion devices, because of their low Z and good high temperature characteristics that are compatible with the operation of nuclear reactors. These materials are often connected (mechanically or by brazing) to metals, that allow the support and the cooling functions (heat sink materials). In the following the experience of Ansaldo Ricerche about the study and the manufacturing of plasma facing components and mockups is described with reference to the influence of the carbon materials in performing brazing junction with metals. It is interesting to observe how the different characteristics of the carbon materials influence the brazing process. ((orig.))

  15. 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.

  16. Magnetic field effects on runaway electron energy deposition in plasma facing materials and components

    International Nuclear Information System (INIS)

    Niemer, K.A.; Gilligan, J.G.

    1992-01-01

    This paper reports magnetic field effects on runaway electron energy deposition in plasma facing materials and components is investigated using the Integrated TIGER Series. The Integrated TIGER Series is a set of time-independent coupled electron/photon Monte Carlo transport codes which perform photon and electron transport, with or without macroscopic electric and magnetic fields. A three-dimensional computational model of 100 MeV electrons incident on a graphite block was used to simulate runawayelectrons striking a plasma facing component at the edge of a tokamak. Results show that more energy from runaway electrons will be deposited in a material that is in the presence of a magnetic field than in a material that is in the presence of no field. For low angle incident runaway electrons in a strong magnetic field, the majority of the increased energy deposition is near the material surface with a higher energy density. Electrons which would have been reflected with no field, orbit the magnetic field lines and are redeposited in the material surface, resulting in a substantial increase in surface energy deposition. Based on previous studies, the higher energy deposition and energy density will result in higher temperatures which are expected to cause more damage to a plasma facing component

  17. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    International Nuclear Information System (INIS)

    Lee, K. H.; Woo, H. K.; Im, K. H.; Cho, S. Y.; Kim, J. B.

    2000-01-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10 -6 ∼10 -7 Pa, to produce clean plasma with low impurity containments. For this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 .deg. C, 350 .deg. C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses

  18. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K.H. [Chungnam National University Graduate School, Taejeon (Korea); Im, K.H.; Cho, S.Y. [Korea Basic Science Institute, Taejeon (Korea); Kim, J.B. [Hyundai Heavy Industries Co., Ltd. (Korea); Woo, H.K. [Chungnam National University, Taejeon (Korea)

    2000-11-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10{sup -6} {approx} 10{sup -7} Pa, to produce clean plasma with low impurity containments. for this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 deg.C, 350 deg.C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses. (author). 9 refs., 11 figs., 1 tab.

  19. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. H.; Woo, H. K. [Chungnam National Univ., Taejon (Korea, Republic of); Im, K. H.; Cho, S. Y. [korea Basic Science Institute, Taejon (Korea, Republic of); Kim, J. B. [Hyundai Heavy Industries Co., Ltd., Ulsan (Korea, Republic of)

    2000-07-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10{sup -6}{approx}10{sup -7}Pa, to produce clean plasma with low impurity containments. For this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 .deg. C, 350 .deg. C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses.

  20. Analysis of singular interface stresses in dissimilar material joints for plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    You, J.H. E-mail: jeong-ha.you@ipp.mpg.de; Bolt, H

    2001-10-01

    Duplex joint structures are typical material combinations for the actively cooled plasma facing components of fusion devices. The structural integrity under the incident heat loads from the plasma is one of the most crucial issues in the technology of these components. The most critical domain in a duplex joint component is the free surface edge of the bond interface between heterogeneous materials. This is due to the fact that the thermal stress usually shows a singular intensification in this region. If the plasma facing armour tile consists of a brittle material, the existence of the stress singularity can be a direct cause of failure. The present work introduces a comprehensive analytical tool to estimate the impact of the stress singularity for duplex PFC design and quantifies the relative stress intensification in various materials joints by use of a model formulated by Munz and Yang. Several candidate material combinations of plasma facing armour and metallic heat sink are analysed and the results are compared with each other.

  1. Analysis of singular interface stresses in dissimilar material joints for plasma facing components

    International Nuclear Information System (INIS)

    You, J.H.; Bolt, H.

    2001-01-01

    Duplex joint structures are typical material combinations for the actively cooled plasma facing components of fusion devices. The structural integrity under the incident heat loads from the plasma is one of the most crucial issues in the technology of these components. The most critical domain in a duplex joint component is the free surface edge of the bond interface between heterogeneous materials. This is due to the fact that the thermal stress usually shows a singular intensification in this region. If the plasma facing armour tile consists of a brittle material, the existence of the stress singularity can be a direct cause of failure. The present work introduces a comprehensive analytical tool to estimate the impact of the stress singularity for duplex PFC design and quantifies the relative stress intensification in various materials joints by use of a model formulated by Munz and Yang. Several candidate material combinations of plasma facing armour and metallic heat sink are analysed and the results are compared with each other

  2. 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

  3. 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

  4. PFMC-16. 16th international conference on plasma-facing materials and components for fusion applications. Abstracts

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2017-07-01

    The performances of fusion devices and of future fusion power plants strongly depend on the plasma-facing materials and components. Resistance to heat and particle loads, compatibility in plasma operations, thermo-mechanical properties, as well as the response to neutron irradiation are critical parameters which need to be understood and tailored from atomistic to component levels. The 16th International Conference on Plasma-Facing Materials and Components for Fusion Applications addresses these issues.

  5. 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.

  6. Development of the plasma facing components for the dome-liner component of the ITER divertor

    International Nuclear Information System (INIS)

    Luconi, U.; Di Marco, M.; Federici, A.; Grattarola, M.; Gualco, G.; Larrea, J.M.; Merola, M.; Ozzano, C.; Pasquale, G.

    2005-01-01

    On the basis of the design and the specification of the dome-liner elaborated by EFDA, a manufacturing route based on high temperature brazing has been developed and proved by means of the fabrication and testing of several samples and mock-ups. The dome is protected with tungsten armour tiles joined onto heat sinks obtained from a bimetallic plate made of precipitation hardened copper-chromium-zirconium alloy and stainless steel realized by explosion bonding. The brazed joint between the tungsten tiles and the heat sink has been qualified by means of thermal fatigue tests on small-scale mock-ups in reactor relevant conditions. The properties of the explosion bonding joint between the front copper alloy plate to the rear steel backing has been assessed by means of an extensive metallurgical and mechanical test program according to the specification provided by EFDA. The dimensional stability during the fabrication route has been investigated by means of the realization of a relevant curved component that has been dimensionally tested after the completion of each step of the manufacturing route. The results of the experimental activity are presented and discussed in this paper

  7. Proceedings of the 4th International Workshop on Tritium Effects in Plasma Facing Components

    International Nuclear Information System (INIS)

    Causey, R. A.

    1999-01-01

    The 4th International Workshop on Tritium Effects in Plasma Facing Components was held in Santa Fe, New Mexico on May 14-15, 1998. This workshop occurs every two years, and has previously been held in Livermore/California, Nagoya/Japan, and the JRC-Ispra Site in Italy. The purpose of the workshop is to gather researchers involved in the topic of tritium migration, retention, and recycling in materials used to line magnetic fusion reactor walls and provide a forum for presentation and discussions in this area. This document provides an overall summary of the workshop, the workshop agenda, a summary of the presentations, and a list of attendees

  8. Simulated plasma facing component measurements for an in situ surface diagnostic on Alcator C-Moda)

    Science.gov (United States)

    Hartwig, Z. S.; Whyte, D. G.

    2010-10-01

    The ideal in situ plasma facing component (PFC) diagnostic for magnetic fusion devices would perform surface element and isotope composition measurements on a shot-to-shot (˜10 min) time scale with ˜1 μm depth and ˜1 cm spatial resolution over large areas of PFCs. To this end, the experimental adaptation of the customary laboratory surface diagnostic—nuclear scattering of MeV ions—to the Alcator C-Mod tokamak is being guided by ACRONYM, a Geant4 synthetic diagnostic. The diagnostic technique and ACRONYM are described, and synthetic measurements of film thickness for boron-coated PFCs are presented.

  9. 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.

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

    International Nuclear Information System (INIS)

    Federici, G.; Raffray, A.R.

    1997-01-01

    For pt.I see ibid., p.85-100, 1997. The transient thermal model RACLETTE (acronym of Rate Analysis Code for pLasma Energy Transfer Transient Evaluation) described in part I of this paper is applied here to analyse the heat transfer and erosion effects of various slow (100 ms-10 s) high power energy transients on the actively cooled plasma facing components (PFCs) of the International Thermonuclear Experimental Reactor (ITER). These have a strong bearing on the PFC design and need careful analysis. The relevant parameters affecting the heat transfer during the plasma excursions are established. The temperature variation with time and space is evaluated together with the extent of vaporisation and melting (the latter only for metals) for the different candidate armour materials considered for the design (i.e., Be for the primary first wall, Be and CFCs for the limiter, Be, W, and CFCs for the divertor plates) and including for certain cases low-density vapour shielding effects. The critical heat flux, the change of the coolant parameters and the possible severe degradation of the coolant heat removal capability that could result under certain conditions during these transients, for example for the limiter, are also evaluated. Based on the results, the design implications on the heat removal performance and erosion damage of the various ITER PFCs are critically discussed and some recommendations are made for the selection of the most adequate protection materials and optimum armour thickness. (orig.)

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

    Science.gov (United States)

    Federici, Gianfranco; Raffray, A. René

    1997-04-01

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

  12. 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

  13. 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

  14. High Heat Flux Interactions and Tritium Removal from Plasma Facing Components by a Scanning Laser

    International Nuclear Information System (INIS)

    Skinner, C.H.; Gentile, C.A.; Hassanein, A.

    2002-01-01

    A new technique for studying high heat flux interactions with plasma facing components is presented. The beam from a continuous wave 300 W neodymium laser was focused to 80 W/mm2 and scanned at high speed over the surface of carbon tiles. These tiles were previously used in the TFTR [Tokamak Fusion Test Reactor] inner limiter and have a surface layer of amorphous hydrogenated carbon that was codeposited during plasma operations. Laser scanning released up to 84% of the codeposited tritium. The temperature rise of the codeposit on the tiles was significantly higher than that of the manufactured material. In one experiment, the codeposit surface temperature rose to 1,770 C while for the same conditions, the manufactured surface increased to only 1,080 C. The peak temperature did not follow the usual square-root dependence on heat pulse duration. Durations of order 100 ms resulted in brittle destruction and material loss from the surface, while a duration of approximately 10 ms showed minimal change. A digital microscope imaged the codeposit before, during, and after the interaction with the laser and revealed hot spots on a 100-micron scale. These results will be compared to analytic modeling and are relevant to the response of plasma facing components to disruptions and vertical displacement events (VDEs) in next-step magnetic fusion devices

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

    Science.gov (United States)

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

    2015-09-01

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

  16. Multi parametric sensitivity study applied to temperature measurement of metallic plasma facing components in fusion devices

    International Nuclear Information System (INIS)

    Aumeunier, M-H.; Corre, Y.; Firdaouss, M.; Gauthier, E.; Loarer, T.; Travere, J-M.; Gardarein, J-L.; EFDA JET Contributor

    2013-06-01

    In nuclear fusion experiments, the protection system of the Plasma Facing Components (PFCs) is commonly ensured by infrared (IR) thermography. Nevertheless, the surface monitoring of new metallic plasma facing component, as in JET and ITER is being challenging. Indeed, the analysis of infrared signals is made more complicated in such a metallic environment since the signals will be perturbed by the reflected photons coming from high temperature regions. To address and anticipate this new measurement environment, predictive photonic models, based on Monte-Carlo ray tracing (SPEOS R CAA V5 Based), have been performed to assess the contribution of the reflective part in the total flux collected by the camera and the resulting temperature error. This paper deals with the effects of metals features, as the emissivity and reflectivity models, on the accuracy of the surface temperature estimation. The reliability of the features models is discussed by comparing the simulation with experimental data obtained with the wide angle IR thermography system of JET ITER like wall. The impact of the temperature distribution is studied by considering two different typical plasma scenarios, in limiter (ITER start-up scenario) and in X-point configurations (standard divertor scenario). The achievable measurement performances of IR system and risks analysis on its functionalities are discussed. (authors)

  17. 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.

  18. 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

  19. Development of bonding techniques of W and Cu-alloys for plasma facing components of fusion reactor with HIP method

    International Nuclear Information System (INIS)

    Saito, S.; Fukaya, K.; Ishiyama, S.; Eto, M.; Sato, K.; Akiba, M.

    1998-01-01

    W (tungsten) and Cu (copper)-alloys, like oxygen free high thermal conductivity (OFHC)-copper or dispersion strengthened (DS)-copper, are candidate materials for plasma facing components(PFC) of TOKAMAK type fusion reactor as armor tile and heat sink, respectively. However, PFC are exposed to cyclic high heat load and heavy irradiation by 14 MeV neutrons. Under these conditions, thermal stresses at bonding interface and irradiation damage will decrease the bonding strength between W and Cu alloys. Therefore, it is necessary to develop a reliable bonding techniques in order to make PFC with enough integrity. We have applied the hot isostatic press (HIP) method to bond W with Cu-alloys. In this experiments, to optimize HIP bonding conditions, four point bending tests were performed for different bonding conditions at temperatures from R.T. to 873 K and we obtained an optimum HIP bonding condition for W and OFHC-Cu as 1273 SK x 2 hours x 98 ∼ 147 MPa. Tensile tests were also performed at the same temperature range. The tensile strength of the bonded W / Cu was almost equal to that of OFHC Cu which was HIPed at the same conditions. Tensile specimens were broken at the bonding interface or OFHC-Cu side. Bonding tests of W and DS-Cu showed that HIP was not successful because tungsten oxide was produced at the bonding interface and residual stresses were not relaxed. Therefore, it was concluded that some insert materials will be needed to bond W and DS-Cu. (author)

  20. Remote Metrology, Mapping, and Motion Sensing of Plasma Facing Components Using FM Coherent Laser Radar

    International Nuclear Information System (INIS)

    Menon, M.M.; Barry, R.E.; Slotwinsky, A.; Kugel, H.W.; Skinner, C.H.

    2000-01-01

    Metrology inside a D/T burning fusion reactor must necessarily be conducted remotely since the in-vessel environment would be highly radioactive due to neutron activation of the torus walls. A technique based on frequency modulated coherent laser radar (FM CLR) for such remote metrology is described. Since the FM CLR relies on frequency shift to measure distances, the results are largely insensitive to surface reflectance characteristics. Results of measurements in TFTR and NSTX fusion devices using a prototype FM CLR unit, capable of remotely measuring distances (range) up to 22 m with better than 0.1-mm precision, are provided. These results illustrate that the FM CLR can be used for precision remote metrology as well as viewing. It is also shown that by conducting Doppler corrected range measurements using the CLR, the motion of objects can be tracked. Thus, the FM CLR has the potential to remotely measure the motion of plasma facing components (PFCs) during plasma disruptions

  1. A possible method of carbon deposit mapping on plasma facing components using infrared thermography

    International Nuclear Information System (INIS)

    Mitteau, R.; Spruytte, J.; Vallet, S.; Travere, J.M.; Guilhem, D.; Brosset, C.

    2007-01-01

    The material eroded from the surface of plasma facing components is redeposited partly close to high heat flux areas. At these locations, the deposit is heated by the plasma and the deposition pattern evolves depending on the operation parameters. The mapping of the deposit is still a matter of intense scientific activity, especially during the course of experimental campaigns. A method based on the comparison of surface temperature maps, obtained in situ by infrared cameras and by theoretical modelling is proposed. The difference between the two is attributed to the thermal resistance added by deposited material, and expressed as a deposit thickness. The method benefits of elaborated imaging techniques such as possibility theory and fuzzy logics. The results are consistent with deposit maps obtained by visual inspection during shutdowns

  2. Safety characteristics of options for plasma-facing components for ITER and beyond

    International Nuclear Information System (INIS)

    Piet, S.J.; McCarthy, K.A.; Holland, D.F.; Longhurst, G.R.; Merrill, B.J.

    1991-01-01

    Plasma-facing components (PFC) likely dominate the safety hazards of the International Thermonuclear Experimental Reactor (ITER) and post-ITER machines. To gain regulatory approval and for fusion energy to fulfill its ultimate attractive safety and environmental potential, safety must be considered when selecting among PFC options. This paper summarizes current PFC safety information. PFC safety issues fall into seven areas: disruption tolerance, disruption severity, tritium inventory and permeation, accidental energy release, activation/toxin hazards, cooling disturbances, and system issues. RFC options include current ITER mainline options (Be or W coating, C tiles), variants on current ITER options, and liquid metal (LM) divertors. No PFC option that we have examined is free of critical safety concerns. There are also innovative ideas that may improve any PFC's performance -- super-permeable vacuum ducts, helium self-pumping, and gaseous divertors. We conclude with recommendations and a future strategy. 17 refs., 1 fig., 3 tabs

  3. 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)

  4. Data merging of infrared and ultrasonic images for plasma facing components inspection

    Energy Technology Data Exchange (ETDEWEB)

    Richou, M. [CEA, IRFM, F-13108 Saint Paul-lez-Durance (France)], E-mail: marianne.richou@cea.fr; Durocher, A. [CEA, IRFM, F-13108 Saint Paul-lez-Durance (France); Medrano, M. [Association EURATOM - CIEMAT, Avda. Complutense 22, 28040 Madrid (Spain); Martinez-Ona, R. [Tecnatom, 28703 S. Sebastian de los Reyes, Madrid (Spain); Moysan, J. [LCND, Universite de la Mediterranee, F-13625 Aix-en-Provence (France); Riccardi, B. [Fusion For Energy, 08019 Barcelona (Spain)

    2009-06-15

    For steady-state magnetic thermonuclear fusion devices which need large power exhaust capability, actively cooled plasma facing components have been developed. In order to guarantee the integrity of these components during the required lifetime, their thermal and mechanical behaviour must be assessed. Before the procurement of the ITER Divertor, the examination of the heat sink to armour joints with non-destructive techniques is an essential topic to be addressed. Defects may be localised at different bonding interfaces. In order to improve the defect detection capability of the SATIR technique, the possibility of merging the infrared thermography test data coming from SATIR results with the ultrasonic test data has been identified. The data merging of SATIR and ultrasonic results has been performed on Carbon Fiber Composite (CFC) monoblocks with calibrated defects, identified by their position and extension. These calibrated defects were realised with machining, with 'stop-off' or by a lack of CFC activation techniques, these last two representing more accurately a real defect. A batch of 56 samples was produced to simulate each possibility of combination with regards to interface location, position and extension and way of realising the defect. The use of a data merging method based on Dempster-Shafer theory improves significantly the detection sensibility and reliability of defect location and size.

  5. Data merging of infrared and ultrasonic images for plasma facing components inspection

    International Nuclear Information System (INIS)

    Richou, M.; Durocher, A.; Medrano, M.; Martinez-Ona, R.; Moysan, J.; Riccardi, B.

    2009-01-01

    For steady-state magnetic thermonuclear fusion devices which need large power exhaust capability, actively cooled plasma facing components have been developed. In order to guarantee the integrity of these components during the required lifetime, their thermal and mechanical behaviour must be assessed. Before the procurement of the ITER Divertor, the examination of the heat sink to armour joints with non-destructive techniques is an essential topic to be addressed. Defects may be localised at different bonding interfaces. In order to improve the defect detection capability of the SATIR technique, the possibility of merging the infrared thermography test data coming from SATIR results with the ultrasonic test data has been identified. The data merging of SATIR and ultrasonic results has been performed on Carbon Fiber Composite (CFC) monoblocks with calibrated defects, identified by their position and extension. These calibrated defects were realised with machining, with 'stop-off' or by a lack of CFC activation techniques, these last two representing more accurately a real defect. A batch of 56 samples was produced to simulate each possibility of combination with regards to interface location, position and extension and way of realising the defect. The use of a data merging method based on Dempster-Shafer theory improves significantly the detection sensibility and reliability of defect location and size.

  6. CFC/Cu bond damage in actively cooled plasma facing components

    International Nuclear Information System (INIS)

    Schlosser, J; Martin, E; Henninger, C; Boscary, J; Camus, G; Escourbiac, F; Leguillon, D; Missirlian, M; Mitteau, R

    2007-01-01

    Carbon fibre composite (CFC) armours have been successfully used for actively cooled plasma facing components (PFCs) of the Tore Supra (TS) tokamak. They were also selected for the divertor of the stellarator W7-X under construction and for the vertical target of the ITER divertor. In TS and W7-X a flat tile design for heat fluxes of 10 MW m -2 has been chosen. To predict the lifetime of such PFCs, it is necessary to analyse the damage mechanisms and to model the damage propagation when the component is exposed to thermal cycling loads. Work has been performed to identify a constitutive law for the CFC and parameters to model crack propagation from the edge singularity. The aim is to predict damage rates and to propose geometric or material improvements to increase the strength and the lifetime of the interfacial bond. For ITER a tube-in-tile concept (monoblock), designed to sustain heat fluxes up to 20 MW m -2 , has been developed. The optimization of the CFC/Cu bond, proposed for flat tiles, could be adopted for the monoblock concept

  7. High heat flux actively cooled plasma facing components development, realization and first results in Tore Supra

    International Nuclear Information System (INIS)

    Grosman, A.

    2004-01-01

    The development, design, manufacture and testing of actively cooled high heat flux plasma facing components (PFC) has been an essential stage towards long powerful tokamak operations for Tore-Supra, it lasted about 10 years. This paper deals with the toroidal pumped limiter (TPL) that is able to sustain up to 10 MW/m 2 of nominal heat flux. This device is based on hardened copper alloy heat sink structures covered by a carbon fiber composite armour, it resulted in the manufacturing of 600 elementary components, called finger elements, to achieve the 7.6 m 2 TPL. This assembly has been operating in Tore-Supra since spring 2002. Some difficulties occurred during the manufacturing phase, the valuable industrial experience is summarized in the section 2. The permanent monitoring of PFC surface temperature all along the discharge is performed by a set of 6 actively cooled infrared endoscopes. The heat flux monitoring and control issue but also the progress made in our understanding of the deuterium retention in long discharges are described in the section 3. (A.C.)

  8. High heat flux actively cooled plasma facing components development, realization and first results in Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Grosman, A. [Association Euratom-CEA, Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee

    2004-07-01

    The development, design, manufacture and testing of actively cooled high heat flux plasma facing components (PFC) has been an essential stage towards long powerful tokamak operations for Tore-Supra, it lasted about 10 years. This paper deals with the toroidal pumped limiter (TPL) that is able to sustain up to 10 MW/m{sup 2} of nominal heat flux. This device is based on hardened copper alloy heat sink structures covered by a carbon fiber composite armour, it resulted in the manufacturing of 600 elementary components, called finger elements, to achieve the 7.6 m{sup 2} TPL. This assembly has been operating in Tore-Supra since spring 2002. Some difficulties occurred during the manufacturing phase, the valuable industrial experience is summarized in the section 2. The permanent monitoring of PFC surface temperature all along the discharge is performed by a set of 6 actively cooled infrared endoscopes. The heat flux monitoring and control issue but also the progress made in our understanding of the deuterium retention in long discharges are described in the section 3. (A.C.)

  9. Alternative electro-chemically based processing routes for joining of plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Krauss, Wolfgang, E-mail: wolfgang.krauss@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Materials Research III, Hermann-von-Helmholtz-Platz 1 76344 Eggenstein-Leopoldshafen (Germany); Lorenz, Julia; Holstein, Nils; Konys, Juergen [Karlsruhe Institute of Technology (KIT), Institute for Materials Research III, Hermann-von-Helmholtz-Platz 1 76344 Eggenstein-Leopoldshafen (Germany)

    2011-10-15

    Tungsten is considered in fusion technology as functional and structural material in the area of blanket and divertor for future application in DEMO. The KIT design of a He-cooled divertor includes joints between W and W-alloys as well as of W with Eurofer-steel. The main challenges range from expansion mismatch problem for tungsten/steel joints over metallurgical reactions with brittle phase formation to crack stopping ability and excellent surface wetting. These requirements were only met partly and insufficiently in the past e.g. by direct Cu-casting of tungsten onto steel. Both, the joining needs and the observed failure scenarios of conventionally joined components initiated the development of improved joining technologies based on electro-chemical processing routes. As electrolytes aqueous and aprotic, water free, system are integrated into this development line. In the first step principle requirements are presented to guarantee a reproducible and adherent deposition of scales based on Ni and Cu acting as inter layers and filler, respectively, to generate a real metallurgical bonding as demonstrate by 1100 deg. C joining tests. The development field aprotic systems based on ionic liquids is discussed with respect to enable development of refractory metal based fillers with focus high temperature W-W brazing.

  10. Improved CuCrZr/316L transition for plasma facing components

    International Nuclear Information System (INIS)

    Tabernig, Bernhard; Rainer, Florian; Scheiber, Karl-Heinz; Schedler, Bertram

    2007-01-01

    Different welding strategies were investigated to improve the tubular transition of CuCrZr to 316L in cooling pipes for actively cooled plasma facing components. Electron beam welding experiments have been carried out on tubular samples using different filler and adapter materials. After non-destructive testing by dye penetrant and He-leak tight testing samples were tensile tested at RT and 400 deg. C to down-select promising candidates. Furthermore samples were taken for a metallographic examination in order to determine the integrity of the welds, the depth of penetration and the hardness profile across the weld. In the scanning electron microscope the weld microstructure and the formation of phases were studied. Good results were obtained by the use of a Ni-filler, an Inconel and explosive welded adapter. The tested samples of these variations fulfilled the strength requirements according to the ITER specification and showed an improved transition compared with the current solution of a pure Ni-adapter. The final down-selection will be based on the results of fatigue and torsion testing

  11. Baking and helium glow discharge cleaning of SST-1 tokamak with graphite plasma facing components

    International Nuclear Information System (INIS)

    Semwal, Pratibha; Khan, Ziauddin; Raval, Dilip

    2015-01-01

    Graphite plasma facing components (PFCs) were installed inside SST-1 vacuum vessel. Prior to installation, all the graphite tiles were baked at 1000 °C in a vacuum furnace operated below 1.0 X 10 -5 mbar. However due to the porous structure of graphite, they absorb a significant amount of water vapour from air during the installation process. Rapid desorption of water vapour requires high temperature bake-out of the PFCs at ≥ 250 °C. In SST-1 the PFCs were baked at 250 °C using hot nitrogen gas facility to remove the absorbed water vapour. Also device with large graphite surface area has the disadvantage that a large quantity of hydrogen gets trapped inside it during plasma discharges which makes density control difficult. Helium (He) glow discharge cleaning (GDC) effectively removes this stored hydrogen as well as other impurities like oxygen and hydrocarbon within few nanometers from the surface by particle induced desorption. Before plasma operation in SST-1 tokamak, both baking of PFCs and He-GDC were carried out so that these impurities were removed effectively. The mean desorption yield of hydrogen was found to be 0.48. In this paper, the results of effect of baking and He-GDC experiments of SST-1 will be presented in detail. (author)

  12. Baking and helium glow discharge cleaning of SST-1 Tokamak with graphite plasma facing components

    International Nuclear Information System (INIS)

    Semwal, P; Khan, Z; Raval, D C; Dhanani, K R; George, S; Paravastu, Y; Prakash, A; Thankey, P; Ramesh, G; Khan, M S; Saikia, P; Pradhan, S

    2017-01-01

    Graphite plasma facing components (PFCs) were installed inside the SST-1 vacuum vessel. Prior to installation, all the graphite tiles were baked at 1000 °C in a vacuum furnace operated below 1.0 × 10 -5 mbar. However due to the porous structure of graphite, they absorb a significant amount of water vapour from air during the installation process. Rapid desorption of this water vapour requires high temperature bake-out of the PFCs at ≥ 250 °C. In SST-1 the PFCs were baked at 250 °C using hot nitrogen gas facility to remove the absorbed water vapour. Also device with large graphite surface area has the disadvantage that a large quantity of hydrogen gets trapped inside it during plasma discharges which makes density control difficult. Helium glow discharge cleaning (He-GDC) effectively removes this stored hydrogen as well as other impurities like oxygen and hydrocarbon within few nano-meters from the surface by particle induced desorption. Before plasma operation in SST-1 tokamak, both baking of PFCs and He-GDC were carried out so that these impurities were removed effectively. The mean desorption yield of hydrogen was found to be 0.24. In this paper the results of baking and He-GDC experiments of SST-1 will be presented in detail. (paper)

  13. Baking and helium glow discharge cleaning of SST-1 Tokamak with graphite plasma facing components

    Science.gov (United States)

    Semwal, P.; Khan, Z.; Raval, D. C.; Dhanani, K. R.; George, S.; Paravastu, Y.; Prakash, A.; Thankey, P.; Ramesh, G.; Khan, M. S.; Saikia, P.; Pradhan, S.

    2017-04-01

    Graphite plasma facing components (PFCs) were installed inside the SST-1 vacuum vessel. Prior to installation, all the graphite tiles were baked at 1000 °C in a vacuum furnace operated below 1.0 × 10-5 mbar. However due to the porous structure of graphite, they absorb a significant amount of water vapour from air during the installation process. Rapid desorption of this water vapour requires high temperature bake-out of the PFCs at ≥ 250 °C. In SST-1 the PFCs were baked at 250 °C using hot nitrogen gas facility to remove the absorbed water vapour. Also device with large graphite surface area has the disadvantage that a large quantity of hydrogen gets trapped inside it during plasma discharges which makes density control difficult. Helium glow discharge cleaning (He-GDC) effectively removes this stored hydrogen as well as other impurities like oxygen and hydrocarbon within few nano-meters from the surface by particle induced desorption. Before plasma operation in SST-1 tokamak, both baking of PFCs and He-GDC were carried out so that these impurities were removed effectively. The mean desorption yield of hydrogen was found to be 0.24. In this paper the results of baking and He-GDC experiments of SST-1 will be presented in detail.

  14. Thermographic analysis of plasma facing components covered by carbon surface layer in tokamaks

    International Nuclear Information System (INIS)

    Gardarein, Jean-Laurent

    2007-01-01

    Tokamaks are reactors based on the thermonuclear fusion energy with magnetic confinement of the plasma. In theses machines, several MW are coupled to the plasma for about 10 s. A large part of this power is directed towards plasma facing components (PFC). For better understanding and control the heat flux transfer from the plasma to the surrounding wall, it is very important to measure the surface temperature of the PFC and to estimate the imposed heat flux. In most of tokamaks using carbon PFC, the eroded carbon is circulating in the plasma and redeposited elsewhere. During the plasma operations, this leads at some locations to the formation of thin or thick carbon layers usually poorly attached to the PFC. These surface layers with unknown thermal properties complicate the calculation of the heat flux from IR surface temperature measurements. To solve this problem, we develop first, inverse method to estimate the heat flux using thermocouple (not sensitive to the carbon surface layers) temperature measurements. Then, we propose a front face pulsed photothermal method allowing an estimation of layers thermal diffusivity, conductivity, effusivity and the thermal contact resistance between the layer and the tile. The principle is to study with an infrared sensor, the cooling of the layer surface after heating by a short laser pulse, this cooling depending on the thermal properties of the successive layers. (author) [fr

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

    Science.gov (United States)

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

    2009-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Rosales, C. [CEIT and Tecnun (University of Navarra), Paseo de Manuel Lardizabal, 15, E-20018 San Sebastian (Spain)], E-mail: cgrosales@ceit.es; Lopez-Galilea, I.; Ordas, N. [CEIT and Tecnun (University of Navarra), Paseo de Manuel Lardizabal, 15, E-20018 San Sebastian (Spain); Adelhelm, C.; Balden, M. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, D-85748 Garching (Germany); Pintsuk, G. [Forschungszentrum Juelich GmbH, EURATOM Association, D-52425 Juelich (Germany); Grattarola, M.; Gualco, C. [Ansaldo Ricerche S.p.A., I-16152 Genoa (Italy)

    2009-04-30

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

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

    International Nuclear Information System (INIS)

    Garcia-Rosales, C.; Lopez-Galilea, I.; Ordas, N.; Adelhelm, C.; Balden, M.; Pintsuk, G.; Grattarola, M.; Gualco, C.

    2009-01-01

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

  18. Direct measurements of particle flux along gap sides in castellated plasma facing component in COMPASS

    International Nuclear Information System (INIS)

    Dejarnac, Renaud; Dimitrova, Miglena; Komm, Michael; Schweer, Bernd; Terra, Alexis; Martin, Aurelien; Boizante, Gontran; Gunn, James P.; Panek, Radomir

    2014-01-01

    Highlights: •We designed a probe to measure plasma deposition into gaps during tokamak discharges. •Isat profiles are measured on both side of the gap for different gap orientations. •Ion current is measured at the bottom of the gap in the toroidal orientation. •Kinetic simulations reproduce well experimental profiles qualitatively. -- Abstract: In this paper, we report results of a dedicated experiment that gives the plasma penetration profiles inside a gap of a tokamak castellated plasma-facing component. A specially designed probe that recreates a gap between two tiles has been built for the purpose of this study. It allows to measure ion saturation profiles along the 2 sides and at the bottom of the gap for both poloidal and toroidal orientations. The novelty of such experiment is the real time measurement of the plasma flux inside the gap during a tokamak D-shaped discharge compared to previous experimental studies which were mainly post-mortem. This experiment was performed in the COMPASS tokamak and results are compared with particle-in-cell simulations. The plasma deposition is found to be asymmetric in both orientations with a stronger effect in poloidal gaps. The Larmor radius of the incoming ions plays a role in the plasma penetration only in poloidal gaps but seems to have little impact in toroidal gaps. Profiles are qualitatively well reproduced by simulations. Ion current is recorded at the bottom of a toroidal gap under certain conditions

  19. Manufacturing study of Be, W and CFC bonded structures for plasma-facing components

    International Nuclear Information System (INIS)

    Onozuka, M.; Hirai, S.; Kikuchi, K.; Oda, Y.; Shimizu, K.

    2004-01-01

    A manufacturing study has been conducted for Be, W, and CFC bonded structures employed in plasma-facing components for the ITER. For Be tiles bonded to the Cu-Cr-Zr alloy heat sink with stainless-steel cooling pipes, a one-axis hot press with two heating processes has been used to bond the three materials. An Al-Si base interlayer has been used to bond Be to the Cu-alloy. The heating processes have been selected to match the required heat treatment conditions for the Cu-alloy. Because of the limited heat processes using a conventional hot press, the manufacturing cost can be minimized. For both the W and CFC tiles, the materials have been brazed at the same time to the Cu-alloy. Ni-Cu-Mn and Cu-Ti brazing materials have been used for the W and CFC tiles, respectively. Using the above bonding techniques, partial mockups of a blanket first-wall panel and divertor target have been successfully manufactured

  20. Electromagnetic and structural analyses of the vacuum vessel and plasma facing components for EAST

    International Nuclear Information System (INIS)

    Xu, Weiwei; Liu, Xufeng; Song, Yuntao; Li, Jun; Lu, Mingxuan

    2013-01-01

    Highlights: • The electromagnetic and structural responses of VV and PFCs for EAST are analyzed. • A detailed finite element model of the VV including PFCs is established. • The two most dangerous scenarios, major disruptions and downward VDEs are considered. • The distribution patterns of eddy currents, EMFs and torques on PFCs are analyzed. -- Abstract: During plasma disruptions, time-varying eddy currents are induced in the vacuum vessel (VV) and Plasma Facing Components (PFCs) of EAST. Additionally, halo currents flow partly through these structures during the vertical displacement events (VDEs). Under the high magnetic field circumstances, the resulting electromagnetic forces (EMFs) and torques are large. In this paper, eddy currents and EMFs on EAST VV, PFCs and their supports are calculated by analytical and numerical methods. ANSYS software is employed to evaluate eddy currents on VV, PFCs and their structural responses. To learn the electromagnetic and structural response of the whole structure more accurately, a detailed finite element model is established. The two most dangerous scenarios, major disruptions and downward VDEs, are examined. It is found that distribution patterns of eddy currents for various PFCs differ greatly, therefore resulting in different EMFs and torques. It can be seen that for certain PFCs the transient reaction force are severe. Results obtained here may set up a preliminary foundation for the future dynamic response research of EAST VV and PFCs which will provide a theoretical basis for the future engineering design of tokamak devices

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

    Energy Technology Data Exchange (ETDEWEB)

    Chevet, G. [Association Euratom-CEA, DSM/DRFC, CEA Cadarache, Saint-Paul-Lez-Durance (France)], E-mail: gaelle.chevet@cea.fr; Schlosser, J. [Association Euratom-CEA, DSM/DRFC, CEA Cadarache, Saint-Paul-Lez-Durance (France); Martin, E.; Herb, V.; Camus, G. [Universite Bordeaux 1, UMR 5801 (CNRS-SAFRAN-CEA-UB1), Laboratoire des Composites Thermostructuraux, F-33600 Pessac (France)

    2009-03-31

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

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

    Science.gov (United States)

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

    2009-03-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  4. Non-destructive testing of bonded structures for plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, M. [Mitsubishi Heavy Industries Ltd., Nuclear Systems Engineering Department, Konan 2-16-5, Minato-ku, Tokyo 108-8215 (Japan)]. E-mail: masanori_onozuka@mhi.co.jp; Kikuchi, K. [Mitsubishi Heavy Industries Ltd., Nuclear Systems Engineering Department, Konan 2-16-5, Minato-ku, Tokyo 108-8215 (Japan); Kirihigashi, A. [Mitsubishi Heavy Industries Ltd., Nuclear Systems Engineering Department, Konan 2-16-5, Minato-ku, Tokyo 108-8215 (Japan); Oda, Y. [Mitsubishi Heavy Industries Ltd., Nuclear Systems Engineering Department, Konan 2-16-5, Minato-ku, Tokyo 108-8215 (Japan); Shimizu, K. [Mitsubishi Heavy Industries Ltd., Nuclear Systems Engineering Department, Konan 2-16-5, Minato-ku, Tokyo 108-8215 (Japan)

    2005-11-15

    A preliminary investigation has been conducted to examine the applicability of the ultrasonic testing (UT) inspection technique for bonded structures in plasma facing components. In this study, existing UT probes have been used. Three test samples to simulate the blanket first-wall panel were fabricated. Artificial defects were applied along the diffusively bonded interfaces of the samples. Three types of UT probes have been tested. A vertical UT probe with 10 MHz, and a phased-array UT probe with 5 MHz, were used to detect defects between the Cu-alloy plates, and between the Cu-alloy plate and the stainless-steel (SS) block. The test results show that defects as small as 2 mm in size could be detected at a signal versus noise (S/N) ratio of more than 2. To detect defects along the SS pipes, a beam-focused-type UT probe with 20 MHz, has been applied. It was found that defects as small as 1 mm were identified at an S/N ratio of more than 2. While the results of the tested techniques were good, optimization of the probe systems is required before it can be concluded that such methods are most applicable for use on the bonded structures.

  5. Manufacturing study of Be, W and CFC bonded structures for plasma-facing components

    Science.gov (United States)

    Onozuka, M.; Hirai, S.; Kikuchi, K.; Oda, Y.; Shimizu, K.

    2004-08-01

    A manufacturing study has been conducted for Be, W, and CFC bonded structures employed in plasma-facing components for the ITER. For Be tiles bonded to the Cu-Cr-Zr alloy heat sink with stainless-steel cooling pipes, a one-axis hot press with two heating processes has been used to bond the three materials. An Al-Si base interlayer has been used to bond Be to the Cu-alloy. The heating processes have been selected to match the required heat treatment conditions for the Cu-alloy. Because of the limited heat processes using a conventional hot press, the manufacturing cost can be minimized. For both the W and CFC tiles, the materials have been brazed at the same time to the Cu-alloy. Ni-Cu-Mn and Cu-Ti brazing materials have been used for the W and CFC tiles, respectively. Using the above bonding techniques, partial mockups of a blanket first-wall panel and divertor target have been successfully manufactured.

  6. Thermal-hydraulic and thermo-mechanical design of plasma facing components for SST-1 tokamak

    International Nuclear Information System (INIS)

    Chaudhuri, Paritosh; Santra, P.; Chenna Reddy, D.; Parashar, S.K.S.

    2014-01-01

    The Plasma Facing Components (PFCs) are one of the major sub-systems of ssT-1 tokamak. PFC of ssT-1 consisting of divertors, passive stabilizers, baffles and limiters are designed to be compatible for steady state operation. The main consideration in the design of the PFC cooling is the steady state heat removal of up to 1 MW/m 2 . The PFC has been designed to withstand the peak heat fluxes and also without significant erosion such that frequent replacement of the armor is not necessary. Design considerations included 2-D steady state and transient tile temperature distribution and resulting thermal loads in PFC during baking, and cooling, coolant parameters necessary to maintain optimum thermal-hydraulic design, and tile fitting mechanism. Finite Element (FE) models using ANSYS have been developed to carry out the heat transfer and stress analyses of the PFC to understand its thermal and mechanical behaviors. The results of the calculation led to a good understanding of the coolant flow behavior and the temperature distribution in the tube wall and the different parts of the PFC. Thermal analysis of the PFC is carried out with the purpose of evaluating the thermal mechanical behavior of PFCs. The detailed thermal-hydraulic and thermo-mechanical designs of PFCs of ssT-1 are discussed in this paper. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Chevet, G., E-mail: gaelle.chevet@cea.fr [Association EURATOM-CEA, DSM/IRFM, CEA Cadarache, F-13108 Saint Paul lez Durance (France); Martin, E., E-mail: martin@lcts.u-bordeaux1.fr [LCTS, CNRS UMR 5801, Universite Bordeaux 1, Bordeaux (France); Boscary, J., E-mail: jean.boscary@ipp.mpg.de [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Camus, G., E-mail: camus@lcts.u-bordeaux1.fr [LCTS, CNRS UMR 5801, Universite Bordeaux 1, Bordeaux (France); Herb, V., E-mail: herb@lcts.u-bordeaux1.fr [LCTS, CNRS UMR 5801, Universite Bordeaux 1, Bordeaux (France); Schlosser, J., E-mail: jacques.schlosser@cea.fr [Association EURATOM-CEA, DSM/IRFM, CEA Cadarache, F-13108 Saint Paul lez Durance (France); Escourbiac, F., E-mail: frederic.escourbiac@cea.fr [Association EURATOM-CEA, DSM/IRFM, CEA Cadarache, F-13108 Saint Paul lez Durance (France); Missirlian, M., E-mail: marc.missirlian@cea.fr [Association EURATOM-CEA, DSM/IRFM, CEA Cadarache, F-13108 Saint Paul lez Durance (France)

    2011-10-01

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

  8. Energy deposition and thermal effects of runaway electrons in ITER-FEAT plasma facing components

    International Nuclear Information System (INIS)

    Maddaluno, G.; Maruccia, G.; Merola, M.; Rollet, S.

    2003-01-01

    The profile of energy deposited by runaway electrons (RAEs) of 10 or 50 MeV in International Thermonuclear Experimental Reactor-Fusion Energy Advanced Tokamak (ITER-FEAT) plasma facing components (PFCs) and the subsequent temperature pattern have been calculated by using the Monte Carlo code FLUKA and the finite element heat conduction code ANSYS. The RAE energy deposition density was assumed to be 50 MJ/m 2 and both 10 and 100 ms deposition times were considered. Five different configurations of PFCs were investigated: primary first wall armoured with Be, with and without protecting CFC poloidal limiters, both port limiter first wall options (Be flat tile and CFC monoblock), divertor baffle first wall, armoured with W. The analysis has outlined that for all the configurations but one (port limiter with Be flat tile) the heat sink and the cooling tube beneath the armour are well protected for both RAE energies and for both energy deposition times. On the other hand large melting (W, Be) or sublimation (C) of the surface layer occurs, eventually affecting the PFCs lifetime

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  10. Energy deposition and thermal effects of runaway electrons in ITER-FEAT plasma facing components

    Science.gov (United States)

    Maddaluno, G.; Maruccia, G.; Merola, M.; Rollet, S.

    2003-03-01

    The profile of energy deposited by runaway electrons (RAEs) of 10 or 50 MeV in International Thermonuclear Experimental Reactor-Fusion Energy Advanced Tokamak (ITER-FEAT) plasma facing components (PFCs) and the subsequent temperature pattern have been calculated by using the Monte Carlo code FLUKA and the finite element heat conduction code ANSYS. The RAE energy deposition density was assumed to be 50 MJ/m 2 and both 10 and 100 ms deposition times were considered. Five different configurations of PFCs were investigated: primary first wall armoured with Be, with and without protecting CFC poloidal limiters, both port limiter first wall options (Be flat tile and CFC monoblock), divertor baffle first wall, armoured with W. The analysis has outlined that for all the configurations but one (port limiter with Be flat tile) the heat sink and the cooling tube beneath the armour are well protected for both RAE energies and for both energy deposition times. On the other hand large melting (W, Be) or sublimation (C) of the surface layer occurs, eventually affecting the PFCs lifetime.

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    Science.gov (United States)

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

    2013-12-01

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

  13. Engineering design and thermal hydraulics of plasma facing components of SST-1

    International Nuclear Information System (INIS)

    Pragash, N. Ravi; Chaudhuri, P.; Santra, P.; Chenna Reddy, D.; Khirwadkar, S.; Saxena, Y.C.

    2001-01-01

    SST-1 is a medium size tokamak with super conducting magnetic field coils. All the subsystems of SST-1 are designed for quasi steady state (∼1000 s) operation. Plasma Facing Components (PFCs) of SST-1 consisting of divertors, passive stabilizers, baffles and poloidal limiters are also designed to be compatible for steady state operation. As SST-1 is designed to run double null divertor plasmas, these components also have up-down symmetry. A closed divertor configuration is chosen to produce high recycling and high pumping speed in the divertor region. All the PFC are made of copper alloys (CuCrZr and CuZr) on which graphite tiles are mechanically attached. These copper alloy back plates are actively cooled with water flowing in the channels grooved on them with the main consideration in the design of PFCs as the steady state heat removal of about 1.0 MW/m 2 . In addition to be able to remove high heat fluxes, the PFCs are also designed to be compatible for baking at 350 degree sign C. Extensive studies, involving different flow parameters and various cooling layouts, have been done to select the final cooling parameters and layout. Thermal response of the PFCs and vacuum vessel during baking, has been calculated using a FORTRAN code and a 2-D finite element analysis. The PFCs and their supports are also designed to withstand large electro-magnetic forces. Finite element analysis using ANSYS software package is used in this and other PFCs design. The engineering design including thermal hydraulics for cooling and baking of all the PFCs is completed. Poloidal limiters are being fabricated. The remaining PFCs, viz. divertors, stabilizers and baffles are likely to go for fabrication in the next few months. The detailed engineering design, the finite element calculations in the structural and thermal designs are presented in this paper

  14. Plasma facing materials and components for future fusion devices - development, characterization and performance under fusion specific loading conditions

    Energy Technology Data Exchange (ETDEWEB)

    Linke, J. [Forschungszentrum Juelich (Germany). Inst. fuer Plasmaphysik

    2006-04-15

    The plasma exposed components in existing and future fusion devices are strongly affected by the plasma material interaction processes. These mechanisms have a strong influence on the plasma performance; in addition they have major impact on the lifetime of the plasma facing armour and the joining interface between the plasma facing material (PFM) and the heat sink. Besides physical and chemical sputtering processes, high heat quasi-stationary fluxes during normal and intense thermal transients are of serious concern for the engineers who develop reliable wall components. In addition, the material and component degradation due to intense fluxes of energetic neutrons is another critical issue in D-T-burning fusion devices which requires extensive RandD. This paper presents an overview on the materials development and joining, the testing of PFMs and components, and the analysis of the neutron irradiation induced degradation.

  15. Plasma facing materials and components for future fusion devices - development, characterization and performance under fusion specific loading conditions

    International Nuclear Information System (INIS)

    Linke, J.

    2006-01-01

    The plasma exposed components in existing and future fusion devices are strongly affected by the plasma material interaction processes. These mechanisms have a strong influence on the plasma performance; in addition they have major impact on the lifetime of the plasma facing armour and the joining interface between the plasma facing material (PFM) and the heat sink. Besides physical and chemical sputtering processes, high heat quasi-stationary fluxes during normal and intense thermal transients are of serious concern for the engineers who develop reliable wall components. In addition, the material and component degradation due to intense fluxes of energetic neutrons is another critical issue in D-T-burning fusion devices which requires extensive RandD. This paper presents an overview on the materials development and joining, the testing of PFMs and components, and the analysis of the neutron irradiation induced degradation

  16. Plasma facing materials and components for future fusion devices—development, characterization and performance under fusion specific loading conditions

    Science.gov (United States)

    Linke, J.

    2006-04-01

    The plasma exposed components in existing and future fusion devices are strongly affected by the plasma material interaction processes. These mechanisms have a strong influence on the plasma performance; in addition they have major impact on the lifetime of the plasma facing armour and the joining interface between the plasma facing material (PFM) and the heat sink. Besides physical and chemical sputtering processes, high heat quasi-stationary fluxes during normal and intense thermal transients are of serious concern for the engineers who develop reliable wall components. In addition, the material and component degradation due to intense fluxes of energetic neutrons is another critical issue in D-T-burning fusion devices which requires extensive R&D. This paper presents an overview on the materials development and joining, the testing of PFMs and components, and the analysis of the neutron irradiation induced degradation.

  17. Heat loads on poloidal and toroidal edges of castellated plasma-facing components in COMPASS

    Science.gov (United States)

    Dejarnac, R.; Corre, Y.; Vondracek, P.; Gaspar, J.; Gauthier, E.; Gunn, J. P.; Komm, M.; Gardarein, J.-L.; Horacek, J.; Hron, M.; Matejicek, J.; Pitts, R. A.; Panek, R.

    2018-06-01

    Dedicated experiments have been performed in the COMPASS tokamak to thoroughly study the power deposition processes occurring on poloidal and toroidal edges of castellated plasma-facing components in tokamaks during steady-state L-mode conditions. Surface temperatures measured by a high resolution infra-red camera are compared with reconstructed synthetic data from a 2D thermal model using heat flux profiles derived from both the optical approximation and 2D particle-in-cell (PIC) simulations. In the case of poloidal leading edges, when the contribution from local radiation is taken into account, the parallel heat flux deduced from unperturbed, upstream measurements is fully consistent with the observed temperature increase at the leading edges of various heights, respecting power balance assuming simple projection of the parallel flux density. Smoothing of the heat flux deposition profile due to finite ion Larmor radius predicted by the PIC simulations is found to be weak and the power deposition on misaligned poloidal edges is better described by the optical approximation. This is consistent with an electron-dominated regime associated with a non-ambipolar parallel current flow. In the case of toroidal gap edges, the different contributions of the total incoming flux along the gap have been observed experimentally for the first time. They confirm the results of recent numerical studies performed for ITER showing that in specific cases the heat deposition does not necessarily follow the optical approximation. Indeed, ions can spiral onto the magnetically shadowed toroidal edge. Particle-in-cell simulations emphasize again the role played by local non-ambipolarity in the deposition pattern.

  18. Design and operation results of nitrogen gas baking system for KSTAR plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang-Tae [National Fusion Research Institute, 113 Gwahang-ro, Yuseong-gu, Daejeon 305-806 (Korea, Republic of); Kim, Young-Jin, E-mail: k43689@nfri.re.kr [National Fusion Research Institute, 113 Gwahang-ro, Yuseong-gu, Daejeon 305-806 (Korea, Republic of); Joung, Nam-Yong; Im, Dong-Seok; Kim, Kang-Pyo; Kim, Kyung-Min; Bang, Eun-Nam; Kim, Yaung-Soo [National Fusion Research Institute, 113 Gwahang-ro, Yuseong-gu, Daejeon 305-806 (Korea, Republic of); Yoo, Seong-Yeon [Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon 305-764 (Korea, Republic of)

    2013-11-15

    Highlights: • Vacuum pressure in a vacuum vessel arrived at 7.24 × 10{sup −8} mbar. • PFC temperature was reached maximum 250 °C by gas temperature at 300 °C. • PFC inlet gas temperature was changed 5 °C per hour during rising and falling. • PFC gas balancing was made temperature difference among them below 8.3 °C. • System has a pre-cooler and a three-way valve to save operation energy. -- Abstract: A baking system for the Korea Superconducting Tokamak Advanced Research (KSTAR) plasma facing components (PFCs) is designed and operated to achieve vacuum pressure below 5 × 10{sup −7} mbar in vacuum vessel with removing impurities. The purpose of this research is to prevent the fracture of PFC because of thermal stress during baking the PFC, and to accomplish stable operation of the baking system with the minimum life cycle cost. The uniformity of PFC temperature in each sector was investigated, when the supply gas temperature was varied by 5 °C per hour using a heater and the three-way valve at the outlet of a compressor. The alternative of the pipe expansion owing to hot gas and the cage configuration of the three-way valve were also studied. During the fourth campaign of the KSTAR in 2011, nitrogen gas temperature rose up to 300 °C, PFC temperature reached at 250 °C, the temperature difference among PFCs was maintained at below 8.3 °C, and vacuum pressure of up to 7.24 × 10{sup −8} mbar was achieved inside the vacuum vessel.

  19. Low cycle thermal fatigue testing of beryllium grades for ITER plasma facing components

    International Nuclear Information System (INIS)

    Watson, R.D.; Youchison, D.L.; Dombrowski, D.E.; Guiniatouline, R.N.; Kupriynov, I.B.

    1996-01-01

    A novel technique has been used to test the relative low cycle thermal fatigue resistance of different grades of US and Russian beryllium, which is proposed as plasma facing armor for fusion reactor first wall, limiter, and divertor components. The 30 kW electron beam test system at Sandia National Laboratories was used to sweep the beam spot along one direction at 1 Hz. This produces a localized temperature ''spike'' of 750 degree C for each pass of the beam. Large thermal stresses in excess of the yield strength are generated due to very high spot heat flux, 250 MW/m 2 . Cyclic plastic strains on the order of 0.6% produced visible cracking on the heated surface in less than 3000 cycles. An in-vacuo fiber optic borescope was used to visually inspect the beryllium surfaces for crack initiation. Grades of US beryllium tested included: S-65C, S- 65H, S-200F, S-200F-H, SR-200, I-400, extruded high purity, HIP'd spherical powder, porous beryllium (94% and 98% dense), Be/30% BeO, Be/60% BeO, and TiBe 12 . Russian grades included: TGP-56, TShGT, DShG-200, and TShG-56. Both the number of cycles to crack initiation, and the depth of crack propagation, were measured. The most fatigue resistant grades were S-65C, DShG-200, TShGT, and TShG-56. Rolled sheet Be (SR-200) showed excellent crack propagation resistance in the plane of rolling, despite early formation of delamination cracks. Only one sample showed no evidence of surface melting, Extruded (T). Metallographic and chemical analyses are provided. Good agreement was found between the measured depth of cracks and a 2-D elastic-plastic finite element stress analysis

  20. Simulation of damage to tokamaks plasma facing components during intense abnormal power deposition

    International Nuclear Information System (INIS)

    Genco, F.; Hassanein, A.

    2014-01-01

    Highlights: • HEIGHTS-PIC a new technique based on particle in cell method to study disruptions events, ELMS and VDE is benchmarked in this paper with the use of the MK-200 experiments. • Disruptions simulations results for erosion and erosion rate are proposed showing good agreement with published experimental available data for such conditions. • Results are also compared with other published results produced by FOREV1/FOREV2 computer package and the original HEIGHTS computer package. • Accuracy of the simulations results is proposed with specific aim to address the use of number of super particles adopted versus computational time. - Abstract: Intense power deposition on plasma facing components (PFC) is expected in tokamaks during loss of confinement events such as disruptions, vertical displacement events (VDE), runaway electrons (RE), or during normal operating conditions such as edge-localized modes (ELM). These highly energetic events are damaging enough to hinder long term operation and may not be easily mitigated without loss of structural or functional performance of the PFC. Surface erosion, melted/ablated-vaporized material splashing, and material transport into the bulk plasma are reliability-threatening for the machine and system performance. A novel particle-in-cell (PIC) technique has been developed and integrated into the existing HEIGHTS package in order to obtain a global view of the plasma evolution upon energy impingement. This newly developed PIC technique is benchmarked against plasma gun experimental data, the original HEIGHTS computer package, and laser experiments. Benchmarking results are shown in this paper for various relevant reactor and experimental devices. The evolution of the plasma vapor cloud is followed temporally and results are explained and commented as a function of the computational time needed and the accuracy of the calculation

  1. An operational non destructive examination for ITER divertor plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Durocher, A.; Escourbiac, F.; Farjon, J.L.; Vignal, N.; Cismondi, F. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Merola, M. [ITER International Team, Cadarache, 13 - St Paul Lez Durance (France); Riccardi, B. [CEFDA CSU-Garching, Garching bei Munchen (Germany)

    2007-07-01

    Full text of publication follows: To meet the power exhaust - heat flux of 20 MW/m{sup 2} - requirements of Plasma Facing Components (PFCs) during plasma operation requires control of their thermal and mechanical integrity. As heat exhaust capability and lifetime of PFCs during in-situ operation are linked to the manufacturing quality, it is an absolute requirement to develop reliable nondestructive examination methods, in particular of the CFC-CuCrZr joint, throughout the manufacturing process. Within the framework of Tokamak Tore Supra upgrade, a pioneering activity has been developed to evaluate the capability of the PFC to be efficiently cooled. In 1998 a test bed - so called SATIR - based on the heat transient method was developed by the CEA and is used today as an inspection tool in order to guarantee the PFCs performances. The technical procurement plan of ITER Divertor targets stated that all Cu cast layers on CFC armour should be subjected to 100% thermographic examination. Each ITER Party should demonstrate its technical capability to carry out the PFC with the required cooling efficiently. The ITER Divertor PFCs pose new challenges especially for the mono-block CFC thickness, and the number of full scale units to be tested which is higher than on any existing or under construction fusion machine. The SATIR method as functional inspection has been identified as the basis test to decide upon the final acceptance of the Divertor PFCs. In order to increase the detection sensitivity of SATIR test bed, several possibilities have been assessed i) the increase of the convective heat transfer coefficient, which improved in a significant way the sensitivity of SATIR diagnostic on ITER components. ii) the installation of a digital infrared camera and the improvement of the thermal signal processing, has led to a considerable increase of performances iii) an innovative process based on spatial image autocorrelation will allow to localize the interlayer defect

  2. 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.

  3. 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

  4. The feasibility of beryllium as structural material for the ITER plasma-facing components (PFC)

    International Nuclear Information System (INIS)

    Vieider, G.; Cardella, A.; Gorenflo, H.

    1993-01-01

    Be as plasma-facing armour has attractive features including excellent plasma compatibility, no T-retention via co-deposition and the potential for in-situ repair via plasma spraying. In order to avoid the bonding of the Be-armour to a heatsink structure in e.g., Cu-alloys, the ITER Joint Central Team (JCT) proposed for the divertor tubular elements with monolithic Be, both as plasma-facing and structural material. The analysis of these Be-tubes with 5 mm wall thickness at a heat load of 5 MW/m 2 showed that even for the most favourable assumptions thermal stresses exceed by far the allowed values according to design codes. Damage by neutrons and disruptions would worsen further the case for Be as monolithic plasma-facing and structural material. For PFC at heat flux significantly above 1 MW/m 2 it appears evident that Be should be used merely as armour bonded to a suitable structural material as heatsink. (orig.)

  5. 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.

  6. Performance Test of Korea Heat Load Test Facility (KoHLT-EB) for the Plasma Facing Components of Fusion Reactor

    International Nuclear Information System (INIS)

    Kim, Suk-Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won; Cho, Seungyon

    2014-01-01

    The main components of the plasma facing components (PFCs) in the tokamak are the blanket first wall and divertor, which include the armour materials, the heat sink with the cooling mechanism, and the diagnostics devices for the temperature measurement. The Korea Heat Load Test facility by using electron beam (KoHLT-EB) has been operating for the plasma facing components to develop fusion engineering. This electron beam facility was constructed using a 300 kW electron gun and a cylindrical vacuum chamber. Performance tests were carried out for the calorimetric calibrations with Cu dummy mockup and for the heat load test of large Cu module. For the simulation of the heat load test of each mockup, the preliminary thermal-hydraulic analyses with ANSYS-CFX were performed. For the development of the plasma facing components in the fusion reactors, test mockups were fabricated and tested in the high heat flux test facility. To perform a beam profile test, an assessment of the possibility of electron beam Gaussian power density profile and the results of the absorbed power for that profile before the test starts are needed. To assess the possibility of a Gaussian profile, for the qualification test of the Gaussian heat load profile, a calorimeter mockup and large Cu module were manufactured to simulate real heat. For this high-heat flux test, the Korean high-heat flux test facility using an electron beam system was constructed. In this facility, a cyclic heat flux test will be performed to measure the surface heat flux, surface temperature profile, and cooling capacity

  7. Performance Test of Korea Heat Load Test Facility (KoHLT-EB) for the Plasma Facing Components of Fusion Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Suk-Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The main components of the plasma facing components (PFCs) in the tokamak are the blanket first wall and divertor, which include the armour materials, the heat sink with the cooling mechanism, and the diagnostics devices for the temperature measurement. The Korea Heat Load Test facility by using electron beam (KoHLT-EB) has been operating for the plasma facing components to develop fusion engineering. This electron beam facility was constructed using a 300 kW electron gun and a cylindrical vacuum chamber. Performance tests were carried out for the calorimetric calibrations with Cu dummy mockup and for the heat load test of large Cu module. For the simulation of the heat load test of each mockup, the preliminary thermal-hydraulic analyses with ANSYS-CFX were performed. For the development of the plasma facing components in the fusion reactors, test mockups were fabricated and tested in the high heat flux test facility. To perform a beam profile test, an assessment of the possibility of electron beam Gaussian power density profile and the results of the absorbed power for that profile before the test starts are needed. To assess the possibility of a Gaussian profile, for the qualification test of the Gaussian heat load profile, a calorimeter mockup and large Cu module were manufactured to simulate real heat. For this high-heat flux test, the Korean high-heat flux test facility using an electron beam system was constructed. In this facility, a cyclic heat flux test will be performed to measure the surface heat flux, surface temperature profile, and cooling capacity.

  8. Hydrodynamic effects of eroded materials on response of plasma-facing component during a tokamak disruption

    International Nuclear Information System (INIS)

    Hassanein, A.; Konkashbaev, I.

    1999-01-01

    Loss of plasma confinement causes surface and structural damage to plasma-facing materials (PFMs) and remains a major obstacle for tokamak reactors. The deposited plasma energy results in surface erosion and structural failure. The surface erosion consists of vaporization, spallation, and liquid splatter of metallic materials, while the structural damage includes large temperature increases in structural materials and at the interfaces between surface coatings and structural members. Comprehensive models (contained in the HEIGHTS computer simulation package) are being used self-consistently to evaluate material damage. Splashing mechanisms occur as a result of volume bubble boiling and liquid hydrodynamic instabilities and brittle destruction mechanisms of nonmelting materials. The effect of macroscopic erosion on total mass losses and lifetime is evaluated. The macroscopic erosion products may further protect PFMs from severe erosion (via the droplet-shielding effect) in a manner similar to that of the vapor shielding concept

  9. 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

  10. 1st IAEA research coordination meeting on tritium retention in fusion reactor plasma facing components. October 5-6, 1995, Vienna, Austria. Summary report

    International Nuclear Information System (INIS)

    Langley, R.A.

    1995-12-01

    The proceedings and results of the 1st IAEA research Coordination Meeting on ''Tritium Retention in Fusion Reactor Plasma Facing Components'' held on October 5 and 6, 1995 at the IAEA Headquarters in Vienna are briefly described. This report includes a summary of presentations made by the meeting participants, the results of a data survey and needs assessment for the retention, release and removal of tritium from plasma facing components, a summary of data evaluation, and recommendations regarding future work. (author). 4 tabs

  11. 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

  12. Proceedings of 2nd Internaitonal workshop on tritium effects in plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Morita, Kenji [Nagoya Univ. (Japan). School of Engineering; Noda, Nobuaki [eds.

    1994-08-01

    This workshop was held at Nagoya University on May 19 and 20, 1994. Approximately 1/3 of the lectures discussed the migration and retention of tritium in graphite and the other forms of carbon. As to this topic, most of the different aspects of the tritium reactions with carbon were generally agreed on. At the temperature lower than 800 K, tritium plasma interacts with graphite by forming a saturated layer on the surface, by forming a codeposited layer of sputtered carbon and tritium, and by allowing tritium diffusion through Pores. At the temperature higher than 800 K, the principal reaction of tritium with carbon is intergranular diffusion with high energy trapping. Because beryllium is the reference plasma-facing material for the ITER, several presentations on the reactions of tritium with beryllium were made. Also the tritium permeation through other metals was the topics. The results of TFTR D-T experiment were reported in the first talk. In this book, the gists of these lectures are collected. (K.I.).

  13. On the generation of runaway electrons and their impact to plasma facing components

    International Nuclear Information System (INIS)

    Kawamura, Takaichi; Obayashi, Haruo; Miyahara, Akira.

    1988-06-01

    Runaway electrons accompanied by inductive or non-inductive plasma currents in a tokamak have severe interactions with plasma facing materials of a first wall, and influence the first wall structure due to activation and damage. In this paper, modelling of runaway electron generation near the wall in a tokamak is carried out. This includes the evaluation of acceleration along magnetic surfaces for relativistic electrons with energies larger than the runaway threshold. Penetration of runaway electrons of energy ranges from a few MeV to several ten MeV leads to gamma ray photon production by bremsstrahlung. One of the specific features of the impact on the first wall technology is that they give rise to activation due to giant resonance of the (γ,n) nuclear reaction and, as a consequence, cause a requirement of remote maintenance. The other is that they bring energy deposition at brazing areas between low Z material and metal, or at a metal itself, and they result in melting, cracking and grain growth. The methods to estimate these effects using nuclear data and material data on the basis of runaway flux modelling are introduced and examples of estimation are given. (author)

  14. Proceedings of 2nd International workshop on tritium effects in plasma facing components

    International Nuclear Information System (INIS)

    Morita, Kenji; Noda, Nobuaki

    1994-08-01

    This workshop was held at Nagoya University on May 19 and 20, 1994. Approximately 1/3 of the lectures discussed the migration and retention of tritium in graphite and the other forms of carbon. As to this topic, most of the different aspects of the tritium reactions with carbon were generally agreed on. At the temperature lower than 800 K, tritium plasma interacts with graphite by forming a saturated layer on the surface, by forming a codeposited layer of sputtered carbon and tritium, and by allowing tritium diffusion through Pores. At the temperature higher than 800 K, the principal reaction of tritium with carbon is intergranular diffusion with high energy trapping. Because beryllium is the reference plasma-facing material for the ITER, several presentations on the reactions of tritium with beryllium were made. Also the tritium permeation through other metals was the topics. The results of TFTR D-T experiment were reported in the first talk. In this book, the gists of these lectures are collected. (K.I.)

  15. Progress in the engineering design and assessment of the European DEMO first wall and divertor plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Barrett, Thomas R., E-mail: tom.barrett@ukaea.uk [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Ellwood, G.; Pérez, G.; Kovari, M.; Fursdon, M.; Domptail, F.; Kirk, S.; McIntosh, S.C.; Roberts, S.; Zheng, S. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Boccaccini, L.V. [KIT, INR, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); You, J.-H. [Max Planck Institute for Plasma Physics, Boltzmannstr. 2, 85748 Garching (Germany); Bachmann, C. [EUROfusion, PPPT, Boltzmann Str. 2, 85748 Garching (Germany); Reiser, J.; Rieth, M. [KIT, IAM, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Visca, E.; Mazzone, G. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Arbeiter, F. [KIT, INR, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Domalapally, P.K. [Research Center Rez, Hlavní 130, 250 68 Husinec – Řež (Czech Republic)

    2016-11-01

    Highlights: • The engineering of the plasma facing components for DEMO is an extreme challenge. • PFC overall requirements, methods for assessment and designs status are described. • Viable divertor concepts for 10 MW/m{sup 2} surface heat flux appear to be within reach. • The first wall PFC concept will need to vary poloidally around the wall. • First wall coolant, structural material and PFC topology are open design choices. - Abstract: The European DEMO power reactor is currently under conceptual design within the EUROfusion Consortium. One of the most critical activities is the engineering of the plasma-facing components (PFCs) covering the plasma chamber wall, which must operate reliably in an extreme environment of neutron irradiation and surface heat and particle flux, while also allowing sufficient neutron transmission to the tritium breeding blankets. A systems approach using advanced numerical analysis is vital to realising viable solutions for these first wall and divertor PFCs. Here, we present the system requirements and describe bespoke thermo-mechanical and thermo-hydraulic assessment procedures which have been used as tools for design. The current first wall and divertor designs are overviewed along with supporting analyses. The PFC solutions employed will necessarily vary around the wall, depending on local conditions, and must be designed in an integrated manner by analysis and physical testing.

  16. 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

    Highlights: ► We analyze the first wall blanket W/EUROFER configuration for DEMO under steady-state normal operation and off-normal conditions, such as vertical displacements events (VDE) and runaway electrons (RE). The main issue is to find the optimal thickness of the W armor which will prevent tungsten surface from evaporation and melting and, on the other hand, will keep EUROFER below the critical thermal stresses. ► 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 W surface temperature and the max. EUROFER temperature are increasing with incoming heat flux. For reference conditions (Dw ∼3 mm, DEUROFER ∼4 mm) the maximum tolerable heat flux which does not causes in thermal stresses in structural material is about ∼13.5 MW/m 2 . ► The RE deposit their energy deeper into W armour and for energies ≥50 MJ/m 2 and deposition times ≤0.1 s, the minimum armor thickness required to prevent EUROFER from thermal distraction is ≥1.4 cm. ► However, at this thickness the W surface melts. For higher RE energy deposition rates (≥100 MJ/m 2 in 0.1 s), the required armor thickness to prevent thermal destruction is even larger so that the bulk of the armor layer will melt and evaporate. - Abstract: We analyze the first wall blanket W/EUROFER configuration for DEMO under steady-state normal operation and off-normal conditions, such as vertical displacement events (VDE) and runaway electrons (RE). The main issue is to find the optimal thickness of the W armor which will prevent tungsten surface from evaporation and melting and, on the other hand, will keep EUROFER below the critical thermal stresses. Under steady-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

  17. 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

    Highlights: Black-Right-Pointing-Pointer We analyze the first wall blanket W/EUROFER configuration for DEMO under steady-state normal operation and off-normal conditions, such as vertical displacements events (VDE) and runaway electrons (RE). The main issue is to find the optimal thickness of the W armor which will prevent tungsten surface from evaporation and melting and, on the other hand, will keep EUROFER below the critical thermal stresses. Black-Right-Pointing-Pointer 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. Black-Right-Pointing-Pointer The W surface temperature and the max. EUROFER temperature are increasing with incoming heat flux. For reference conditions (Dw {approx}3 mm, DEUROFER {approx}4 mm) the maximum tolerable heat flux which does not causes in thermal stresses in structural material is about {approx}13.5 MW/m{sup 2}. Black-Right-Pointing-Pointer The RE deposit their energy deeper into W armour and for energies {>=}50 MJ/m{sup 2} and deposition times {<=}0.1 s, the minimum armor thickness required to prevent EUROFER from thermal distraction is {>=}1.4 cm. Black-Right-Pointing-Pointer However, at this thickness the W surface melts. For higher RE energy deposition rates ({>=}100 MJ/m{sup 2} in 0.1 s), the required armor thickness to prevent thermal destruction is even larger so that the bulk of the armor layer will melt and evaporate. - Abstract: We analyze the first wall blanket W/EUROFER configuration for DEMO under steady-state normal operation and off-normal conditions, such as vertical displacement events (VDE) and runaway electrons (RE). The main issue is to find the optimal thickness of the W armor which will prevent tungsten surface from evaporation and melting and, on the other hand, will keep EUROFER below the critical thermal stresses. Under steady

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

    International Nuclear Information System (INIS)

    Federici, G.; Raffray, A.R.; Chiocchio, S.; Esser, B.; Dietz, J.; Igitkhanov, Y.; Janeschitz, G.

    1995-01-01

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

  19. 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.

  20. 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.

  1. Laser-induced breakdown spectroscopy for the analysis of plasma facing components of tokamaks: parametric study and calibration-free measurements

    International Nuclear Information System (INIS)

    Mercadier, L.

    2011-09-01

    During the operation of a nuclear fusion device like the future reactor ITER, a fraction of tritium is trapped in the plasma facing components and has to be measured in order to fulfill nuclear safety requirements. Laser-induced breakdown spectroscopy (LIBS) is proposed to achieve this measurement. The laser plasma produced on carbon fibre composite tiles from the Tore Supra reactor is analyzed via a parametric study: it has to have a temperature over 10000 K and an electron density over 10 17 cm -3 to optimize the application. A calibration-free procedure that takes into account self-absorption is proposed to determine the relative concentration of hydrogen from the experimental spectra. The time- and space-resolved spectral emission of the plasma plume is investigated and reveals the presence of a temperature gradient from the core towards the periphery. This gradient is taken into account and the H/C concentration ratio is deduced. The accuracy of the results is evaluated and discussed. The study of the D/H isotopic ratio under low pressure argon reveals the presence of plume segregation that leads to an error of about 50%, error that can partially be reduced. Tungsten materials are investigated and difficulties related to spectroscopic databases are discussed. Finally, the feasibility of LIBS analysis with depth resolution is validated for multilayered metallic samples. (author)

  2. 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)

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

    Science.gov (United States)

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

    2017-11-01

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

  4. 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.

  5. Hot radial pressing: An alternative technique for the manufacturing of plasma-facing components

    International Nuclear Information System (INIS)

    Visca, E.; Libera, S.; Mancini, A.; Mazzone, G.; Pizzuto, A.; Testani, C.

    2005-01-01

    The Hot radial pressing (HRP) manufacturing technique is based on the radial diffusion bonding principle performed between the cooling tube and the armour tile. The bonding is achieved by pressurizing the cooling tube while the joining interface is kept at the vacuum and temperature conditions. This technique has been used for the manufacturing of relevant mock-ups of the ITER divertor vertical target. Tungsten monoblock mock-ups were successfully tested to high heat flux thermal fatigue (20 MW/m 2 of absorbed heat flux for 1000 cycles). After these good results the activity is now focused on the developing of a manufacturing process suitable also for the CFC monoblock mock-ups. A FE calculation was performed to investigate the stress involved in the CFC tiles during the process and to avoid the CFC fracture. The results obtained by the FE calculation and by the test performed in air simulating a HRP manufacturing process for a CFC monoblock mock-ups is reported in the paper

  6. Exploring liquid metal plasma facing component (PFC) concepts-Liquid metal film flow behavior under fusion relevant magnetic fields

    International Nuclear Information System (INIS)

    Narula, M.; Abdou, M.A.; Ying, A.; Morley, N.B.; Ni, M.; Miraghaie, R.; Burris, J.

    2006-01-01

    The use of fast moving liquid metal streams or 'liquid walls' as a plasma contact surface is a very attractive option and has been looked upon with considerable interest over the past several years, both by the plasma physics and fusion engineering programs. Flowing liquid walls provide an ever replenishing contact surface to the plasma, leading to very effective particle pumping and surface heat flux removal. A key feasibility issue for flowing liquid metal plasma facing component (PFC) systems, pertains to their magnetohydrodynamic (MHD) behavior under the spatially varying magnetic field environment, typical of a fusion device. MHD forces hinder the development of a smooth and controllable liquid metal flow needed for PFC applications. The present study builds up on the ongoing research effort at UCLA, directed towards providing qualitative and quantitative data on liquid metal free surface flow behavior under fusion relevant magnetic fields

  7. Results of high heat flux testing of W/CuCrZr multilayer composites with percolating microstructure for plasma-facing components

    International Nuclear Information System (INIS)

    Greuner, Henri; Zivelonghi, Alessandro; Böswirth, Bernd; You, Jeong-Ha

    2015-01-01

    Highlights: • Improvement of the performance of plasma-facing components made of W and CuCrZr. • Functionally graded composite at the interface of W and CuCrZr to mitigate the CTE. • A three-layer composite system (W volume fraction: 70/50/30%) was developed. • Design of water-cooled divertor components up to 20 MW/m"2 heat load for e.g. DEMO. • HHF tests up to 20 MW/m"2 were successfully performed. - Abstract: Reliable joining of tungsten to copper is a major issue in the design of water-cooled divertor components for future fusion reactors. One of the suggested advanced engineering solutions is to use functionally graded composite interlayers. Recently, the authors have developed a novel processing route for fabricating multi-layer graded W/CuCrZr composites. Previous characterization confirmed that the composite materials possess enhanced strength compared to the matrix alloy and shows reasonable ductility up to 300 °C indicating large potential to extend the operation temperature limit. Furthermore, a three-layer composite system (W volume fraction: 70/50/30%) was developed as a graded interlayer between the W armour and CuCrZr heat sink. In this study, we investigated the structural performance of the graded joint. Three water-cooled mock-ups of a flat tile type component were fabricated using electron beam welding and thermally loaded at the hydrogen neutral beam test facility GLADIS. Cycling tests at 10 MW/m"2 and screening tests up to 20 MW/m"2 were successfully performed and confirmed the expected thermal performance of the compound. The measured temperature values were in good agreement with the prediction of finite element analysis. Microscopic investigation confirmed the structural integrity of the newly developed functionally graded composite after these tests.

  8. Results of high heat flux testing of W/CuCrZr multilayer composites with percolating microstructure for plasma-facing components

    Energy Technology Data Exchange (ETDEWEB)

    Greuner, Henri, E-mail: henri.greuner@ipp.mpg.de; Zivelonghi, Alessandro; Böswirth, Bernd; You, Jeong-Ha

    2015-10-15

    Highlights: • Improvement of the performance of plasma-facing components made of W and CuCrZr. • Functionally graded composite at the interface of W and CuCrZr to mitigate the CTE. • A three-layer composite system (W volume fraction: 70/50/30%) was developed. • Design of water-cooled divertor components up to 20 MW/m{sup 2} heat load for e.g. DEMO. • HHF tests up to 20 MW/m{sup 2} were successfully performed. - Abstract: Reliable joining of tungsten to copper is a major issue in the design of water-cooled divertor components for future fusion reactors. One of the suggested advanced engineering solutions is to use functionally graded composite interlayers. Recently, the authors have developed a novel processing route for fabricating multi-layer graded W/CuCrZr composites. Previous characterization confirmed that the composite materials possess enhanced strength compared to the matrix alloy and shows reasonable ductility up to 300 °C indicating large potential to extend the operation temperature limit. Furthermore, a three-layer composite system (W volume fraction: 70/50/30%) was developed as a graded interlayer between the W armour and CuCrZr heat sink. In this study, we investigated the structural performance of the graded joint. Three water-cooled mock-ups of a flat tile type component were fabricated using electron beam welding and thermally loaded at the hydrogen neutral beam test facility GLADIS. Cycling tests at 10 MW/m{sup 2} and screening tests up to 20 MW/m{sup 2} were successfully performed and confirmed the expected thermal performance of the compound. The measured temperature values were in good agreement with the prediction of finite element analysis. Microscopic investigation confirmed the structural integrity of the newly developed functionally graded composite after these tests.

  9. Data combination of infrared thermography images and lock-in thermography images for NDE of plasma facing components

    International Nuclear Information System (INIS)

    Moysan, J.; Gueudre, C.; Corneloup, G.; Durocher, A.

    2006-01-01

    A pioneering activity has been developed by CEA and the European industry in the field of actively cooled high heat flux plasma facing components (PFC) from the very beginning of Tore Supra project. These components have been developed in order to enable a large power exhaust capability. The goal of this study is to improve the Non Destructive Evaluation (NDE) of these components. The difficulty encountered is the evaluation of the junction between a carbon and a metallic substrate. This was even more difficult when complex designs have to be implemented. A first NDE solution was based on the so called SATIR test. The method is based on infrared measurements of tile surface temperatures during a thermal transient produced by hot/cold water flowing in the heat sink cooling channel. In order to improve the definition of acceptance rules for the PFCs, a second NDE method based on Lock-in Thermography is developed. In this work we present how we can combine the two resulting images in order to accept or to reject a component. This prospective study allows improving the experimental setup and the definition of acceptance criteria. The experimental study was conducted on trial components for the Wendelstein 7X stellarator. The conclusions will also influence future non destructive projects dedicated to the ITER project. (orig.)

  10. 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.

  11. Beryllium processing technology review for applications in plasma-facing components

    International Nuclear Information System (INIS)

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

    1993-07-01

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

  12. Beryllium processing technology review for applications in plasma-facing components

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-07-01

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

  13. Development of an original active thermography method adapted to ITER plasma facing components control

    Energy Technology Data Exchange (ETDEWEB)

    Durocher, A.; Vignal, N.; Escourbiac, F.; Farjon, J.L.; Schlosser, J. [CEA Cadarache, Dept. de Recherches sur la Fusion Controlee, 13 - Saint-Paul-lez-Durance (France); Cismondi, F. [Toulon Univ., 83 - La Garde (France)

    2004-07-01

    Among all Non-Destructive Examinations (NDE), active infrared thermography is becoming recognised as a technique available today for improving quality control of many materials and structures involved in heat transfer. The infrared thermography allows to characterise the bond between two materials having different thermal physical properties. In order to increase the defect detection limit of the SATIR test bed, several possibilities have been evaluated to improve the infrared thermography inspection. The implementation in 2003 of a micro-bolometer camera and the improving of the thermo-signal process allowed to increase considerably the detection sensitivity of the SATIR facility. The quality, the spatial stability of infrared image and the detection of edge defect have been also improved. The coupling on the same test bed of SATIR method with a lock-in thermography will be evaluated in this paper. An improvement of the global reliability is expected by data merging produced by the two thermal excitation sources. A new enhanced facility named SATIRPACA has been designed for the full Non Destructive Examination of the High Heat Flux ITER components taking into account these main improvements. These systematic acceptance tests obviously need tools for quality control of critical parts. (authors)

  14. Development of an original active thermography method adapted to ITER plasma facing components control

    International Nuclear Information System (INIS)

    Durocher, A.; Vignal, N.; Escourbiac, F.; Farjon, J.L.; Schlosser, J.; Cismondi, F.

    2004-01-01

    Among all Non-Destructive Examinations (NDE), active infrared thermography is becoming recognised as a technique available today for improving quality control of many materials and structures involved in heat transfer. The infrared thermography allows to characterise the bond between two materials having different thermal physical properties. In order to increase the defect detection limit of the SATIR test bed, several possibilities have been evaluated to improve the infrared thermography inspection. The implementation in 2003 of a micro-bolometer camera and the improving of the thermo-signal process allowed to increase considerably the detection sensitivity of the SATIR facility. The quality, the spatial stability of infrared image and the detection of edge defect have been also improved. The coupling on the same test bed of SATIR method with a lock-in thermography will be evaluated in this paper. An improvement of the global reliability is expected by data merging produced by the two thermal excitation sources. A new enhanced facility named SATIRPACA has been designed for the full Non Destructive Examination of the High Heat Flux ITER components taking into account these main improvements. These systematic acceptance tests obviously need tools for quality control of critical parts. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Rubel, Marek, E-mail: Marek.Rubel@ee.kth.se [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Fusion Plasma Physics, Royal Institute of Technology (KTH), 100 44 Stockholm (Sweden); Petersson, Per [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Fusion Plasma Physics, Royal Institute of Technology (KTH), 100 44 Stockholm (Sweden); Alves, Eduardo [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Instituto Superior Técnico, Universidade de Lisboa, 1049-001 Lisbon (Portugal); Brezinsek, Sebastijan [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Institut für Klima- und Energieforschung, Forschungszentrum Jülich, D-52425 Jülich (Germany); Coad, Joseph Paul [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Heinola, Kalle [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); University of Helsinki, 00014 Helsinki (Finland); Mayer, Matej [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Max-Planck-Institut für Plasmaphysik, 85478 Garching (Germany); Widdowson, Anna [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)

    2016-03-15

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

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

    Science.gov (United States)

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

    2016-03-01

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

  17. Design of Plasma Facing Components for Superconducting Modification of JT-60

    International Nuclear Information System (INIS)

    Shinji Sakurai; Kei Masaki; Yusuke-Kudo Shibama; Hiroshi Tamai; Makoto Matsukawa; Cordier, J.J.

    2006-01-01

    remote handling capability for in-vessel components should be required due to the increase in the neutron budget by an order of magnitude with respect to the original design. Upper and lower divertor cassettes and inboard first wall units should be designed to be exchangeable by the ITER-like remote handling system. Design modification for the increase of heating power and neutron budget will be completed in the end of 2006 under the conceptual design activity in the collaboration with EU and Japan. (author)

  18. Final Report: Safety of Plasma-Facing Components and Aerosol Transport During Hard Disruptions and Accidental Energy Release in Fusion Reactor

    International Nuclear Information System (INIS)

    Bourham, Mohamed A.; Gilligan, John G.

    1999-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jaworski, M.A., E-mail: mjaworsk@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Brooks, A.; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Lopes-Cardozo, N. [TU/Eindhoven, Eindhoven (Netherlands); Menard, J.; Ono, M. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Rindt, P. [TU/Eindhoven, Eindhoven (Netherlands); Tresemer, K. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

    2016-11-15

    Highlights: • An upgrade path for the NSTX-U tokamak is proposed that maintains scientific productivity while enabling exploration of novel, liquid metal PFC. • Pre-filled liquid metal divertor targets are proposed as an intermediate step that mitigates technical and scientific risks associated with liquid metal PFC. • Analysis of leading edge features show a strong link between engineering design considerations and expected performance as a PFC. • A method for optimizing porous liquid metal targets restrained by capillary forces is provided indicating pore-sizes well within current technical capabilities. - Abstract: Liquid metal plasma-facing components (PFCs) provide numerous potential advantages over solid-material components. One critique of the approach is the relatively less developed technologies associated with deploying these components in a fusion plasma-experiment. Exploration of the temperature limits of liquid lithium PFCs in a tokamak divertor and the corresponding consequences on core operation are a high priority informing the possibilities for future liquid lithium PFCs. An all-metal NSTX-U is envisioned to make direct comparison between all high-Z wall operation and liquid lithium PFCs in a single device. By executing the all-metal upgrades incrementally, scientific productivity will be maintained while enabling physics and engineering-science studies to further develop the solid- and liquid-metal components. Six major elements of a flowing liquid-metal divertor system are described and a three-step program for implementing this system is laid out. The upgrade steps involve the first high-Z divertor target upgrade in NSTX-U, pre-filled liquid metal targets and finally, an integrated, flowing liquid metal divertor target. Two example issues are described where the engineering and physics experiments are shown to be closely related in examining the prospects for future liquid metal PFCs.

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

    Science.gov (United States)

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

    2016-02-01

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

  1. Lithium Coatings on NSTX Plasma Facing Components and Its Effects On Boundary Control, Core Plasma Performance, and Operation

    Energy Technology Data Exchange (ETDEWEB)

    H.W.Kugel, M.G.Bell, H.Schneider, J.P.Allain, R.E.Bell, R Kaita, J.Kallman, S. Kaye, B.P. LeBlanc, D. Mansfield, R.E. Nygen, R. Maingi, J. Menard, D. Mueller, M. Ono, S. Paul, S.Gerhardt, R.Raman, S.Sabbagh, C.H.Skinner, V.Soukhanovskii, J.Timberlake, L.E.Zakharov, and the NSTX Research Team

    2010-01-25

    NSTX high-power divertor plasma experiments have used in succession lithium pellet injection (LPI), evaporated lithium, and injected lithium powder to apply lithium coatings to graphite plasma facing components. In 2005, following wall conditioning and LPI, discharges exhibited edge density reduction and performance improvements. Since 2006, first one, and now two lithium evaporators have been used routinely to evaporate lithium onto the lower divertor region at total rates of 10-70 mg/min for periods 5-10 min between discharges. Prior to each discharge, the evaporators are withdrawn behind shutters. Significant improvements in the performance of NBI heated divertor discharges resulting from these lithium depositions were observed. These evaporators are now used for more than 80% of NSTX discharges. Initial work with injecting fine lithium powder into the edge of NBI heated deuterium discharges yielded comparable changes in performance. Several operational issues encountered with lithium wall conditions, and the special procedures needed for vessel entry are discussed. The next step in this work is installation of a Liquid Lithium Divertor surface on the outer part of the lower divertor.

  2. Lithium coatings on NSTX plasma facing components and its effects on boundary control, core plasma performance, and operation

    Energy Technology Data Exchange (ETDEWEB)

    Kugel, H.W., E-mail: hkugel@pppl.gov [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Bell, M.G.; Schneider, H. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Allain, J.P. [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47907 (United States); Bell, R.E.; Kaita, R.; Kallman, J.; Kaye, S.; LeBlanc, B.P.; Mansfield, D. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Nygren, R.E. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Maingi, R. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Menard, J.; Mueller, D.; Ono, M.; Paul, S.; Gerhardt, S. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Raman, R. [University of Washington, Seattle, WA 98195 (United States); Sabbagh, S. [Columbia University, New York, NY 10027 (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States)

    2010-11-15

    NSTX high power divertor plasma experiments have used in succession lithium pellet injection (LPI), evaporated lithium, and injected lithium powder to apply lithium coatings to graphite plasma facing components. In 2005, following the wall conditioning and LPI, discharges exhibited edge density reduction and performance improvements. Since 2006, first one, and now two lithium evaporators have been used routinely to evaporate lithium onto the lower divertor region at total rates of 10-70 mg/min for periods 5-10 min between discharges. Prior to each discharge, the evaporators are withdrawn behind shutters. Significant improvements in the performance of NBI heated divertor discharges resulting from these lithium depositions were observed. These evaporators are now used for more than 80% of NSTX discharges. Initial work with injecting fine lithium powder into the edge of NBI heated deuterium discharges yielded comparable changes in performance. Several operational issues encountered with lithium wall conditions, and the special procedures needed for vessel entry are discussed. The next step in this work is installation of a liquid lithium divertor surface on the outer part of the lower divertor.

  3. Critical plasma-wall interaction issues for plasma-facing materials and components in near-term fusion devices

    International Nuclear Information System (INIS)

    Federici, G.; Coad, J.P.; Haasz, A.A.; Janeschitz, G.; Noda, N.; Philipps, V.; Roth, J.; Skinner, C.H.; Tivey, R.; Wu, C.H.

    2000-01-01

    The increase in pulse duration and cumulative run-time, together with the increase of the plasma energy content, will represent the largest changes in operation conditions in future fusion devices such as the International Thermonuclear Experimental Reactor (ITER) compared to today's experimental facilities. These will give rise to important plasma-physics effects and plasma-material interactions (PMIs) which are only partially observed and accessible in present-day experiments and will open new design, operation and safety issues. For the first time in fusion research, erosion and its consequences over many pulses (e.g., co-deposition and dust) may determine the operational schedule of a fusion device. This paper identifies the most critical issues arising from PMIs which represent key elements in the selection of materials, the design, and the optimisation of plasma-facing components (PFCs) for the first-wall and divertor. Significant advances in the knowledge base have been made recently, as part of the R and D supporting the engineering design activities (EDA) of ITER, and some of the most relevant data are reviewed here together with areas where further R and D work is urgently needed

  4. Damage prediction of carbon fibre composite armoured actively cooled plasma-facing components under cycling heat loads

    International Nuclear Information System (INIS)

    Chevet, G; Schlosser, J; Courtois, X; Escourbiac, F; Missirlian, M; Herb, V; Martin, E; Camus, G; Braccini, M

    2009-01-01

    In order to predict the lifetime of carbon fibre composite (CFC) armoured plasma-facing components in magnetic fusion devices, it is necessary to analyse the damage mechanisms and to model the damage propagation under cycling heat loads. At Tore Supra studies have been launched to better understand the damage process of the armoured flat tile elements of the actively cooled toroidal pump limiter, leading to the characterization of the damageable mechanical behaviour of the used N11 CFC material and of the CFC/Cu bond. Up until now the calculations have shown damage developing in the CFC (within the zone submitted to high shear stress) and in the bond (from the free edge of the CFC/Cu interface). Damage is due to manufacturing shear stresses and does not evolve under heat due to stress relaxation. For the ITER divertor, NB31 material has been characterized and the characterization of NB41 is in progress. Finite element calculations show again the development of CFC damage in the high shear stress zones after manufacturing. Stresses also decrease under heat flux so the damage does not evolve. The characterization of the CFC/Cu bond is more complex due to the monoblock geometry, which leads to more scattered stresses. These calculations allow the fabrication difficulties to be better understood and will help to analyse future high heat flux tests on various mock-ups.

  5. 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.)

  6. Lithium Coatings on NSTX Plasma Facing Components and Its Effects On Boundary Control, Core Plasma Performance, and Operation

    International Nuclear Information System (INIS)

    Kugel, H.W.; Bell, M.G.; Schneider, H.; Allain, J.P.; Bell, R.E.; Kaita, R.; Kallman, J.; Kaye, S.; LeBlanc, B.P.; Mansfield, D.; Nygen, R.E.; Maingi, R.; Menard, J.; Mueller, D.; Ono, M.; Paul, S.; Gerhardt, S.; Raman, R.; Sabbagh, S.; Skinner, C.H.; Soukhanovskii, V.; Timberlake, J.; Zakharov, L.E.; NSTX Research Team

    2010-01-01

    NSTX high-power divertor plasma experiments have used in succession lithium pellet injection (LPI), evaporated lithium, and injected lithium powder to apply lithium coatings to graphite plasma facing components. In 2005, following wall conditioning and LPI, discharges exhibited edge density reduction and performance improvements. Since 2006, first one, and now two lithium evaporators have been used routinely to evaporate lithium onto the lower divertor region at total rates of 10-70 mg/min for periods 5-10 min between discharges. Prior to each discharge, the evaporators are withdrawn behind shutters. Significant improvements in the performance of NBI heated divertor discharges resulting from these lithium depositions were observed. These evaporators are now used for more than 80% of NSTX discharges. Initial work with injecting fine lithium powder into the edge of NBI heated deuterium discharges yielded comparable changes in performance. Several operational issues encountered with lithium wall conditions, and the special procedures needed for vessel entry are discussed. The next step in this work is installation of a Liquid Lithium Divertor surface on the outer part of the lower divertor.

  7. Leak tightness tests on actively cooled plasma facing components: Lessons learned from Tore Supra experience and perspectives for the new fusion machines

    Energy Technology Data Exchange (ETDEWEB)

    Chantant, M., E-mail: michel.chantant@cea.fr; Lambert, R.; Gargiulo, L.; Hatchressian, J.-C.; Guilhem, D.; Samaille, F.; Soler, B.

    2015-10-15

    Highlights: • Test procedures for the qualification of the tightness of actively cooled plasma facing components were defined. • The test is performed after the component manufacturing and before its set-up in the vacuum vessel. • It allows improving the fusion machine availability. • The lessons of tests over 20 years at Tore Supra are presented. - Abstract: The fusion machines under development or construction (ITER, W7X) use several hundreds of actively cooled plasma facing components (ACPFC). They are submitted to leak tightness requirements in order to get an appropriate vacuum level in the vessel to create the plasma. During the ACPFC manufacturing and before their installation in the machine, their leak tightness performance must be measured to check that they fulfill the vacuum requirements. A relevant procedure is needed which allows to segregate potential defects. It must also be optimized in terms of test duration and costs. Tore Supra, as an actively cooled Tokamak, experienced several leaks on ACPFCs during the commissioning and during the operation of the machine. A test procedure was then defined and several test facilities were set-up. Since 1990 the tightness of all the new ACPFCs is systematically tested before their installation in Tore Supra. During the qualification test, the component is set up in a vacuum test tank, and its cooling circuits are pressurized with helium. It is submitted to 3 temperature cycles from room temperature up to the baking temperature level in Tore Supra (200 °C) and two pressurization tests are performed (6 MPa at room temperature and 4 MPa at 200 °C) at each stage. At the end of the last cycle when the ACPFC is at room temperature and pressurized with helium at 6 MPa, the measured leak rate must be lower than 5 × 10{sup −11} Pa m{sup 3} s{sup −1}, the pressure in the test tank being <5 × 10{sup −5} Pa. A large experience has been gained on ACPFCs with carbon parts on stainless steel and Cu

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

    Science.gov (United States)

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

    2017-11-01

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

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

    Science.gov (United States)

    Genco, Filippo

    Damage to plasma-facing components (PFC) due to various plasma instabilities is still a major concern for the successful development of fusion energy and represents a significant research obstacle in the community. It is of great importance to fully understand the behavior and lifetime expectancy of PFC under both low energy cycles during normal events and highly energetic events as disruptions, Edge-Localized Modes (ELM), Vertical Displacement Events (VDE), and Run-away electron (RE). The consequences of these high energetic dumps with energy fluxes ranging from 10 MJ/m2 up to 200 MJ/m 2 applied in very short periods (0.1 to 5 ms) can be catastrophic both for safety and economic reasons. Those phenomena can cause a) large temperature increase in the target material b) consequent melting, evaporation and erosion losses due to the extremely high heat fluxes c) possible structural damage and permanent degradation of the entire bulk material with probable burnout of the coolant tubes; d) plasma contamination, transport of target material into the chamber far from where it was originally picked. The modeling of off-normal events such as Disruptions and ELMs requires the simultaneous solution of three main problems along time: a) the heat transfer in the plasma facing component b) the interaction of the produced vapor from the surface with the incoming plasma particles c) the transport of the radiation produced in the vapor-plasma cloud. In addition the moving boundaries problem has to be considered and solved at the material surface. Considering the carbon divertor as target, the moving boundaries are two since for the given conditions, carbon doesn't melt: the plasma front and the moving eroded material surface. The current solution methods for this problem use finite differences and moving coordinates system based on the Crank-Nicholson method and Alternating Directions Implicit Method (ADI). Currently Particle-In-Cell (PIC) methods are widely used for solving

  10. Experimental results of near real-time protection system for plasma facing components in Wendelstein 7-X at GLADIS

    Science.gov (United States)

    Ali, A.; Jakubowski, M.; Greuner, H.; Böswirth, B.; Moncada, V.; Sitjes, A. Puig; Neu, R.; Pedersen, T. S.; the W7-X Team

    2017-12-01

    One of the aims of stellarator Wendelstein 7-X (W7-X), is to investigate steady state operation, for which power exhaust is an important issue. The predominant fraction of the energy lost from the confined plasma region will be absorbed by an island divertors, which is designed for 10 {{MWm}}-2 steady state operation. In order to protect the divertor targets from overheating, 10 state-of-the-art infrared endoscopes will be installed at W7-X. In this work, we present the experimental results obtained at the high heat flux test facility GLADIS (Garching LArge DIvertor Sample test facility in IPP Garching) [1] during tests of a new plasma facing components (PFCs) protection algorithm designed for W7-X. The GLADIS device is equipped with two ion beams that can generate a heat load in the range from 3 MWm-2 to 55 MWm-2. The algorithms developed at W7-X to detect defects and hot spots are based on the analysis of surface temperature evolution and are adapted to work in near real-time. The aim of this work was to test the near real-time algorithms in conditions close to those expected in W7-X. The experiments were performed on W7-X pre-series tiles to detect CFC/Cu delaminations. For detection of surface layers, carbon fiber composite (CFC) blocks from the divertor of the Wendelstein 7-AS stellarator were used to observe temporal behavior of fully developed surface layers. These layers of re-deposited materials, like carbon, boron, oxygen and iron, were formed during the W7-AS operation. A detailed analysis of the composition and their thermal response to high heat fluxes (HHF) are described in [2]. The experiments indicate that the automatic detection of critical events works according to W7-X PFC protection requirements.

  11. Heat transfer characteristics of rectangular coolant channels with various aspect ratios in the plasma-facing components under fully developed MHD laminar flow

    International Nuclear Information System (INIS)

    Takase, K.; Hasan, M.Z.

    1995-01-01

    Convective heat transfer in MHD laminar flow through rectangular channels in the plasma-facing components of a fusion reactor has been analyzed numerically to investigate the effects of channel aspect ratio, defined as the ratio of the lengths of the plasma-facing side to the other side. The adverse effect of the nonuniformity of surface heat flus on Nusselt number (Nu) at the plasma-facing side can be alleviated by increasing the aspect ratio of a rectangular duct. At the center and corner of the plasma-facing side of a square duct, the Nu of non-MHD flow are 6.8 and 2.2, respectively, for uniform surface heat flux. In the presence of a strong magnetic field, Nu at the center and corner increases to 22 and 3.6, respectively. However, when the heat flux is highly nonuniform, as in the plasma-facing components, Nu decreases from 22 to 3.1 at the center and from 3.6 to 3.1 at the corner. When the aspect ratio is increased to 4, Nu at the center and corner increase to 5 and 4.7. Along the circumference of a rectangular channel, there are locations where the wall temperature is equal to or less than the bulk coolant temperature, thus making the Nu with conventional definition infinity or negative. The ratio between Nu of MHD flow and Nu of non-MHD flow for various aspect ratios is constant in the region of Hartmann number of more than 200 at least. On the other hand, its ratio increases monotonously with increasing the aspect ratio

  12. 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

  13. 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

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

    Science.gov (United States)

    Jacob, Wolfgang; Linsmeier, Christian; Rubel, Marek

    2011-12-01

    subjected to extremely high heat loads and very high particle and neutron fluxes. They must have high thermal conductivity for efficient heat transport, high cohesive energy for low erosion by particle bombardment and low atomic number to minimize plasma cooling. These contradictory requirements make the development of plasma-facing materials one of the greatest challenges ever faced by materials scientists. The erosion of plasma-facing materials is one of the main factors influencing the operational schedule of experimental fusion reactors and future power plants. A number of materials selected for current designs cannot withstand the presently foreseen plasma scenarios of a power plant for a commercially viable period of time. Therefore, further coordinated development of plasma scenarios and materials is essential for the realization of fusion as an energy source. The design and development of plasma-facing materials requires a detailed understanding of the processes that occur when a material surface is bombarded with an intense flux of heat, particles and neutrons simultaneously. These materials-related topics are the focus of this series of workshops which has established itself as a discussion forum for experts from research institutions and industry dealing with materials for plasma-facing components in present and future thermonuclear fusion devices. During the joint conference PFMC-13/FEMaS-1 recent developments and research results in the following fields were addressed: carbon, beryllium, and tungsten based materials mixed materials erosion and redeposition high heat flux component development benchmarking of radiation damage modelling synchrotron and neutron based characterization techniques application of advanced transmission electron microscopy and micro-/nano-mechanical testing. With the approaching technical realization of ITER, the ITER-related PFMC topics are naturally the main focus of research. In this respect the start of the ITER-like wall experiment

  15. 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

  16. 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

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

    Science.gov (United States)

    Raffray, A. René; Federici, Gianfranco

    1997-04-01

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

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

    International Nuclear Information System (INIS)

    Raffray, A.R.; Federici, G.

    1997-01-01

    For pt.II see ibid., p.101-30, 1997. RACLETTE (Rate Analysis Code for pLasma Energy Transfer Transient Evaluation), a comprehensive but relatively simple and versatile model, was developed to help in the design analysis of plasma facing components (PFCs) under 'slow' high power transients, such as those associated with plasma vertical displacement events. The model includes all the key surface heat transfer processes such as evaporation, melting, and radiation, and their interaction with the PFC block thermal response and the coolant behaviour. This paper represents part I of two sister and complementary papers. It covers the model description, calibration and validation, and presents a number of parametric analyses shedding light on and identifying trends in the PFC armour block response to high plasma energy deposition transients. Parameters investigated include the plasma energy density and deposition time, the armour thickness and the presence of vapour shielding effects. Part II of the paper focuses on specific design analyses of ITER plasma facing components (divertor, limiter, primary first wall and baffle), including improvements in the thermal-hydraulic modeling required for better understanding the consequences of high energy deposition transients in particular for the ITER limiter case. (orig.)

  19. Use of EPICS and Python technology for the development of a computational toolkit for high heat flux testing of plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Sugandhi, Ritesh, E-mail: ritesh@ipr.res.in; Swamy, Rajamannar, E-mail: rajamannar@ipr.res.in; Khirwadkar, Samir, E-mail: sameer@ipr.res.in

    2016-11-15

    Highlights: • An integrated approach to software development for computational processing and experimental control. • Use of open source, cross platform, robust and advanced tools for computational code development. • Prediction of optimized process parameters for critical heat flux model. • Virtual experimentation for high heat flux testing of plasma facing components. - Abstract: The high heat flux testing and characterization of the divertor and first wall components are a challenging engineering problem of a tokamak. These components are subject to steady state and transient heat load of high magnitude. Therefore, the accurate prediction and control of the cooling parameters is crucial to prevent burnout. The prediction of the cooling parameters is based on the numerical solution of the critical heat flux (CHF) model. In a test facility for high heat flux testing of plasma facing components (PFC), the integration of computations and experimental control is an essential requirement. Experimental physics and industrial control system (EPICS) provides powerful tools for steering controls, data simulation, hardware interfacing and wider usability. Python provides an open source alternative for numerical computations and scripting. We have integrated these two open source technologies to develop a graphical software for a typical high heat flux experiment. The implementation uses EPICS based tools namely IOC (I/O controller) server, control system studio (CSS) and Python based tools namely Numpy, Scipy, Matplotlib and NOSE. EPICS and Python are integrated using PyEpics library. This toolkit is currently under operation at high heat flux test facility at Institute for Plasma Research (IPR) and is also useful for the experimental labs working in the similar research areas. The paper reports the software architectural design, implementation tools and rationale for their selection, test and validation.

  20. Use of EPICS and Python technology for the development of a computational toolkit for high heat flux testing of plasma facing components

    International Nuclear Information System (INIS)

    Sugandhi, Ritesh; Swamy, Rajamannar; Khirwadkar, Samir

    2016-01-01

    Highlights: • An integrated approach to software development for computational processing and experimental control. • Use of open source, cross platform, robust and advanced tools for computational code development. • Prediction of optimized process parameters for critical heat flux model. • Virtual experimentation for high heat flux testing of plasma facing components. - Abstract: The high heat flux testing and characterization of the divertor and first wall components are a challenging engineering problem of a tokamak. These components are subject to steady state and transient heat load of high magnitude. Therefore, the accurate prediction and control of the cooling parameters is crucial to prevent burnout. The prediction of the cooling parameters is based on the numerical solution of the critical heat flux (CHF) model. In a test facility for high heat flux testing of plasma facing components (PFC), the integration of computations and experimental control is an essential requirement. Experimental physics and industrial control system (EPICS) provides powerful tools for steering controls, data simulation, hardware interfacing and wider usability. Python provides an open source alternative for numerical computations and scripting. We have integrated these two open source technologies to develop a graphical software for a typical high heat flux experiment. The implementation uses EPICS based tools namely IOC (I/O controller) server, control system studio (CSS) and Python based tools namely Numpy, Scipy, Matplotlib and NOSE. EPICS and Python are integrated using PyEpics library. This toolkit is currently under operation at high heat flux test facility at Institute for Plasma Research (IPR) and is also useful for the experimental labs working in the similar research areas. The paper reports the software architectural design, implementation tools and rationale for their selection, test and validation.

  1. Design, fabrication and testing of an improved high heat flux element, experience feedback on steady state plasma facing components in Tore Supra

    International Nuclear Information System (INIS)

    Schlosser, J.; Chappuis, P.; Chatelier, M.; Durocher, A.; Guilheim, D.; Lipa, M.; Mitteau, R.; Tonon, G.; Tsitrone, E.

    1998-01-01

    Actively cooled plasma facing components (PFC) have been developed and used in Tore Supra since 1985. One of the main technological problem is due to the expansion mismatch between graphite armour and metallic heat sink material. A first technology used graphite tiles with or without a reinforcement and a compliant layer, brazed with titanium copper-silver (TiCuAg) alloy. The next technology used carbon fiber material (CFC) tiles with a 2 mm pure copper compliant layer, since the good mechanical strength of the CFC allowed the reinforcement layer to be suppressed. No destructive inspection during the manufacturing procedure was found to be essential to insure a good reliability of the elements. (orig.)

  2. Feasibility of arc-discharge and plasma-sputtering methods in cleaning plasma-facing and diagnostics components of fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hakola, Antti, E-mail: antti.hakola@vtt.fi [VTT Technical Research Centre of Finland, VTT (Finland); Likonen, Jari [VTT Technical Research Centre of Finland, VTT (Finland); Karhunen, Juuso; Korhonen, Juuso T. [Department of Applied Physics, Aalto University (Finland); Aints, Märt; Laan, Matti; Paris, Peeter [Department of Physics, University of Tartu (Estonia); Kolehmainen, Jukka; Koskinen, Mika; Tervakangas, Sanna [DIARC-Technology Oy, Espoo (Finland)

    2015-10-15

    Highlights: • Feasibility of the arc-discharge and plasma-sputtering techniques in removing deposited layers from ITER-relevant samples demonstrated. • Samples with the size of an A4 paper can be cleaned from 1-μm thick deposited layers in 10–20 minutes by the arc-discharge method. • The plasma-sputtering method is 5–10 times slower but the resulting surfaces are very smooth. • Arc-discharge method could be used for rapid cleaning of plasma-facing components during maintenance shutdowns of ITER, plasma sputtering is preferred for diagnostics mirrors. - Abstract: We have studied the feasibility of arc-discharge and plasma-sputtering methods in removing deposited layers from ITER-relevant test samples. Prototype devices have been designed and constructed for the experiments and the cleaning process is monitored by a spectral detection system. The present version of the arc-discharge device is capable of removing 1-μm thick layers from 350-mm{sup 2} areas in 4–8 s, but due to the increased roughness of the cleaned surfaces and signs of local melting, mirror-like surfaces cannot be treated by this technique. The plasma-sputtering approach, for its part, is some 5–10 times slower in removing the deposited layers but no changes in surface roughness or morphology of the samples could be observed after the cleaning phase. The arc-discharge technique could therefore be used for rapid cleaning of plasma-facing components during maintenance shutdowns of ITER while in the case of diagnostics mirrors plasma sputtering is preferred.

  3. Study of heat fluxes on plasma facing components in a tokamak from measurements of temperature by infrared thermography

    International Nuclear Information System (INIS)

    Daviot, R.

    2010-05-01

    The goal of this thesis is the development of a method of computation of those heat loads from measurements of temperature by infrared thermography. The research was conducted on three issues arising in current tokamaks but also future ones like ITER: the measurement of temperature on reflecting walls, the determination of thermal properties for deposits observed on the surface of tokamak components and the development of a three-dimensional, non-linear computation of heat loads. A comparison of several means of pyrometry, monochromatic, bi-chromatic and photothermal, is performed on an experiment of temperature measurement. We show that this measurement is sensitive to temperature gradients on the observed area. Layers resulting from carbon deposition by the plasma on the surface of components are modeled through a field of equivalent thermal resistance, without thermal inertia. The field of this resistance is determined, for each measurement points, from a comparison of surface temperature from infrared thermographs with the result of a simulation, which is based on a mono-dimensional linear model of components. The spatial distribution of the deposit on the component surface is obtained. Finally, a three-dimensional and non-linear computation of fields of heat fluxes, based on a finite element method, is developed here. Exact geometries of the component are used. The sensitivity of the computed heat fluxes is discussed regarding the accuracy of the temperature measurements. This computation is applied to two-dimensional temperature measurements of the JET tokamak. Several components of this tokamak are modeled, such as tiles of the divertor, upper limiter and inner and outer poloidal limiters. The distribution of heat fluxes on the surface of these components is computed and studied along the two main tokamak directions, poloidal and toroidal. Toroidal symmetry of the heat loads from one tile to another is shown. The influence of measurements spatial resolution

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

    International Nuclear Information System (INIS)

    Di Pietro, E.; Visca, E.; Orsini, A.; Sacchetti, M.; Borruto, T.M.R.; Varone, P.; Vesprini, R.

    1995-01-01

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

  5. The Plasma-Facing Components Transporter (PFCT) : a Prototype System for PFC Replacement on the new ITER 2001 Cassette Mock-up

    International Nuclear Information System (INIS)

    Micciche, G.; Lorenzelli, L.; Muro, L.; Irving, M.

    2006-01-01

    The remote maintainability of the early ITER divertor cassette (based on the ITER 1998 design) was successfully proved during test campaigns carried out in the Divertor Refurbishment Platform (DRP) at the ENEA research centre at Brasimone over the period 1999-2003. Due to subsequent major modifications in the ITER divertor cassette design, the main focus over the past few years has been on the design and manufacture of the various components, devices and tools needed for refurbishment of the new ITER 2001 Divertor Cassette. The design of this new cassette differs substantially from the earlier version: in particular the shape, weight and attachment system of the Plasma Facing Components (PFC's) has been completely revised, and this also entailed a review of the procedures adopted for its refurbishment. One of the major requirements of the cassette refurbishment process is removal and replacement of the three PFC's. In the old cassette concept, target replacement was performed by means of a purpose-built '' C '' frame slung from a standard bridge crane. The 2001 cassette design precludes such handling methods for a number of reasons, notably because of the extremely tight inter-PFC clearances, and the need for controlled inclination of the target in addition to normal translational movements, both impossible with a simple Cartesian crane. To demonstrate the refurbishment feasibility operations for the new ITER Divertor 2001 cassettes, an experimental machine known as the Plasma-Facing Component Transporter (PFCT) has been designed, fabricated and commissioned in the years 2004-5. This full six degree-of-freedom system has been designed to handle payloads of up to 5 tonnes with good positional accuracy, and axes capable of very low joint velocities, including inclination of the PFC's over the range of ± 10 o in both horizontal axes, and controlled rotation about the vertical axis. Preliminary trials carried out during the commissioning phase have proved its

  6. A mature industrial solution for ITER divertor plasma facing components: hypervapotron cooling concept adapted to Tore Supra flat tile technology

    Energy Technology Data Exchange (ETDEWEB)

    Escourbiac, F.; Missirlian, M.; Schlosser, J. [Association EURATOM-CEA Cadarache, Departement de Recherches sur la Fusion Controlee, 13 - Saint Paul lez Durance (France); Bobin-Vastra, I. [AREVA Centre Technique de Framatome, 71 - Le Creusot (France); Kuznetsov, V. [Efremov Institute, Doroga na Metallostroy, St. Petersburg (Russian Federation); Schedler, B. [Plansee AG, Reutte (Austria)

    2004-07-01

    The use of flat tile technology to handle heat fluxes in the range of 20 MW/m{sup 2} with components relevant for fusion experiment applications is technically possible with the hypervapotron cooling concept. This paper deals with recent high heat flux performances operated with success on 2 identical mock-ups, based on this concept, that were tested in 2 different electron gun facilities. Each mock-up consisted of a CuCrZr heat sink armored with 25 flat tiles of the 3D carbon fibre composite material SEPcarb NS31 assembled with pure copper by active metal casting (AMC). The AMC tiles were electron beam welded on the CuCrZr bar, fins and slots on the neutral beam JET design were machined into the bar, then the bar was closed with a thick CuCrZr rear plug including hydraulic connections then the bar was electron beam welded onto the sidewalls. The testing results show that full ITER design specifications were achieved with margins, the critical heat flux limit was even higher than 30 MW/m{sup 2}. These results highlight the high potential of this technology for ITER divertor application.

  7. A mature industrial solution for ITER divertor plasma facing components: hypervapotron cooling concept adapted to Tore Supra flat tile technology

    International Nuclear Information System (INIS)

    Escourbiac, F.; Missirlian, M.; Schlosser, J.; Bobin-Vastra, I.; Kuznetsov, V.; Schedler, B.

    2004-01-01

    The use of flat tile technology to handle heat fluxes in the range of 20 MW/m 2 with components relevant for fusion experiment applications is technically possible with the hypervapotron cooling concept. This paper deals with recent high heat flux performances operated with success on 2 identical mock-ups, based on this concept, that were tested in 2 different electron gun facilities. Each mock-up consisted of a CuCrZr heat sink armored with 25 flat tiles of the 3D carbon fibre composite material SEPcarb NS31 assembled with pure copper by active metal casting (AMC). The AMC tiles were electron beam welded on the CuCrZr bar, fins and slots on the neutral beam JET design were machined into the bar, then the bar was closed with a thick CuCrZr rear plug including hydraulic connections then the bar was electron beam welded onto the sidewalls. The testing results show that full ITER design specifications were achieved with margins, the critical heat flux limit was even higher than 30 MW/m 2 . These results highlight the high potential of this technology for ITER divertor application

  8. Steady-state operation of magnetic fusion devices: Plasma control and plasma facing components. Report on the IAEA technical committee meeting held at Fukuoka, 25-29 October 1999

    International Nuclear Information System (INIS)

    Engelmann, F.

    2000-01-01

    An IAEA Technical Committee Meeting on Steady-State Operation of Magnetic Fusion Devices - Plasma Control and Plasma Facing Components was held at Fukuoka, Japan, from 25 to 29 October 1999. The meeting was the second IAEA Techical Committee Meeting on the subject, following the one held at Hefei, China, a year earlier. The meeting was attended by over 150 researchers from 10 countries

  9. Processing of W-Cu functionally graded materials (FGM) through the powder metallurgy route: application as plasma facing components for ITER-like thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Raharijaona, J.J.

    2009-11-01

    The aim of this study was to study and optimize the sintering of W-Cu graded composition materials, for first wall of ITER-like thermonuclear reactor application. The graded composition in the material generates graded functional properties (Functionally Graded Materials - FGM). Rough thermomechanical calculations have shown the interest of W-Cu FGM to improve the lifetime of Plasma Facing Components (PFC). To process W-Cu FGM, powder metallurgy route was analyzed and optimized from W-CuO powder mixtures. The influence of oxide reduction on the sintering of powder mixtures was highlighted. An optimal heating treatment under He/H 2 atmosphere was determined. The sintering mechanisms were deduced from the analysis of the effect of the Cu-content. Sintering of W-Cu materials with a graded composition and grain size has revealed two liquid migration steps: i) capillary migration, after the Cu-melting and, ii) expulsion of liquid, at the end of sintering, from the dense part to the porous part, due to the continuation of W-skeleton sintering. These two steps were confirmed by a model based on capillary pressure calculation. In addition, thermal conductivity measurements were conducted on sintered parts and showed values which gradually increase with the Cu-content. Hardness tests on a polished cross-section in the bulk are consistent with the composition profiles obtained and the differential grain size. (author)

  10. The Effects of Temperature and Oxidation on Deuterium Retention in Solid and Liquid Lithium Films on Molybdenum Plasma-Facing Components

    Science.gov (United States)

    Capece, Angela

    2014-10-01

    Liquid metal plasma-facing components (PFCs) enable in-situ renewal of the surface, thereby offering a solution to neutron damage, erosion, and thermal fatigue experienced by solid PFCs. Lithium in particular has a high chemical affinity for hydrogen, which has resulted in reduced recycling and enhanced plasma performance on many fusion devices including TFTR, T11-M, FTU, CDX-U, LTX, TJ-II, and NSTX. A key component to the improvement in plasma performance is deuterium retention in Li; however, this process is not well understood in the complex tokamak environment. Recent surface science experiments conducted at the Princeton Plasma Physics Laboratory have used electron spectroscopy and temperature programmed desorption to understand the mechanisms for D retention in Li coatings on Mo substrates. The experiments were designed to give monolayer-control of Li films and were conducted in ultrahigh vacuum under controlled environments. An electron cyclotron resonance plasma source was used to deliver a beam of deuterium ions to the surface over a range of ion energies. Our work shows that D is retained as LiD in metallic Li films. However, when oxygen is present in the film, either by diffusion from the subsurface at high temperature or as a contaminant during the deposition process, Li oxides are formed that retain D as LiOD. Experiments indicate that LiD is more thermally stable than LiOD, which decomposes to liberate D2 gas and D2O at temperatures 100 K lower than the LiD decomposition temperature. Other experiments show how D retention varies with substrate temperature to provide insight into the differences between solid and liquid lithium films. This work was supported by DOE Contract No. DE AC02-09CH11466.

  11. 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.

  12. 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.

  13. Characterization of a segmented plasma torch assisted High Heat Flux (HHF) system for performance evaluation of plasma facing components in fusion devices

    International Nuclear Information System (INIS)

    Ngangom, Aomoa; Sarmah, Trinayan; Sah, Puspa; Kakati, Mayur; Ghosh, Joydeep

    2015-01-01

    A wide variety of high heat and particle flux test facilities are being used by the fusion community to evaluate the thermal performance of plasma facing materials/components, which includes electron beam, ion beam, neutral beam and thermal plasma assisted sources. In addition to simulate heat loads, plasma sources have the additional advantage of reproducing exact fusion plasma like conditions, in terms of plasma density, temperature and particle flux. At CPP-IPR, Assam, we have developed a high heat and particle flux facility using a DC, non-transferred, segmented thermal plasma torch system, which can produce a constricted, stabilized plasma jet with high ion density. In this system, the plasma torch exhausts into a low pressure chamber containing the materials to be irradiated, which produces an expanded plasma jet with more uniform profiles, compared to plasma torches operated at atmospheric pressure. The heat flux of the plasma beam was studied by using circular calorimeters of different diameters (2 and 3 cm) for different input power (5-55 kW). The effect of the change in gas (argon) flow rate and mixing of gases (argon + hydrogen) was also studied. The heat profile of the plasma beam was also studied by using a pipe calorimeter. From this, the radial heat flux was calculated by using Abel inversion. It is seen that the required heat flux of 10 MW/m 2 is achievable in our system for pure argon plasma as well as for plasma with gas mixtures. The plasma parameters like the temperature, density and the beam velocity were studied by using optical emission spectroscopy. For this, a McPherson made 1.33 meter focal length spectrometer; model number 209, was used. A plane grating with 1800 g/mm was used which gave a spectral resolution of 0.007 nm. A detailed characterization with respect to these plasma parameters for different gas (argon) flow rate and mixing of gases (argon+hydrogen) for different input power will be presented in this paper. The plasma

  14. 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

  15. ITER plasma facing materials. Some critical considerations

    International Nuclear Information System (INIS)

    Barabash, V.; Dietz, K.J.; Federici, G.; Janeschitz, G.; Matera, R.; Tanaka, S.

    1995-01-01

    The description of current status with the choice of materials for ITER plasma facing components is presented. The main problem with lifetime of divertor elements is the particle and energy-induced erosion of armour materials. A solution for the first operation phase consists in using Be as an armour for the first wall and the divertor, however other possible materials (e.g. W) could be considered. (orig.)

  16. 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

  17. 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

  18. 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.

  19. General directions and recently test modelling results of lithium capillary-pore systems as plasma facing components for tokamak-reactor

    International Nuclear Information System (INIS)

    Evtikhin, V.A.; Lyublinski, I.E.; Vertkov, A.V.; Azizov, E.A.; Mirnov, S.V.; Lazaret, V.B.; Safronov, V.M.

    2003-01-01

    Full text: At present the most promising principal solution of the divertor problem appears to be the use of liquid metals and primarily of lithium Capillary-Pore Systems (CPS) as of plasma facing material. A solid CPS filled with liquid lithium will have high resistance to surface and volume damage because of neutron radiation effects, melting, splashing and thermal stress induced cracking in steady state and during plasma transitions (disruptions, ELMs, VDEs, runaways) to provide the normal operation of divertor target plates and first wall protection elements. These materials would not be the sources of impurities inducing the raise of Z eff and they will not be collected as dust in the divertor area and in ducts. The key directions of experimental investigation of lithium CPS behaviour in first wall and divertor operation simulating conditions are considered. Experiments with lithium CPS in plasma disruption simulation conditions on the hydrogen plasma accelerator MK-200UG (∼10-15 MJ/m 2 , ∼50 μs) have been performed. Shielding lithium plasma layer formation and high stability of these systems have been shown. The new lithium limiter with a thermal regulation system tests on up graded T-11M tokamak (plasma current up to 100 kA, pulse length ∼0.3 s) have been performed. Sorption and desorption of plasma-forming gas, lithium emission into discharge, lithium erosion, limiter deposited power are investigated in this tests

  20. Tungsten

    International Nuclear Information System (INIS)

    Eschnauer, H.

    1978-01-01

    There is no substitute for tungsten in its main field of application so that the demand will not decrease, but there is a need for further important applications. If small variations are left out of account, a small but steady increase in the annual tungsten consumption can be expected. The amount of tungsten available will increase due to the exploritation of new deposits and the extension of existing mines. This tendency will probably be increased by the world-wide prospection. It is hard to make an assessment of the amount of tungsten are obtained in the People's Republic of china, the purchases of Eastern countries in the West, and the sales policy of the USA; pice forecasts are therefore hard to make. A rather interesting subject with regard to the tungsten cycle as a whole is the reprocessing of tungsten-containing wastes. (orig.) [de

  1. 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.

  2. Application of tungsten-fibre-reinforced copper matrix composites to a high-heat-flux component: A design study by dual scale finite element analysis

    International Nuclear Information System (INIS)

    Jeong-Ha You

    2006-01-01

    According to the European Power Plant Conceptual Study, actively cooled tungsten mono-block is one of the divertor design options for fusion reactors. In this study the coolant tube acts as a heat sink and the tungsten block as plasma-facing armour. A key material issue here is how to achieve high temperature strength and high heat conductivity of the heat sink tube simultaneously. Copper matrix composite reinforced with continuous strong fibres has been considered as a candidate material for heat sink of high-heat-flux components. Refractory tungsten wire is a promising reinforcement material due to its high strength, winding flexibility and good interfacial wetting with copper. We studied the applicability of tungsten-fibre-reinforced copper matrix composite heat sink tubes for the tungsten mono-block divertor by means of dual-scale finite element analysis. Thermo-elasto-plastic micro-mechanics homogenisation technique was applied. A heat flux of 15 MW/m 2 with cooling water temperature of 320 o C was considered. Effective stress-free temperature was assumed to be 500 o C. Between the tungsten block and the composite heat sink tube interlayer (1 mm thick) of soft Cu was inserted. The finite element analysis yields the following results: The predicted maximum temperature at steady state is 1223 o C at the surface and 562 o C at the interface between tube and copper layer. On the macroscopic scale, residual stress is generated during fabrication due to differences in thermal expansion coefficients of the materials. Strong compressive stress occurs in the tungsten block around the tube while weak tensile stress is present in the interlayer. The local and global probability of brittle failure of the tungsten block was also estimated using the probabilistic failure theories. The thermal stresses are significantly decreased upon subsequent heat flux loading. Resolving the composite stress on microscopic scale yields a maximum fibre axial stress of 3000 MPa after

  3. 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

  4. 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.

  5. 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

  6. 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.

  7. 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.)

  8. Tritium permeation model for plasma facing components

    Science.gov (United States)

    Longhurst, G. R.

    1992-12-01

    This report documents the development of a simplified one-dimensional tritium permeation and retention model. The model makes use of the same physical mechanisms as more sophisticated, time-transient codes such as implantation, recombination, diffusion, trapping and thermal gradient effects. It takes advantage of a number of simplifications and approximations to solve the steady-state problem and then provides interpolating functions to make estimates of intermediate states based on the steady-state solution. The model is developed for solution using commercial spread-sheet software such as Lotus 123. Comparison calculations are provided with the verified and validated TMAP4 transient code with good agreement. Results of calculations for the ITER CDA diverter are also included.

  9. Tritium permeation model for plasma facing components

    International Nuclear Information System (INIS)

    Longhurst, G.R.

    1992-12-01

    This report documents the development of a simplified one-dimensional tritium permeation and retention model. The model makes use of the same physical mechanisms as more sophisticated, time-transient codes such as implantation, recombination, diffusion, trapping and thermal gradient effects. It takes advantage of a number of simplifications and approximations to solve the steady-state problem and then provides interpolating functions to make estimates of intermediate states based on the steady-state solution. The model is developed for solution using commercial spread-sheet software such as Lotus 123. Comparison calculations are provided with the verified and validated TMAP4 transient code with good agreement. Results of calculations for the ITER CDA diverter are also included

  10. 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

  11. 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.

  12. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kaita, R., E-mail: kaita@pppl.gov [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Lucia, M. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Allain, J.P.; Bedoya, F. [Department of Nuclear, Plasma, & Radiological Engineering, University of Illinois at Urbana-Champaign, Urbana, IL (United States); Bell, R.; Boyle, D. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Capece, A. [Department of Physics, The College of New Jersey, Ewing, NJ (United States); Jaworski, M. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Koel, B.E. [Department of Chemical & Biological Engineering, Princeton University, Princeton, NJ (United States); Majeski, R. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Roszell, J. [Department of Chemical & Biological Engineering, Princeton University, Princeton, NJ (United States); Schmitt, J. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Scotti, F. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Soukhanovskii, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States)

    2017-04-15

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

  14. 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

  15. 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.)

  16. 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.)

  17. 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

  18. 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.

  19. Recovery of Tungsten Surface with Fiber-Form Nanostructure by Plasmas Exposures

    International Nuclear Information System (INIS)

    Miyamoto, Takanori; Takamura, Shuichi; Kurishita, Hiroaki

    2013-01-01

    One of the serious concerns for tungsten materials in fusion devices is the radiation defects caused by helium plasma irradiation since helium is a fusion product. The fiber-formed nanostructure is thought to have a possible weakness against the plasma heat flux on the plasma-facing component and also may destroy the reflectivity of optical mirrors. In this paper an interesting method for the recovery of such tungsten surfaces is shown. The recovery process depends on the grade and manufacturing process of tungsten materials. (fusion engineering)

  20. 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

  1. 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.)

  2. 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

  3. 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

  4. 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%

  5. Hydrogen permeation properties of plasma-sprayed tungsten

    International Nuclear Information System (INIS)

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

    1994-01-01

    Tungsten has been proposed as a plasma-facing component material for advanced fusion facilities. This paper reports on laboratory-scale studies that were done to assess the hydrogen permeation properties of plasma-sprayed tungsten for such applications. The work entailed deuterium permeation measurements for plasma-sprayed (PS) tungsten coatings, sputter-deposited (SP) tungsten coatings, and steel substrate material using a mass-analyzed, 3 keV D + 3 ion beam with fluxes of similar 6.5x10 19 D/m 2 s. Extensive characterization analyses for the plasma-sprayed tungsten coatings were made using Auger spectrometry and scanning electron microscopy (SEM). Observed permeation rates through composite PS-tungsten/steel specimens were several orders of magnitude below the permeation levels observed for SP-tungsten/steel composite specimens and pure steel specimens. Characterization analyses indicated that the plasma-sprayed tungsten coating had a nonhomogeneous microstructure that consisted of splats with columnar solidification, partially-melted particles with grain boundaries, and void regions. Reduced permeation levels can be attributed to the complex microstructure and a substantial surface-connected porosity. ((orig.))

  6. Hydrogen permeation properties of plasma-sprayed tungsten

    Energy Technology Data Exchange (ETDEWEB)

    Anderl, R.A. (Idaho National Engineering Lab., EG and G Idaho Inc., Idaho Falls, ID (United States)); Pawelko, R.J. (Idaho National Engineering Lab., EG and G Idaho Inc., Idaho Falls, ID (United States)); Hankins, M.R. (Idaho National Engineering Lab., EG and G Idaho Inc., Idaho Falls, ID (United States)); Longhurst, G.R. (Idaho National Engineering Lab., EG and G Idaho Inc., Idaho Falls, ID (United States)); Neiser, R.A. (Sandia National Laboratories, Albuquerque, NM 87185 (United States))

    1994-09-01

    Tungsten has been proposed as a plasma-facing component material for advanced fusion facilities. This paper reports on laboratory-scale studies that were done to assess the hydrogen permeation properties of plasma-sprayed tungsten for such applications. The work entailed deuterium permeation measurements for plasma-sprayed (PS) tungsten coatings, sputter-deposited (SP) tungsten coatings, and steel substrate material using a mass-analyzed, 3 keV D[sup +][sub 3] ion beam with fluxes of similar 6.5x10[sup 19] D/m[sup 2] s. Extensive characterization analyses for the plasma-sprayed tungsten coatings were made using Auger spectrometry and scanning electron microscopy (SEM). Observed permeation rates through composite PS-tungsten/steel specimens were several orders of magnitude below the permeation levels observed for SP-tungsten/steel composite specimens and pure steel specimens. Characterization analyses indicated that the plasma-sprayed tungsten coating had a nonhomogeneous microstructure that consisted of splats with columnar solidification, partially-melted particles with grain boundaries, and void regions. Reduced permeation levels can be attributed to the complex microstructure and a substantial surface-connected porosity. ((orig.))

  7. Hydrogen permeation properties of plasma-sprayed tungsten*1

    Science.gov (United States)

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

    1994-09-01

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

  8. Electro-chemically-based technologies for processing of tungsten components in fusion technology

    International Nuclear Information System (INIS)

    Holstein, N.; Konys, J.; Krauss, W.; Lorenz, J.

    2010-01-01

    In fusion technology layers and bulk components fabricated from tungsten and W-alloys are used as functional materials, e.g. as coatings of blanket modules or T-permeation barriers and also as structural components in a He-cooled divertor. Their application under high heat loads and temperatures is besides manufacturing, also challenging regarding joining, caused e.g. by expansion mismatches in combination with steel or other diffusion issues. Driven by these needs, electro-chemically-based technologies were analyzed concerning their advantages in processing in the fields of soft structuring of tungsten alloys and in deposition of functional scales. The Electro-Chemistry (EC) of tungsten is characterized by its affection to build up passivation layers in aqueous media during the initial oxidation, which is the result of an unavoidable basic electrochemical reaction with water (W + 3H 2 O → WO 3 + 3H 2 ), although the element standard potential is situated between common EC material like iron and copper. (orig.)

  9. 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.

  10. 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

  11. 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

  12. A new fully automatic PIM tool to replicate two component tungsten DEMO divertor parts

    International Nuclear Information System (INIS)

    Antusch, Steffen; Commin, Lorelei; Heneka, Jochen; Piotter, Volker; Plewa, Klaus; Walter, Heinz

    2013-01-01

    Highlights: • Development of a fully automatic 2C-PIM tool. • Replicate fusion relevant components in one step without additional brazing. • No cracks or gaps in the seam of the joining zone visible. • For both material combinations a solid bond of the material interface was achieved. • PIM is a powerful process for mass production as well as for joining even complex shaped parts. -- Abstract: At Karlsruhe Institute of Technology (KIT), divertor design concepts for future nuclear fusion power plants beyond ITER are intensively investigated. One promising KIT divertor design concept for the future DEMO power reactor is based on modular He-cooled finger units. The manufacturing of such parts by mechanical machining such as milling and turning, however, is extremely cost and time intensive because tungsten is very hard and brittle. Powder Injection Molding (PIM) has been adapted to tungsten processing at KIT since a couple of years. This production method is deemed promising in view of large-scale production of tungsten parts with high near-net-shape precision, hence, offering an advantage of cost-saving process compared to conventional machining. The properties of the effectively and successfully manufactured divertor part tile consisting only of pure tungsten are a microstructure without cracks and a high density (>98% T.D.). Based on the achieved results a new fully automatic multicomponent PIM tool was developed and allows the replication and joining without brazing of fusion relevant components of different materials in one step and the creation of composite materials. This contribution describes the process route to design and engineer a new fully automatic 2C-PIM tool, including the filling simulation and the implementing of the tool. The complete technological fabrication process of tungsten 2C-PIM, including material and feedstock (powder and binder) development, injection molding, and heat-treatment of real DEMO divertor parts is outlined

  13. 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)

  14. 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)

  15. 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.

  16. The Role of the Component Metals in the Toxicity of Military-Grade Tungsten Alloy

    Directory of Open Access Journals (Sweden)

    Christy A. Emond

    2015-12-01

    Full Text Available Tungsten-based composites have been recommended as a suitable replacement for depleted uranium. Unfortunately, one of these mixtures composed of tungsten (W, nickel (Ni and cobalt (Co induced rhabdomyosarcomas when implanted into the leg muscle of laboratory rats and mice to simulate a shrapnel wound. The question arose as to whether the neoplastic effect of the mixture could be solely attributed to one or more of the metal components. To investigate this possibility, pellets with one or two of the component metals replaced with an identical amount of the biologically-inert metal tantalum (Ta were manufactured and implanted into the quadriceps of B6C3F1 mice. The mice were followed for two years to assess potential adverse health effects. Implantation with WTa, CoTa or WNiTa resulted in decreased survival, but not to the level reported for WNiCo. Sarcomas in the implanted muscle were found in 20% of the CoTa-implanted mice and 5% of the WTa- and WCoTa-implanted rats and mice, far below the 80% reported for WNiCo-implanted mice. The data obtained from this study suggested that no single metal is solely responsible for the neoplastic effects of WNiCo and that a synergistic effect of the three metals in tumor development was likely.

  17. 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)

  18. 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.

  19. Conditionings for boron-carbon plasma facing wall

    International Nuclear Information System (INIS)

    Hino, Tomoaki; Yamauchi, Yuji; Yamashina, Toshiro

    1994-01-01

    For plasma facing material with components of boron and carbon, the method of conditionings due to He discharge cleaning and baking is considered. The conditioning time required to suppress the hydrogen recycling is discussed. It is shown that the hydrogen trapped by the boron can be relatively easily removed only by the baking at 300degC or only by He discharge cleaning with current density of 0.1 mA/cm 2 . It is not easy to remove the hydrogen trapped by the carbon by the baking since the temperature required becomes 500degC. The current density required also becomes high, 1 mA/cm 2 , for the reduction of the hydrogen trapped by the carbon. (author)

  20. Plasma facing parts and repairing method

    International Nuclear Information System (INIS)

    Fuse, Toshiaki; Tachikawa, Nobuo.

    1994-01-01

    Plasma facing parts of the present invention are constituted by joining an armour comprising a material having a high melting point and a cooling member comprising copper or the like. A metal member having good solderability with the cooling member is disposed on the joined surface of the armor member. In addition, the joined surface of the cooling member is provided with a barrier layer for preventing invasion of a solder. A solder having a low melting point is interposed between the armour and the cooling member. If they are heated entirely, the solder having low melting point is melted, so that the metal member having good solderability disposed on the armor member is soldered with the barrier layer for the cooling member. Upon exchange of the armour, the joint is heated again. Then, the solder having a low melting point is melted and the armour member and the cooling member are separated. If a solder is put on the cooling member and a new armour is placed and then heated, repairing is completed. (I.S.)

  1. 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...

  2. 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.

  3. New electron beam facility for irradiated plasma facing materials testing in hot cell

    International Nuclear Information System (INIS)

    Sakamoto, N.; Kawamura, H.; Akiba, M.

    1995-01-01

    Since plasma facing components such as the first wall and the divertor for the next step fusion reactors are exposed to high heat loads and high energy neutron flux generated by the plasma, it is urgent to develop of plasma facing components which can resist these. Then, we have established electron beam heat facility (open-quotes OHBISclose quotes, Oarai Hot-cell electron Beam Irradiating System) at a hot cell in JMTR (Japan Materials Testing Reactor) hot laboratory in order to estimate thermal shock resistivity of plasma facing materials and heat removal capabilities of divertor elements under steady state heating. In this facility, irradiated plasma facing materials (beryllium, carbon based materials and so on) and divertor elements can be treated. This facility consists of an electron beam unit with the maximum beam power of 50kW and the vacuum vessel. The acceleration voltage and the maximum beam current are 30kV (constant) and 1.7A, respectively. The loading time of electron beam is more than 0.1ms. The shape of vacuum vessel is cylindrical, and the mainly dimensions are 500mm in inner diameter, 1000mm in height. The ultimate vacuum of this vessel is 1 x 10 -4 Pa. At present, the facility for thermal shock test has been established in a hot cell. And performance estimation on the electron beam is being conducted. Presently, the devices for heat loading tests under steady state will be added to this facility

  4. New electron beam facility for irradiated plasma facing materials testing in hot cell

    International Nuclear Information System (INIS)

    Shimakawa, S.; Akiba, M.; Kawamura, H.

    1996-01-01

    Since plasma facing components such as the first wall and the divertor for the next step fusion reactors are exposed to high heat loads and high energy neutron flux generated by the plasma, it is urgent to develop plasma facing components which can resist these. We have established electron beam heat facility ('OHBIS', Oarai hot-cell electron beam irradiating system) at a hot cell in JMTR (Japan materials testing reactor) hot laboratory in order to estimate thermal shock resistivity of plasma facing materials and heat removal capabilities of divertor elements under steady state heating. In this facility, irradiated plasma facing materials (beryllium, carbon based materials and so on) and divertor elements can be treated. This facility consists of an electron beam unit with the maximum beam power of 50 kW and the vacuum vessel. The acceleration voltage and the maximum beam current are 30 kV (constant) and 1.7 A, respectively. The loading time of the electron beam is more than 0.1 ms. The shape of vacuum vessel is cylindrical, and the main dimensions are 500 mm in inside diameter, 1000 mm in height. The ultimate vacuum of this vessel is 1 x 10 -4 Pa. At present, the facility for the thermal shock test has been established in a hot cell. The performance of the electron beam is being evaluated at this time. In the future, the equipment for conducting static heat loadings will be incorporated into the facility. (orig.)

  5. 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)

  6. 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

  7. 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.

  8. 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

  9. 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.

  10. 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

  11. Development and fabrication aspects regarding tungsten components for a He-cooled divertor

    International Nuclear Information System (INIS)

    Krauss, W.; Holstein, N.; Konys, J.

    2005-01-01

    Under the EU framework of power plant conceptual study (PPCS), a modular He-cooled divertor concept is investigated, which is projected to remove high heat loads of up to 15 MW/m 2 . This design is based on a modular arrangement of cooling fingers consisting of a tile acting as sacrificial layer, a thimble through-flowed by high pressurized He and special micro-structured components for enhanced heat transfer. The success of this design is strongly correlated to the availability of special tungsten alloys and for the pin/slot option efficient micro-structuring of W or W-1% La 2 O 3 arrays. An evaluation of shaping technologies for array manufacturing under consideration of applicability, degree of development status, expected effectiveness and economy was performed and the most promising methods were tested. Based on the today's knowledge, electrical discharge machining (EDM) and laser etching (LE) allow the shaping of slot arrays; however, an impact on microstructure was detected. Technologies like powder injection moulding (PIM) or electro-chemically assisted machining processes (ECM) need further development and testing to be applied as reliable fabrication processes in structuring of W-alloys

  12. Development of beryllium bonds for plasma-facing components

    International Nuclear Information System (INIS)

    Franconi, E.; Ceccotti, G.C.; Magnoli, L.

    1992-01-01

    This study concerns the techniques of bonding beryllium to both structural material (AISI 316 SS) and heat sink material (copper and DS-copper) plates, and the characterization of the bonding material obtained. Conventional bonding techniques for joining Be to SS and copper using brazing alloys were first investigated. The best result was obtained using a silver-copper eutetic alloy as a brazing alloy. However, the high-temperature capability of the materials prepared by this method is limited by the performance of brazing alloys at the operating temperature. To avoid this problem, we are developing a joining process known as solid-state reaction bonding that improves the capability at the operating temperature. (orig.)

  13. Development of plasma facing components for fusion experimental reactors

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, M.; Fujiya, Y.; Inoue, M.; Morimoto, M. [Mitsubishi Heavy Industries, Ltd., Yokohama (Japan)

    1995-12-31

    The divertor structure and fabrication process have been investigated, including the structures of the divertor elements and support, fundamental brazing techniques, brazing of large divertor tiles and fabrication method of large divertor modules. Using direct brazing, a partial divertor module with large CFC tiles was fabricated and tested. It was shown that the model had sufficient structural integrity against thermal shocks of {approximately}17MW/m{sup 2} {times} 4 sec for up to 1,600 times. A fabrication technique for large and complex-shaped divertor module has been developed and successfully applied to a 1m-long linear and 0.8m-long curved divertor modules. In addition, preliminary investigation of direct brazing of beryllium to the copper substrate has been conducted. It was found that the bending strength of the bonded materials was around 40 MPa. Furthermore, boron coating on the CFC and Mo has been examined. Using the boron ion implantation technique, boron ions were implanted to the CFC and Mo plates prior to the boron atoms deposition. The samples fabricated with this method were found to have a sufficient thermal shock resistance.

  14. The plasma facing components of the Tore Supra ICRF antenna

    International Nuclear Information System (INIS)

    Beaumont, B.; Agarici, G.; Gauthier, E.; Kuus, H.; Schlosser, J.

    1994-01-01

    Two generations of Faraday shields for the Tore Supra ICRH antennas interacting with the edge plasma are presented. The last one, using a film of boron carbide as protective material performs well, proving the relevance of this technique for in vessel equipment submitted to low power fluxes. The different lateral protections used on Tore Supra are submitted to high power fluxes. Finite element calculations allow to assess their performances. One type, using Boron Carbide, can be used to measure the local heat flux. The estimation of this flux confirm the specificity of the edge/RF interaction, which is more than one order of magnitude above the exponential decay observed in ohmic plasmas. (author) 11 refs.; 1 fig

  15. Thermal shock problems of bonded structure for plasma facing components

    International Nuclear Information System (INIS)

    Shibui, M.; Kuroda, T.; Kubota, Y.

    1991-01-01

    Thermal shock tests have been performed on W(Re)/Cu and Mo/Cu duplex structures with a particular emphasis on two failure modes: failure on the heated surface and failure near the bonding interface. The results indicate that failure of the duplex structure largely depends on the constraint of thermal strain on the heated surface and on the ductility changes of armour materials. Rapid debonding of the bonding interface may be attributed to the yielding of armour materials. This leads to a residual bending deformation when the armour cools down. Arguments are also presented in this paper on two parameter characterization of the failure of armour materials and on stress distribution near the free edge of the bonding interface. (orig.)

  16. 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.

  17. 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.

  18. 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.

  19. 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

  20. 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%.

  1. 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)

  2. 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.)

  3. 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

  4. 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

  5. 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

  6. Investigation of the interaction between the components of a Nichrome-tungsten composite

    Energy Technology Data Exchange (ETDEWEB)

    Prokopov, I.P.; Logvinova, T.N.

    1980-01-01

    Experimental results are presented on the effect of Nichrome melting on tungsten in the case of different rates of solidification of the composite. Consideration is given to the effect of the volume fraction of reinforced materials on the size of the transition zone between the fibers and the die and on the microhardness distribution in the composite system.

  7. 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.

  8. The impact of transient thermal loads on beryllium as plasma facing material

    Energy Technology Data Exchange (ETDEWEB)

    Spilker, Benjamin Christof

    2017-01-24

    The rising global energy consumption requires a broad research and development approach in the field of energy technology. Besides renewables, nuclear fusion promises an efficient, CO{sub 2} free, no long-term radioactive waste producing, and safe energy source using only deuterium and lithium as primary resources, which are widely abundant. However, several technical challenges have to be overcome before a nuclear fusion power plant can be built. For this purpose, the experimental reactor ITER is currently under construction in France. ITER is intended to demonstrate the scientific and technological feasibility of net energy generation via nuclear fusion. The most heavily loaded components inside a fusion reactor, which are directly facing the fusion plasma, have to be armoured with well suited materials, which need to be able to withstand the high thermal and particle loads for an economically reasonable lifetime. For ITER, beryllium is chosen as plasma facing material for the largest fraction of the inner vacuum vessel, the so called first wall. Tungsten will be applied in the bottom region of the vacuum vessel, the so called divertor, which acts as the exhaust system of the machine. The choice of beryllium as plasma facing material was driven by its outstanding advantages, e.g. the low atomic number assures that eroded wall material does not strongly decrease the fusion plasma performance, while it combines a high thermal conductivity with low chemical sputtering characteristics. However, the relatively low melting temperature of beryllium of 1287 C comprises the risk of amour damage by melting during transient plasma events, such as edge localized modes or plasma disruptions. Even when mitigated, these events put tremendous power densities in the GW m{sup -2} range with durations in the ms scale onto the plasma facing materials. Hence, the performance of the ITER reference beryllium grade S-65 under transient thermal loads was studied within this work. Thereby

  9. The impact of transient thermal loads on beryllium as plasma facing material

    International Nuclear Information System (INIS)

    Spilker, Benjamin Christof

    2017-01-01

    The rising global energy consumption requires a broad research and development approach in the field of energy technology. Besides renewables, nuclear fusion promises an efficient, CO_2 free, no long-term radioactive waste producing, and safe energy source using only deuterium and lithium as primary resources, which are widely abundant. However, several technical challenges have to be overcome before a nuclear fusion power plant can be built. For this purpose, the experimental reactor ITER is currently under construction in France. ITER is intended to demonstrate the scientific and technological feasibility of net energy generation via nuclear fusion. The most heavily loaded components inside a fusion reactor, which are directly facing the fusion plasma, have to be armoured with well suited materials, which need to be able to withstand the high thermal and particle loads for an economically reasonable lifetime. For ITER, beryllium is chosen as plasma facing material for the largest fraction of the inner vacuum vessel, the so called first wall. Tungsten will be applied in the bottom region of the vacuum vessel, the so called divertor, which acts as the exhaust system of the machine. The choice of beryllium as plasma facing material was driven by its outstanding advantages, e.g. the low atomic number assures that eroded wall material does not strongly decrease the fusion plasma performance, while it combines a high thermal conductivity with low chemical sputtering characteristics. However, the relatively low melting temperature of beryllium of 1287 C comprises the risk of amour damage by melting during transient plasma events, such as edge localized modes or plasma disruptions. Even when mitigated, these events put tremendous power densities in the GW m"-"2 range with durations in the ms scale onto the plasma facing materials. Hence, the performance of the ITER reference beryllium grade S-65 under transient thermal loads was studied within this work. Thereby, the

  10. 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

  11. 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

  12. 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.

  13. 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.

  14. Tungsten transport in the plasma edge at ASDEX upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Janzer, Michael Arthur

    2015-04-30

    The Plasma Facing Components (PFC) will play a crucial role in future deuterium-tritium magnetically confined fusion power plants, since they will be subject to high energy and particle loads, but at the same time have to ensure long lifetimes and a low tritium retention. These requirements will most probably necessitate the use of high-Z materials such as tungsten for the wall materials, since their erosion properties are very benign and, unlike carbon, capture only little tritium. The drawback with high-Z materials is, that they emit strong line radiation in the core plasma, which acts as a powerful energy loss mechanism. Thus, the concentration of these high-Z materials has to be controlled and kept at low levels in order to achieve a burning plasma. Understanding the transport processes in the plasma edge is essential for applying the proper impurity control mechanisms. This control can be exerted either by enhancing the outflux, e.g. by Edge Localized Modes (ELM), since they are known to expel impurities from the main plasma, or by reducing the influx, e.g. minimizing the tungsten erosion or increasing the shielding effect of the Scrape Off Layer (SOL). ASDEX Upgrade (AUG) has been successfully operating with a full tungsten wall for several years now and offers the possibility to investigate these edge transport processes for tungsten. This study focused on the disentanglement of the frequency of type-I ELMs and the main chamber gas injection rate, two parameters which are usually linked in H-mode discharges. Such a separation allowed for the first time the direct assessment of the impact of each parameter on the tungsten concentration. The control of the ELM frequency was performed by adjusting the shape of the plasma, i.e. the upper triangularity. The radial tungsten transport was investigated by implementing a modulated tungsten source. To create this modulated source, the linear dependence of the tungsten erosion rate at the Ion Cyclotron Resonance

  15. 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.

  16. Remote-LIBS characterization of ITER-like plasma facing materials

    International Nuclear Information System (INIS)

    Almaviva, S.; Caneve, L.; Colao, F.; Fantoni, R.; Maddaluno, G.

    2012-01-01

    Graphical abstract: Display Omitted Highlights: ► Description of a LIBS set-up as remote diagnostics in new generation fusion machines. ► Identification of the atomic composition of samples simulating plasma facing components. ► Submicrometric resolution in depth profiling the elemental composition of the samples. ► Identification of elements present in traces or as impurities on the sample surface. ► Discussion on the applicability of the Calibration Free method for quantitative analysis. - Abstract: The occurrence of several plasma-wall interaction processes, eventually affecting the overall system performances, is expected in a working fusion device chamber. Monitoring the changes in the composition of the plasma facing component (PFC) surface layer, as a result of erosion and redeposition mechanisms, can provide useful information on the possible plasma pollution and fuel retention. To this aim, suitable diagnostic techniques able to perform depth profiling analysis of the superficial layers on the PFCs have been developed. Due to the constraints commonly found in fusion devices, the measuring apparatus must be non invasive, remote and sensitive to light elements. These requirements make LIBS (Laser Induced Breakdown Spectroscopy) an ideal candidate for on-line monitoring the walls of current and of next generation (as ITER) fusion devices. LIBS is a well established tool for qualitative, semi-quantitative and quantitative analysis of surfaces, with micro-destructive characteristics and some capabilities for stratigraphy. In this work, LIBS depth profiling capability has been verified for the determination of the composition of multilayer structures simulating plasma facing components covered with deposited impurity layers. A new experimental setup has been designed and realized in order to optimize the characteristics of a LIBS system working in vacuum conditions and remotely, two noticeable properties for an ITER-relevant diagnostics. A quantitative

  17. 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.

  18. 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.

  19. 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.

  20. 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

  1. 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#

  2. 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.

  3. 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

  4. EU Development of High Heat Flux Components

    International Nuclear Information System (INIS)

    Linke, J.; Lorenzetto, P.; Majerus, P.; Merola, M.; Pitzer, D.; Roedig, M.

    2005-01-01

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

  5. 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.

  6. 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

  7. Behavior of liquid Li-Sn alloy as plasma facing material on ISTTOK

    Energy Technology Data Exchange (ETDEWEB)

    Loureiro, J.P.S., E-mail: jpsloureiro@ipfn.tecnico.ulisboa.pt [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, 1049-001 Lisboa (Portugal); Tabarés, F.L. [Laboratorio Nacional de Fusion, Ciemat, Avenida Complutense 22, E-28040 Madrid (Spain); Fernandes, H.; Silva, C.; Gomes, R.; Alves, E.; Mateus, R.; Pereira, T.; Alves, H.; Figueiredo, H. [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, 1049-001 Lisboa (Portugal)

    2017-04-15

    The high power loads impinging on the first wall and particularly the divertor of fusion reactors is a decisive factor to the success of nuclear fusion. An alternative to solid plasma facing components is the use of liquid metals such as lithium or tin due to the regenerative properties of the liquid surface. Another suitable candidate is the eutectic lithium tin alloy (30 at.% Li) which is suggested to display beneficial properties of both its constituent elements. The application of these materials as liquid metal plasma facing components depends on several factors such as their affinity to retain hydrogenic isotopes and the discharge performance degradation induced by the enhanced impurity contamination, among others. An experimental setup has been developed to produce and expose samples to ISTTOK plasmas on both liquid and solid states. Samples of Li-Sn alloy were exposed at ISTTOK to deuterium plasmas. Post-mortem analysis of the samples was performed by means of ion beam diagnostics. To quantify the fuel retention on the samples the nuclear reaction analysis (NRA) technique was applied. Complementary, Rutherford backscattering spectrometry (RBS) was used for determination material composition, particularly of impurities, on the samples. Regardless of the high sensitivity of these techniques no deuterium was detected in the samples. Emission of the Li-I 670.7 nm line indicates that there was interaction of the plasma with the samples. Alternative reasons for the low retention of this material are discussed. Lithium segregation to the surface of the sample was observed.

  8. 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.

  9. 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).

  10. 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

  11. 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.

  12. 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

  13. 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.

  14. 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

  15. Investigation of cascade effect failure for tungsten armour

    International Nuclear Information System (INIS)

    Makhankov, A.; Barabash, V.; Berkhov, N.; Divavin, V.; Giniatullin, R.; Grigoriev, S.; Ibbott, C.; Komarov, V.; Labusov, A.; Mazul, I.; McDonald, J.; Tanchuk, V.; Youchison, D.

    2001-01-01

    The glancing angle of incident power on the target of a tokamak divertor results in doubled and highly peaked heat flux onto adjacent downstream tile in the case of lost of tile event (LOTE). As a result downstream tile has higher probability to fail resulting in triple loads to the next downstream tile and so on (cascade effect). This paper devoted to analytical and experimental investigation of the cascade effect failure for the flat tile option of tungsten armoured plasma facing components. Armour geometry resistant to the cascade effect failure was selected on the base of thermal and stress analyses. Experimental investigation of the LOTE has been performed also. Small size W/Cu mock-up withstood not only LOTE simulation load, but also survived afterwards for 1500 cycles at 26-28 MW/m 2 without damage in joint

  16. 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)

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

    Czech Academy of Sciences Publication Activity Database

    Tiron, V.; Velicu, I. L.; Porosnicu, C.; Burducea, I.; Dinca, P.; Malinský, Petr

    Roč. 416, SEP (2017), s. 878-884 ISSN 0169-4332 R&D Projects: GA ČR(CZ) GBP108/12/G108; GA MŠk LM2015056 Institutional support: RVO:61389005 Keywords : Tugensten nitride layers * m-HIPIMS * deuterium retention * deuterium plasma jet * thermal desorption spectrometry Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders OBOR OECD: Nuclear physics Impact factor: 3.387, year: 2016

  18. Rapid determination of main components by means of flame-atomic-absorption spectrometry for chromium, silicon and tungsten in CrSiW materials

    International Nuclear Information System (INIS)

    Mueller, E.; Stahlberg, R.

    1985-01-01

    The application of Flame-Atomic-Absorption Spectrometry (FAAS) for determining chromium, silicon and tungsten in CrSiW materials is described. The FAAS determinations of the main components are shown under optimum conditions. Sufficient precision and reliability have been achieved for routine analysis. The application of a mixture of acids for preparing CrSiW solutions is proposed. The preparation of samples is discussed in detail. Optimum conditions are recommended for determining chromium, silicon and tungsten using one solution only. (orig.) [de

  19. 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.

  20. Hydrogen isotope retention in beryllium for tokamak plasma-facing applications

    Energy Technology Data Exchange (ETDEWEB)

    Anderl, R.A.; Longhurst, G.R. [Lockheed Martin Idaho Technol. Co., Idaho Falls, ID (United States). Idaho Nat. Eng. and Environ. Lab.; Causey, R.A.; Wampler, W.R.; Wilson, K.L. [Sandia National Laboratories, Livermore, CA (United States)]|[Sandia National Labs., Albuquerque, NM (United States); Davis, J.W.; Haasz, A.A. [Institute for Aerospace Studies, University of Toronto, Toronto (Canada); Doerner, R.P. [California Univ., San Diego, La Jolla, CA (United States). Center for Magnetic Recording Research; Federici, G. [ITER JWS Garching Co-center, Garching (Germany)

    1999-06-01

    Beryllium has been used as a plasma-facing material to effect substantial improvements in plasma performance in the Joint European Torus (JET), and it is planned as a plasma-facing material for the first wall (FW) and other components of the International Thermonuclear Experimental Reactor (ITER). The interaction of hydrogenic ions, and charge-exchange neutral atoms from plasmas, with beryllium has been studied in recent years with widely varying interpretations of results. In this paper we review experimental data regarding hydrogenic atom inventories in experiments pertinent to tokamak applications and show that with some very plausible assumptions, the experimental data appear to exhibit rather predictable trends. A phenomenon observed in high ion-flux experiments is the saturation of the beryllium surface such that inventories of implanted particles become insensitive to increased flux and to continued implantation fluence. Methods for modeling retention and release of implanted hydrogen in beryllium are reviewed and an adaptation is suggested for modeling the saturation effects. The TMAP4 code used with these modifications has succeeded in simulating experimental data taken under saturation conditions where codes without this feature have not. That implementation also works well under more routine conditions where the conventional recombination-limited release model is applicable. Calculations of tritium inventory and permeation in the ITER FW during the basic performance phase (BPP) using both the conventional recombination model and the saturation effects assumptions, show a difference of several orders of magnitude in both inventory and permeation rate to the coolant. (orig.) 78 refs.

  1. Tritium recycling and inventory in eroded debris of plasma-facing materials

    International Nuclear Information System (INIS)

    Hassanein, A.

    1999-01-01

    Damage to plasma-facing components (PFCs) and structural materials due to loss of plasma confinement in magnetic fusion reactors remains one of the most serious concerns for safe, successful, and reliable tokamak operation. High erosion losses due to surface vaporization, spallation, and melt-layer splashing are expected during such an event. The eroded debris and dust of the PFCs, including trapped tritium, will be contained on the walls or within the reactor chamber therefore, they can significantly influence plasma behavior and tritium inventory during subsequent operations. Tritium containment and behavior in PFCS and in the dust and debris is an important factor in evaluating and choosing the ideal plasma-facing materials (PFMs). Tritium buildup and release in the debris of candidate materials is influenced by the effect of material porosity on diffusion and retention processes. These processes have strong nonlinear behavior due to temperature, volubility, and existing trap sites. A realistic model must therefore account for the nonlinear and multidimensional effects of tritium diffusion in the porous-redeposited and neutron-irradiated materials. A tritium-transport computer model, TRAPS (Tritium Accumulation in Porous Structure), was developed and used to evaluate and predict the kinetics of tritium transport in porous media. This model is coupled with the TRICS (Tritium In Compound Systems) code that was developed to study the effect of surface erosion during normal and abnormal operations on tritium behavior in PFCS

  2. In vitro profiling of epigenetic modifications underlying heavy metal toxicity of tungsten-alloy and its components

    International Nuclear Information System (INIS)

    Verma, Ranjana; Xu, Xiufen; Jaiswal, Manoj K.; Olsen, Cara; Mears, David; Caretti, Giuseppina; Galdzicki, Zygmunt

    2011-01-01

    Tungsten-alloy has carcinogenic potential as demonstrated by cancer development in rats with intramuscular implanted tungsten-alloy pellets. This suggests a potential involvement of epigenetic events previously implicated as environmental triggers of cancer. Here, we tested metal induced cytotoxicity and epigenetic modifications including H3 acetylation, H3-Ser10 phosphorylation and H3-K4 trimethylation. We exposed human embryonic kidney (HEK293), human neuroepithelioma (SKNMC), and mouse myoblast (C2C12) cultures for 1-day and hippocampal primary neuronal cultures for 1-week to 50-200 μg/ml of tungsten-alloy (91% tungsten/6% nickel/3% cobalt), tungsten, nickel, and cobalt. We also examined the potential role of intracellular calcium in metal mediated histone modifications by addition of calcium channel blockers/chelators to the metal solutions. Tungsten and its alloy showed cytotoxicity at concentrations > 50 μg/ml, while we found significant toxicity with cobalt and nickel for most tested concentrations. Diverse cell-specific toxic effects were observed, with C2C12 being relatively resistant to tungsten-alloy mediated toxic impact. Tungsten-alloy, but not tungsten, caused almost complete dephosphorylation of H3-Ser10 in C2C12 and hippocampal primary neuronal cultures with H3-hypoacetylation in C2C12. Dramatic H3-Ser10 dephosphorylation was found in all cobalt treated cultures with a decrease in H3 pan-acetylation in C2C12, SKNMC and HEK293. Trimethylation of H3-K4 was not affected. Both tungsten-alloy and cobalt mediated H3-Ser10 dephosphorylation were reversed with BAPTA-AM, highlighting the role of intracellular calcium, confirmed with 2-photon calcium imaging. In summary, our results for the first time reveal epigenetic modifications triggered by tungsten-alloy exposure in C2C12 and hippocampal primary neuronal cultures suggesting the underlying synergistic effects of tungsten, nickel and cobalt mediated by changes in intracellular calcium homeostasis and

  3. Plasma facing surface composition during NSTX Li experiments

    Energy Technology Data Exchange (ETDEWEB)

    Skinner, C.H., E-mail: cskinner@pppl.gov [Princeton Plasma Physics Laboratory, POB 451, Princeton, NJ 08543 (United States); Sullenberger, R. [Department of Mechanical and Aerospace Engineering, Princeton University, NJ 08540 (United States); Koel, B.E. [Department of Chemical and Biological Engineering, Princeton University, NJ 08540 (United States); Jaworski, M.A.; Kugel, H.W. [Princeton Plasma Physics Laboratory, POB 451, Princeton, NJ 08543 (United States)

    2013-07-15

    Lithium conditioned plasma facing surfaces have lowered recycling and enhanced plasma performance on many fusion devices. However, the nature of the plasma–lithium surface interaction has been obscured by the difficulty of in-tokamak surface analysis. We report laboratory studies of the chemical composition of lithium surfaces exposed to typical residual gases found in tokamaks. Solid lithium and a molybdenum alloy (TZM) coated with lithium have been examined using X-ray photoelectron spectroscopy, temperature programmed desorption, and Auger electron spectroscopy both in ultrahigh vacuum conditions and after exposure to trace gases. Lithium surfaces near room temperature were oxidized after exposure to 1–2 Langmuirs of oxygen or water vapor. The oxidation rate by carbon monoxide was four times less. Lithiated PFC surfaces in tokamaks will be oxidized in about 100 s depending on the tokamak vacuum conditions.

  4. 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.

  5. 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.

  6. 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

  7. 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.

  8. 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.

  9. 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.

  10. Experimental estimation of tungsten impurity sputtering due to Type i ELMs in JET-ITER-like wall using pedestal electron cyclotron emission and target Langmuir probe measurements

    Czech Academy of Sciences Publication Activity Database

    Guillemaut, C.; Jardin, A.; Horáček, Jan; Borodkina, I.; Autricque, A.; Arnoux, G.; Boom, J.; Brezinsek, S.; Coenen, J.W.; De La Luna, E.; Devaux, S.; Eich, T.; Harting, D.; Kirschner, A.; Lipschultz, B.; Matthews, G. F.; Meigs, A.; Moulton, D.; O'Mullane, M.; Stamp, M.

    T167, February (2016), s. 014005 ISSN 0031-8949. [International Conference on Plasma-Facing Materials and Components for Fusion Applications, PFMC 2015/15./. Aix-en-Provence, 18.05.2015-22.05.2015] R&D Projects: GA MŠk LG14002 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : tungsten sputtering * edge localized mode * magnetic confinement fusion * ITER * H-mode * ELMs * Langmuir Probes (LP) Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: 1.3 Physical sciences Impact factor: 1.280, year: 2016 http://iopscience.iop.org/issue/1402-4896/2016/T167

  11. Selection, development and characterisation of plasma facing materials for ITER

    International Nuclear Information System (INIS)

    Barabash, V.; Akiba, M.; Ulrickson, M.; Vieider, G.

    1996-01-01

    The current status of the selection of the armour materials for first wall, limiters and divertor are presented. The candidate armour materials are beryllium, tungsten and carbon base materials (mainly carbon fiber composites). The selection of the references grades from these material classes is discussed and the candidate grades are described. The main reasons for the selection of the reference grades are also discussed. The urgent materials R and D needs for the development of the design are described briefly. (orig.)

  12. Interaction of candidate plasma facing materials with tokamak plasma in COMPASS

    Czech Academy of Sciences Publication Activity Database

    Matějíček, Jiří; Weinzettl, Vladimír; Macková, Anna; Malinský, Petr; Havránek, Vladimír; Naydenkova, Diana; Klevarová, Veronika; Petersson, P.; Gasior, P.; Hakola, A.; Rubel, M.; Fortuna, E.; Kolehmainen, J.; Tervakangas, S.

    2017-01-01

    Roč. 493, September (2017), s. 102-119 ISSN 0022-3115. [International Conference on Plasma-Facing Materials and Components for Fusion Applications/15./. Aix-en-Provence, 18.05.2015-22.05.2015] R&D Projects: GA ČR(CZ) GA14-12837S; GA ČR(CZ) GA15-10723S; GA MŠk(CZ) LM2015045; GA MŠk LM2015056 Institutional support: RVO:61389021 ; RVO:61389005 Keywords : erosion * COMPASS tokamak * plasma-material interaction * ion beam analysis Subject RIV: JF - Nuclear Energetics; JF - Nuclear Energetics (UJF-V) OBOR OECD: Nuclear related engineering ; Nuclear related engineering (UJF-V) Impact factor: 2.048, year: 2016 http://www.sciencedirect.com/science/ article /pii/S0022311517301708

  13. 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).

  14. 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).

  15. Development of Si–W transient tolerant plasma facing material

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.C., E-mail: wongc@fusion.gat.com [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Chen, B. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Hollmann, E.M.; Rudakov, D.L. [University of California, San Diego, 9500 Gilman Dr., La Jolla, CA 92093-0417 (United States); Wall, D.; Tao, R. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Wright, M. [Ultramet, 12173 Montague Street, Pacoima, CA 91331 (United States)

    2013-07-15

    Solid W is projected as the preferred plasma facing material. Unfortunately, W surfaces could suffer radiation damage under DT operation and will melt under Type-I edge localized modes and disruption events. A possible approach is the use of a low-Z sacrificial material, like Si deposited on the W-surface to withstand a few type-I ELMs and/or disruptions via the vapor shielding effect. Accordingly, sets of Si–W test buttons were fabricated and exposed in the DIII-D lower divertor. We found that when the Si–W buttons were exposed to a few DIII-D vertical displacement event disruptions, tungsten–silicide was formed which melts at 1414 °C. This clearly indicates that the Si–W combination cannot be used as a transient tolerance surface material, since the W surface can be damaged. Even when Si is used as a wall conditioning material the Si–W surface temperature should be operated at much lower than 1400 °C.

  16. Supply of a prototype component for the ITER divertor baffle

    International Nuclear Information System (INIS)

    Bobin-Vastra, I.; Febvre, M.; Schedler, B.; Ploechl, L.; Bouveret, Y.; Cauvin, D.; Raisson, G.; Merola, M.

    2001-01-01

    The ITER divertor baffle is one of the Plasma facing components which are developed in the frame of the ITER concept. The supply consisted in the manufacturing of four panels with four First Wall geometries using macroblock or heat sink+armour concepts. DS-Copper, and CuCrZr were the materials for the heat sink, and CFC or Tungsten Plasma spray were the armour. The panels included two Copper-based tubes each. The final purpose is the comparison of the fabricability of each type and the performances of each panel under heat fluxes

  17. 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.

  18. 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

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

    Science.gov (United States)

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

    2018-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Stone, Howard A. [Princeton Univ., NJ (United States); Koel, Bruce E. [Princeton Univ., NJ (United States); Bernasek, Steven L. [Princeton Univ., NJ (United States); Carter, Emily A. [Princeton Univ., NJ (United States); Debenedetti, Pablo G. [Princeton Univ., NJ (United States); Panagiotopoulos, Athanassios Z. [Princeton Univ., NJ (United States)

    2017-06-23

    included (i) quantum mechanical calculations that allow inclusion of many thousands of atoms for the characterization of the interface of liquid metals exposed to continuous bombardment by deuterium and tritium as expected in fusion, (ii) molecular dynamics studies of the phase behavior of liquid metals, which (a) utilize thermodynamic properties computed using our quantum mechanical calculations and (b) establish material and wetting properties of the liquid metals, including relevant eutectics, (iii) experimental investigations of the surface science of liquid metals, interacting both with the solid substrate as well as gaseous species, and (iv) fluid dynamical studies that incorporate the material and surface science results of (ii) and (iii) in order to characterize flow in capillary porous materials and the thin-film flow along curved boundaries, both of which are potentially major components of plasma-facing materials. The outcome of these integrated studies was new understanding that enables developing design rules useful for future developments of the plasma-facing components critical to the success of fusion energy systems.

  1. 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

  2. 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

  3. 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.

  4. Radiation damage and redeposited-layer formation on plasma facing materials in the TRIAM-1M

    International Nuclear Information System (INIS)

    Hirai, Takeshi; Tokunaga, Kazutoshi; Fujiwara, Tadashi; Yoshida, Naoaki; Itoh, Satoshi

    1997-01-01

    As an aim to obtain some informations of material damage at long time discharge and redeposited-layer formed by scrape off layer (SOL), two collector probe experiments were conducted by using Tokamak of Research Institute for Applied Mechanics (TRIAM-IM). As a result, radiation damage due to charge exchange neutral particles of more than 2 MeV high energy component flying from plasma was observed. And in either experiment, redeposited-layer formation due to deposite of impurity atoms in the plasma could be observed. In the first experiment, a redeposited-layer with fine crystalline particles was observed, which was formed to contain multi-component system of Fe, Cr and Ni and light elements O and C. And, in the second experiment, a redeposited-layer grain-grown in which main component was Mo was observed. Surface modification of plasma facing material such as above-mentioned damage induction, redeposited-layer formation, and so on, was thought to much affect deterioration of materials and recycling of hydrogen. (G.K.)

  5. Tungsten isotopes in bulk meteorites and their inclusions-Implications for processing of presolar components in the solar protoplanetary disk

    DEFF Research Database (Denmark)

    Holst, Jesper Christian; Paton, Chad; Wielandt, Daniel Kim Peel

    2015-01-01

    We present high precision, low- and high-resolution tungsten isotope measurements of iron meteorites Cape York (IIIAB), Rhine Villa (IIIE), Bendego (IC), and the IVB iron meteorites Tlacotepec, Skookum, and Weaver Mountains, as well as CI chondrite Ivuna, a CV3 chondrite refractory inclusion (CAI...

  6. Results of high heat flux qualification tests of W monoblock components for WEST

    Science.gov (United States)

    Greuner, H.; Böswirth, B.; Lipa, M.; Missirlian, M.; Richou, M.

    2017-12-01

    One goal of the WEST project (W Environment in Steady-state Tokamak) is the manufacturing, quality assessment and operation of ITER-like actively water-cooled divertor plasma facing components made of tungsten. Six W monoblock plasma facing units (PFUs) from different suppliers have been successfully evaluated in the high heat flux test facility GLADIS at IPP. Each PFU is equipped with 35 W monoblocks of an ITER-like geometry. However, the W blocks are made of different tungsten grades and the suppliers applied different bonding techniques between tungsten and the inserted Cu-alloy cooling tubes. The intention of the HHF test campaign was to assess the manufacturing quality of the PFUs on the basis of a statistical analysis of the surface temperature evolution of the individual W monoblocks during thermal loading with 100 cycles at 10 MW m-2. These tests confirm the non-destructive examinations performed by the manufacturer and CEA prior to the installation of the WEST platform, and no defects of the components were detected.

  7. Results of high heat flux qualification tests of W monoblock components for WEST

    International Nuclear Information System (INIS)

    Greuner, H; Böswirth, B; Lipa, M; Missirlian, M; Richou, M

    2017-01-01

    One goal of the WEST project (W Environment in Steady-state Tokamak) is the manufacturing, quality assessment and operation of ITER-like actively water-cooled divertor plasma facing components made of tungsten. Six W monoblock plasma facing units (PFUs) from different suppliers have been successfully evaluated in the high heat flux test facility GLADIS at IPP. Each PFU is equipped with 35 W monoblocks of an ITER-like geometry. However, the W blocks are made of different tungsten grades and the suppliers applied different bonding techniques between tungsten and the inserted Cu-alloy cooling tubes. The intention of the HHF test campaign was to assess the manufacturing quality of the PFUs on the basis of a statistical analysis of the surface temperature evolution of the individual W monoblocks during thermal loading with 100 cycles at 10 MW m −2 . These tests confirm the non-destructive examinations performed by the manufacturer and CEA prior to the installation of the WEST platform, and no defects of the components were detected. (paper)

  8. 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.)

  9. 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)

  10. 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.

  11. The tritium confinement and surface chemistry of plasma facing materials in controlled D-T fusion devices

    International Nuclear Information System (INIS)

    Wu, C.H.

    1987-01-01

    Tritium permeation through first walls, limiters or divertors subjected to energetic tritium charge exchange neutral bombardment is a potentially serious problem area for advanced D-T reactors operating at elevated temperatures. High concentrations of tritium in the near surface region can be reached by implantation of the charge neutral flux combined with a relatively slow recombination of these atoms into molecules at the plasma/ first wall interface. A concentration gradient is established, causing tritium to diffuse into the bulk and essentially to the outer wall surface where it can enter the first wall coolant. Since tritium separation from cooling water is very costly, release of even a small fraction of tritium to the environment could pose undesirable safety problems. Therefore, it is necessary to reduce the tritium permeation. An analysis of the way of inhibition has been made. The tritium interacts with the solid surface of the plasma facing components, resulting in trapping and material erosion, and posing problems with respect to plasma density control. The erosion of the plasma facing component materials is mainly caused by physical and chemical erosion. A detailed analysis of chemical erosion by tritium has been performed and the results are described. (author)

  12. Irradiation-induced structure and property changes in tokamak plasma-facing, carbon-carbon composites

    International Nuclear Information System (INIS)

    Burchell, T.D.

    1994-01-01

    Carbon-carbon composites are an attractive choice for fusion reactor plasma-facing components because of their low atomic number, superior thermal shock resistance, and low neutron activation. Next generation plasma fusion reactors, such as the International Thermonuclear Experimental Reactor (ITER), will require advanced carbon-carbon composite materials possessing high thermal conductivity to manage the anticipated severe heat loads. Moreover, ignition machines such as ITER will produce large neutron fluxes. Consequently, the influence of neutron damage on the structure and properties of carbon-carbon composite materials must be evaluated. Data from two irradiation experiments are reported and discussed here. Carbon-carbon composite materials were irradiated in target capsules in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). A peak damage dose of 4.7 displacements per atom (dpa) at 600 degree C was attained. The carbon materials irradiated included uni-directional, two-directional, and three-directional carbon-carbon composites. Dimensional changes are reported for the composite materials and are related to single crystal dimensional changes through fiber and composite structural models. Moreover, the irradiation-induced dimensional changes are reported and discussed in terms of their architecture, fiber type, and graphitization temperature. The effect of neutron irradiation on thermal conductivity of two three-directional, carbon-carbon composites is reported and the recovery of thermal conductivity due to thermal annealing is discussed

  13. A bulk tungsten tile for JET: Heat flux tests in the MARION facility on the power-handling performance and validation of the thermal model

    International Nuclear Information System (INIS)

    Mertens, Ph.; Altmann, H.; Chaumet, P.; Joffrin, E.; Knaup, M.; Matthews, G.F.; Neubauer, O.; Nicolai, D.; Riccardo, V.; Tanchuk, V.; Thompson, V.; Uhlemann, R.; Samm, U.

    2011-01-01

    In the frame of the ITER-like Wall (ILW) for the JET tokamak, a divertor row made of bulk tungsten material has been developed for the position where the outer strike point is located in most of the foreseen plasma configurations. In the absence of active cooling, this represents a formidable challenge when one considers the temperature reached by tungsten (T W,surf > 2000 deg. C) and the vertical gradient ∂T/∂z = 5 x 10 4 K/m. As the development is drawing to an end and most components are in production, actual 1:1 prototypes are exposed to an ion beam with a power density around 7 MW/m 2 on the plasma-facing surface. Advantage is taken of the flexibility of the MARION facility to bombard the tungsten stack under shallow angles of incidence (∼6 o ) with a powerful beam of ions and neutrals (>70 MW/m 2 on axis). The shallow angles are important, with respect to the toroidal wetted surface, for properly simulating the expected performance under actual tokamak conditions. The MARION tests have been used to validate for a few typical cases the thermal calculations that were steadily developed along with the tungsten tile and, at the same time, to gather information on the actual temperatures of individual components. The latter is an important factor to a finer estimation of the power handling capabilities.

  14. 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.

  15. 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.)

  16. 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)

  17. 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

  18. US-Japan workshop Q-181 on high heat flux components and plasma-surface interactions for next devices: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    McGrath, R.T. [ed.] [Sandia National Labs., Albuquerque, NM (United States); Yamashina, T. [ed.] [Hokkadio Univ. (Japan)

    1994-04-01

    This report contain viewgraphs of papers from the following sessions: plasma facing components issues for future machines; recent PMI results from several tokamaks; high heat flux technology; plasma facing components design and applications; plasma facing component materials and irradiation damage; boundary layer plasma; plasma disruptions; conditioning and tritium; and erosion/redeposition.

  19. US-Japan workshop Q-181 on high heat flux components and plasma-surface interactions for next devices: Proceedings

    International Nuclear Information System (INIS)

    McGrath, R.T.; Yamashina, T.

    1994-04-01

    This report contain viewgraphs of papers from the following sessions: plasma facing components issues for future machines; recent PMI results from several tokamaks; high heat flux technology; plasma facing components design and applications; plasma facing component materials and irradiation damage; boundary layer plasma; plasma disruptions; conditioning and tritium; and erosion/redeposition

  20. Room and ultrahigh temperature structure-mechanical property relationships of tungsten alloys formed by field assisted sintering technique (FAST)

    Energy Technology Data Exchange (ETDEWEB)

    Browning, Paul N.; Alagic, Sven [Pennsylvania State University, Department of Materials Science and Engineering, State College, PA-16801 (United States); Pennsylvania State University, Applied Research Laboratory, State College, PA-16801 (United States); Kulkarni, Anil [Pennsylvania State University, Department of Nuclear and Mechanical Engineering, State College, PA-16801 (United States); Matson, Lawrence [Materials and Manufacturing Directorate, Wright Patterson Air Force Base, Dayton, OH (United States); Singh, Jogender, E-mail: jxs46@arl.psu.edu [Pennsylvania State University, Department of Materials Science and Engineering, State College, PA-16801 (United States); Pennsylvania State University, Applied Research Laboratory, State College, PA-16801 (United States)

    2016-09-30

    Tungsten based alloys have become of critical importance in a number of applications including plasma-facing materials in nuclear fusion reactors, rocket nozzles for aerospace applications, and in kinetic energy penetrators in the defense industry. Formation of components for these uses by powder metallurgical techniques has proven challenging, due to tungsten's relatively poor sinterability. Here we report the use of field assisted sintering technique (FAST) to produce high density, fine grain alloys with mechanical properties comparable or superior to that of components produced by conventional techniques. Alloys of pure tungsten, W-3 vol%TiC, W-5 vol%TiC, and W-10 vol%Ta were synthesized at 2100 °C, 35 MPa for 25 min using FAST. Microstructural characterization revealed effective reduction of grain size with TiC addition and preferential diffusion of oxygen into the center of tantalum particles in tantalum containing alloys. Tensile testing of alloys revealed TiC addition to W resulted in substantially improved ultimate tensile strength at the cost of ductility in comparison at temperatures up to 1926 °C (3500 °F) however this strengthening effect was lost at 2204 °C (4000 °F). Addition of 10 vol%Ta to W resulted in reduced hardness at room temperature, but substantially increased yield strength at the cost of slightly reduced ductility at 1926 °C and 2204 °C.

  1. 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.)

  2. 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.

  3. Tritium loading in ITER plasma-facing surfaces and its release under accident conditions

    International Nuclear Information System (INIS)

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

    1996-01-01

    Plasma-facing surfaces of the International Thermonuclear Experimental Reactor (ITER) will take up tritium from the plasma. These surfaces will probably consist of matures of Be, C, and possibly W together with other impurities. Recent experimental results have suggested mechanisms, not previously considered in analyses, by which tritium and other hydrogen isotopes are retained in Be. This warrants revised modeling and estimation of the amount of tritium that will be deposited in ITER beryllium plasma-facing surfaces and the rates at which it can be released under postulated accident scenarios. In this paper we describe improvements in modeling and experiments planned at the Idaho National Engineering Laboratory (INEL) to investigate the tritium uptake and thermal release behavior for mixed plasma- facing materials. TMAP4 calculations were made using recent data to estimate first-wall tritium inventories in ITER. 16 refs., 1 fig

  4. 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.

  5. Report of a technical evaluation panel on the use of beryllium for ITER plasma facing material and blanket breeder material

    Energy Technology Data Exchange (ETDEWEB)

    Ulrickson, M.A. [ed.] [Sandia National Labs., Albuquerque, NM (United States); Manly, W.D. [Oak Ridge National Lab., TN (United States); Dombrowski, D.E. [Brush Wellman, Inc., Cleveland, OH (United States)] [and others

    1995-08-01

    Beryllium because of its low atomic number and high thermal conductivity, is a candidate for both ITER first wall and divertor surfaces. This study addresses the following: why beryllium; design requirements for the ITER divertor; beryllium supply and unirradiated physical/mechanical property database; effects of irradiation on beryllium properties; tritium issues; beryllium health and safety; beryllium-coolant interactions and safety; thermal and mechanical tests; plasma erosion of beryllium; recommended beryllium grades for ITER plasma facing components; proposed manufacturing methods to produce beryllium parts for ITER; emerging beryllium materials; proposed inspection and maintenance techniques for beryllium components and coatings; time table and costs; and the importance of integrating materials and manufacturing personnel with designers.

  6. Report of a technical evaluation panel on the use of beryllium for ITER plasma facing material and blanket breeder material

    International Nuclear Information System (INIS)

    Ulrickson, M.A.; Manly, W.D.; Dombrowski, D.E.

    1995-08-01

    Beryllium because of its low atomic number and high thermal conductivity, is a candidate for both ITER first wall and divertor surfaces. This study addresses the following: why beryllium; design requirements for the ITER divertor; beryllium supply and unirradiated physical/mechanical property database; effects of irradiation on beryllium properties; tritium issues; beryllium health and safety; beryllium-coolant interactions and safety; thermal and mechanical tests; plasma erosion of beryllium; recommended beryllium grades for ITER plasma facing components; proposed manufacturing methods to produce beryllium parts for ITER; emerging beryllium materials; proposed inspection and maintenance techniques for beryllium components and coatings; time table and costs; and the importance of integrating materials and manufacturing personnel with designers

  7. 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.

  8. Experimental studies of lithium-based surface chemistry for fusion plasma-facing materials applications

    International Nuclear Information System (INIS)

    Allain, J.P.; Rokusek, D.L.; Harilal, S.S.; Nieto-Perez, M.; Skinner, C.H.; Kugel, H.W.; Heim, B.; Kaita, R.; Majeski, R.

    2009-01-01

    Lithium has enhanced the operational performance of fusion devices such as: TFTR, CDX-U, FTU, T-11 M, and NSTX. Lithium in the solid and liquid state has been studied extensively in laboratory experiments including its erosion and hydrogen-retaining properties. Reductions in physical sputtering up to 40-60% have been measured for deuterated solid and liquid lithium surfaces. Computational modeling indicates that up to a 1:1 deuterium volumetric retention in lithium is possible. This paper presents the results of systematic in situ laboratory experimental studies on the surface chemistry evolution of ATJ graphite under lithium deposition. Results are compared to post-mortem analysis of similar lithium surface coatings on graphite exposed to deuterium discharge plasmas in NSTX. Lithium coatings on plasma-facing components in NSTX have shown substantial reduction of hydrogenic recycling. Questions remain on the role lithium surface chemistry on a graphite substrate has on particle sputtering (physical and chemical) as well as hydrogen isotope recycling. This is particularly due to the lack of in situ measurements of plasma-surface interactions in tokamaks such as NSTX. Results suggest that the lithium bonding state on ATJ graphite is lithium peroxide and with sufficient exposure to ambient air conditions, lithium carbonate is generated. Correlation between both results is used to assess the role of lithium chemistry on the state of lithium bonding and implications on hydrogen pumping and lithium sputtering. In addition, reduction of factors between 10 and 30 reduction in physical sputtering from lithiated graphite compared to pure lithium or carbon is also measured.

  9. Mechanical characterization and modeling of brazed tungsten and Cu-Cr-Zr alloy using stress relief interlayers

    Science.gov (United States)

    Qu, Dandan; Zhou, Zhangjian; Yum, Youngjin; Aktaa, Jarir

    2014-12-01

    A rapidly solidified foil-type Ti-Zr based amorphous filler with a melting temperature of 850 °C was used to braze tungsten to Cu-Cr-Zr alloy for water cooled divertors and plasma facing components application. Brazed joints of dissimilar materials suffer from a mismatch in coefficients of thermal expansion. In order to release the residual stress caused by the mismatch, brazed joints of tungsten and Cu-Cr-Zr alloy using different interlayers were studied. The shear strength tests of brazed W/Cu joints show that the average strength of the joint with a W70Cu30 composite plate interlayer reached 119.8 MPa, and the average strength of the joint with oxygen free high conductivity copper (OFHC Cu)/Mo multi-interlayers reached 140.8 MPa, while the joint without interlayer was only 16.6 MPa. Finite element method (FEM) has been performed to investigate the stress distribution and effect of stress relief interlayers. FEM results show that the maximum von Mises stress occurs in the tungsten/filler interface and that the filler suffers the peak residual stresses and becomes the weakest zone. And the use of OFHC Cu/Mo multi-interlayers can reduce the residual stress significantly, which agrees with the mechanical experiment data.

  10. Mechanical characterization and modeling of brazed tungsten and Cu–Cr–Zr alloy using stress relief interlayers

    Energy Technology Data Exchange (ETDEWEB)

    Qu, Dandan, E-mail: dandan.qu@partner.kit.edu [School of Materials Science and Engineering, University of Science and Technology Beijing, 100083 Beijing (China); Karlsruhe Institute of Technology, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Zhou, Zhangjian, E-mail: zhouzhangjianustb@163.com [School of Materials Science and Engineering, University of Science and Technology Beijing, 100083 Beijing (China); Yum, Youngjin [School of Mechanical Engineering, University of Ulsan, Ulsan 680-749 (Korea, Republic of); Aktaa, Jarir [Karlsruhe Institute of Technology, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2014-12-15

    A rapidly solidified foil-type Ti–Zr based amorphous filler with a melting temperature of 850 °C was used to braze tungsten to Cu–Cr–Zr alloy for water cooled divertors and plasma facing components application. Brazed joints of dissimilar materials suffer from a mismatch in coefficients of thermal expansion. In order to release the residual stress caused by the mismatch, brazed joints of tungsten and Cu–Cr–Zr alloy using different interlayers were studied. The shear strength tests of brazed W/Cu joints show that the average strength of the joint with a W70Cu30 composite plate interlayer reached 119.8 MPa, and the average strength of the joint with oxygen free high conductivity copper (OFHC Cu)/Mo multi-interlayers reached 140.8 MPa, while the joint without interlayer was only 16.6 MPa. Finite element method (FEM) has been performed to investigate the stress distribution and effect of stress relief interlayers. FEM results show that the maximum von Mises stress occurs in the tungsten/filler interface and that the filler suffers the peak residual stresses and becomes the weakest zone. And the use of OFHC Cu/Mo multi-interlayers can reduce the residual stress significantly, which agrees with the mechanical experiment data.

  11. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Marian, Jaime; Becquart, Charlotte S.; Domain, Christophe; Dudarev, Sergei L.; Gilbert, Mark R.; Kurtz, Richard J.; Mason, Daniel R.; Nordlund, Kai; Sand, Andrea E.; Snead, Lance L.; Suzudo, Tomoaki; Wirth, Brian D.

    2017-06-09

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

  13. Beryllium-metals joints for application in the plasma-facing components

    International Nuclear Information System (INIS)

    Barabash, V.R.; Gitarsky, L.S.; Ignakovskaya, G.S.; Prokofiev, Yu.G.

    1994-01-01

    The results of the technological development for Be joining with other metals for high heat flux application are presented. The different types of joining technology - high temperature brazing by using the different brazing alloys and solid state diffusion bonding are compared. The results of diffusion bonding technology development for Be-Cu and Be-dispersion strengthened copper alloys joinings are presented. It was shown that for the joining of Be with austenitic stainless steels, the vacuum high temperature brazing using the ternary brazing alloy of Ag-Cu-Me system is more preferable than common eutectic Ag-Cu alloy. The high temperature brazing technology for joining Be-Be using the Al brazing alloys was also analyzed. The problems of nondestructive examination of Be joints, the data on mechanical properties, microhardness testing and results of microstructural examination of Be joint are presented. ((orig.))

  14. NIFS joint research meeting on plasma facing components, PSI, and heat/particle control

    International Nuclear Information System (INIS)

    Yamashina, T.

    1997-10-01

    The LHD collaboration has been started in 1996. Particle and heat control is one of the categories for the collaboration, and a few programs have been nominated in these two years. A joint research meeting on PFC, PSI, heat and particle meeting was held at NIFS on June 27, 1997, in which present status of these programs were reported. This is a collection of the notes and view graphs presented in this meeting. Brief reviews and research plan of each program are included in relation to divertor erosion and sputtering, impurity generation, hydrogen recycling, edge plasma structure, edge transport and its control, heat removal, particle exhaust, wall conditioning etc. (author)

  15. Effect of heat loads on the plasma facing components of demo

    Energy Technology Data Exchange (ETDEWEB)

    Igitkhanov, Yu., E-mail: juri.igitkhanov@partner.kit.edu [ITEP, Karlsruhe Institute of Technology (Germany); Fetzer, R. [IHM, Karlsruhe Institute of Technology (Germany); Bazylev, B. [INR, Karlsruhe Institute of Technology (Germany)

    2016-11-01

    Highlights: • Under the DEMO1 stationary operation the nominal power fluxes along the magnetic field at the FW blanket modules is expected about 50 MW/m{sup 2}. • In the current design and averaged incident angle about 3–4.5° (similar to ITER) the engineering power load to the FW is expected within 2.5÷3.9 MW/m{sup 2}. • In the case of the unmitigated Type I ELMs unavoidable in the higher confinement H-mode of operation energy load per ELM is about 20 MJ/m{sup 2} along the field line, arriving at a frequency of 0.8 Hz with deposition time of 0.6 ms per each ELM. - Abstract: In this paper we analyse a thermo-hydraulic performance of the first wall blanket module during the stationary DEMO operation with the edge localized mode (ELM). Heat loads are estimated based on scaling arguments and predictions from the peeling-ballooning ELM model. Effect of parallel heat fluxes intersecting with the first wall panels and avoidance of overheating by inclination of the panels are considered. The material temperatures of the W/EUROFER sandwich type module with water cooling stainless steel tube and Cu alloy compliance embedded into EUROFER is calculated by using the MEMOS code. The calculations were carried out indicating the required geometric parameters as well as the cooling conditions which allow keeping materials temperatures within allowable engineering limits. Effect of inclination of the first wall plates to avoid the misalignment problems is considered.

  16. Experimental activity on the definition of acceptance criteria for the ITER divertor plasma facing components

    International Nuclear Information System (INIS)

    Escourbiac, F.; Constans, S.; Vignal, N.; Cantone, V.; Richou, M.; Durocher, A.; Riccardi, B.; Bobin, I.; Jouvelot, J.L.; Merola, M.

    2009-01-01

    Tens of thousands of armor/heat sink joints will be produced by the industry during the manufacturing of ITER divertor PFC, statistically, there is a probability that joints with defects be delivered. The purpose of this paper is to study the detection and evolution during operation of calibrated defects artificially implemented on samples, as an experimental basis for the definition of acceptance criteria for the bond armor/heat sink in the frame of industrial manufacturing conditions.It was found that current CFC monoblock design option was compatible with the heat loads specified at the lower part of the vertical target (up to 20 MW/m 2 ), including the presence of armor/heat sink defects (up to 50 deg. extension for a location at 0 deg. or 45 deg.) detectable with NDE techniques developed in Europe (US, SATIR). The current W monoblock design appeared suitable for the upper part of the vertical target with defects extension up to 50 deg. but is not adapted for heat flux of 20 MW/m 2 . The studied W flat tile design proved to be compatible with fluxes of 5 MW/m 2 but unable to sustain cycling fluxes of 10 MW/m 2 .

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

    Science.gov (United States)

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

    2017-12-01

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

  18. Heat loads on JET plasma facing components from ICRF and LH wave absorption in the SOL

    Czech Academy of Sciences Publication Activity Database

    Jacquet, P.; Colas, L.; Mayoral, M.-L.; Arnoux, G.; Bobkov, V.; Brix, M.; Coad, P.; Czarnecka, A.; Dodt, D.; Durodie, F.; Ekedahl, A.; Frigione, D.; Fursdon, M.; Gauthier, E.; Goniche, M.; Graham, M.; Joffrin, E.; Korotkov, A.; Lerche, E.; Mailloux, J.; Monakhov, I.; Noble, C.; Ongena, J.; Petržílka, Václav; Portafaix, C.; Rimini, F.; Sirinelli, A.; Riccardo, V.; Vizvary, Z.; Widdowson, A.; Zastrow, K.-D.

    2011-01-01

    Roč. 51, č. 10 (2011), s. 103018-103018 ISSN 0029-5515 R&D Projects: GA ČR GA202/07/0044 Institutional research plan: CEZ:AV0Z20430508 Keywords : LH wave * plasma Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.090, year: 2011 http://iopscience.iop.org/0029-5515/51/10/103018/pdf/0029-5515_51_10_103018.pdf

  19. 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.

  20. 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.

  1. ELM elimination with Li powder injection in EAST discharges using the tungsten upper divertor

    Science.gov (United States)

    Maingi, R.; Hu, J. S.; Sun, Z.; Tritz, K.; Zuo, G. Z.; Xu, W.; Huang, M.; Meng, X. C.; Canik, J. M.; Diallo, A.; Lunsford, R.; Mansfield, D. K.; Osborne, T. H.; Gong, X. Z.; Wang, Y. F.; Li, Y. Y.; EAST Team

    2018-02-01

    We report the first successful use of lithium (Li) to eliminate edge-localized modes (ELMs) with tungsten divertor plasma-facing components in the EAST device. Li powder injected into the scrape-off layer of the tungsten upper divertor successfully eliminated ELMs for 3-5 s in EAST. The ELM elimination became progressively more effective in consecutive discharges at constant lithium delivery rates, and the divertor D α baseline emission was reduced, both signatures of improved wall conditioning. A modest decrease in stored energy and normalized energy confinement was also observed, but the confinement relative to H98 remained well above 1, extending the previous ELM elimination results via Li injection into the lower carbon divertor in EAST (Hu et al 2015 Phys. Rev. Lett. 114 055001). These results can be compared with recent observations with lithium pellets in ASDEX-Upgrade that failed to mitigate ELMs (Lang et al 2017 Nucl. Fusion 57 016030), highlighting one comparative advantage of continuous powder injection for real-time ELM elimination.

  2. 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)

  3. 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

  4. 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.

  5. Molecular dynamics simulations of interactions between energetic dust and plasma-facing materials

    International Nuclear Information System (INIS)

    Niu, Guo-jian; Li, Xiao-chun; Xu, Qian; Yang, Zhong-shi; Luo, Guang-nan

    2015-01-01

    The interactions between dust and plasma-facing material (PFM) relate to the lifetime of PFM and impurity production. Series results have been obtained theoretically and experimentally but more detailed studies are needed. In present research, we investigate the evolution of kinetic, potential and total energy of plasma-facing material (PFM) in order to understand the dust/PFM interaction process. Three typical impacting energy are selected, i.e., 1, 10 and 100 keV/dust for low-, high- and hyper-energy impacting cases. For low impacting energy, dust particles stick on PFM surface without damaging it. Two typical time points exist and the temperature of PFM grows all the time but PFM structure experience a modifying process. Under high energy case, three typical points appear. The temperature curve fluctuates in the whole interaction process which indicates there are dust/PFM and kinetic/potential energy exchanges. In the hyper-energy case in present simulation, the violence dust/PFM interactions cause sputtering and crater investigating on energy evolution curves. We further propose the statistics of energy distribution. Results show that about half of impacting energy consumes on heating plasma-facing material meanwhile the other half on PFM structure deformation. Only a small proportion becomes kinetic energy of interstitial or sputtering atoms.

  6. Molecular dynamics simulations of interactions between energetic dust and plasma-facing materials

    Energy Technology Data Exchange (ETDEWEB)

    Niu, Guo-jian, E-mail: niugj@ipp.ac.cn [Institute of Plasma Physics Chinese Academy of Sciences, Hefei (China); Li, Xiao-chun; Xu, Qian; Yang, Zhong-shi [Hefei Center Physical Science and Technology, Hefei (China); Luo, Guang-nan [Institute of Plasma Physics Chinese Academy of Sciences, Hefei (China); Hefei Center Physical Science and Technology, Hefei (China); Hefei Science Center of CAS, Hefei (China)

    2015-11-15

    The interactions between dust and plasma-facing material (PFM) relate to the lifetime of PFM and impurity production. Series results have been obtained theoretically and experimentally but more detailed studies are needed. In present research, we investigate the evolution of kinetic, potential and total energy of plasma-facing material (PFM) in order to understand the dust/PFM interaction process. Three typical impacting energy are selected, i.e., 1, 10 and 100 keV/dust for low-, high- and hyper-energy impacting cases. For low impacting energy, dust particles stick on PFM surface without damaging it. Two typical time points exist and the temperature of PFM grows all the time but PFM structure experience a modifying process. Under high energy case, three typical points appear. The temperature curve fluctuates in the whole interaction process which indicates there are dust/PFM and kinetic/potential energy exchanges. In the hyper-energy case in present simulation, the violence dust/PFM interactions cause sputtering and crater investigating on energy evolution curves. We further propose the statistics of energy distribution. Results show that about half of impacting energy consumes on heating plasma-facing material meanwhile the other half on PFM structure deformation. Only a small proportion becomes kinetic energy of interstitial or sputtering atoms.

  7. 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

  8. 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

  9. 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

  10. 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

  11. Micro creep mechanisms of tungsten

    International Nuclear Information System (INIS)

    Levoy, R.; Hugon, I.; Burlet, H.; Baillin, X.; Guetaz, L.

    2000-01-01

    Due to its high melting point (3410 deg C), tungsten offers good mechanical properties at elevated temperatures for several applications in non-oxidizing environment. The creep behavior of tungsten is well known between 1200 and 2500 deg C and 10 -3 to 10 -1 strain. However, in some applications when dimensional stability of components is required, these strains are excessive and it is necessary to know the creep behavior of the material for micro-strains (between 10 -4 and 10 -6 ). Methods and devices used to measure creep micro-strains are presented, and creep equations (Norton and Chaboche laws) were developed for wrought, annealed and recrystallized tungsten. The main results obtained on tungsten under low stresses are: stress exponent 1, symmetry of micro-strains in creep-tension and creep-compression, inverse creep (threshold stress), etc. TEM, SEM and EBSD studies allow interpretation of the micro-creep mechanism of tungsten under low stresses and low temperature (∼0.3 K) like the Harper-Dorn creep. In Harper-Dorn creep, micro-strains are associated with the density and the distribution of dislocations existing in the crystals before creep. At 975 deg C, the initial dislocation structure moves differently whether or not a stress is applied. To improve the micro-creep behavior of tungsten, a heat treatment is proposed to create the optimum dislocation structure. (authors)

  12. Analysis of the quasi-continuum band emitted by highly ionised tungsten atoms in the 4-7 nm range

    International Nuclear Information System (INIS)

    Madeira, T.I.; Amorim, P.; Marques, J.P.; Parente, F.; Indelicato, P.

    2013-01-01

    For the next upcoming generation of fusion experiments, such as ITER,Tungsten has been chosen as the materials for plasma facing components. Spectra emitted by highly ionized tungsten atoms from magnetically confined plasmas show a common feature: a narrow structured quasi-continuum emission band most prominent in the range 4-7 nm, which accounts for 40-80% of the radiated power. This band has been fairly well explained by unresolved transitions from groups 4d-4p, 4f-4d (Δn = 0) and 5d-4f, 5g-4f and 5p-4d (Δn = 1). In this work we use a Multi-Configuration Dirac-Fock code in Breit self-consistent field mode to compute level energies and transition probabilities for W 27+ to W 37+ ions contributing to this emission band. Intra-shell correlation was introduced in the calculation for both initial and final states and all dipole and quadrupole radiative transitions have been considered. The wavefunctions in the initial and final states are optimized separately and the resulting non-orthogonality effect is fully taken into account. The importance of some satellite lines was assessed. Together with the ionic distributions obtained by using the FLYCHK application and assuming that the initial states population depends statistically on the temperature we were able to synthesize plasma emission spectrum profiles for several electron temperatures. (authors)

  13. 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.

  14. Fabrication and characterization of tungsten and graphite based PFC for divertor target elements of ITER like tokamak application

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-15

    The development of the fabrication technology of macro-brush configuration of tungsten (W) and carbon (graphite and CFC) plasma facing components (PFCs) for ITER like tokamak application is presented. The fabrication of qualified joint of PFC is a requirement for fusion tokamak. Vacuum brazing method has been employed for joining of W/CuCrZr and C/CuCrZr. Oxygen free high conductivity (OFHC) copper casting on W tiles was performed followed by machining, polishing and ultrasonic cleaning of the samples prior to vacuum brazing. The W/CuCrZr and graphite/CuCrZr based test mockups were vacuum brazed using silver free alloys. The mechanical shear and tensile strengths were evaluated for the W/CuCrZr and graphite/CuCrZr brazed joint samples. The micro-structural examination of the joints showed smooth interface. The details of fabrication and characterization procedure for macro-brush tungsten and carbon based PFC test mockups are presented.

  15. 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)

  16. 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.

  17. Boron carbide-coated carbon material, manufacturing method therefor and plasma facing material

    International Nuclear Information System (INIS)

    Suzuki, Takayuki; Kikuchi, Yoshihiro; Hyakki, Yasuo.

    1997-01-01

    The present invention concerns a plasma facing material suitable to a thermonuclear device. The material comprises a carbon material formed by converting the surface of a carbon fiber-reinforced carbon material comprising a carbon matrix and carbon fibers to a boron carbide, the material has a surface comprising vertically or substantially vertically oriented carbon fibers, and the thickness of the surface converted to boron carbide is reduced in the carbon fiber portion than in the carbon matrix portion. Alternatively, a carbon fiber-reinforced carbon material containing carbon fibers having a higher graphitizing degree than the carbon matrix is converted to boron carbide on the surface where the carbon fibers are oriented vertically or substantially vertically. The carbon fiber-reinforced material is used as a base material, and a resin material impregnated into a shaped carbon fiber product is carbonized or thermally decomposed carbon is filled as a matrix. The material of the present invention has high heat conduction and excellent in heat resistance thereby being suitable to a plasma facing material for a thermonuclear device. Electric specific resistivity of the entire coating layer can be lowered, occurrence of arc discharge is prevented and melting can be prevented. (N.H.)

  18. 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

  19. Micro-powder injection moulding of tungsten

    International Nuclear Information System (INIS)

    Zeep, B.

    2007-12-01

    For He-cooled Divertors as integral components of future fusion power plants, about 300000 complex shaped tungsten components are to be fabricated. Tungsten is the favoured material because of its excellent properties (high melting point, high hardness, high sputtering resistance, high thermal conductivity). However, the material's properties cause major problems for large scale production of complex shaped components. Due to the resistance of tungsten to mechanical machining, new fabrication technologies have to be developed. Powder injection moulding as a well established shaping technology for a large scale production of complex or even micro structured parts might be a suitable method to produce tungsten components for fusion applications but is not yet commercially available. The present thesis is dealing with the development of a powder injection moulding process for micro structured tungsten components. To develop a suitable feedstock, the powder particle properties, the binder formulation and the solid load were optimised. To meet the requirements for a replication of micro patterned cavities, a special target was to define the smallest powder particle size applicable for micro-powder injection moulding. To investigate the injection moulding performance of the developed feedstocks, experiments were successfully carried out applying diverse cavities with structural details in micro dimension. For debinding of the green bodies, a combination of solvent debinding and thermal debinding has been adopted for injection moulded tungsten components. To develop a suitable debinding strategy, a variation of the solvent debinding time, the heating rate and the binder formulation was performed. For investigating the thermal consolidation behaviour of tungsten components, sinter experiments were carried out applying tungsten powders suitable for micro-powder injection moulding. First mechanical tests of the sintered samples showed promising material properties such as a

  20. 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

  1. Repair of manufacturing defects in the armor of plasma facing units of the ITER Divertor Dome

    International Nuclear Information System (INIS)

    Litunovsky, Nikolay; Alekseenko, Evgeny; Kuznetsov, Vladimir; Lyanzberg, Dmitriy; Makhankov, Aleksey; Rulev, Roman

    2013-01-01

    Highlights: • Sporadic manufacturing defects in ITER Divertor Dome PFUs may be repaired. • We have developed a repair technique for ITER Divertor Dome PFUs. • Armor repair technique for ITER Divertor Dome PFUs is successfully tested. -- Abstract: The paper describes the repair procedure developed for removal of manufacturing defects occurring sporadically during armoring of plasma facing units (PFUs) of the ITER Divertor Dome. Availability of armor repair technique is prescribed by the procurement arrangement for the ITER Divertor Dome concluded in 2009 between the ITER Organization and the ITER Domestic Agency of Russia. The paper presents the detailed description of the procedure, data on its effect on the joints of the rest part of the armor and on the grain structure of the PFU heat sink. The results of thermocycling of large-scale Dome PFU mock-ups manufactured with demonstration of armor repair are also given

  2. Evaluation of surface fractal dimension of carbon for plasma-facing material damaged by hydrogen plasma

    International Nuclear Information System (INIS)

    Nishino, Nobuhiro

    1997-01-01

    The surface structure of the plasma facing materials (PFM) changes due to plasma-surface interaction in a nuclear fusion reactor. Usually B 4 C coated graphite block are used as PFM. In this report, the surface fractal was applied to study the surface structure of plasma-damaged PFM carbon. A convenient flow-type adsorption apparatus was developed to evaluate the surface fractal dimension of materials. Four branched alkanol molecules with different apparent areas were used as the probe adsorbates. The samples used here were B 4 C coated isotopic graphite which were subjected to hydrogen plasma for various periods of exposure. The monolayer capacities of these samples for alkanols were determined by applying BET theory. The surface fractal dimension was calculated using the monolayer capacities and molecular areas for probe molecules and was found to increase from 2 to 3 with the plasma exposure time. (author)

  3. Observation of plasma-facing-wall via high dynamic range imaging

    International Nuclear Information System (INIS)

    Villamayor, Michelle Marie S.; Rosario, Leo Mendel D.; Viloan, Rommel Paulo B.

    2013-01-01

    Pictures of plasmas and deposits in a discharge chamber taken by varying shutter speeds have been integrated into high dynamic range (HDR) images. The HDR images of a graphite target surface of a compact planar magnetron (CPM) discharge device have clearly indicated the erosion pattern of the target, which are correlated to the light intensity distribution of plasma during operation. Based upon the HDR image technique coupled to colorimetry, a formation history of dust-like deposits inside of the CPM chamber has been recorded. The obtained HDR images have shown how the patterns of deposits changed in accordance with discharge duration. Results show that deposition takes place near the evacuation ports during the early stage of the plasma discharge. Discoloration of the plasma-facing-walls indicating erosion and redeposition eventually spreads at the periphery after several hours of operation. (author)

  4. Repair of manufacturing defects in the armor of plasma facing units of the ITER Divertor Dome

    Energy Technology Data Exchange (ETDEWEB)

    Litunovsky, Nikolay, E-mail: nlitunovsky@sintez.niiefa.spb.su; Alekseenko, Evgeny; Kuznetsov, Vladimir; Lyanzberg, Dmitriy; Makhankov, Aleksey; Rulev, Roman

    2013-10-15

    Highlights: • Sporadic manufacturing defects in ITER Divertor Dome PFUs may be repaired. • We have developed a repair technique for ITER Divertor Dome PFUs. • Armor repair technique for ITER Divertor Dome PFUs is successfully tested. -- Abstract: The paper describes the repair procedure developed for removal of manufacturing defects occurring sporadically during armoring of plasma facing units (PFUs) of the ITER Divertor Dome. Availability of armor repair technique is prescribed by the procurement arrangement for the ITER Divertor Dome concluded in 2009 between the ITER Organization and the ITER Domestic Agency of Russia. The paper presents the detailed description of the procedure, data on its effect on the joints of the rest part of the armor and on the grain structure of the PFU heat sink. The results of thermocycling of large-scale Dome PFU mock-ups manufactured with demonstration of armor repair are also given.

  5. 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.

  6. 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.

  7. The effect of low energy helium ion irradiation on tungsten-tantalum (W-Ta) alloys under fusion relevant conditions

    Science.gov (United States)

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

    2017-08-01

    Currently, tungsten remains the best candidate for plasma-facing components (PFCs) for future fusion devices because of its high melting point, low erosion, and strong mechanical properties. However, continued investigation has shown tungsten to undergo severe morphology changes under fusion-like conditions. These results motivate the study of innovative PFC materials which are resistant to surface morphology evolution. The goal of this work is to examine tungsten-tantalum (W-Ta) alloys, a potential PFC material, and their response to low energy helium ion irradiation. Specifically, W-Ta samples are exposed to 100 eV helium irradiations with a flux of 1.15 × 1021 ions m-2 s-1, at 873 K, 1023 K, and 1173 K for 1 h duration. Scanning electron microscopy (SEM) reveals significant changes in surface deterioration due to helium ion irradiation as a function of both temperature and tantalum concentration in W-Ta samples. X-Ray Diffraction (XRD) studies show a slight lattice parameter expansion in W-Ta alloy samples compared to pure W samples. The observed lattice parameter expansion in W-Ta alloy samples (proportional to increasing Ta wt.% concentrations) reflect significant differences observed in the evolution of surface morphology, i.e., fuzz development processes for both increasing Ta wt.% concentration and target temperature. These results suggest a correlation between the observed morphology differences and the induced crystal structure change caused by the presence of tantalum. Shifts in the XRD peaks before and after 100 eV helium irradiation with a flux of 1.15 × 1021 ions m-2 s-1, 1023 K, for 1 h showed a significant difference in the magnitude of the shift. This has suggested a possible link between the atomic spacing of the material and the accumulated damage. Ongoing research is needed on W-Ta alloys and other innovative materials for their application as irradiation resistant materials in future fusion or irradiation environments.

  8. Behavior of tungsten coatings on CuCrZr heat sink with the different interlayers under high heat flux

    International Nuclear Information System (INIS)

    Chong, F.L.; Chen, J.L.; Li, J.G.; Zheng, X.B.; Hu, D.Y.; Ding, C.X.

    2007-01-01

    In recent years, tungsten coated CuCrZr by means of vacuum plasma spraying technology was studied at Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). Plasma spraying technology is a good integration way of armor material and heat sink, which overcomes the disadvantage of heavy weight and poor workability of tungsten, and offers the ability to coat large area, even complex shapes and in situ repair of damaged parts. But tungsten coated CuCrZr is a challenge due to the larger mismatch of their thermal expansion coefficients (CTE), which will induce the stress concentration on the joint interface of plasma facing component. In order to enhance the adhesion of W coating on CuCrZr substrate and avoid the thermal stress concentration, it is necessary to use a compliant interlayer. At present, titanium (Ti), nickel-chromium-aluminum (NiCrAl) alloys and W/Cu mixtures were chosen as the compliant layers to insert between W coating and CuCrZr substrate. The adhesion strength was performed at RT. The behaviors of W/Cu mock up under high heat flux were carried out by means of the electron beam facility with actively cooling. The results indicated that the mock-ups with the interlayer architectures can withstand the higher heat flux compared to that with the sharp interface, which exhibited the effect of interlayers on reducing the maximum stress and enhancing the properties of resistant heat flux load, though the maximum surface temperature increased due to inserting the interlayers. Among three interlayers, W/Cu interlayer was much better due to its good heat removal capability and flexible W/Cu ratios. Meanwhile, the behaviors of W/Cu mock-ups with the different interlayers were analyzed and optimized by ANSYS finite element code. (authors)

  9. Design, manufacture and initial operation of the beryllium components of the JET ITER-like wall

    Energy Technology Data Exchange (ETDEWEB)

    Riccardo, V., E-mail: valeria.riccardo@ccfe.ac.uk [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Lomas, P.; Matthews, G.F. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Nunes, I. [Associação EURATOM-IST, IPFN – Laboratório Associado, IST, Lisbon (Portugal); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Thompson, V. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Villedieu, E. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom)

    2013-10-15

    Highlights: ► 40 m{sup 2} of plasma facing surface covered with bulk Be re-using existing supports, designed for C-based tiles (hence for much lower disruption loads). ► Optimization of power handling to allow compatibility with higher (×1.5) and longer (×2) neutral beam power. ► Beryllium re-cycling. ► Machining and cleaning to ultra high vacuum standards of <350 μm thin castellations in Be. ► Quality control to minimize installation problems (proto-types, full scale jigs, inspections). -- Abstract: The aim of the JET ITER-like wall project was to provide JET with the plasma facing material combination now selected for the DT phase of ITER (bulk beryllium main chamber limiters and a full tungsten divertor) and, in conjunction with the upgraded neutral beam heating system, to achieve ITER relevant conditions. The design of the bulk Be plasma facing components had to be compatible with increased heating power and pulse length, as well as to reuse the existing tile supports originally designed to cope with disruption loads from carbon based tiles and be installed by remote handling. Risk reduction measures (prototypes, jigs, etc.) were implemented to maximize efficiency during the shutdown. However, a large number of clashes with existing components not fully captured by the configuration model occurred. Restarting the plasma on the ITER-like Wall proved much easier than for the carbon wall and no deconditioning by disruptions was observed. Disruptions have been more threatening than expected due to the reduced radiative losses compared to carbon, leaving most of the plasma magnetic energy to be conducted to the wall and requiring routine disruption mitigation. The main chamber power handling has achieved and possibly exceeded the design targets.

  10. Design, manufacture and initial operation of the beryllium components of the JET ITER-like wall

    International Nuclear Information System (INIS)

    Riccardo, V.; Lomas, P.; Matthews, G.F.; Nunes, I.; Thompson, V.; Villedieu, E.

    2013-01-01

    Highlights: ► 40 m 2 of plasma facing surface covered with bulk Be re-using existing supports, designed for C-based tiles (hence for much lower disruption loads). ► Optimization of power handling to allow compatibility with higher (×1.5) and longer (×2) neutral beam power. ► Beryllium re-cycling. ► Machining and cleaning to ultra high vacuum standards of <350 μm thin castellations in Be. ► Quality control to minimize installation problems (proto-types, full scale jigs, inspections). -- Abstract: The aim of the JET ITER-like wall project was to provide JET with the plasma facing material combination now selected for the DT phase of ITER (bulk beryllium main chamber limiters and a full tungsten divertor) and, in conjunction with the upgraded neutral beam heating system, to achieve ITER relevant conditions. The design of the bulk Be plasma facing components had to be compatible with increased heating power and pulse length, as well as to reuse the existing tile supports originally designed to cope with disruption loads from carbon based tiles and be installed by remote handling. Risk reduction measures (prototypes, jigs, etc.) were implemented to maximize efficiency during the shutdown. However, a large number of clashes with existing components not fully captured by the configuration model occurred. Restarting the plasma on the ITER-like Wall proved much easier than for the carbon wall and no deconditioning by disruptions was observed. Disruptions have been more threatening than expected due to the reduced radiative losses compared to carbon, leaving most of the plasma magnetic energy to be conducted to the wall and requiring routine disruption mitigation. The main chamber power handling has achieved and possibly exceeded the design targets

  11. Tungsten and tungsten alloys by powder metallurgy

    International Nuclear Information System (INIS)

    Belhadjhamida, A.; German, R.M.

    1991-01-01

    Tungsten has a historical link with powder metallurgy and there is continued progress in expanding the available compositions and processing options. This paper starts with an introduction to the history of tungsten powder metallurgy and use this as a basis for analyzing some of the current trends. The literature base in tungsten processing is expanding and includes new alloys, microstructures, and processing routes. A few examples will be emphasize here to produce a frame work for this program, including description of sintering mechanisms for tungsten, liquid phase sintering advances, hot consolidation fundamentals, and options for complex shaping using powder injection modeling. For this base, subsequent presentations will expand on these fundamental advances

  12. Thermal resistivity of tungsten grades under fusion relevant conditions

    Energy Technology Data Exchange (ETDEWEB)

    Wirtz, M.; Linke, J.; Pintsuk, G. [Forschungszentrum Juelich (Germany). EURATOM Association

    2010-05-15

    Controlled nuclear fusion on earth is a very promising but also a very challenging task. Fusion devices like ITER and DEMO are major steps on the way of solving the energy problems of the future. However, the realisation of such thermonuclear fusion reactors still needs high efforts in many areas of research. One of the most critical issues is the field of in - vessel materials and components and in particular the plasma facing material (PFM). This not only has to be compatible to the heat sink material being able to withstand thermal fatigue loading conditions during steady state heat loading (up to 20 MW/m{sup 2}) but also has to withstand extreme thermal loads during transient events. The latter are divided into normal and off normal events, such as plasma disruptions or vertical displacement events (VDEs), resulting in irreversible damage of the material. Therefore they have to be avoided in future fusion devices by an improved plasma control. In contrast, edge localized modes (ELMs) occur during normal operation and are the result of complex plasma configuration. In the next step experiment ITER they are generated with a frequency of {>=} 1 Hz and a duration of 200 - 500 {mu}s depositing energies of {<=} 1 MJ/m{sup 2}. One of the most promising materials for the application as PFM in particular in the divertor region is tungsten. Its main advantages are a high thermal conductivity, a high melting temperature, a low tritium inventory and a low erosion rate. However there are some drawbacks like a high ductile to brittle transitions temperature (DBTT), its high atomic number Z and the remarkable neutron irradiation induced activation and degradation of its mechanical properties. The main aim of future R and D will be to understand the mechanisms of thermal induced damages and subsequently to minimize these types of damages. Therefore various tungsten grades have to be tested under fusion relevant conditions, e.g. by electron, ion or plasma beam exposure; the

  13. Effect of collision cascades on dislocations in tungsten: A molecular dynamics study

    Energy Technology Data Exchange (ETDEWEB)

    Fu, B.Q., E-mail: bqfu@scu.edu.cn [Key Laboratory for Radiation Physics and Technology, Institute of Nuclear Science and Technology, Sichuan University, Chengdu 610065 (China); Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); Fitzgerald, S.P. [Department of Applied Mathematics, University of Leeds, Leeds LS2 9JT (United Kingdom); Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); Hou, Q.; Wang, J.; Li, M. [Key Laboratory for Radiation Physics and Technology, Institute of Nuclear Science and Technology, Sichuan University, Chengdu 610065 (China)

    2017-02-15

    Highlights: • A cascde near a dislocation promotes climb motion. • Kinks induced by cascade facilitate the dipoles motion toward the cascade. • Shearing of dipole is dependent on PKA energy, position, direction, and dipole width. - Abstract: Tungsten (W) is the prime candidate material for the divertor and other plasma-facing components in DEMO. The point defects (i.e. vacancies and self-interstitials) produced in collision cascades caused by incident neutrons aggregate into dislocation loops (and voids), which strongly affect the mechanical properties. The point defects also interact with existing microstructural features, and understanding these processes is crucial for modelling the long term microstructural evolution of the material under fusion conditions. In this work, we performed molecular dynamics simulations of cascades interacting with initially straight edge dislocation dipoles. It was found that the residual vacancy number usually exceeds the residual interstitial number for cascades interacting with vacancy type dipoles, but for interstitial type dipoles these are close. We observed that a cascade near a dislocation promotes climb, i.e. it facilitates the movement of point defects along the climb direction. We also observed that the dislocations move easily along the glide direction, and that kinks are formed near the centre of the cascade, which then facilitate the movement of the dipoles. Some dipoles are sheared off by the cascade, and this is dependent on PKA energy, position, direction, and the width of dipole.

  14. Estimation of the tritium retention in ITER tungsten divertor target using macroscopic rate equations simulations

    Science.gov (United States)

    Hodille, E. A.; Bernard, E.; Markelj, S.; Mougenot, J.; Becquart, C. S.; Bisson, R.; Grisolia, C.

    2017-12-01

    Based on macroscopic rate equation simulations of tritium migration in an actively cooled tungsten (W) plasma facing component (PFC) using the code MHIMS (migration of hydrogen isotopes in metals), an estimation has been made of the tritium retention in ITER W divertor target during a non-uniform exponential distribution of particle fluxes. Two grades of materials are considered to be exposed to tritium ions: an undamaged W and a damaged W exposed to fast fusion neutrons. Due to strong temperature gradient in the PFC, Soret effect’s impacts on tritium retention is also evaluated for both cases. Thanks to the simulation, the evolutions of the tritium retention and the tritium migration depth are obtained as a function of the implanted flux and the number of cycles. From these evolutions, extrapolation laws are built to estimate the number of cycles needed for tritium to permeate from the implantation zone to the cooled surface and to quantify the corresponding retention of tritium throughout the W PFC.

  15. 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.

  16. 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.

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

    Science.gov (United States)

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

    2017-12-01

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

  18. Investigations on in situ diagnostics by an infrared camera to distinguish between the plasma facing tiles with carbonaceous surface layer and defect in the underneath junction

    International Nuclear Information System (INIS)

    Cai, Laizhong; Gauthier, Eric; Corre, Yann; Liu, Jian

    2013-01-01

    Both a deposition surface layer and a delamination underneath junction existing on plasma facing components (PFCs) can result in abnormal high surface temperature under normal heating conditions. The tile with delamination has to be replaced to prevent from a critical failure (complete delamination) during plasma operation while the carbon deposit can be removed without any repairing. Therefore, distinguishing in situ deposited tiles and junction defect tiles is crucial to avoid the critical failure without unwanted shutdown. In this paper, the thermal behaviors of junction defect tiles and carbon deposit tiles are simulated numerically. A modified time constant method is then introduced to analyze the thermal behaviors of deposited tiles and junction defect tiles. The feasibility of