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Sample records for divertor plasma facing

  1. 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.

  2. 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

  3. 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

  4. 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

  5. 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.

  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. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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

  13. 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

  14. 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.

  15. 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.

  16. 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.

  17. First wall and divertor plate disposed facing to plasma of thermonuclear device

    International Nuclear Information System (INIS)

    Araki, Masanori; Suzuki, Satoshi; Akiba, Masato; Hayata, Yoshiho; Inoue, Taiji; Hayashi, Yukihiro; Kude, Yukinori

    1998-01-01

    In order to make the most of characteristics of each ingredient of carbon fiber-reinforced composite materials, carbon fiber unidirectionally reinforced materials and a carbon fiber three-directionally reinforced material are laminated in the direction of the thickness to form a carbon fiber-reinforced carbon composite material. In this case, the carbon fibers are continuously oriented in the direction of the thickness to constitute the carbon fiber reinforced carbon composite materials integrally. In addition, a carbon fiber-reinforced carbon composite material prepared by bonding a metal on one surface in adjacent with the unidirectional carbon fiber reinforced portion and substantially in perpendicular to the direction of the thickness of the unidirectional carbon fiber reinforced portion is used as a main constitutional material. Further, a metal tube is buried in the carbon fiber three-directionally reinforced carbon composite material. Then, a first wall and a divertor plate excellent in thermal impact resistance to be disposed facing to plasmas of a thermonuclear device can be provided. (N.H.)

  18. 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.

  19. 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)

  20. 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

  1. 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.

  2. 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.

  3. 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.)

  4. 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.

  5. High temperature divertor plasma operation

    International Nuclear Information System (INIS)

    Ohyabu, Nobuyoshi.

    1991-02-01

    High temperature divertor plasma operation has been proposed, which is expected to enhance the core energy confinement and eliminates the heat removal problem. In this approach, the heat flux is guided through divertor channel to a remote area with a large target surface, resulting in low heat load on the target plate. This allows pumping of the particles escaping from the core and hence maintaining of the high divertor temperature, which is comparable to the core temperature. The energy confinement is then determined by the diffusion coefficient of the core plasma, which has been observed to be much lower than the thermal diffusivity. (author)

  6. 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.

  7. 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.

  8. 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 .

  9. Detached divertor plasmas in JET

    Energy Technology Data Exchange (ETDEWEB)

    Horton, L D; Borrass, K; Corrigan, G; Gottardi, N; Lingertat, J; Loarte, A; Simonini, R; Stamp, M F; Taroni, A [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking; Stangeby, P C [Toronto Univ., ON (Canada). Inst. for Aerospace Studies

    1994-07-01

    In simulations with high radiated power fractions, it is possible to produce the drop in ion current to the divertor targets typical of detached plasmas. Despite the fact that these experiments are performed on beryllium target tiles, radiation from deuterium and beryllium cannot account for the measured power losses. The neutral deuterium levels in the SOL in these plasmas are higher than the model predicts. This may be due to leakage from the divertor or to additional wall sources related to the non-steady nature of these plasmas. In contrast, a surprisingly high level of carbon is present in these discharges; higher even than would be predicted are the divertor target tiles pure carbon. This level may well be large enough to produce the measured radiation. (authors). 6 refs., 2 figs., 1 tab.

  10. NSTX Plasma Response to Lithium Coated Divertor

    Energy Technology Data Exchange (ETDEWEB)

    H.W. Kugel, M.G. Bell, J.P. Allain, R.E. Bell, S. Ding, S.P. Gerhardt, M.A. Jaworski, R. Kaita, J. Kallman, S.M. Kaye, B.P. LeBlanc, R. Maingi, R. Majeski, R. Maqueda, D.K. Mansfield, D. Mueller, R. Nygren, S.F. Paul, R. Raman, A.L. Roquemore, S.A. Sabbagh, H. Schneider, C.H. Skinner, V.A. Soukhanovskii, C.N. Taylor, J.R. Timberlak, W.R. Wampler, L.E. Zakharov, S.J. Zweben, and the NSTX Research Team

    2011-01-21

    NSTX experiments have explored lithium evaporated on a graphite divertor and other plasma facing components in both L- and H- mode confinement regimes heated by high-power neutral beams. Improvements in plasma performance have followed these lithium depositions, including a reduction and eventual elimination of the HeGDC time between discharges, reduced edge neutral density, reduced plasma density, particularly in the edge and the SOL, increased pedestal electron and ion temperature, improved energy confinement and the suppression of ELMs in the H-mode. However, with improvements in confinement and suppression of ELMs, there was a significant secular increase in the effective ion charge Zeff and the radiated power in H-mode plasmas as a result of increases in the carbon and medium-Z metallic impurities. Lithium itself remained at a very low level in the plasma core, <0.1%. Initial results are reported from operation with a Liquid Lithium Divertor (LLD) recently installed.

  11. NSTX plasma response to lithium coated divertor

    International Nuclear Information System (INIS)

    Kugel, H.W.; Bell, M.G.; Allain, J.P.; Bell, R.E.; Ding, S.; Gerhardt, S.P.; Jaworski, M.A.; Kaita, R.; Kallman, J.; Kaye, S.M.; LeBlanc, B.P.; Maingi, Rajesh; Majeski, R.; Maqueda, R.J.; Mansfield, D.K.; Mueller, D.; Nygren, R.E.; Paul, S.F.; Raman, R.; Roquemore, A.L.; Sabbagh, S.A.; Schneider, H.; Skinner, C.H.; Soukhanovskii, V.A.; Taylor, C.N.; Timberlake, J.; Wampler, W.R.; Zakharov, L.E.; Zweben, S.J.

    2011-01-01

    NSTX experiments have explored lithium evaporated on a graphite divertor and other plasma-facing components in both L- and H- mode confinement regimes heated by high-power neutral beams. Improvements in plasma performance have followed these lithium depositions, including a reduction and eventual elimination of the HeGDC time between discharges, reduced edge neutral density, reduced plasma density, particularly in the edge and the SOL, increased pedestal electron and ion temperature, improved energy confinement and the suppression of ELMs in the H-mode. However, with improvements in confinement and suppression of ELMs, there was a significant secular increase in the effective ion charge Z(eff) and the radiated power in H-mode plasmas as a result of increases in the carbon and medium-Z metallic impurities. Lithium itself remained at a very low level in the plasma core, < 0.1%. Initial results are reported from operation with a Liquid Lithium Divertor (LLD) recently installed.

  12. Snowflake divertor plasmas on TCV

    International Nuclear Information System (INIS)

    Piras, F; Coda, S; Furno, I; Moret, J-M; Sauter, O; Turri, G; Bencze, A; Duval, B P; Felici, F; Pochelon, A; Zucca, C; Pitts, R A; Tal, B

    2009-01-01

    Starting from a standard single null X-point configuration, a second order null divertor (snowflake (SF)) has been successfully created on the Tokamak a Configuration Variable (TCV) tokamak. The magnetic properties of this innovative configuration have been analysed and compared with a standard X-point configuration. For the SF divertor, the connection length and the flux expansion close to the separatrix exceed those of the standard X-point by more than a factor of 2. The magnetic shear in the plasma edge is also larger for the SF configuration.

  13. Controlling marginally detached divertor plasmas

    Science.gov (United States)

    Eldon, D.; Kolemen, E.; Barton, J. L.; Briesemeister, A. R.; Humphreys, D. A.; Leonard, A. W.; Maingi, R.; Makowski, M. A.; McLean, A. G.; Moser, A. L.; Stangeby, P. C.

    2017-06-01

    A new control system at DIII-D has stabilized the inter-ELM detached divertor plasma state for H-mode in close proximity to the threshold for reattachment, thus demonstrating the ability to maintain detachment with minimal gas puffing. When the same control system was instead ordered to hold the plasma at the threshold (here defined as T e  =  5 eV near the divertor target plate), the resulting T e profiles separated into two groups with one group consistent with marginal detachment, and the other with marginal attachment. The plasma dithers between the attached and detached states when the control system attempts to hold at the threshold. The control system is upgraded from the one described in Kolemen et al (2015 J. Nucl. Mater. 463 1186) and it handles ELMing plasmas by using real time D α measurements to remove during-ELM slices from real time T e measurements derived from divertor Thomson scattering. The difference between measured and requested inter-ELM T e is passed to a PID (proportional-integral-derivative) controller to determine gas puff commands. While some degree of detachment is essential for the health of ITER’s divertor, more deeply detached plasmas have greater radiative losses and, at the extreme, confinement degradation, making it desirable to limit detachment to the minimum level needed to protect the target plate (Kolemen et al 2015 J. Nucl. Mater. 463 1186). However, the observed bifurcation in plasma conditions at the outer strike point with the ion B   ×  \

  14. 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

  15. Divertor plasma physics experiments on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Mahdavi, M.A.; Allen, S.L.; Evans, T.E.

    1996-10-01

    In this paper we present an overview of the results and conclusions of our most recent divertor physics and development work. Using an array of new divertor diagnostics we have measured the plasma parameters over the entire divertor volume and gained new insights into several divertor physics issues. We present direct experimental evidence for momentum loss along the field lines, large heat convection, and copious volume recombination during detachment. These observations are supported by improved UEDGE modeling incorporating impurity radiation. We have demonstrated divertor exhaust enrichment of neon and argon by action of a forced scrape off layer (SOL) flow and demonstrated divertor pumping as a substitute for conventional wall conditioning. We have observed a divertor radiation zone with a parallel extent that is an order of magnitude larger than that estimated from a 1-D conduction limited model of plasma at coronal equilibrium. Using density profile control by divertor pumping and pellet injection we have attained H-mode confinement at densities above the Greenwald limit. Erosion rates of several candidate ITER plasma facing materials are measured and compared with predictions of a numerical model

  16. Multiple equilibria of divertor plasmas

    International Nuclear Information System (INIS)

    Vu, H.X.; Prinja, A.K.

    1993-01-01

    A one-dimensional, two-fluid transport model with a temperature-dependent neutral recycling coefficient is shown to give rise to multiple equilibria of divertor plasmas (bifurcation). Numerical techniques for obtaining these multiple equilibria and for examining their stability are presented. Although these numerical techniques have been well known to the scientific community, this is the first time they have been applied to divertor plasma modeling to show the existence of multiple equilibria as well as the stability of these solutions. Numerical and approximate analytical solutions of the present one-dimensional transport model both indicate that there exists three steady-state solutions corresponding to (1) a high-temperature, low-density equilibrium, (2) a low-temperature, high-density equilibrium, and (3) an intermediate-temperature equilibrium. While both the low-temperature and the high-temperature equilibria are stable, with respect to small perturbations in the plasma conditions, the intermediate-temperature equilibrium is physically unstable, i.e., any small perturbation about this equilibrium will cause a transition toward either the high-temperature or low-temperature equilibrium

  17. 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.

  18. The isotope effect on divertor conditions and neutral pumping in horizontal divertor configurations in JET-ILW Ohmic plasmas

    Directory of Open Access Journals (Sweden)

    J. Uljanovs

    2017-08-01

    Full Text Available Understanding the impact of isotope mass and divertor configuration on the divertor conditions and neutral pressures is critical for predicting the performance of the ITER divertor in DT operation. To address this need, ohmically heated hydrogen and deuterium plasma experiments were conducted in JET with the ITER-like wall in varying divertor configurations. In this study, these plasmas are simulated with EDGE2D-EIRENE outfitted with a sub-divertor model, to predict the neutral pressures in the plenum with similar fashion to the experiments. EDGE2D-EIRENE predictions show that the increased isotope mass results in up to a 25% increase in peak electron densities and 15% increase in peak ion saturation current at the outer target in deuterium when compared to hydrogen for all horizontal divertor configurations. Indicating that a change from hydrogen to deuterium as main fuel decreases the neutral mean free path, leading to higher neutral density in the divertor. Consequently, this mechanism also leads to higher neutral pressures in the sub-divertor. The experimental data provided by the hydrogen and deuterium ohmic discharges shows that closer proximity of the outer strike point to the pumping plenum results in a higher neutral pressure in the sub-divertor. The diaphragm capacitance gauge pressure measurements show that a two to three-fold increase in sub-divertor pressure was achieved in the corner and nearby horizontal configurations compared to the far-horizontal configurations, likely due to ballistic transport (with respect to the plasma facing components of the neutrals into the sub-divertor. The corner divertor configuration also indicates that a neutral expansion occurs during detachment, resulting in a sub-divertor neutral density plateau as a function of upstream density at the outer-mid plane.

  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. 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.

  1. 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.

  2. 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)

  3. Divertor plasma modification by divertor biasing and edge ergodization in JFT-2M

    International Nuclear Information System (INIS)

    Shoji, T.; Nagashima, K.; Tamai, H.; Ohdachi, S.; Miura, Y.; Ohasa, K.; Maeda, H.; Ohyabu, N.; Leonard, A.W.; Aikawa, H.; Fujita, T.; Hoshino, K.; Kawashima, H.; Matsuda, T.; Maeno, M.; Mori, M.; Ogawa, H.; Shimada, M.; Uehara, K.; Yamauchi, T.

    1995-01-01

    The effects of divertor biasing and edge ergodization on the divertor plasma have been investigated in the JFT-2M tokamak. Experimental results show; (1) The differential divertor biasing can change the in/out asymmetry of the divertor plasma. It especially changes the density on the ion side divertor plasma. The in/out electron pressure difference has a good correlation with the biasing current. (2) The unipolar divertor biasing can change the density profile of divertor plasma. The radial electric field and shear flow are the cause for this change. (3) The electron temperature of the divertor plasma in the H-mode with frequent ELMs induced by edge ergodization is lower than that of usual H-mode. That is due to the enhancement of the radial particle flux by frequent ELMs, ((orig.))

  4. Comparison between FEM and high heat flux thermal fatigue testing results of ITER divertor plasma facing mock-ups

    Energy Technology Data Exchange (ETDEWEB)

    Crescenzi, F., E-mail: fabio.crescenzi@enea.it; Roccella, S.; Visca, E.; Moriani, A.

    2014-10-15

    Highlights: • Divertor is an important part of the ITER machine. • Finite element analysis allows designers to explore multiple design options, reducing physical prototypes and optimizing design performance. • The hydraulic thermal-mechanical analysis performed by ANSYS and the test results on small-scale mock-ups manufactured by HRP were compared. • FEA results confirmed many experimental data, then it could be very useful for next design optimization. - Abstract: The divertor is one of the most challenging components of “DEMO” the next step ITER machine, so many tasks regarding modeling and experiments have been made in the past years to assess manufacturing processes, materials and thus the life-time of the components. In this context the finite element analysis (FEA) allows designers to explore multiple design options, to reduce physical prototypes and to optimize design performance. The comparison between the hydraulic thermal-mechanical analysis performed by ANSYS WORKBENCH 14.5 and the test results [1] on small-scale mock-ups manufactured with the Hot Radial Pressing (HRP) [2] technology is presented in this paper. During the thermal fatigue testing in the Efremov TSEFEY facility to assess the heat flux load-carrying capability of the mock-ups, only the surface temperature was measured, so the FEA was important because it allowed to know any other information (temperature inside the materials, local water temperature, local stress, etc.). FEA was performed coupling the thermal-hydraulic analysis, that calculated the temperature distributions on the components and the heat transfer coefficient (HTC) between water and heat sink tube, with the mechanical analysis. The comparison between analysis and testing results was based on the temperature maps of the loaded surface and on number of the cycles supported during the testing and those predicted by the mechanical analysis using the experimental fatigue curves for CuCrZr-IG, that is the structural

  5. 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.

  6. 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

  7. Model of divertor biasing and control of scrape-off layer and divertor plasmas

    International Nuclear Information System (INIS)

    Nagasaki, K.; Itoh, K.; Itoh, S.

    1991-02-01

    Analytic model of the divertor biasing is described. For the given plasma and energy sources from the core plasma, the heat and particle flux densities on the divertor plate as well as scrape-off-layer (SOL)/divertor plasmas are analyzed in a slab model. Using a two-dimensional model, the effects of the divertor biasing and SOL current are studied. The conditions to balance the plasma temperature or sheath potential on different divertor plates are obtained. Effect of the SOL current on the heat channel width is also discussed. (author)

  8. Atomic and molecular processes in JT-60U divertor plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Takenaga, H.; Shimizu, K.; Itami, K. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    1997-01-01

    Atomic and molecular data are indispensable for the understanding of the divertor characteristics, because behavior of particles in the divertor plasma is closely related to the atomic and molecular processes. In the divertor configuration, heat and particles escaping from the main plasma flow onto the divertor plate along the magnetic field lines. In the divertor region, helium ash must be effectively exhausted, and radiation must be enhanced for the reduction of the heat load onto the divertor plate. In order to exhaust helium ash effectively, the difference between behavior of neutral hydrogen (including deuterium and tritium) and helium in the divertor plasma should be understood. Radiation from the divertor plasma generally caused by the impurities which produced by the erosion of the divertor plate and/or injected by gas-puffing. Therefore, it is important to understand impurity behavior in the divertor plasma. The ions hitting the divertor plate recycle through the processes of neutralization, reflection, absorption and desorption at the divertor plates and molecular dissociation, charge-exchange reaction and ionization in the divertor plasma. Behavior of hydrogen, helium and impurities in the divertor plasmas can not be understood without the atomic and molecular data. In this report, recent results of the divertor study related to the atomic and molecular processes in JT-60U were summarized. Behavior of neural deuterium and helium was discussed in section 2. In section 3, the comparisons between the modelling of the carbon impurity transport and the measurements of C II and C IV were discussed. In section 4, characteristics of the radiative divertor using Ne puffing were reported. The new diagnostic method for the electron density and temperature in the divertor plasmas using the intensity ratios of He I lines was described in section 5. (author)

  9. Boundary plasma control with the ergodic divertor

    International Nuclear Information System (INIS)

    Ghendrih, Ph.; Becoulet, M.; Beyer, P.

    1999-01-01

    Ergodic divertor experiments on Tore Supra provide evidence of significant control of plasma-wall interaction. Theoretical investigation of the laminar region (i.e. governed by parallel transport) indicates that control of the plasma state at the target plate can be achieved with plasma states similar to that observed with the axisymmetric divertor. Analysis of the temperature field with a 2-D test particle code allows one to recover the observed spatial modulation and shows that an intrinsic barrier appears to develop at the separatrix. Energy deposition peaking, analysed with a 3-D code, is strongly reduced when moderate transverse transport is considered. Possible control of upstream parameters can thus be achieved in the ergodic region, for instance a lowering of the parallel energy flux by cross field transport. (author)

  10. Boundary plasma control with the ergodic divertor

    International Nuclear Information System (INIS)

    Ghendrih, Ph.; Becoulet, M.; Beyer, P.

    2001-01-01

    Ergodic divertor experiments on Tore Supra provide evidence of significant control of plasma-wall interaction. Theoretical investigation of the laminar region (i.e. governed by parallel transport) indicates that control of the plasma state at the target plate can be achieved with plasma states similar to that observed with the axisymmetric divertor. Analysis of the temperature field with a 2-D test particle code allows one to recover the observed spatial modulation and shows that an intrinsic barrier appears to develop at the separatrix. Energy deposition peaking, analysed with a 3-D code, is strongly reduced when moderate transverse transport is considered. Possible control of upstream parameters can thus be achieved in the ergodic region, for instance a lowering of the parallel energy flux by cross field transport. (author)

  11. NSTX plasma operation with a Liquid Lithium Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Kugel, H.W., E-mail: hkugel@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Allain, J.P. [Purdue University, West Lafayette, IN 47907 (United States); Bell, M.G.; Bell, R.E.; Diallo, A.; Ellis, R.; Gerhardt, S.P. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Heim, B. [Purdue University, West Lafayette, IN 47907 (United States); Jaworski, M.A.; Kaita, R.; Kallman, J.; Kaye, S.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Maingi, R.; McLean, A. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Menard, J.; Mueller, D. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Nygren, R. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Ono, M.; Paul, S.F. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); and others

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer NSTX 2010 experiments tested the effectiveness of maintaining the deuterium retention properties of a static liquid lithium molybdenum divertor surface when refreshed by lithium evaporation as an approximation to a flowing liquid lithium surface. Black-Right-Pointing-Pointer Noteworthy improvements in plasma performance with the plasma strike point on the liquid lithium molybdenum divertor were obtained similar to those obtained previously with lithiated graphite. The role of lithium impurities in this result is discussed. Black-Right-Pointing-Pointer Inspection of the liquid lithium molybdenum divertor after the Campaign indicated mechanical damage to supports, and other hardware resulting from forces following plasma current disruptions. - Abstract: NSTX 2010 experiments were conducted using a molybdenum Liquid Lithium Divertor (LLD) surface installed on the outer part of the lower divertor. This tested the effectiveness of maintaining the deuterium retention properties of a static liquid lithium surface when refreshed by lithium evaporation as an approximation to a flowing liquid lithium surface. The LLD molybdenum front face has a 45% porosity to provide sufficient wetting to spread 37 g of lithium, and to retain it in the presence of magnetic forces. Lithium Evaporators were used to deposit lithium on the LLD surface. At the beginning of discharges, the LLD lithium surface ranged from solid to liquefied depending on the amount of applied and plasma heating. Noteworthy improvements in plasma performance were obtained similar to those obtained previously with lithiated graphite, e.g., ELM-free, quiescent edge, H-modes. During these experiments with the plasma outer strike point on the LLD, the rate of deuterium retention in the LLD, as indicated by the fueling needed to achieve and maintain stable plasma conditions, was the about the same as that for solid lithium coatings on the graphite prior to the installation of the

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-15

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

  13. Role of molecular effects in divertor plasma recombination

    Directory of Open Access Journals (Sweden)

    A.S. Kukushkin

    2017-08-01

    Full Text Available Molecule-Activated Recombination (MAR effect is re-considered in view of divertor plasma conditions. A strong isotopic effect is demonstrated. In deuterium plasmas, the reaction chain through D2+ formation, usually considered dominant and included in 2D edge plasma models, is negligible. However, in this case the other branch, through D−, usually neglected in modelling, becomes relatively strong. The overall share of MAR in divertor plasma recycling stays within 20%. The operational parameters of the divertor plasmas, such as the peak power loading on the divertor targets or the pressure limit for partial detachment of the divertor plasma, are insensitive to the presence of MAR, although the latter may be important for correct interpretation of the divertor diagnostics.

  14. Divertor, thermonuclear device and method of neutralizing high temperature plasma

    International Nuclear Information System (INIS)

    Ikegami, Hideo.

    1995-01-01

    The thermonuclear device comprises a thermonuclear reactor for taking place fusion reactions to emit fusion plasmas, and a divertor made of a hydrogen occluding material, and the divertor is disposed at a position being in contact with the fusion plasmas after nuclear fusion reaction. The divertor is heated by fusion plasmas after nuclear fusion reaction, and hydrogen is released from the hydrogen occluding material as a constituent material. A gas blanket is formed by the released hydrogen to cool and neutralize the supplied high temperature nuclear fusion plasmas. This prevents the high temperature plasmas from hitting against the divertor, elimination of the divertor by melting and evaporation, and solve a problem of processing a divertor activated by neutrons. In addition, it is possible to utilize hydrogen isotopes of fuels effectively and remove unnecessary helium. Inflow of impurities from out of the system can also be prevented. (N.H.)

  15. Divertor plasma studies on DIII-D: Experiment and modeling

    International Nuclear Information System (INIS)

    West, W.P.; Brooks, N.H.; Allen, S.L.

    1996-09-01

    In a magnetically diverted tokamak, the scrape-off layer (SOL) and divertor plasma provides separation between the first wall and the core plasma, intercepting impurities generated at the wall before they reach the core plasma. The divertor plasma can also serve to spread the heat and particle flux over a large area of divertor structure wall using impurity radiation and neutral charge exchange, thus reducing peak heat and particle fluxes at the divertor strike plate. Such a reduction will be required in the next generation of tokamaks, for without it, the divertor engineering requirements are very demanding. To successfully demonstrate a radiative divertor, a highly radiative condition with significant volume recombination must be achieved in the divertor, while maintaining a low impurity content in the core plasma. Divertor plasma properties are determined by a complex interaction of classical parallel transport, anomalous perpendicular transport, impurity transport and radiation, and plasma wall interaction. In this paper the authors describe a set of experiments on DIII-D designed to provide detailed two dimensional documentation of the divertor and SOL plasma. Measurements have been made in operating modes where the plasma is attached to the divertor strike plate and in highly radiating cases where the plasma is detached from the divertor strike plate. They also discuss the results of experiments designed to influence the distribution of impurities in the plasma using enhanced SOL plasma flow. Extensive modeling efforts will be described which are successfully reproducing attached plasma conditions and are helping to elucidate the important plasma and atomic physics involved in the detachment process

  16. Study of the radiation in divertor plasmas; Etude du rayonnement dans les plasmas de divertor

    Energy Technology Data Exchange (ETDEWEB)

    Laugier, F

    2000-10-19

    We have studied the cooling of the edge plasma by radiation in the divertor volume, in order to optimize the extraction of power in tokamaks and to limit the wall erosion. In attached divertor plasmas experiments, the concentration of intrinsic impurities at the edge is related to the response of the wall to the incident energy flow of plasma, depending on a phenomenological law. We carried out an analysis of the radiation according to this law and to the control parameters of the discharges. The largest radiated fraction and best synergy are obtained when the concentration of intrinsic impurities strongly increases with the energy of incident plasma. On the other hand, the erosion of the wall is stronger. In detached plasmas, we proved that the performances in terms of incident plasma energy loss and pressure loss are optimal when the density of the slowest neutrals is strong at the edge and when their radial penetration is small. On Tore Supra, we highlighted the correlations between the maximum Mach number of incident plasma flow, the radiation front and the penetration of the neutrals. A simple diagnostic based on the localization of the maximum Mach number proves that detached mode is not optimal on Tore Supra, because the radial penetration of the slowest neutrals is not sufficiently small. In the last part, we obtained the three-dimensional topology of the radiation in the ergodic divertor using a spectral analysis code and boundary conditions consistent with the temperature distribution on the wall. The radiation is maximum in front of the divertor modules. As a consequence, radiated power is underestimated by standards measurements of Tore Supra that are located between the modules. We finally showed that the profiles of temperature along the field lines are modulated, this is specific to the ergodic divertor. (author)

  17. 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.)

  18. Plasma flow in the DIII-D divertor

    International Nuclear Information System (INIS)

    Boedo, J.A.; Porter, G.D.; Schaffer, M.J.

    1998-07-01

    Indications that flows in the divertor can exhibit complex behavior have been obtained from 2-D modeling but so far remain mostly unconfirmed by experiment. An important feature of flow physics is that of flow reversal. Flow reversal has been predicted analytically and it is expected when the ionization source arising from neutral or impurity ionization in the divertor region is large, creating a high pressure zone. Plasma flows arise to equilibrate the pressure. A radiative divertor regime has been proposed in order to reduce the heat and particle fluxes to the divertor target plates. In this regime, the energy and momentum of the plasma are dissipated into neutral gas introduced in the divertor region, cooling the plasma by collisional, radiative and other atomic processes so that the plasma becomes detached from the target plates. These regimes have been the subject of extensive studies in DIII-D to evaluate their energy and particle transport properties, but only recently it has been proposed that the energy transport over large regions of the divertor must be dominated by convection instead of conduction. It is therefore important to understand the role of the plasma conditions and geometry on determining the region of convection-dominated plasma in order to properly control the heat and particle fluxes to the target plates and hence, divertor performance. The authors have observed complex structures in the deuterium ion flows in the DIII-D divertor. Features observed include reverse flow, convective flow over a large volume of the divertor and stagnant flow. They have measured large gradients in the plasma potential across the separatrix in the divertor and determined that these gradients induce poloidal flows that can potentially affect the particle balance in the divertor

  19. The magnetic vapour shield effect at divertor plates during plasma disruptions

    International Nuclear Information System (INIS)

    Piazza, G.; Goel, B.; Hoebel, W.; Wuerz, H.; Landman, I.

    1995-01-01

    Hard disruptions in a TOKAMAK cause a large thermal load on the divertor plates with an instantaneous ablation of a part of the heated material. The produced vapour cloud screens the plasma facing component from the direct interaction with the disrupting plasma (vapour shield effect). In order to quantify the damage to the divertor the magneto-hydrodynamic behaviour of the expanding vapour cloud has been investigated using an extended version of the 1-dimensional Lagrangian hydrodynamic code KATACO. Modelling of the magnetic field effects on the expanding plasma takes into account that the magnetic field is oblique to the divertor (1 1/2 dimensional model). The ''Radiation Heat Conduction Approximation'' has been used for describing the radiative energy transport. In this paper results are presented assuming graphite as divertor material, irradiated with a proton beam of an energy density of 12MJ/m 2 and a duration of 100μs. (orig.)

  20. Divertor heat flux control and plasma-material interaction

    International Nuclear Information System (INIS)

    Kikuchi, Yusuke; Nagata, Masayoshi; Sawada, Keiji; Takamura, Shuichi; Ueda, Yoshio

    2014-01-01

    Development of reliable radiative-cooling divertors is essential in DEMO reactor because it uses low-activation materials with low heat removal and the plasma heat flux exhausted from the confined region is 5 times as large as in ITER. It is important to predict precisely the heat and particle flux toward the divertor plate by simulation. In this present article, theoretical and experimental data of the reflection, secondary emission and surface recombination coefficients of the divertor plate by ion bombardment are given and their effects on the power transmission coefficient are discussed. In addition, some topics such as the erosion process of the divertor plate by ELM and the plasma disruption, the thermal shielding due to the vapor layer on the divertor plate and the formation of fuzz structure on W by helium plasma irradiation, are described. (author)

  1. U.S. Assessment of advanced limiter-divertor plasma-facing systems (ALPS) design, analysis, and R and D needs

    International Nuclear Information System (INIS)

    Mattas, R. F.

    1998-01-01

    The purpose of the ALPS program is to identify and evaluate advanced limiter/diverter systems that will enhance the attractiveness of fusion power. The highest priority goals at present are achieving high power density, up to 50 MW/m 2 , and showing compatibility of plasma-facing surfaces with plasma operation. Personnel representing a wide range of disciplines from a number of institutions are engaged in the program, where an evaluation phase of the program is planned for three years. Successful identification of promising concepts in the evaluation phase should lead to an R and D phase that includes proof-of-principle experiments

  2. The edge plasma and divertor in TIBER

    Energy Technology Data Exchange (ETDEWEB)

    Barr, W.L.

    1987-10-16

    An open divertor configuration has been adopted for TIBER. Most recent designs, including DIII-D, NET and CIT use open configurations and rely on a dense edge plasma to shield the plasma from the gas produced at the neutralizer plate. Experiments on ASDEX, PDX, D-III, and recently on DIII-D have shown that a dense edge plasma can be produced by re-ionizing most of the gas produced at the plate. This high recycling mode allows a large flux of particles to carry the heat to the plate, so that the mean energy per particle can be low. Erosion of the plate can be greatly reduced if the average impact energy of the ions at the plate can be reduced to near or below the threshold for sputtering of the plate material. The present configuration allows part of the flux of edge plasma ions to be neutralized at the entrance to the pumping duct so that helium is pumped as well as hydrogen. 7 refs., 3 figs.

  3. The edge plasma and divertor in TIBER

    International Nuclear Information System (INIS)

    Barr, W.L.

    1987-01-01

    An open divertor configuration has been adopted for TIBER. Most recent designs, including DIII-D, NET and CIT use open configurations and rely on a dense edge plasma to shield the plasma from the gas produced at the neutralizer plate. Experiments on ASDEX, PDX, D-III, and recently on DIII-D have shown that a dense edge plasma can be produced by re-ionizing most of the gas produced at the plate. This high recycling mode allows a large flux of particles to carry the heat to the plate, so that the mean energy per particle can be low. Erosion of the plate can be greatly reduced if the average impact energy of the ions at the plate can be reduced to near or below the threshold for sputtering of the plate material. The present configuration allows part of the flux of edge plasma ions to be neutralized at the entrance to the pumping duct so that helium is pumped as well as hydrogen. 7 refs., 3 figs

  4. 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

  5. Plasma parameters in the COMPASS divertor during Ohmic plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Dimitrova, M. [Institute of Plasma Physics, Academy of Sciences of the Czech Republic v.v.i., Prague (Czech Republic); Emil Djakov Institute of Electronics, Bulgarian Academy of Sciences, Sofia (Bulgaria); Dejarnac, R.; Stoeckel, J.; Havlicek, J.; Janky, F.; Panek, R. [Institute of Plasma Physics, Academy of Sciences of the Czech Republic v.v.i., Prague (Czech Republic); Popov, Ts.K. [Faculty of Physics, St. Kl. Ohridski University of Sofia (Bulgaria); Ivanova, P.; Vasileva, E. [Emil Djakov Institute of Electronics, Bulgarian Academy of Sciences, Sofia (Bulgaria); Kovacic, J. [Jozef Stefan Institute, Ljubljana (Slovenia)

    2014-04-15

    This paper reports on probe measurements of the electron energy distribution function and plasma potential in the divertor region of the COMPASS tokamak during D-shaped plasmas. The probe data have been processed using the novel first-derivative technique. A comparison with the results obtained by processing the same data with the classical probe technique, which assumes Maxwellian electron energy distribution functions is presented and discussed. In the vicinity of the inner and outer strike points of the divertor the electron energy distribution function can be approximated by a bi-Maxwellian, with a dominating low-energy electron population (4-7 eV) and a minority of higher energy electrons (12-25 eV). In the private flux region between the two strike points the electron energy distribution function is found to be Maxwellian with temperatures in the range of 7-10 eV. The comparative analysis using both techniques has allowed a better insight into the underlying physical processes at the divertor region of the COMPASS tokamak. (copyright 2014 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  6. Experimental demonstration of vector E x vector B plasma divertor

    International Nuclear Information System (INIS)

    Strait, E.J.; Kerst, D.W.; Sprott, J.C.

    1977-01-01

    The vector E x vector B drift due to an applied radial electric field in a tokamak with poloidal divertor can speed the flow of plasma out of the scrape-off region, and provide a means of externally controlling the flow rate and thus the width of the density fall-off. An experiment in the Wisconsin levitated toroidal octupole, using vector E x vector B drifts alone, demonstrates divertor-like behavior, including 70% reduction of plasma density near the wall and 40% reduction of plasma flux to the wall, with no adverse effects on confinement of the main plasma

  7. Charge exchange in a divertor plasma with excited particles

    International Nuclear Information System (INIS)

    Krasheninnikov, S.I.; Lisitsa, V.S.; Pigarov, A.Y.

    1988-01-01

    A model is constructed for the dynamics of neutral atoms and multicharged ions in a tokamak plasma. The influence of cascade excitation on charge exchange and ionization is taken into account. The effective rates of the resonant charge exchange of a proton with a hydrogen atom, the nonresonant charge exchange of a helium atom with a proton, and that of an α particle with atomic hydrogen are calculated as functions of the parameters of the divertor plasma in a tokamak. The charge exchange H + +He→H+He + can represent a significant fraction (∼30%) of the total helium ionization rate. Incorporating the charge exchange of He 2+ with atomic hydrogen under the conditions prevailing in the divertor plasma of the INTOR reactor can lead to substantial He 2+ →He + conversion and thereby reduce the sputtering of the divertor plates by helium ions

  8. Turbulence studies in tokamak boundary plasmas with realistic divertor geometry

    International Nuclear Information System (INIS)

    Xu, X.Q.; Cohen, R.H.; Porter, G.D.; Rognlien, T.; Ryutov, D.D.; Myra, J.R.; D'Ippolito, D.A.; Moyer, R.; Groebner, R.J.

    2001-01-01

    Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT and the linearized shooting code BAL for studies of turbulence in tokamak boundary plasmas and its relationship to the L-H transition, in a realistic divertor plasma geometry. The key results include: (1) the identification of the dominant resistive X-point mode in divertor geometry and (2) turbulence suppression in the L-H transition by shear in the ExB drift speed, ion diamagnetism and nite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters. (author)

  9. Turbulence studies in tokamak boundary plasmas with realistic divertor geometry

    International Nuclear Information System (INIS)

    Xu, X.Q.; Cohen, R.H.; Por, G.D. ter; Rognlien, T.D.; Ryutov, D.D.; Myra, J.R.; D'Ippolito, D.A.; Moyer, R.; Groebner, R.J.

    1999-01-01

    Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT and the linearized shooting code BAL for studies of turbulence in tokamak boundary plasmas and its relationship to the L-H transition, in a realistic divertor plasma geometry. The key results include: (1) the identification of the dominant resistive X-point mode in divertor geometry and (2) turbulence suppression in the L-H transition by shear in the E x B drift speed, ion diamagnetism and finite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters. (author)

  10. Radiative divertor plasmas with convection in DIII-D

    International Nuclear Information System (INIS)

    Leornard, A.W.; Porter, G.D.; Wood, R.D.

    1998-01-01

    The radiation of divertor heat flux on DIII-D is shown to greatly exceed the limits imposed by assumptions of energy transport dominated by electron thermal conduction parallel to the magnetic field. Approximately 90% of the power flowing into the divertor is dissipated through low Z radiation and plasma recombination. The dissipation is made possible by an extended region of low electron temperature in the divertor. A one-dimensional analysis of the parallel heat flux finds that the electron temperature profile is incompatible with conduction dominated parallel transport. Plasma flow at up to the ion acoustic speed, produced by upstream ionization, can account for the parallel heat flux. Modeling with the two-dimensional fluid code UEDGE has reproduced many of the observed experimental features

  11. Electron and molecular ion collisions relevant to divertor plasma

    International Nuclear Information System (INIS)

    Takagi, H.

    2005-01-01

    We introduce the concept of the multi-channel quantum defect theory (MQDT) and show the outline of the MQDT newly extended to include the dissociative states. We investigate some molecular processes relevant to the divertor plasma by using the MQDT: the dissociative recombination, dissociative excitation, and rotation-vibrational transition in the hydrogen molecular ion and electron collisions. (author)

  12. Heat and particle transport of sol/divertor plasma in the W-shaped divertor on JT-60U

    International Nuclear Information System (INIS)

    Asakura, N.; Sakurai, S.; Hosogane, N.

    1999-01-01

    The plasma profile and parallel flow in the scrape-off layer (SOL) were systematically measured using Mach probes installed at the midplane and the divertor x-point. Quantitative evaluation of a parallel flow: naturally produced in a torus to keep the pressure constant along the field line, was consistent with the measurement. Geometry effects of the W-shaped divertor on the divertor plasma and particle recycling at the newly installed baffle plates were evaluated quantitatively using the edge plasma data. (author)

  13. An analytic model for flow reversal in divertor plasmas

    International Nuclear Information System (INIS)

    Cooke, P.I.H.; Prinja, A.K.

    1987-04-01

    An analytic model is developed and used to study the phenomenon of flow reversal which is observed in two-dimensional simulations of divertor plasmas. The effect is shown to be caused by the radial spread of neutral particles emitted from the divertor target which can lead to a strong peaking of the ionization source at certain radial locations. The results indicate that flow reversal over a portion of the width of the scrape-off layer is inevitable in high recycling conditions. Implications for impurity transport and particle removal in reactors are discussed

  14. A survey of problems in divertor and edge plasma theory

    International Nuclear Information System (INIS)

    Boozer, A.; Braams, B.; Weitzner, H.; Hazeltine, R.; Houlberg, W.; Oktay, E.; Sadowski, W.; Wootton, A.

    1992-01-01

    Theoretical physics problems related to divertor design are presented, organized by the region in which they occur. Some of the open questions in edge physics are presented from a theoretician's point of view. After a cursory sketch of the fluid models of the edge plasma and their numerical realization, the following topics are taken up: time-dependent problems, non-axisymmetric effects, anomalous transport in the scrape-off layer, edge kinetic theory, sheath effects and boundary conditions in divertors, electric field effects, atomic and molecular data issues, impurity transport in the divertor region, poloidally localized power dissipation (MARFEs and dense gas targets), helium ash removal, and neutral transport. The report ends with a summary of selected problems of particular significance and a brief bibliography of survey articles and related conference proceedings

  15. Parametric analysis of the thermal effects on the divertor in tokamaks during plasma disruptions

    International Nuclear Information System (INIS)

    Bruhn, M.L.

    1988-04-01

    Plasma disruptions are an ever present danger to the plasma-facing components in today's tokamak fusion reactors. This threat results from our lack of understanding and limited ability to control this complex phenomenon. In particular, severe energy deposition occurs on the divertor component of the double-null configured tokamak reactor during such disruptions. A hybrid computational model developed to estimate and graphically illustrate global thermal effects of disruptions on the divertor plates is described in detail. The quasi-two-dimensional computer code, TADDPAK (Thermal Analysis Divertor during Disruptions PAcKage), is used to conduct parametric analysis for the TIBER II Tokamak Engineering Test Reactor Design. The dependence of these thermal effects on divertor material choice, disruption pulse length, disruption pulse shape, and the characteristic thickness of the plasma scrape-off layer is investigated for this reactor design. Results and conclusions from this analysis are presented. Improvements to this model and issues that require further investigation are discussed. Cursory analysis for ITER (International Thermonuclear Experimental Reactor) is also presented in the appendix. 75 refs., 49 figs., 10 tabs

  16. Plasma shape control calculations for BPX divertor design

    International Nuclear Information System (INIS)

    Strickler, D.J.; Neilson, G.H.; Jardin, S.C.; Pomphrey, N.

    1991-01-01

    The Burning Plasma Experiment (BPX) divertor is to be capable of withstanding heat loads corresponding to ignited operation and 500 MW of fusion power for a current rise time and flattop lasting several seconds. The poloidal field (PF), diagnostic, and feedback equilibrium control systems must provide precise X-point position control in order to sweep the separatrices across the divertor target surface and optimally distribute the heat loads. A control matrix MHD equilibrium code, BEQ, and the Tokamak Simulation Code (TSC) are used to compute preprogrammed double-null (DN) divertor sweep trajectories that maximize sweep distance while simultaneously satisfying a set of strict constraints: minimum lengths of the field lines between the X-point and strike points, minimum spacing between the inboard plasma edge and the limiter, maximum spacing between the outboard plasma edge and the ICRF antennas, minimum safety factor, and linked poloidal flux. A sequence of DN diverted equilibria and a consistent TSC fiducial discharge simulation are used in evaluating the performance of the BPX divertor shape and possible modifications. 5 refs., 10 figs

  17. 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.)

  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. Plasma/neutral gas transport in divertors and limiters

    International Nuclear Information System (INIS)

    Gierszewski, P.J.

    1983-09-01

    The engineering design of the divertor and first wall region of fusion reactors requires accurate knowledge of the energies and particle fluxes striking these surfaces. Simple calculations indicate that approx. 10 MW/m 2 heat fluxes and approx. 1 cm/yr erosion rates are possible, but there remain fundamental physics questions that bear directly on the engineering design. The purpose of this study was to treat hydrogen plasma and neutral gas transport in divertors and pumped limiters in sufficient detail to answer some of the questions as to the actual conditions that will be expected in fusion reactors. This was accomplished in four parts: (1) a review of relevant atomic processes to establish the dominant interactions and their data base; (2) a steady-state coupled O-D model of the plasma core, scrape-off layer and divertor exhaust to determine gross modes of operation and edge conditions; (3) a 1-D kinetic transport model to investigate the case of collisionless divertor exhaust, including non-Maxwellian ions and neutral atoms, highly collisional electrons, and a self-consistent electric field; and (4) a 3-D Monte Carlo treatment of neutral transport to correctly account for geometric effects

  20. OEDGE modeling of plasma contamination efficiency of Ar puffing from different divertor locations in EAST

    Science.gov (United States)

    Pengfei, ZHANG; Ling, ZHANG; Zhenwei, WU; Zong, XU; Wei, GAO; Liang, WANG; Qingquan, YANG; Jichan, XU; Jianbin, LIU; Hao, QU; Yong, LIU; Juan, HUANG; Chengrui, WU; Yumei, HOU; Zhao, JIN; J, D. ELDER; Houyang, GUO

    2018-04-01

    Modeling with OEDGE was carried out to assess the initial and long-term plasma contamination efficiency of Ar puffing from different divertor locations, i.e. the inner divertor, the outer divertor and the dome, in the EAST superconducting tokamak for typical ohmic plasma conditions. It was found that the initial Ar contamination efficiency is dependent on the local plasma conditions at the different gas puff locations. However, it quickly approaches a similar steady state value for Ar recycling efficiency >0.9. OEDGE modeling shows that the final equilibrium Ar contamination efficiency is significantly lower for the more closed lower divertor than that for the upper divertor.

  1. Comparison of Ne and Ar seeded radiative divertor plasmas in JT-60U

    Energy Technology Data Exchange (ETDEWEB)

    Nakano, T., E-mail: nakano.tomohide@jaea.go.jp

    2015-08-15

    In H-mode plasmas with Ne, Ar and a mixture of Ne and Ar injection, the divertor radiation power fractions amongst these impurities in addition to an intrinsic impurity, C, are investigated. In plasmas with the inner divertor plasma attached, carbon is the biggest radiator, whichever impurity, Ne, Ar or a mixture of Ar and Ne is injected. In contrast, in plasmas with the inner divertor plasma detached, Ne is the biggest radiator due to a significantly high recombination radiation from Ne VIII. Ar is always a minor contributor in plasmas with the inner divertor both attached and detached.

  2. Comparison of 2D simulations of detached divertor plasmas with divertor Thomson measurements in the DIII-D tokamak

    Directory of Open Access Journals (Sweden)

    T.D. Rognlien

    2017-08-01

    Full Text Available A modeling study is reported using new 2D data from DIII-D tokamak divertor plasmas and improved 2D transport model that includes large cross-field drifts for the numerically difficult low anomalous transport regime associated with the H-mode. The data set, which spans a range of plasma densities for both forward and reverse toroidal magnetic field (Bt, is provided by divertor Thomson scattering (DTS. Measurements utilizing X-point sweeping give corresponding 2D profiles of electron temperature (Te and density (ne across both divertor legs for individual discharges. The simulations focus on the open magnetic field-line regions, though they also include a small region of closed field lines. The calculations show the same features of in/out divertor plasma asymmetries as measured in the experiment, with the normal Bt direction (ion ∇B drift toward the X-point having higher ne and lower Te in the inner divertor leg than outer. Corresponding emission data for total radiated power shows a strong inner-divertor/outer-divertor asymmetry that is reproduced by the simulations. These 2D UEDGE transport simulations are enabled for steep-gradient H-mode conditions by newly implemented algorithms to control isolated grid-scale irregularities.

  3. Plasma-neutral gas interaction in a tokamak divertor: effects of hydrogen molecules and plasma recombination

    International Nuclear Information System (INIS)

    Krasheninnikov, S.I.; Pigarov, A.Yu.; Soboleva, T.K.; Sigmar, D.J.

    1997-01-01

    We investigate the influence of hydrogen molecules on plasma recombination using a collisional-radiative model for multispecies hydrogen plasmas and tokamak detached divertor parameters. The rate constant found for molecular activated recombination of a plasma can be as high as 2 x 10 -10 cm 3 /s, confirming our pervious estimates. We investigate the effects of hydrogen molecules and plasma recombination on self-consistent plasma-neutral gas interactions in the recycling region of a tokamak divertor. We treat the plasma flow in a fluid approximation retaining the effects of plasma recombination and employing a Knudsen neutral transport model for a 'gas box' divertor geometry. For the model of plasma-neutral interactions we employ we find: (a) molecular activated recombination is a dominant channel of divertor plasma recombination; and (b) plasma recombination is a key element leading to a decrease in the plasma flux onto the target and substantial plasma pressure drop which are the main features of detached divertor regimes. (orig.)

  4. Edge plasma control: Particle channeling in Tore Supra pump limiter and ergodic divertor

    International Nuclear Information System (INIS)

    Ghendrih, P.; Samain, A.; Grosman, A.; Capes, H.; Morera, J.P.

    1989-01-01

    Improved pumping efficiency can be achieved on Tore Supra by channeling process for particles, i.e. channeling of neutrals in the throat of pump limiters and channeling of plasma towards neutralizer plates in the ergodic divertor. The plugging length for the pump limiter throat is computed and numerical evidence of plasma flux channeling between the conductor bars of the ergodic divertor is presented. The effect of the Tore Supra ergodic divertor on edge plasma state and edge plasma transport is discussed. (orig.)

  5. Diagnosing transient plasma status: from solar atmosphere to tokamak divertor

    International Nuclear Information System (INIS)

    Giunta, A.S.; Henderson, S.; O'Mullane, M.; Summers, H.P.; Harrison, J.; Doyle, J.G.

    2016-01-01

    This work strongly exploits the interdisciplinary links between astrophysical (such as the solar upper atmosphere) and laboratory plasmas (such as tokamak devices) by sharing the development of a common modelling for time-dependent ionisation. This is applied to the interpretation of solar flare data observed by the UVSP (Ultraviolet Spectrometer and Polarimeter), on-board the Solar Maximum Mission and the IRIS (Interface Region Imaging Spectrograph), and also to data from B2-SOLPS (Scrape Off Layer Plasma Simulations) for MAST (Mega Ampère Spherical Tokamak) Super-X divertor upgrade. The derived atomic data, calculated in the framework of the ADAS (Atomic Data and Analysis Structure) project, allow equivalent prediction in non-stationary transport regimes and transients of both the solar atmosphere and tokamak divertors, except that the tokamak evolution is about one thousand times faster.

  6. Hybrid formulation of radiation transport in optically thick divertor plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Rosato, J.; Marandet, Y.; Bufferand, H.; Stamm, R. [PIIM, UMR 7345 Aix-Marseille Universite / CNRS, Centre de St-Jerome, Marseille (France); Reiter, D. [IEK-4 Plasmaphysik, Forschungszentrum Juelich GmbH, Juelich (Germany)

    2016-08-15

    Kinetic Monte Carlo simulations of coupled atom-radiation transport in optically thick divertor plasmas can be computationally very demanding, in particular in ITER relevant conditions or even larger devices, e.g. for power plant divertor studies. At high (∝ 10{sup 15} cm{sup -3}) atomic densities, it can be shown that sufficiently large divertors behave in certain areas like a black body near the first resonance line of hydrogen (Lyman α). This suggests that, at least in part, the use of continuum model (radiation hydrodynamics) can be sufficiently accurate, while being less time consuming. In this work, we report on the development of a hybrid model devoted to switch automatically between a kinetic and a continuum description according to the plasma conditions. Calculations of the photo-excitation rate in a homogeneous slab are performed as an illustration. The outlined hybrid concept might be also applicable to neutral atom transport, due to mathematical analogy of transport equations for neutrals and radiation. (copyright 2016 The Authors. Contributions to Plasma Physics published by Wiley-VCH Verlag GmbH and Co. KGaA Weinheim. This)

  7. Radiative divertor plasmas with convection in DIII-D

    International Nuclear Information System (INIS)

    Leonard, A.W.; Porter, G.D.; Wood, R.D.; Allen, S.L.; Boedo, J.; Brooks, N.H.; Evans, T.E.; Fenstermacher, M.E.; Hill, D.N.; Isler, R.C.; Lasnier, C.J.; Lehmer, R.D.; Mahdavi, M.A.; Maingi, R.; Moyer, R.A.; Petrie, T.W.; Schaffer, M.J.; Wade, M.R.; Watkins, J.G.; West, W.P.; Whyte, D.G.

    1998-01-01

    The radiation of divertor heat flux on DIII-D [J. Luxon et al., in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion (International Atomic Energy Agency, Vienna, 1987), p. 159] is shown to greatly exceed the limits imposed by assumptions of energy transport dominated by electron thermal conduction parallel to the magnetic field. Approximately 90% of the power flowing into the divertor is dissipated through low-Z radiation and plasma recombination. The dissipation is made possible by an extended region of low electron temperature in the divertor. A one-dimensional analysis of the parallel heat flux finds that the electron temperature profile is incompatible with conduction-dominated parallel transport. Plasma flow at up to the ion acoustic speed, produced by upstream ionization, can account for the parallel heat flux. Modeling with the two-dimensional fluid code UEDGE [T. Rognlien, J. L. Milovich, M. E. Rensink, and G. D. Porter, J. Nucl. Mater. 196 endash 198, 347 (1992)] has reproduced many of the observed experimental features. copyright 1998 American Institute of Physics

  8. Innovative Divertor Development to Solve the Plasma Heat-Flux Problem

    International Nuclear Information System (INIS)

    Rognlien, T.; Ryutov, D.; Makowski, M.; Soukhanovskii, V.; Umansky, M.; Cohen, R.; Hill, D.; Joseph, I.

    2009-01-01

    Large, localized plasma heat exhaust continues to be one of the critical problems for the development of tokamak fusion reactors. Excessive heat flux erodes and possibly melts plasma-facing materials, thereby dramatically shortening their lifetime and increasing the impurity contamination of the core plasma. A detailed assessment by the ITER team for their divertor has revealed substantial limitations on the operational space imposed by the divertor performance. For a fusion reactor, the problem becomes worse in that the divertor must accommodate 20% of the total fusion power (less any broadly radiated loss), while not allowing excess buildup of tritium in the walls nor excessive impurity production. This is an extremely challenging set of problems that must be solved for fusion to succeed as a power source; it deserves a substantial research investment. Material heat-flux constraints: Results from present-day tokamaks show that there are two major limitations of peak plasma heat exhaust. The first is the continuous flow of power to the divertor plates and nearby surfaces that, for present technology, is limited to 10-20 MW/m 2 . The second is the transient peak heat-flux that can be tolerated in a short time, τ m , before substantial ablation and melting of the surface occurs; such common large transient events are Edge Localized Mode (ELMs) and disruptions. The material limits imposed by these events give a peak energy/τ m 1/2 parameter of ∼ 40 MJ/m 2 s 1/2 (1). Both the continuous and transient limits can be approached by input powers in the largest present-day devices, and future devices are expected to substantially exceed the limits unless a solution can be found. Since the early 90's LLNL has developed the analytic and computational foundation for analyzing divertor plasmas, and also suggested and studied a number of solid and liquid material concepts for improving divertor/wall performance, with the most recent being the Snowflake divertor concept (2

  9. 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)

  10. Modelling of neutral particle transport in divertor plasma

    International Nuclear Information System (INIS)

    Kakizuka, Tomonori; Shimizu, Katsuhiro

    1995-01-01

    An outline of the modelling of neutral particle transport in the diverter plasma was described in the paper. The characteristic properties of divertor plasma were largely affected by interaction between neutral particles and divertor plasma. Accordingly, the behavior of neutral particle should be investigated quantitatively. Moreover, plasma and neutral gas should be traced consistently in the plasma simulation. There are Monte Carlo modelling and the neutral gas fluid modelling as the transport modelling. The former need long calculation time, but it is able to make the physical process modelling. A ultra-large parallel computer is good for the former. In spite of proposing some kinds of models, the latter has not been established. At the view point of reducing calculation time, a work station is good for the simulation of the latter, although some physical problems have not been solved. On the Monte Carlo method particle modelling, reducing the calculation time and introducing the interaction of particles are important subjects to develop 'the evolutional Monte Carlo Method'. To reduce the calculation time, two new methods: 'Implicit Monte Carlo method' and 'Free-and Diffusive-Motion Hybrid Monte-Carlo method' have been developing. (S.Y.)

  11. Emission spectroscopy of hydrogen molecules in technical and divertor plasmas

    International Nuclear Information System (INIS)

    Fantz, U.

    2002-01-01

    The paper gives an overview of the diagnostics of hydrogen molecules in technical plasmas (MW and RF discharges) and in divertor plasmas of fusion experiments (ASDEX Upgrade / Tokamak at the Max-Planck-Institut fuer Plasmaphysik in Garching near Munich, Germany). The Fulcher transition in the visible spectral range was chosen for analysis since this is the most prominent band in the spectrum of molecular hydrogen. Examples for diagnostics of molecular densities will be given, and the problems arising in the interpretation of spectra will be discussed. In divertor plasmas the diagnostics of molecular.uxes will be introduced and the contribution of molecules to the plasma recombination will be discussed. Results for vibrational populations in the ground state and the correlation to the upper Fulcher state will be given, providing an electron temperature diagnostic. Finally, the in.uence of surfaces (high-grade steel and graphite) on vibrational populations and on re.ection coe.cients of atoms will be shown. Special attention is given on a comparison of the isotopes hydrogen and deuterium. (Abstract Copyright [2002], Wiley Periodicals, Inc.)

  12. Self-sustained oscillations in the divertor plasma

    International Nuclear Information System (INIS)

    Krasheninnikov, S.I.; Kukushkin, A.S.; Pistunovich, V.I.; Pozharov, V.A.

    1985-01-01

    A simple analytical model of the edge plasma with high recycling, which relays on the presence of a small parameter - the ratio of the particle flows crossing the magnetic field to those impinging onto the divertor target, is proposed. A concept of the one-dimensional steady state (OSS) is introduced as the zero approximation in the small parameter. The mean number density N-tilde of the particles - ions plus neutrals - in the magnetic flux tube is choosen as the most representative and convenient parameter of the problem. The OSS are shown to be ambiguous in some N-tilde range for sufficiently high values of the energy flow entering the scrape-off layer from the bulk plasma. An equation, that describes a quasi-steady variations in OSS, is derived and a mechanism of exciting the self-sustained oscillations is developed. Results of simulation of the edge plasma oscillations are found to be in a good agreement with this mechanism, which could be responsible for the H-mode oscillations observed in the divertor experiments

  13. Analysis of noble gas recycling at a fusion plasma divertor

    International Nuclear Information System (INIS)

    Brooks, J.N.

    1996-01-01

    Near-surface recycling of neon and argon atoms and ions at a divertor has been studied using impurity transport and surface interaction codes. A fixed background deuterium endash tritium plasma model is used corresponding to the International Thermonuclear Experimental Reactor (ITER) [ITER EDA Agreement and Protocol 2, ITER EDA Documentation Series No. 5 (International Atomic Energy Agency, Vienna, 1994)] radiative plasma conditions (T e ≤10 eV). The noble gas transport depends critically on the divertor surface material. For low-Z materials (Be and C) both neon and argon recycle many (e.g., ∼100) times before leaving the near-surface region. This is also true for an argon on tungsten combination. For neon on tungsten, however, there is low recycling. These variations are due to differences in particle and energy reflection coefficients, mass, and ionization rates. In some cases a high flux of recycling atoms is ionized within the magnetic sheath and this can change local sheath parameters. Due to inhibited backflow, high recycling, and possibly high sputtering, noble gas seeding (for purposes of enhancing radiation) may be incompatible with Be or C surfaces, for fusion reactor conditions. On the other hand, neon use appears compatible with tungsten. copyright 1996 American Institute of Physics

  14. Magnetic Fluctuations during plasma current rise of divertor discharge in JT-60

    International Nuclear Information System (INIS)

    Ushigusa, Kenkichi; Kikuchi, Mitsuru; Hosogane, Nobuyuki; Tsuji, Syunji; Hayashi, Kazuo.

    1986-03-01

    During a current rise phase in the JT-60 divertor discharge, a series of magnetic fluctuations which do not rotate poloidally (phase-locking) is observed. They cause a cooling of plasma periphery and an enhancement of H α emission in the divertor chamber. A significant increase in β P + 1 i /2 with minor disruptions during the phase-locked magnetic fluctuation suggests a relaxation of the current profile in the current rise phase of the divertor discharge. (author)

  15. Facilities for technology testing of ITER divertor concepts, models, and prototypes in a plasma environment

    International Nuclear Information System (INIS)

    Cohen, S.A.

    1991-12-01

    The exhaust of power and fusion-reaction products from ITER plasma are critical physics and technology issues from performance, safety, and reliability perspectives. Because of inadequate pulse length, fluence, flux, scrape-off layer plasma temperature and density, and other parameters, the present generation of tokamaks, linear plasma devices, or energetic beam facilities are unable to perform adequate technology testing of divertor components, though they are essential contributors to many physics issues such as edge-plasma transport and disruption effects and control. This Technical Requirements Documents presents a description of the capabilities and parameters divertor test facilities should have to perform accelerated life testing on predominantly technological divertor issues such as basic divertor concepts, heat load limits, thermal fatigue, tritium inventory and erosion/redeposition. The cost effectiveness of such divertor technology testing is also discussed

  16. Towards fully authentic modelling of ITER divertor plasmas

    International Nuclear Information System (INIS)

    Maddison, G.P.; Hotston, E.S.; Reiter, D.; Boerner, P.

    1991-01-01

    Ignited next step tokamaks such as NET or ITER are expected to use a poloidal magnetic divertor to facilitate exhaust of plasma particles and energy. We report a development coupling together detailed computational models for both plasma and recycled neutral particle transport processes, to produce highly detailed and consistent design solutions. A particular aspect is involvement of an accurate specification of edge magnetic geometries, determined by an original equilibrium discretisation code, named LINDA. Initial results for a prototypical 22MA ITER double-null configuration are presented. Uncertainties in such modelling are considered, especially with regard to intrinsic physical scale lengths. Similar results produced with a simple, analytical treatment of recycling are also compared. Finally, a further extension allowing true oblique target sections is anticipated. (author) 8 refs., 5 figs

  17. 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

  18. Interferometric density measurements in the divertor and edge plasma regions for the additionally heated JT-60 plasmas

    International Nuclear Information System (INIS)

    Fukuda, T.; Yoshida, H.; Nagashima, A.; Ishida, S.; Kikuchi, M.; Yokomizo, H.

    1989-01-01

    The first divertor plasma density measurement and the interferometric edge plasma density measurement with boundary condition preserving millimeter waveguides were demonstrated to elucidate the mutual correlation among the divertor plasma, scrape-off layer plasma and the bulk plasma properties in the additionally heated JT-60 plasmas. The electron density in the divertor region exhibited a nonlinear dependence on the bulk plasma density for the joule-heated plasmas. When neutral beam heating is applied on the plasmas with the electron density above 2x10 19 /m 3 , however, the bulk plasma density is scraped off from the outer region to lead to density clamping, and the electron density in the divertor region rapidly increases over 1x10 20 /m 3 , from which we can deduce that the particle flow along the magnetic field is dominant, resulting in the apparent degradation of the particle confinement time. As for the case when neutral beam injection is applied to low-density plasmas, the bulk plasma electron density profile becomes flattened to yield a smaller density increase in the divertor region and no density clamping of the bulk plasma was observed. Simulation analysis which correlates the transport of the divertor plasma and the scrape-off layer plasma was also carried out to find the consistency with the experimental results. (orig.)

  19. Divertor power load studies for attached L-mode single-null plasmas in TCV

    NARCIS (Netherlands)

    Maurizio, R.; Elmore, S.; Fedorczak, N.; Gallo, A.; Reimerdes, H.; Labit, B.; Theiler, C.; Tsui, C. K.; Vijvers, W. A. J.; TCV team,; MST1 Team,

    2018-01-01

    This paper investigates the power loads at the inner and outer divertor targets of attached, Ohmic L-mode, deuterium plasmas in the TCV tokamak, in various experimental situations using an Infrared thermography system. The study comprises variations of the outer divertor leg length and target flux

  20. The effect of charge exchange with neutral deuterium on carbon emission in JET divertor plasmas

    International Nuclear Information System (INIS)

    Maggi, C.; Horton, L.; Summers, H.

    1999-11-01

    High density, low temperature divertor plasma operation in tokamaks results in large neutral deuterium concentrations in the divertor volume. In these conditions, low energy charge transfer reactions between neutral deuterium and the impurity ions can in principle enhance the impurity radiative losses and thus help to reduce the maximum heat load to the divertor target. A quantitative study of the effect of charge exchange on carbon emission is presented, applied to the JET divertor. Total and state selective effective charge exchange recombination rate coefficients were calculated in the collisional radiative picture. These coefficients were coupled to divertor and impurity transport models to study the effect of charge exchange on the measured carbon spectral emission in JET divertor discharges. The sensitivity of the effect of charge exchange to the assumptions in the impurity transport model was also investigated. A reassessment was made of fundamental charge exchange cross section data in support of this study. (author)

  1. Effect of low density H-mode operation on edge and divertor plasma parameters

    International Nuclear Information System (INIS)

    Maingi, R.; Mioduszewski, P.K.; Cuthbertson, J.W.

    1994-07-01

    We present a study of the impact of H-mode operation at low density on divertor plasma parameters on the DIII-D tokamak. The line-average density in H-mode was scanned by variation of the particle exhaust rate, using the recently installed divertor cryo-condensation pump. The maximum decrease (50%) in line-average electron density was accompanied by a factor of 2 increase in the edge electron temperature, and 10% and 20% reductions in the measured core and divertor radiated power, respectively. The measured total power to the inboard divertor target increased by a factor of 3, with the major contribution coming from a factor of 5 increase in the peak heat flux very close to the inner strike point. The measured increase in power at the inboard divertor target was approximately equal to the measured decrease in core and divertor radiation

  2. Comment on “Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake” [Phys. Plasmas 20, 102507 (2013)

    International Nuclear Information System (INIS)

    Ryutov, D. D.; Cohen, R. H.; Rognlien, T. D.; Soukhanovskii, V. A.; Umansky, M. V.

    2014-01-01

    In the recently published paper “Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake” [Phys. Plasmas 20, 102507 (2013)], the authors raise interesting and important issues concerning divertor physics and design. However, the paper contains significant errors: (a) The conceptual framework used in it for the evaluation of divertor “quality” is reduced to the assessment of the magnetic field structure in the outer Scrape-Off Layer. This framework is incorrect because processes affecting the pedestal, the private flux region and all of the divertor legs (four, in the case of a snowflake) are an inseparable part of divertor operation. (b) The concept of the divertor index focuses on only one feature of the magnetic field structure and can be quite misleading when applied to divertor design. (c) The suggestion to rename the divertor configurations experimentally realized on NSTX (National Spherical Torus Experiment) and DIII-D (Doublet III-D) from snowflakes to X-divertors is not justified: it is not based on comparison of these configurations with the prototypical X-divertor, and it ignores the fact that the NSTX and DIII-D poloidal magnetic field geometries fit very well into the snowflake “two-null” prescription

  3. Fabrication of divertor cassette for ITER

    International Nuclear Information System (INIS)

    Sanguinetti, G.P.

    2008-01-01

    The Divertor is the component located on the bottom of the ITER vacuum vessel, whose main function is to adsorb the high thermal flux generated by the plasma whilst keeping the plasma impurity at a reasonable low level. The divertor consist of 54 units, each comprising outer components, facing the plasma and a component supporting the plasma facing components (PFC) and providing coolant distribution to them (divertor cassette). The divertor cassette is a box structure, butt welded and machined, made from plates and forgins of austenitic stainless steels. The cassette fabrication, which is in detail described, includes manufacturing of the attachments of the PFC to the cassette, the coolant distribution channels, and the cassette to vacuum vessel locking system. The divertor cassette is a pressure component (the cooling water runs at 40 bar) and therefore divertor cassette design, fabrication and service shall comply with the European PED and the applicable French law for the ITER. (orig.)

  4. Erosion products of ITER divertor materials under plasma disruption simulation

    Energy Technology Data Exchange (ETDEWEB)

    Guseva, M.I.; Gureev, V.M.; Kolbasov, B.N.; Korshunov, S.N.; Martynenko, Yu.V. E-mail: martyn@nfi.kiae.ru; Stolyarova, V.G.; Strunnikov, V.M.; Vasiliev, V.I

    2003-09-01

    Candidate ITER divertor armor materials: carbon-fiber-composite and four tungsten grades/alloys as well as mixed re-deposited W+Be and W+C layers were exposed in electrodynamic plasma accelerator MKT which provided a pulsed deuterium plasma flux simulating plasma disruptions with maximum ion energy of 1-2 keV, an energy density of 300 kJ/m{sup 2} per shot and a pulse duration of {approx}60 {mu}s. The number of pulses was from 2 to 10. The resultant erosion products were collected on a basalt filter and Si-collectors and studied in terms of morphology and size distribution using both scanning and transmission electron microscopy. Metal erosion products usually occurred in the form of spherical droplets, sometimes flakes. Their size distribution depended on the positioning of the collector. Simultaneously irradiated W, CFC and mixed W+Be targets appeared to have undergone a greater erosion than the same targets irradiated individually. Particles sized from 0.01 to 30 {mu}m were found on collectors and on a molten W-surface. A model of droplet emission and behavior in shielding plasma is provided.

  5. Effects of divertor geometry and pumping on plasma performance on DIII-D

    International Nuclear Information System (INIS)

    Allen, S.L.; Hill, D.N.; Porter, G.D.

    1997-06-01

    This paper reports the status of an ongoing investigation to discern the influence of the divertor and plasma geometry on the confinement of both ELM-free and ELMing discharges in DIII-D. The ultimate goal is to achieve a high-performance core plasma which coexists with an advanced divertor plasma. The divertor plasma must reduce the heat flux to acceptable levels; the current technique disperses the heat flux over a wide area by radiation (a radiative divertor). To date, we have obtained our best performance in double-null (DN) high-triangularity (δ ∼ 0.8) ELM-free discharges. As discussed in detail elsewhere, there are several advantages for both the core and divertor plasma with highly-shaped DN operation. Previous radiative-divertor experiments with D 2 injection in DN high-δ ELMing H-mode have shown that this configuration is more sensitive to gas puffing (τ decreases). Moving the X-point away from the target plate (to ∼15 cm above the plate) decreases this sensitivity. Preliminary measurements also indicate that gas puffing reduces the divertor heat flux but does not reduce the plasma pressure along the field line. The up/down heat flux balance can be varied magnetically (by changing the distance between the separatrices), with a slight magnetic imbalance required to balance the heat flux. The overall mission of the Radiative Divertor Project (RDP) is to install a fully pumped and baffled high-δ DN divertor. To date, however, both the DIII-D divertor diagnostics and pump were optimized for lower single-null (LSN) low-δ (δ∼ 0.4) plasmas, so much of the divertor physics has been performed in LSN; these results are discussed in Section 2. As part of the first phase of the RDP, we have installed a new high-δ USN divertor baffle and pump; these results are discussed in Section 3. Both divertor and core parameters are discussed in each case

  6. Plasma decontamination during ergodic divertor experiments in TORE SUPRA

    International Nuclear Information System (INIS)

    Monier-Garbet, P.; DeMichelis, C.; Fall, T.; Ghendrih, Ph.; Goniche, M.; Grosman, A.; Hess, W.; Mattioli, M.

    1991-01-01

    In Tore Supra an ergodic divertor (ED) has been integrated in the machine design and successfully operated, as already reported. This paper analyses the decontamination effect resulting from the creation of an ergodic boundary zone. Two plasma geometrical configurations (outboard and inboard) are studied, the plasma being limited respectively either, on the low field side (lfs), by an outboard limiter (3 to 5 cm ahead of the ED modules) or, on the high field side (hfs), by the graphite inner wall. Strong decontamination effects have already been reported for the first configuration by observing line emission of the intrinsic (carbon and oxygen) and purposely injected (nitrogen) impurities. When limited by the inner wall, the plasma is several centimeters farther from the ED modules than in the lfs configuration. The magnetic perturbation is then greatly reduced, and much smaller decontamination effects should be expected. In this paper, the hfs configuration data is compared with that from the lfs configuration. Preliminary experiments combining lower hybrid current drive and ED operation in the hfs configuration are also reported. (author) 5 refs., 4 figs

  7. Plasma diagnostics for the DIII-D divertor upgrade (abstract)

    International Nuclear Information System (INIS)

    Hill, D.N.; Futch, A.; Buchenauer, D.; Doerner, R.; Lehmer, R.; Schmitz, L.; Klepper, C.C.; Menon, M.; Leikind, B.; Lippmann, S.; Mahdavi, M.A.; Schaffer, M.; Smith, J.; Salmonson, J.; Watkins, J.

    1990-01-01

    The DIII-D tokamak is being upgraded to allow for divertor biasing, baffling, and pumping experiments. This paper gives an overview of the new diagnostics added to DIII-D as part of this advanced divertor program. They include tile current monitors, fast reciprocating Langmuir probes, a fixed probe array in the divertor, fast neutral pressure gauges, and H α measurements with TV cameras and fiber optics coupled to a high-resolution spectrometer

  8. Role of the pump limiter throat-ergodic divertor effect on edge plasma

    International Nuclear Information System (INIS)

    Grosman, A.; Samain, A.; Ghendrih, P.; Capes, H.; Morera, J.P.

    1988-01-01

    A large part of the Tore Supra programme is devoted to plasma edge studies. Two types of such density control apparatus have been implemented, a set of pumps limiters and the ergodic divertor. The goal of the present paper is to investigate the effect of the pump limiter throat on pumping efficiency. We present also the possibilities of the ergodic divertor device to facilitate plasma pumping and power exhaust

  9. One dimensional simulation on stability of detached plasma in a tokamak divertor

    International Nuclear Information System (INIS)

    Nakazawa, Shinji; Nakajima, Noriyoshi; Okamoto, Masao; Ohyabu, Nobuyoshi

    1999-06-01

    The stability of radiation front in the Scrape-Off-Layer (SOL) of a tokamak is studied with a one dimensional fluid code; the time-dependent transport equations are solved in the direction parallel to a magnetic field line. The simulation results show that stable detached solutions exist, where the plasma temperature near the divertor target is ∼2 eV. It is found that whenever such stable detached states are attained, the strong radiation front is contact with or at a small distance from the divertor target. When the energy externally injected into the SOL is decreased below a critical value, the radiation front starts to move towards the X-point, cooling the SOL plasma. In such cases, no stationary solutions such that the radiation front rests in the divertor channel are observed in our parameter space. This qualitatively corresponds to the results of tokamak divertor experiments which show the movement of radiation front. (author)

  10. 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

  11. Detached divertor plasmas in Alcator C-Mod: A study of the role of atomic physics

    International Nuclear Information System (INIS)

    Lipschultz, B.; Boswell, C.; Goetz, J.A.

    1999-01-01

    Detailed profiles of the volumetric recombination occurring in Alcator C-Mod plasmas are presented. During detachment the recombination sink is compared to the divertor plate sink as well as the divertor ion source. Depending on plasma conditions, volume recombination removes between 10 and 75% of the ions before they reach the plates. A second, equally important process that leads to a drop in plate ion current is inferred to be a reduction in divertor ion source, which is correlated with a drop in power flowing into the ionization region and the pressure loss of detachment. For high n e the divertor recombination can cross the separatrix near the x-point, cool the core and lead to a disruption. Experimental measurements show a difference in ion and neutral velocities for H-mode detached plasmas. The resulting ion-neutral collisions are found to be more efficacious than recombination in removing momentum from the ions. The neutral component of volumetric power emission from the divertor has been measured by means of a novel filtering technique to be substantial (∼ 20% of the total divertor volumetric emission). (author)

  12. Exposures of tungsten nanostructures to divertor plasmas in DIII-D

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  13. Bursty fluctuation characteristics in SOL/divertor plasmas of Large Helical Device

    International Nuclear Information System (INIS)

    Ohno, N.; Masuzaki, S.; Morisaki, T.; Ohyabu, N.; Komori, A.; Budaev, V.P.; Miyoshi, H.; Takamura, S.

    2006-10-01

    Bursty electrostatic fluctuation in the scrape off layer (SOL) and the divertor region of the Large Helical Device (LHD) have been investigated by using a Langmuir probe array on a divertor plate and a reciprocating Langmuir probe. Large positive bursty events were often observed in the ion saturation current measured with a divertor probe near the divertor leg at which the magnetic line of force connected to the area of a low-field side with a short connection length. Condition averaging result of the positive bursty events indicates the intermittent feature with a rapid increase and a slow decay is similar to that of plasma blobs observed in tokamaks. On the other hand, at a striking point with a long connection length, negative spikes were observed. Statistical analysis based on probability distribution function (PDF) was employed to investigate the bursty fluctuation property. The observed scaling exponents disagree with the predictions for the self-organized criticality (SOC) paradigm. (author)

  14. Modeling of thermal effects on TIBER II divertor during plasma disruptions

    International Nuclear Information System (INIS)

    Bruhn, M.L.; Perkins, L.J.

    1987-01-01

    Mapping the disruption power flow from the mid-plane of the TIBER Engineering Test Reactor to its divertor and calculating the resulting thermal effects are accomplished through the modification and coupling of three presently existing computer codes. The resulting computer code TADDPAK (Thermal Analysis Divertor during Disruption PAcKage) provides three-dimensional graphic presentations of time and positional dependent thermal effects on a poloidal cross section of the double-null-divertor configured reactor. These thermal effects include incident heat flux, surface temperature, vaporization rate, total vaporization, and melting depth. The dependence of these thermal effects on material choice, disruption pulse shape, and the characteristic thickness of the plasma scrape-off layer is determined through parametric analysis with TADDPAK. This computer code is designed to be a convenient, rapid, and user-friendly modeling tool which can be easily adapted to most tokamak double-null-divertor reactor designs

  15. Particle and power deposition on divertor targets in EAST H-mode plasmas

    International Nuclear Information System (INIS)

    Wang, L.; Xu, G.S.; Guo, H.Y.; Chen, R.; Ding, S.; Gan, K.F.; Gao, X.; Gong, X.Z.; Jiang, M.; Liu, P.; Liu, S.C.; Luo, G.N.; Ming, T.F.; Wan, B.N.; Wang, D.S.; Wang, F.M.; Wang, H.Q.; Wu, Z.W.; Yan, N.; Zhang, L.

    2012-01-01

    The effects of edge-localized modes (ELMs) on divertor particle and heat fluxes were investigated for the first time in the Experimental Advanced Superconducting Tokamak (EAST). The experiments were carried out with both double null and lower single null divertor configurations, and comparisons were made between the H-mode plasmas with lower hybrid current drive (LHCD) and those with combined ion cyclotron resonance heating (ICRH). The particle and heat flux profiles between and during ELMs were obtained from Langmuir triple-probe arrays embedded in the divertor target plates. And isolated ELMs were chosen for analysis in order to reduce the uncertainty resulting from the influence of fast electrons on Langmuir triple-probe evaluation during ELMs. The power deposition obtained from Langmuir triple probes was consistent with that from the divertor infra-red camera during an ELM-free period. It was demonstrated that ELM-induced radial transport predominantly originated from the low-field side region, in good agreement with the ballooning-like transport model and experimental results of other tokamaks. ELMs significantly enhanced the divertor particle and heat fluxes, without significantly broadening the SOL width and plasma-wetted area on the divertor target in both LHCD and LHCD + ICRH H-modes, thus posing a great challenge for the next-step high-power, long-pulse operation in EAST. Increasing the divertor-wetted area was also observed to reduce the peak heat flux and particle recycling at the divertor target, hence facilitating long-pulse H-mode operation. The particle and heat flux profiles during ELMs appeared to exhibit multiple peak structures, and were analysed in terms of the behaviour of ELM filaments and the flux tubes induced by modified magnetic topology during ELMs. (paper)

  16. A numerical study of plasma detachment conditions in JET divertor plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Simonini, R; Corrigan, G; Radford, G; Spence, J; Taroni, A; Weber, S [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

    1994-07-01

    Simulation results obtained with the EDGE2D/U code confirm that for a given particle inventory in the SOL (including the divertor), the main parameter determining whether or not particle, momentum and energy detachment occurs, is the residual power P - P{sub lost}, where P is the total power entering the SOL and P{sub lost} is the power lost by transport to walls and by volume losses in the SOL outside the region where detachment takes place. For particle contents leading to reasonable values of the separatrix mid-plane density, detachment is found if the residual power is low enough. Typically the residual power must be inferior to 3 MW for good detachment, with the exact value depending on the geometry of the divertor, the transport assumptions and the neutral recirculation scheme. The results show that divertor plasma conditions relevant for the study of power exhaust and impurity control problems are possible in JET. 9 refs., 2 figs., 1 tab.

  17. 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

  18. 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

  19. 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.

  20. 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.

  1. Numerical study of divertor plasma transport with thermal force due to temperature gradient

    International Nuclear Information System (INIS)

    Ohtsu, Shigeki; Tanaka, Satoru; Yamawaki, Michio

    1992-01-01

    A one-dimensional, steady state divertor plasma model is developed in order to study the carbon impurity transport phenomena considering thermal force. The divertor plasma is composed of four regions in terms of momentum transport between hydrogen and carbon impurity: Momentum transferring region, equilibrium region, hydrogen recycling region and carbon recycling region. In the equilibrium region where the friction force is counterbalanced by the thermal force, the localization of carbon impurity occurs. The sufficient condition to avoid the reverse of carbon velocity due to the thermal force is evaluated. (orig.)

  2. A 3D Monte Carlo code for plasma transport in island divertors

    International Nuclear Information System (INIS)

    Feng, Y.; Sardei, F.; Kisslinger, J.; Grigull, P.

    1997-01-01

    A fully 3D self-consistent Monte Carlo code EMC3 (edge Monte Carlo 3D) for modelling the plasma transport in island divertors has been developed. In a first step, the code solves a simplified version of the 3D time-independent plasma fluid equations. Coupled to the neutral transport code EIRENE, the EMC3 code has been used to study the particle, energy and neutral transport in W7-AS island divertor configurations. First results are compared with data from different diagnostics (Langmuir probes, H α cameras and thermography). (orig.)

  3. Taming the plasma-material interface with the snowflake divertor.

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V A

    2015-04-24

    Experiments in several tokamaks have provided increasing support for the snowflake configuration as a viable tokamak heat exhaust concept. This white paper summarizes the snowflake properties predicted theoretically and studied experimentally, and identifies outstanding issues to be resolved in existing and future facilities before the snowflake divertor can qualify for the reactor interface.

  4. Divertor detachment

    Science.gov (United States)

    Krasheninnikov, Sergei

    2015-11-01

    The heat exhaust is one of the main conceptual issues of magnetic fusion reactor. In a standard operational regime the large heat flux onto divertor target reaches unacceptable level in any foreseeable reactor design. However, about two decades ago so-called ``detached divertor'' regimes were found. They are characterized by reduced power and plasma flux on divertor targets and look as a promising solution for heat exhaust in future reactors. In particular, it is envisioned that ITER will operate in a partly detached divertor regime. However, even though divertor detachment was studied extensively for two decades, still there are some issues requiring a new look. Among them is the compatibility of detached divertor regime with a good core confinement. For example, ELMy H-mode exhibits a very good core confinement, but large ELMs can ``burn through'' detached divertor and release large amounts of energy on the targets. In addition, detached divertor regimes can be subject to thermal instabilities resulting in the MARFE formation, which, potentially, can cause disruption of the discharge. Finally, often inner and outer divertors detach at different plasma conditions, which can lead to core confinement degradation. Here we discuss basic physics of divertor detachment including different mechanisms of power and momentum loss (ionization, impurity and hydrogen radiation loss, ion-neutral collisions, recombination, and their synergistic effects) and evaluate the roles of different plasma processes in the reduction of the plasma flux; detachment stability; and an impact of ELMs on detachment. We also evaluate an impact of different magnetic and divertor geometries on detachment onset, stability, in- out- asymmetry, and tolerance to the ELMs. Supported by the U.S. Department of Energy Office of Science, Office of Fusion Energy Sciences under Award Number DE-DE-FG02-04ER54739 at UCSD.

  5. Advanced divertor configurations with large flux expansion

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V.A., E-mail: vlad@llnl.gov [Lawrence Livermore National Laboratory, Livermore, CA (United States); Bell, R.E.; Diallo, A.; Gerhardt, S.; Kaye, S.; Kolemen, E.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); McLean, A. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Menard, J.E.; Paul, S.F.; Podesta, M. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Raman, R. [University of Washington, Seattle, WA (United States); Ryutov, D.D. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Scotti, F.; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Maingi, R. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mueller, D.M.; Roquemore, A.L. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Reimerdes, H.; Canal, G.P. [Ecole Polytechnique Fédérale de Lausanne, Centre de Recherches en Physique des Plasmas, Association Euratom Confédération Suisse, Lausanne (Switzerland); and others

    2013-07-15

    Experimental studies of the novel snowflake divertor concept (D. Ryutov, Phys. Plasmas 14 (2007) 064502) performed in the NSTX and TCV tokamaks are reviewed in this paper. The snowflake divertor enables power sharing between divertor strike points, as well as the divertor plasma-wetted area, effective connection length and divertor volumetric power loss to increase beyond those in the standard divertor, potentially reducing heat flux and plasma temperature at the target. It also enables higher magnetic shear inside the separatrix, potentially affecting pedestal MHD stability. Experimental results from NSTX and TCV confirm the predicted properties of the snowflake divertor. In the NSTX, a large spherical tokamak with a compact divertor and lithium-coated graphite plasma-facing components (PFCs), the snowflake divertor operation led to reduced core and pedestal impurity concentration, as well as re-appearance of Type I ELMs that were suppressed in standard divertor H-mode discharges. In the divertor, an otherwise inaccessible partial detachment of the outer strike point with an up to 50% increase in divertor radiation and a peak divertor heat flux reduction from 3–7 MW/m{sup 2} to 0.5–1 MW/m{sup 2} was achieved. Impulsive heat fluxes due to Type-I ELMs were significantly dissipated in the high magnetic flux expansion region. In the TCV, a medium-size tokamak with graphite PFCs, several advantageous snowflake divertor features (cf. the standard divertor) have been demonstrated: an unchanged L–H power threshold, enhanced stability of the peeling–ballooning modes in the pedestal region (and generally an extended second stability region), as well as an H-mode pedestal regime with reduced (×2–3) Type I ELM frequency and slightly increased (20–30%) normalized ELM energy, resulting in a favorable average energy loss comparison to the standard divertor. In the divertor, ELM power partitioning between snowflake divertor strike points was demonstrated. The NSTX

  6. 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.

  7. Two-dimensional analysis of limiter/divertor transition in scrape-off layer plasmas

    International Nuclear Information System (INIS)

    Ueda, N.; Itoh, K.; Itoh, S.I.

    1989-01-01

    The structures of scrape-off layer and divertor plasmas have been studied numerically with a neutral code and a two-dimensional fluid code. Doublet-III is taken as an example for an open divertor configuration. A decisive parameter is the distance between the plasma surface (determined by the magnetic separatrix) and the limiter, which is varied in order to assess the interaction of the plasma with the limiter as well as the effect of neutrals on the main plasma. The minimum value of the limiter clearance needed to prevent plasma-limiter interaction is determined. The scaling of the edge temperature and the dependence of the e-folding length of the scrape-off layer plasma on the heating power are obtained. (author). 16 refs, 17 figs

  8. Transport simulation analysis of peripheral plasma with the open and the closed LHD divertor

    International Nuclear Information System (INIS)

    Kawamura, G.; Kobayashi, M.; Shoji, M.; Morisaki, T.; Masuzaki, S.; Feng, Y.

    2014-10-01

    Simulation modeling of the ergodic and divertor plasmas of the Large Helical Device (LHD) and its application to analysis of neutral particles, plasma, and impurity transport is presented. EMC3-EIRENE simulation with a new calculation mesh system is employed to evaluate effects of different divertor configurations: the open and the closed divertor. Qualitatively good agreement of neutral gas pressure with measurements was obtained, where the closed configuration causes roughly 20 times higher pressure under a dome structure than the open configuration. Effects of different configurations and gas pumping were investigated to understand recycling. Impurity accumulation and impurity screening in the ergodic region were investigated and differences caused by the configurations are evaluated. The closed configuration causes large impurity accumulation but the impurity screening effect suppress the accumulation at the same level of as the open configuration. (author)

  9. The ‘churning mode’ of plasma convection in the tokamak divertor region

    International Nuclear Information System (INIS)

    Ryutov, D D; Cohen, R H; Farmer, W A; Rognlien, T D; Umansky, M V

    2014-01-01

    The churning mode can arise in a toroidally-symmetric plasma where it causes convection in the vicinity of the poloidal magnetic field null. The mode is driven by the toroidal curvature of magnetic field lines coupled with a pressure gradient. The toroidal equilibrium conditions cannot be satisfied easily in the virtual absence of the poloidal field (PF)—hence the onset of this mode, which ‘churns’ the plasma around the PF null without perturbing the strong toroidal field. We find the conditions under which this mode can be excited in magnetic configurations with first-, second-, and third-order PF nulls (i.e., in the geometry of standard, snowflake and cloverleaf divertors). The size of the affected zone in second- and third-order-null divertors is much larger than in a standard first-order-null divertor. The proposed phenomenological theory allows one to evaluate observable characteristics of the mode, in particular the frequency and amplitude of the PF perturbations. The mode spreads the tokamak heat exhaust between multiple divertor legs and may lead to a broadening of the plasma width in each leg. The mode causes much more intense plasma convection in the poloidal plane than the classical plasma drifts. (invited comment)

  10. Driving mechanism of SOL plasma flow and effects on the divertor performance in JT-60U

    International Nuclear Information System (INIS)

    Asakura, N.

    2002-01-01

    SOL plasma flow plays an important role in the plasma transport along the field lines, and influences control of the divertor plasma and impurity ions. Recently, mechanisms producing the SOL flow such as drifts produced by electric field and pressure gradient are pointed out. In JT-60U, three reciprocating Mach probes were installed at the high-field-side (HFS) baffle, low-field-side (LFS) midplane and just below the X-point. The measurements of the SOL flow and plasma profiles both at the HFS and LFS, for the first time, found out the SOL flow pattern and its driving mechanism. 'Flow reversal' was found near the separatrix of the HFS and LFS. Radial profiles of the SOL flow were similar to those calculated numerically using the UEDGE code with the plasma drifts included. SOL particle fluxes towards the HFS and LFS divertors were, for the first time, evaluated. Important physics issues for the divertor design and operation, such as in-out asymmetries of the heat and particle fluxes, and control of impurity ions with intense gas puff and divertor pump (puff and pump), were investigated. (author)

  11. 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)

  12. Simulation of an ITER-like dissipative divertor plasma with a combined edge plasma Navier-Stokes neutral model

    International Nuclear Information System (INIS)

    Knoll, D.A.; McHugh, P.R.; Krasheninnikov, S.I.; Sigmar, D.J.

    1996-01-01

    A combined edge plasma/Navier-Stokes neutral transport model is used to simulate dissipative divertor plasmas in the collisional limit for neutrals on a simplified two-dimensional slab geometry with ITER-like plasma conditions and scale lengths. The neutral model contains three momentum equations which are coupled to the plasma through ionization, recombination, and ion-neutral elastic collisions. The neutral transport coefficients are evaluated including both ion-neutral and neutral-neutral collisions. (orig.)

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

    Directory of Open Access Journals (Sweden)

    Yuki Homma

    2017-08-01

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

  14. Advanced divertor concepts

    International Nuclear Information System (INIS)

    Ohyabu, N.; Komori, A.; Sagara, A.; Suzuki, H.; Morisaki, T.; Masuzaki, S.; Watanabe, T.; Noda, N.; Motojima, O.

    1996-01-01

    LHD divertor development program has generated various innovative divertor concepts and technologies which will help to improve the plasma performance in both helical and tokamak devices. They are two divertor operational scenarios (confinement improvement by generating high temperature divertor plasma and simultaneous achievement of radiative cooling and H-mode-like confinement improvement). Local island divertor geometry has also been proposed. This new divertor has been successfully tested in the CHS device and is planned to be installed in the LHD device. In addition, technological development of new efficient hydrogen pumping schemes (carbon sheet pump and membrane pump) are being pursued for enhancement of the divertor control capability. 17 refs., 8 figs

  15. Divertor experiments in a toroidal plasma, with E x B drift due to an applied radial electric field

    International Nuclear Information System (INIS)

    Strait, E.J.

    1979-09-01

    It is proposed that the E x B drift arising from an externally applied electric field could be used in a tokamak or other toroidal magnetic plasma confinement device to remove plasma and impurities from the region near the wall and reduce the amount of plasma striking the wall. This could either augment or replace a conventional magnetic field divertor. Among the possible advantages of this scheme are easy external control over the rate of removal of plasma, more rapid removal than the naturally occurring rate in a magnetic divertor, and simplification of construction if the magnetic divertor is eliminated. Results of several related experiments performed in the Wisconsin Levitated Octupole are presented

  16. 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.)

  17. Method of plasma impurity control without magnetic divertor

    International Nuclear Information System (INIS)

    Schivell, J.F.

    1977-06-01

    A method is proposed for controlling impurity generation in a tokomak by skimming and pumping the scrape-off. This method avoids many of the complications of a magnetic divertor, such as specially configured magnetic fields, toroidal symmetry, and inefficient use of toroidal field volume. Estimates are given for operating parameters. Impurity reductions of as much as a factor of 10 should be achievable. The necessary high-capacity pump would employ either titanium gettering or cryocondensation

  18. A mechanism for large divertor plasma energy loss via lithium radiation in tokamaks

    Science.gov (United States)

    Rognlien, T. D.; Meier, E. T.; Soukhanovskii, V. A.

    2012-10-01

    Lithium has been used as a wall-conditioning element in a number of tokamaks over the years, including TFTR, FTU, and NSTX, where core plasma energy confinement and particle control are often found to improve following such conditioning. Here the possible role of Li in providing substantial energy loss for divertor plasmas via line radiation is reported. A multi-charge-state 2D UEDGE fluid model is used where the hydrogenic and Li ions and neutrals are each evolved as separate species and separate equations are solved for the electron and ion temperatures. It is shown that a sufficient level of Li neutrals evolving from the divertor surface via sputtering or evaporation can induce energy detachment of the divertor plasma, yielding a strongly radiating zone near the divertor where ionization and recombination from/to neutral Li can radiate most of the power flowing into the scrape-off layer while maintaining low core contamination. A local peaking of Li emissivity for electron temperatures near 1 eV appears to play an important role in the detachment of the mixed deuterium/Li plasma. Evidence of such behavior from NSTX discharges will be discussed.

  19. Effects of the New Island Divertor on the Plasma Performance in the W7-AS Stellarator

    International Nuclear Information System (INIS)

    Grigull, P.; McCormick, K.; Baldzuhn, J.; Burhenn, R.; Brakel, R.; Ehmler, H.; Feng, Y.; Gadelmeier, F.; Giannone, L.; Hartmann, D.; Hildebrandt, D.; Hirsch, M.; Jaenicke, R.; Kisslinger, J.; Klinger, T.; Knauer, J.; Koenig, R.; Naujoks, D.; Niedermeyer, H.; Pasch, E.

    2003-01-01

    The island divertor in the W7-AS stellarator enables access to a new NBI-heated, high density operating regime with promising confinement properties. This regime -- the High Density H-Mode -- displays no evident mode activity, is extant above a threshold density and characterized by flat density profiles, high energy- and low impurity-confinement times and edge localized radiation. Impurity accumulation, normally associated with ELM-free H-modes, is avoided. Quasi steady-state discharges with n e up to 4 1020 m-3, edge radiation levels up to 90%, and partial plasma detachment at the divertor targets can be simultaneously realized

  20. Magnetic field structure near the plasma boundary in helical systems and divertor tokamaks

    International Nuclear Information System (INIS)

    Nagasaki, Kazunobu; Itoh, Kimitaka

    1990-02-01

    Magnetic field structure of the scrape off layer (SOL) region in both helical systems and divertor tokamaks is studied numerically by using model fields. The connection length of the field line to the wall is calculated. In helical systems, the connection length, L, has a logarithmic dependence on the distance from the outermost magnetic surface or that from the residual magnetic islands. The effect of axisymmetric fields on the field structure is also determined. In divertor tokamaks, the connection length also has logarithmic properties near the separatrix. Even when the perturbations, which resonate to rational surfaces near the plasma boundary, are added, logarithmic properties still remain. We compare the connection length of torsatron/helical-heliotron systems with that of divertor tokamaks. It is found that the former is shorter than the latter by one order magnitude with similar aspect ratio. (author)

  1. Driving mechanism of SOL plasma flow and effects on the divertor performance in JT-60U

    International Nuclear Information System (INIS)

    Asakura, N.; Takenaga, H.; Sakurai, S.

    2003-01-01

    The measurements of the SOL flow and plasma profiles both at the high-field-side (HFS) and low field- side (LFS), for the first time, identified the SOL flow pattern and its driving mechanism. 'Flow reversal' was found near the HFS and LFS separatrix of the main plasma for the ion ∇β drift direction towards the divertor. Radial profiles of the SOL flow were similar to those calculated numerically using the UEDGE code with the plasma drifts included although Mach numbers in measurements were greater than those obtained numerically. Particle fluxes towards the HFS and LFS divertors produced by the parallel SOL flow and E r xB drift flow were evaluated. The particle flux for the case of intense gas puff and divertor pump (puff and pump) was investigated, and it was found that both the Mach number and collisionality were enhanced, in particular, at HFS. Drift flux in the private flux region was also evaluated, and important physics issues for the divertor design and operation, such as in-out asymmetries of the heat and particle fluxes, and control of impurity ions were investigated. (author)

  2. Driving mechanism of SOL plasma flow an effects on the divertor performance in JT-60U

    International Nuclear Information System (INIS)

    Asakura, Nobuyuki; Takenaga, H.; Sakurai, S.

    2003-01-01

    The measurements of the scrape-off layer(SOL) flow and plasma profiles both at the high-field-side (HFS) and low-field-side (LFS), for the first time, identified the SOL flow pattern and its driving mechanism. 'Flow reversal' was found near the HFS and LFS separatrix of the main plasma for the ion ∇B drift direction towards the divertor, Radial profiles of the SOL flow were similar to those calculated numerically using the UEDGE code with the plasma drifts included although Mach numbers in measurements were greater than those obtained numerically. Particle fluxes towards the HFS and LFS divertors produced by the parallel SOL flow and E r xB drift flow were evaluated. The particle flux for the case of intense gas puff and divertor pump (puff and pump) was investigated, and it was found that both the Mach number and collisionality were enhanced, in particular, at HFS. Drift flux in the private flux region was also evaluated, and important physics issues for the divertor design and operation, such as in-out asymmetries of the heat and particle fluxes, and control of impurity ions were investigated. (author)

  3. 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.

  4. 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.)

  5. 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

  6. Charge-exchange processes in a divertor plasma with account for excited particles

    International Nuclear Information System (INIS)

    Krasheninnikov, S.I.; Lisitsa, V.S.; Pigarov, A.Yu.

    1988-01-01

    A model describing dynamics of neutral atoms and multicharge ions in tokamak plasma, taking account of cascade excitation effect on charge exchange and ionization processes, is constructed. Dependences of effective rate of processes of proton charge exchange on hydrogen atom and non-resonance helium atom charge exchange on proton and α-particle- on atomic hydrogen on tokamak divertor plasma parameters are calculated. It is shown that H + +He→H-He + charge exchange can make up a notable shave (∼30%) in full helium ionization rate. Accounting for Ge 2+ charge exchange on atomic hydrogen under INTOR reactor divertor plasma conditions can lead to substantial He 2+ →He + conversion and thus increase diverter plate sputtering by helium ions

  7. Divertor modeling for the design of the National Centralized Tokamak with high beta steady-state plasmas

    International Nuclear Information System (INIS)

    Kawashima, H.; Sakurai, S.; Shimizu, K.; Takizuka, T.; Tamai, H.; Matsukawa, M.; Fujita, T.; Miura, Y.

    2006-01-01

    The modification of the JT-60U to a fully superconducting coil tokamak, National Centralized Tokamak (NCT) facility, has been programmed to accomplish the high beta steady-state plasma research. A 2D divertor simulation code, SOLDOR/NEUT2D, is applied to the construction of a database for optimum design of the divertor. A semi-closed divertor configuration with vertical target is adopted as the first conceptual divertor design on NCT. With an anticipated SOL power flux of 12 MW at the high beta steady-state operation, the peak heat load on the divertor target is evaluated to be ∼16 MW/m 2 . Effects of divertor geometry and intense gas puffing are demonstrated with a view to reduce the heat load. For the simulation of divertor pumping, we find that the pumping efficiency increases by a factor of 2∼3 by narrowing the divertor gap from 20 to 5 cm. An attractive feature in reducing the heat load and improving the particle controllability has been obtained for a new divertor design due to a recent progress in NCT design

  8. Plasma density control with ergodic divertor on Tore Supra; Controle de la densite du plasma en presence du divertor ergodique dans le tokamak Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Meslin, B

    1998-04-30

    Plasma density control on the tokamak Tore Supra is important for the optimization of every experimental scenario dealing with the improvement of plasma performances. Specific conditions are required both in the plasma bulk and at the edge. Within the framework of the present study, a magnetic configuration is used in the e plasma edge of Tore Supra: the ergodic divertor configuration. A magnetic perturbation which is resonant with the permanent field destroys the plasma confinement locally, opening the field lines onto the material components. They aim of the study is the characterization of the edge density in every relevant scenario for Tore Supra. The first part of this work is dedicated to density and temperature measurements by a series of fixed Langmuir probes located at the very edge of the plasma. Thanks to them, density regimes have been put in evidence during experiments where the volume averaged density , an usual control parameter of the plasma, was varied. The analysis of heat and particle transport through the plasma edge region explains the mechanisms leading to those regimes. The essential factor in our analysis is the dependence of the electron conductivity and ionization depth on temperature. While heat conduction governs the heat transport, the edge density varies linearly according to . Below a critical temperature, reached when the ion flux amplification at constant power density is large enough, a parallel temperature gradient appears leading to a density gradient in the opposite direction in order to maintain the pressure constant along the field lines. A high recycling regime is obtained and the edge density varies like {sup 3}. The pressure conservation is no more satisfied during the detachment of the plasma, which is characterized by a high neutral density at low temperatures leading to a ion momentum loss by friction against the neutrals. The edge density drops in those conditions. These regimes are similar

  9. 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

  10. 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.)

  11. 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

  12. 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.

  13. 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.

  14. 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.

  15. Experimental simulation and numerical modeling of vapor shield formation and divertor material erosion for ITER typical plasma disruptions

    International Nuclear Information System (INIS)

    Wuerz, H.; Arkhipov, N.I.; Bakhtin, V.P.; Konkashbaev, I.; Landman, I.; Safronov, V.M.; Toporkov, D.A.; Zhitlukhin, A.M.

    1995-01-01

    The high divertor heat load during a tokamak plasma disruption results in sudden evaporation of a thin layer of divertor plate material, which acts as vapor shield and protects the target from further excessive evaporation. Formation and effectiveness of the vapor shield are theoretically modeled and are experimentally analyzed at the 2MK-200 facility under conditions simulating the thermal quench phase of ITER tokamak plasma disruptions. ((orig.))

  16. Effects of low-Z and high-Z impurities on divertor detachment and plasma confinement

    Directory of Open Access Journals (Sweden)

    H.Q. Wang

    2017-08-01

    Full Text Available The impurity-seeded detached divertor is essential for heat exhaust in ITER and other reactor-relevant devices. Dedicated experiments with injection of N2, Ne and Ar have been performed in DIII-D to assess the impact of the different impurities on divertor detachment and confinement. Seeding with N2, Ne and Ar all promote divertor detachment, greatly reducing heat flux near the strike point. The upstream plasma density at the onset of detachment decreases with increasing impurity-puffing flow rates. For all injected impurity species, the confinement and pedestal pressure are correlated with the impurity content and the ratio of separatrix loss power to the l-H transition threshold power. As the divertor plasma approaches detachment, the high-Z impurity seeding tends to degrade the core confinement owing to the increased core radiation. In particular, Ar injection with up to 50% of the injected power radiating in the core cools the pedestal and core plasmas, thus significantly degrading the confinement. As for Ne seeding, medium confinement with H98∼0.8 can be maintained during the detachment phase with the pedestal temperature being reduced by about 50%. In contrast, in the N2 seeded plasmas, radiation is predominately confined in the boundary plasma, which leads to less effect on the confinement and pedestal. In the case of strong N2 gas puffing, the confinement recovers during the detachment, from ∼20% reduction at the onset of the detachment to greater than unity comparable to that before the seeding. The core and pedestal temperatures feature a reduction of 30% from the initial attached phase and remain nearly constant during the detachment phase. The improvement in confinement appears to arise from the increase in pedestal and core density despite the temperature reduction.

  17. 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

  18. The divertor remote maintenance project

    International Nuclear Information System (INIS)

    Maisonnier, D.; Martin, E.; Akou, K.

    2001-01-01

    Remote replacement of the ITER divertor will be required several times during the life of ITER. To facilitate its regular exchange, the divertor is assembled in the ITER vacuum vessel from 60 cassettes. Radial movers transport each cassette along radial rails through the handling ports and into the vessel where a toroidal mover lifts and transports the cassette around a pair of toroidal rails. Once at its final position the cassette is locked to the toroidal rails and is accurately aligned in both poloidal and toroidal directions. A further requirement on the divertor is to minimise the amount of activated waste to be sent to a repository. To this end the cassettes have been designed to allow the remote replacement, in a hot cell, of their plasma facing components. The paper describes the two facilities built at ENEA Brasimone, Italy, whose aim is to demonstrate the reliable remote maintenance of the divertor cassettes. (author)

  19. The divertor remote maintenance project

    International Nuclear Information System (INIS)

    Maisonnier, D.; Martin, E.; Akou, K.

    1999-01-01

    Remote replacement of the ITER divertor will be required several times during the life of ITER. To facilitate its regular exchange, the divertor is assembled in the ITER vacuum vessel from 60 cassettes. Radial movers transport each cassette along radial rails through the handling ports and into the vessel where a toroidal mover lifts and transports the cassette around a pair of toroidal rails. Once at its final position the cassette is locked to the toroidal rails and is accurately aligned in both poloidal and toroidal directions. A further requirement on the divertor is to minimise the amount of activated waste to be sent to a repository. To this end the cassettes have been designed to allow the remote replacement, in a hot cell, of their plasma facing components. The paper describes the two facilities built at ENEA Brasimone, Italy, whose aim is to demonstrate the reliable remote maintenance of the divertor cassettes. (author)

  20. End loss analyzer system for measurements of plasma flux at the C-2U divertor electrode

    Energy Technology Data Exchange (ETDEWEB)

    Griswold, M. E., E-mail: mgriswold@trialphaenergy.com; Korepanov, S.; Thompson, M. C. [Tri Alpha Energy, P.O. Box 7010, Rancho Santa Margarita, California 92688 (United States)

    2016-11-15

    An end loss analyzer system consisting of electrostatic, gridded retarding-potential analyzers and pyroelectric crystal bolometers was developed to characterize the plasma loss along open field lines to the divertors of C-2U. The system measures the current and energy distribution of escaping ions as well as the total power flux to enable calculation of the energy lost per escaping electron/ion pair. Special care was taken in the construction of the analyzer elements so that they can be directly mounted to the divertor electrode. An attenuation plate at the entrance to the gridded retarding-potential analyzer reduces plasma density by a factor of 60 to prevent space charge limitations inside the device, without sacrificing its angular acceptance of ions. In addition, all of the electronics for the measurement are isolated from ground so that they can float to the bias potential of the electrode, 2 kV below ground.

  1. 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.

  2. 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.)

  3. Understanding the SOL flow in L-mode plasma on divertor tokamaks, and its influence on the plasma transport

    International Nuclear Information System (INIS)

    Asakura, Nobuyuki

    2007-01-01

    Significant progress has been made in understanding the driving mechanisms in SOL mass transport along the magnetic field lines (SOL flow). SOL flow measurements by Mach probes and impurity plume have been performed in L-mode plasma at various poloidal locations in divertor tokamaks. All results showed common SOL flow patterns: subsonic flow with parallel Mach number (M parallel ) of 0.2-1 was generated from the Low-Field-Side (LFS) SOL to the High-Field-Side (HFS) divertor for the ion ∇B drift towards the divertor. The SOL flow pattern was formed mainly by LFS-enhanced asymmetry in diffusion and by classical drifts. In addition, divertor detachment and/or intense puffing-and-pump enhanced the HFS SOL flow. Most codes have incorporated drift effects, and asymmetric diffusion was modelled to simulate the fast SOL flow. Influences of the fast SOL flow on the impurity flow in the SOL, shielding from core plasma, and deposition profile, were directly observed in experiments

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  5. Divertor scenario development for NSTX Upgrade

    Science.gov (United States)

    Soukhanovskii, V. A.; McLean, A. G.; Meier, E. T.; Rognlien, T. D.; Ryutov, D. D.; Bell, R. E.; Diallo, A.; Gerhardt, S. P.; Kaita, R.; Kolemen, E.; Leblanc, B. P.; Menard, J. E.; Podesta, M.; Scotti, F.

    2012-10-01

    In the NSTX-U tokamak, initial plans for divertor plasma-facing components (PFCs) include lithium and boron coated graphite, with a staged transition to molybdenum. Steady-state peak divertor heat fluxes are projected to reach 20-30 MW/m^2 in 2 MA, 12 MW NBI-heated discharges of up to 5 s duration, thus challenging PFC thermal limits. Based on the recent NSTX divertor experiments and modeling with edge transport code UEDGE, a favorable basis for divertor power handling in NSTX-U is developed. The snowflake divertor geometry and feedback-controlled divertor impurity seeding applied to the lower and upper divertors are presently envisioned. In the NSTX snowflake experiments with lithium-coated graphite PFCs, the peak divertor heat fluxes from Type I ELMs and between ELMs were significantly reduced due to geometry effects, increased volumetric losses and null-point convective redistribution between strike points. H-mode core confinement was maintained at H98(y,2)<=1 albeit the radiative detachment. Additional CD4 seeding demonstrated potential for a further increase of divertor radiation.

  6. The MAST improved divertor

    International Nuclear Information System (INIS)

    Darke, A.C.; Hayward, R.J.; Counsell, G.F.; Hawkins, K.

    2005-01-01

    The Mega Amp Spherical Tokamak (MAST) at Culham is one of the leading world machines studying the spherical tokamak (ST) concept. At the time of the initial construction in 1998 little was known about the sort of divertor structures that would be required in an ST. The machine was therefore provided with relatively rudimentary structures that were designed mostly to protect important components from the hot plasma. While these have served the machine well it was accepted that they might not be suitable when operating MAST to its full potential. The years of experience of operating MAST have led to the design, manufacture and now installation of a new divertor, the MAST improved divertor (MID), that should be able to cope with the full performance of the machine. The design is based on imbricated (fan-shaped) disks of tiles at the top and bottom of the machine for the outer strike points, giving an excellent compromise between power handling and diagnostic access, with substantial new centre column strike point armour and a shaped plate in between. High purity graphite is chosen as the plasma facing material in preference to CFC since in this case it has a better balance of performance and cost. The lower imbricated disk is insulated in alternate sectors for studies of divertor biasing and extensive diagnostics and additional inboard gas injection are included

  7. New Modular Heliotron system compatible with closed helical divertor and good plasma confinement

    International Nuclear Information System (INIS)

    Yamazaki, K.; Watanabe, K.Y.

    1994-04-01

    A new helical system ('Modular Heliotron') with improved modular coils compatible with efficient closed helical divertor and good plasma confinement property is proposed based on a Heliotron system with continuous helical coils and one pair of poloidal coils. The physics optimization of this system as a function of the gap angle between adjacent modular coils has been carried out by means of vacuum magnetic surface calculations and finite-beta plasma analyses, and a new improved coil system is invented by combining sectored helical field coils with sectored returning poloidal field coils. The Modular Heliotron with standard coil winding law (reference Modular Heliotron) was previously proposed, but it is found that this is not appropriate to keep clean helical divertor and high beta configuration when the coil gap becomes large. By modulating the modular coil winding with outside-plus and inside-minus pitch modulation, almost the same good magnetic configuration as that of a conventional Heliotron can be produced. The optimal gap angle is determined as a function of the modulation parameter. This improved Modular Heliotron permits larger gap angle between adjacent modules and produces more clean helical divertor configuration than the reference Modular Heliotron. All these helical system are created by only modular coils without poloidal coils. (author)

  8. New modular heliotron system compatible with closed helical divertor and good plasma confinement

    International Nuclear Information System (INIS)

    Yamazaki, K.; Watanabe, K.Y.

    1995-01-01

    A new helical system ('modular heliotron') with improved modular coils compatible with an efficient closed helical divertor and a good plasma confinement property is proposed, based on a heliotron system with continuous helical coils and one pair of poloidal coils. The physics optimization of this system as a function of the gap angle between adjacent modular coils has been carried out by means of vacuum magnetic surface calculations and finite-beta plasma analyses, and a new improved coil system is invented by combining sectored helical field coils with sectored returning poloidal field coils. A modular heliotron with standard coil winding law (the reference modular heliotron) was previously proposed, but it is found that this was not appropriate to keep a clean helical divertor and high beta configuration when the coil gap becomes large. By modulating the modular coil winding with outside-plus and inside-minus pitch modulation, almost the same good magnetic configuration as that of a conventional heliotron can be produced. The optimal gap angle is determined as a function of the modulation parameter. This improved modular heliotron permits a larger gap angle between adjacent modules and produces a cleaner helical divertor configuration than the reference modular heliotron. All these helical systems are created by only modular coils without poloidal coils. (author). Letter-to-the-editor. 11 refs, 7 figs

  9. Plasma performance of Wendelstein 7-AS with the new boundary-island divertor modules

    International Nuclear Information System (INIS)

    McCormick, K.; Grigull, P.; Burhenn, R.; Brakel, R.; Ehmler, H.; Feng, Y.; Gadelmeier, F.; Giannone, L.; Hildebrandt, D.; Hirsch, M.; Jaenicke, R.; Kisslinger, J.; Klinger, T.; Klose, S.; Knauer, J.P.; Konig, R.; Kuhner, G.; Laqua, H.P.; Naujoks, D.; Niedermeyer, H.; Pasch, E.; Ramasubramanian, N.; Rust, N.; Sardei, F.; Wagner, F.; Weller, A.; Wenzel, U.; Werner, A.

    2002-01-01

    A promising new plasma operational regime on the Wendelstein stellarator W7-AS has been discovered. It is extant above a threshold density and characterized by flat density profiles, high energy- and low impurity-confinement times and edge-localized radiation. Impurity accumulation is avoided. Quasi-stationary discharges with line-averaged densities n e to 4x10 20 m -3 , radiation levels to 90%, and partial plasma detachment at the divertor target plates can be simultaneously realized. Energy confinement is up to twice that predicted by a conventional scaling. Copyright (2002) Australian National University- Research School of Physical Sciences and Engineering

  10. A supersonic gas target for a bundle divertor plasma

    International Nuclear Information System (INIS)

    Chang, F.R.; Fisher, J.L.

    1982-01-01

    A novel gas target concept for recovering both energy and particles from a high-energy plasma stream is presented. This concept includes the maintenance of a pressure discontinuity by a normal shock and a very high mass flow rate in a relatively small system. The pressure discontinuity allows the exhaust plasma stream to minimize backflow into the plasma, by interacting with the target in a low-pressure region; the high mass flow rate allows exit temperatures that are reasonable from a materials viewpoint and suitable for energy recovery. (author)

  11. 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.

  12. The effects of particle recycling on the divertor plasma: A particle-in-cell with Monte Carlo collision simulation

    Science.gov (United States)

    Chang, Mingyu; Sang, Chaofeng; Sun, Zhenyue; Hu, Wanpeng; Wang, Dezhen

    2018-05-01

    A Particle-In-Cell (PIC) with Monte Carlo Collision (MCC) model is applied to study the effects of particle recycling on divertor plasma in the present work. The simulation domain is the scrape-off layer of the tokamak in one-dimension along the magnetic field line. At the divertor plate, the reflected deuterium atoms (D) and thermally released deuterium molecules (D2) are considered. The collisions between the plasma particles (e and D+) and recycled neutral particles (D and D2) are described by the MCC method. It is found that the recycled neutral particles have a great impact on divertor plasma. The effects of different collisions on the plasma are simulated and discussed. Moreover, the impacts of target materials on the plasma are simulated by comparing the divertor with Carbon (C) and Tungsten (W) targets. The simulation results show that the energy and momentum losses of the C target are larger than those of the W target in the divertor region even without considering the impurity particles, whereas the W target has a more remarkable influence on the core plasma.

  13. Dynamic behavior of detached recombining plasmas during ELM-like plasma heat pulses in the divertor plasma simulator NAGDIS-II

    International Nuclear Information System (INIS)

    Uesugi, Y.; Hattori, N.; Nishijima, D.; Ohno, N.; Takamura, S.

    2001-01-01

    It has been recognized that the ELMs associated with a good confinement at the edge, such as H-mode, must bring an enormous energy to the divertor target plate through SOL and detached plasmas. The understanding of the ELM energy transport through SOL to the divertor target is rather poor at the moment, which leads to an ambiguous estimation of the deposited heat load on the divertor target in ITER. In the present work the ELM-like plasma heat pulse is generated by rf heating in a linear divertor plasma simulator. Energetic electrons with an energy range 10-40 eV are effectively generated by rf heating in low temperature plasmas with (T e )< ∼1 eV. It is observed experimentally that the energetic electrons ionize the highly excited Rydberg atoms quickly, bringing a rapid increase of the ion particle flux to the target, and make the detached plasmas attached to the target. Detailed physical processes about the interaction between the heat pulse with conduction and convection, and detached recombining plasmas are discussed

  14. Evaluation of the plasma parameters in COMPASS tokamak divertor area

    Czech Academy of Sciences Publication Activity Database

    Dimitrova, M.; Ivanova, P.; Kotseva, I.; Popov, Tsv.K.; Benova, E.; Bogdanov, T.; Stöckel, Jan; Dejarnac, Renaud

    2012-01-01

    Roč. 356, č. 1 (2012), s. 012007 ISSN 1742-6588. [InternationalSummerSchoolonVacuum,Electron, and IonTechnologies(VEIT2011)/17./. Sunny Beach, 19.09.2011-23.09.2011] Institutional research plan: CEZ:AV0Z20430508 Keywords : Plasma * tokamak * diagnostics * electric probe * magnetic-field * Langmuir probe * intermediate * pressures Subject RIV: BL - Plasma and Gas Discharge Physics http://iopscience.iop.org/1742-6596/356/1/012007/pdf/1742-6596_356_1_012007.pdf

  15. 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

  16. 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

  17. SOLPS-ITER Study of neutral leakage and drift effects on the alcator C-Mod divertor plasma

    Directory of Open Access Journals (Sweden)

    W. Dekeyser

    2017-08-01

    Full Text Available As part of an effort to validate the edge plasma model in the SOLPS-ITER code suite under ITER-relevant divertor plasma and neutral conditions, we report on progress in the modeling of the Alcator C-Mod divertor plasma with the new code. We perform simulations with a complete drifts model and kinetic neutrals, including effects of neutral viscosity, ion-molecule collisions and Lyα-opaque conditions, but assuming a pure deuterium plasma. Through a series of simulations with varying divertor geometries, we show the importance of including neutal leakage paths through the divertor substructure on the divertor plasma solution. Moreover, the impact of drifts on inner-outer target asymmetries is assessed. Including both effects, we achieve excellent agreement between simulations and upstream and outer target Langmuir Probe data. In absence of strong volumetric losses due to e.g. impurity radiation in our simulations, the strong inner target detachment observed experimentally remains elusive in our modeling at present.

  18. Characteristics of divertor plasma and scrape-off layer in JT-60U

    International Nuclear Information System (INIS)

    Itami, K.; Shimada, M.; Hosogane, N.

    1992-01-01

    Heat flux to the divertor is measured by thermography and the heat transport in the scrape-off layer is studied in beam heated discharges of JT-60U. The heat flux onto the divertor is ∝50% of total beam power at maximum. The in-out asymmetry of the heat flux P HEAT in /P HEAT out is as large as 20-40% when the ion grad-B drift is toward the divertor. Differences in P HEAT in /P HEA T out due to the direction of ion grad-B drift are as large as large as ∝40%. A scaling of the peaking factor Y of heat flux, defined by Y=2πRfq max /P HEAT , is obtained for beam heated discharges in JT-60U with a wide range of plasma parameters. The Y corresponds to the inverse of the thickness of the scrape-off layer. From a statistical analysis, it is found that the peaking factor Y of heat flux scales as P HEAT 0.49±0.18 anti n e -0.45±0.22 q eff -0.67±0.18 . (orig.)

  19. LHD helical divertor

    International Nuclear Information System (INIS)

    Ohyabu, N.; Watanabe, T.; Ji Hantao

    1993-07-01

    The Large Helical Device (LHD) now under construction is a heliotron/torsatron device with a closed divertor system. The edge LHD magnetic structure has been studied in detail. A peculiar feature of the configuration is existence of edge surface layers, a complicated three dimensional magnetic structure which does not, however, seem to hamper the expected divertor functions. Two divertor operational modes are being considered for the LHD experiment, high density, cold radiative divertor operation as a safe heat removal scheme and high temperature divertor plasma operation. In the latter operation, a divertor plasma with temperature of a few kev, generated by efficient pumping, expects to lead to significant improvement in core plasma confinement. Conceptual designs of the LHD divertor components are under way. (author)

  20. Modeling of thermal effects on TIBER II [Tokamak Ignition/Burn Experimental Reactor] divertor during plasma disruption

    International Nuclear Information System (INIS)

    Bruhn, M.L.; Perkins, L.J.

    1987-01-01

    Mapping the disruption power flow from the mid-plane of the TIBER Engineering Test Reactor to its divertor and calculating the resulting thermal effects are accomplished through the modification and coupling of three presently existing computer codes. The resulting computer code TADDPAK (Thermal Analysis Divertor during Disruption PAcKage) provides three-dimensional graphic presentations of time and positional dependent thermal effects on a poloidal cross section of the double-null-divertor configured reactor. These thermal effects include incident heat flux, surface temperature, vaporization rate, total vaporization, and melting depth. The dependence of these thermal effects on material choice, disruption pulse shape, and the characteristic thickness of the plasma scrape-off layer is determined through parametric analysis with TADDPAK. This computer code is designed to be a convenient, rapid, and user-friendly modeling tool which can be easily adapted to most tokamak double-null-divertor reactor designs. 14 refs

  1. Investigation of the influence of divertor recycling on global plasma confinement in JET ITER-like wall

    NARCIS (Netherlands)

    Tamain, P.; Joffrin, E.; Bufferand, H.; Jarvinen, A.; Brezinsek, S.; Ciraolo, G.; Delabie, E.; Frassinetti, L.; Giroud, C.; Groth, M.; Lipschultz, B.; Lomas, P.; Marsen, S.; Menmuir, S.; Oberkofler, M.; Stamp, M.; Wiesen, S.; JET-EFDA Contributors,

    2015-01-01

    Abstract The impact of the divertor geometry on global plasma confinement in type I ELMy H-mode has been investigated in the JET tokamak equipped with ITER-Like Wall. Discharges have been performed in which the position of the strike-points was changed while keeping the bulk plasma equilibrium

  2. Engineering solutions for components facing the plasma in experimental power reactors

    International Nuclear Information System (INIS)

    Casini, G.; Farfaletti-Casali, F.

    1985-01-01

    A review of the engineering problems related to the structures in front of the plasma of experimental Tokamak-type reactors is made. Attention is focused on the so-named ''first wall'', i.e. the wall side of the blanket segments facing the plasma, and on the collector plates of the impurity control system, in particular for the case of the single-null poloidal divertor. Even if the uncertainties related to the plasma-wall interaction are stil relevant, some engineering solutions which look manageable are identified and described. (orig.)

  3. Integrated ELM simulation with edge MHD stability and transport of SOL-divertor plasmas

    International Nuclear Information System (INIS)

    Hayashi, Nobuhiko; Takizuka, Tomonori; Aiba, Nobuyuki; Ozeki, Takahisa; Oyama, Naoyuki

    2007-07-01

    The effect of the pressure profile on the energy loss caused by edge localized modes (ELMs) has been investigated by using an integrated simulation code TOPICS-IB based on a core transport code with a stability code for the peeling-ballooning modes and a transport model for scrape-off-layer and divertor plasmas. The steep pressure gradient inside the pedestal top is found to broaden the region of the ELM enhanced transport through the broadening of eigenfunctions and enhance the ELM energy loss. The ELM energy loss in the simulation becomes larger than 15% of the pedestal energy, as is shown in the database of multi-machine experiments. (author)

  4. Numerical study of the ITER divertor plasma with the B2-EIRENE code package

    Energy Technology Data Exchange (ETDEWEB)

    Kotov, V.; Reiter, D. [Forschungszentrum Juelich (DE). Inst. fuer Energieforschung (IEF), Plasmaphysik (IEF-4); Kukushkin, A.S. [ITER International Team, Cadarache (France)

    2007-11-15

    The problem of plasma-wall interaction and impurity control is one of the remaining critical issues for development of an industrial energy source based on nuclear fusion of light isotopes. In this field sophisticated integrated numerical tools are widely used both for the analysis of current experiments and for predictions guiding future device design. The present work is dedicated to the numerical modelling of the edge plasma region in divertor configurations of large-scale tokamak fusion devices. A well established software tool for this kind of modelling is the B2-EIRENE code. It was originally developed for a relatively hot (>> 10 eV) ''high recycling divertor''. It did not take into account a number of physical effects which can be potentially important for ''detached conditions'' (cold, - several eV, - high density, - {approx} 10{sup 21} m{sup -3}, - plasma) typical for large tokamak devices. This is especially critical for the modelling of the divertor plasma of ITER: an international project of an experimental tokamak fusion reactor to be built in Cadarache, France by 2016. This present work is devoted to a major upgrade of the B2-EIRENE package, which is routinely used for ITER modelling, essentially with a significantly revised version of EIRENE: the Monte-Carlo neutral transport code. The main part of the thesis address three major groups of the new physical effects which have been added to the model in frame of this work: the neutral-neutral collisions, the up-to date hydrogen molecular reaction kinetics and the line radiation transport. The impact of the each stage of the upgrade on the self-consistent (between plasma, the neutral gas and the radiation field) solution for the reference ITER case is analysed. The strongest effect is found to be due to the revised molecular collision kinetics, in particular due to hitherto neglected elastic collisions of hydrogen molecules with ions. The newly added non

  5. Numerical study of the ITER divertor plasma with the B2-EIRENE code package

    International Nuclear Information System (INIS)

    Kotov, V.; Reiter, D.; Kukushkin, A.S.

    2007-11-01

    The problem of plasma-wall interaction and impurity control is one of the remaining critical issues for development of an industrial energy source based on nuclear fusion of light isotopes. In this field sophisticated integrated numerical tools are widely used both for the analysis of current experiments and for predictions guiding future device design. The present work is dedicated to the numerical modelling of the edge plasma region in divertor configurations of large-scale tokamak fusion devices. A well established software tool for this kind of modelling is the B2-EIRENE code. It was originally developed for a relatively hot (>> 10 eV) ''high recycling divertor''. It did not take into account a number of physical effects which can be potentially important for ''detached conditions'' (cold, - several eV, - high density, - ∼ 10 21 m -3 , - plasma) typical for large tokamak devices. This is especially critical for the modelling of the divertor plasma of ITER: an international project of an experimental tokamak fusion reactor to be built in Cadarache, France by 2016. This present work is devoted to a major upgrade of the B2-EIRENE package, which is routinely used for ITER modelling, essentially with a significantly revised version of EIRENE: the Monte-Carlo neutral transport code. The main part of the thesis address three major groups of the new physical effects which have been added to the model in frame of this work: the neutral-neutral collisions, the up-to date hydrogen molecular reaction kinetics and the line radiation transport. The impact of the each stage of the upgrade on the self-consistent (between plasma, the neutral gas and the radiation field) solution for the reference ITER case is analysed. The strongest effect is found to be due to the revised molecular collision kinetics, in particular due to hitherto neglected elastic collisions of hydrogen molecules with ions. The newly added non-linear effects (neutral-neutral collisions, radiation opacity

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

    Science.gov (United States)

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

    2011-10-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  8. Effect of transport on MAR in detached divertor plasma

    International Nuclear Information System (INIS)

    Miyamoto, Kenji; Hatayama, A.; Ishii, Y.; Miyamoto, T.; Fukano, A.

    2003-01-01

    The effect of H 2 transport on the onset of MAR in the relatively lower plasma parameter regime of a detached state (n e =1x10 19 m -3 , T e =1 eV) is investigated theoretically. The vibrationally excited molecular densities and the degree of MAR are evaluated by using a 1-D Monte Carlo method (with transport effect), and by solving time-dependent 0-D rate equations without the transport term (without transport effect), respectively. It is found that the degree of MAR with transport is smaller than that without transport under the same H 2 flow rate. Especially, the degree of MAR is negligible near the gas inlet. This smaller degree of MAR with transport is due to the lack of highly excited vibrational molecules which contribute to MAR. The hydrogen molecular density available for MAR is determined by the external hydrogen molecular source and the outflow due to transport, i.e., a 'net' confinement time

  9. 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.

  10. Simulation of dense recombining divertor plasmas with a Navier endash Stokes neutral transport model

    International Nuclear Information System (INIS)

    Knoll, D.A.; McHugh, P.R.; Krasheninnikov, S.I.; Sigmar, D.J.

    1996-01-01

    A two-dimensional combined edge plasma Navier endash Stokes neutral transport model is presented for the simulation of dense recombining divertor plasmas. This model includes ions, electrons, and neutral atoms which undergo Coulomb collisions, electron impact ionization, ion endash neutral elastic collisions, three-body and radiative recombination, and neutral endash neutral collisions. The advanced fully implicit solution algorithm is briefly described and a variety of results on a model geometry are presented. It is shown that interesting neutral flow patterns can exist and that these flows can convect significant energy. A solution that ignores neutral endash neutral collisions is shown to be quantitatively different from one that includes neutral endash neutral collisions. Solutions are also shown to be sensitive to the plasma opacity for Lyman α radiation. copyright 1996 American Institute of Physics

  11. Characterization of fueling NSTX H-mode plasmas diverted to a liquid lithium divertor

    Energy Technology Data Exchange (ETDEWEB)

    Kaita, R., E-mail: kaita@pppl.gov [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Kugel, H.W.; Abrams, T. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Allain, J.P. [Purdue University, West Lafayette, IN 47907 (United States); Bell, M.G.; Bell, R.E.; Diallo, A.; Gerhardt, S.P. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Heim, B. [Purdue University, West Lafayette, IN 47907 (United States); Jaworski, M.A., E-mail: mjaworsk@pppl.gov [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Kallman, J. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Kaye, S.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Maingi, R. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Mansfield, D. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); McLean, A. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Menard, J.; Mueller, D. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Nygren, R. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Ono, M. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); and others

    2013-07-15

    Deuterium fueling experiments were conducted with the NSTX Liquid Lithium Divertor (LLD). Lithium evaporation recoated the LLD surface to approximate flowing liquid Li to sustain D retention. In the first experiment with the diverted outer strike point on the LLD, the difference between the applied D gas input and the plasma D content reached very high values without disrupting the plasma, as would normally occur in the absence of Li pumping, and there was also little change in plasma D content. In the second experiment, constant fueling was applied, as the LLD temperature was varied to change the surface from solid to liquid. The D retention was relatively constant, and about the same as that for solid Li coatings on graphite, or twice that achieved without Li PFC coatings. Contamination of the LLD surface was also possible due to compound formation and erosion and redeposition from carbon PFCs.

  12. 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.)

  13. 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.

  14. 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

  15. Interaction of ICRF power and edge plasma in Tore Supra ergodic divertor configuration

    International Nuclear Information System (INIS)

    Nguyen, F.; Grosman, A.; Basiuk, V.; Fraboulet, D.; Beaumont, B.; Becoulet, A.; Ghendrih, Ph.; Ladurelle, L.; Meslin, B.

    2000-01-01

    The coupling of ICRF power to plasma is a crucial problem in Tore Supra for high power and long pulse operations and depends greatly on the edge parameters, in particular on the edge density. Conversely, the behaviour of the bulk plasma is related to the edge conditions and the injection of RF power also induces major modifications on the edge plasma. Moreover, the Ergodic Divertor (ED) of Tore Supra imposes a complex configuration at the edge due to the presence of the magnetic perturbation. Several diagnostics are available to study the interaction of ICRF power with the edge plasma: Langmuir probes on the ED modules, infra red (IR) cameras, charge exchange neutral analysers. In minority heating scheme, the edge density is very sensitive to any perturbation in the high recycling regime which is always found in the ED configuration for relevant plasma parameters. Partially detached regimes, with or without inhomogeneities of density and temperature induced by the flux tubes of the laminar layer, are obtained for high resistance coupling values. The coupling is then not very robust and feedback control or antenna automatic matching techniques are developed. In fast wave electron heating scheme with ED, various fast wave absorption mechanisms (minority heating, Mode Conversion, Alfven resonance) are present at the plasma edge due to the large size of the plasma. The ICRF coupling is difficult due to the low fast wave direct electron damping, even with high hydrogen minority scheme. An increase of the injected ICRF power could improve this situation

  16. Response to “Comment on ‘Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake’ ” [Phys. Plasmas 21, 054701 (2014)

    International Nuclear Information System (INIS)

    Kotschenreuther, Mike; Valanju, Prashant; Covele, Brent; Mahajan, Swadesh

    2014-01-01

    Relying on coil positions relative to the plasma, the “Comment on ‘Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake’ ” [Phys. Plasmas 21, 054701 (2014)], emphasizes a criterion for divertor characterization that was critiqued to be ill posed [M. Kotschenreuther et al., Phys. Plasmas 20, 102507 (2013)]. We find that no substantive physical differences flow from this criteria. However, using these criteria, the successful NSTX experiment by Ryutov et al. [Phys. Plasmas 21, 054701 (2014)] has the coil configuration of an X-divertor (XD), rather than a snowflake (SF). On completing the divertor index (DI) versus distance graph for this NSTX shot (which had an inexplicably missing region), we find that the DI is like an XD for most of the outboard wetted divertor plate. Further, the “proximity condition,” used to define an SF [M. Kotschenreuther et al., Phys. Plasmas 20, 102507 (2013)], does not have a substantive physics basis to override metrics based on flux expansion and line length. Finally, if the criteria of the comment are important, then the results of NSTX-like experiments could have questionable applicability to reactors

  17. Response to “Comment on ‘Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake’ ” [Phys. Plasmas 21, 054701 (2014)

    Energy Technology Data Exchange (ETDEWEB)

    Kotschenreuther, Mike; Valanju, Prashant; Covele, Brent; Mahajan, Swadesh [Institute for Fusion Studies, The University of Texas at Austin, Austin, Texas 78712 (United States)

    2014-05-15

    Relying on coil positions relative to the plasma, the “Comment on ‘Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake’ ” [Phys. Plasmas 21, 054701 (2014)], emphasizes a criterion for divertor characterization that was critiqued to be ill posed [M. Kotschenreuther et al., Phys. Plasmas 20, 102507 (2013)]. We find that no substantive physical differences flow from this criteria. However, using these criteria, the successful NSTX experiment by Ryutov et al. [Phys. Plasmas 21, 054701 (2014)] has the coil configuration of an X-divertor (XD), rather than a snowflake (SF). On completing the divertor index (DI) versus distance graph for this NSTX shot (which had an inexplicably missing region), we find that the DI is like an XD for most of the outboard wetted divertor plate. Further, the “proximity condition,” used to define an SF [M. Kotschenreuther et al., Phys. Plasmas 20, 102507 (2013)], does not have a substantive physics basis to override metrics based on flux expansion and line length. Finally, if the criteria of the comment are important, then the results of NSTX-like experiments could have questionable applicability to reactors.

  18. Bursty fluctuation characteristics in SOL/divertor plasmas of large helical device

    International Nuclear Information System (INIS)

    Ohno, N.

    2006-01-01

    Full text: Fluctuation properties in the SOL plasmas were intensively studied to understand the crossfield plasma transport, which determines the SOL structure and heat/particle deposition onto the first wall. Recent studies in tokamaks showed that the SOL density fluctuation is highly intermittent. Convective cross-field transport associated with the intermittent events would have strong influence on recycling processes and impurity generation from the first wall. On the other hand, in helical devices, there are few systematic studies on the SOL fluctuation property focusing on the intermittent bursty fluctuations related to plasma blob transport. Recent theory predicts that the blobs propagate toward a low field side in tokamaks. On the other hand, in the Large Helical Device (LHD), the direction of the gradient in B is not uniform because the high-field and the low-field sides rotates poloidally along the torus in the helical system. Comparison between the intermittent bursty fluctuations in the edge plasma of tokamaks and helical devices makes it possible to understand the essential physics of the blob transport. Recently, fast camera observation showed the radial motion of filaments in the edge of the LHD, suggesting the convective cross-field transport. In this paper, bursty fluctuation properties in the edge of the LHD have been investigated by analyzing the ion saturation currents measured with a probe array embedded in an outboard divertor plate. Statistical analysis based on probability distribution function was employed to determine the intermittent evens in the density fluctuation. Large positive bursty events were often observed in the ion saturation current measured with a divertor probe near a divertor leg at which the magnetic line of force connected to the area of a low-field side with a short connection length. Condition averaging result of the positive bursty events indicates the intermittent feature with a rapid increase and a slow decay is

  19. Three-dimensional modeling of plasma edge transport and divertor fluxes during application of resonant magnetic perturbations on ITER

    Czech Academy of Sciences Publication Activity Database

    Schmitz, O.; Becoulet, M.; Cahyna, Pavel; Evans, T.E.; Feng, Y.; Frerichs, H.; Loarte, A.; Pitts, R.A.; Reiser, D.; Fenstermacher, M.E.; Harting, D.; Kirschner, A.; Kukushkin, A.; Lunt, T.; Saibene, G.; Reiter, D.; Samm, U.; Wiesen, S.

    2016-01-01

    Roč. 56, č. 6 (2016), č. článku 066008. ISSN 0029-5515 Institutional support: RVO:61389021 Keywords : resonant magnetic perturbations * plasma edge physics * 3D modeling * neutral particle physics * ITER * divertor heat and particle loads * ELM control Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 3.307, year: 2016 http://iopscience.iop.org/article/10.1088/0029-5515/56/6/066008/meta

  20. Experimental study on highly collisional edge plasmas in W7-AS island divertor configurations

    International Nuclear Information System (INIS)

    Grigull, P.; Hildebrandt, D.; Sardei, F.; Feng, Y.; Herre, G.; Herrmann, A.; Hofmann, J.V.; Kisslinger, J.; Kuehner, G.; Niedermeyer, H.; Schneider, R.; Verbeek, H.; Wagner, F.; Wolf, R.; Zhang, X.D.

    1997-01-01

    Edge plasma scenarios in island divertor configurations ('natural' magnetic islands intersected by targets) are studied by comparing data from moderate to high density NBI discharges with 3D code (EMC3/EIRENE) results. The data strongly indicate that high recycling with significant particle flux enhancement was achieved in this geometry. But, plasma pressure losses towards the targets are relatively strong, and high recycling sets in only at n e >10 20 m -3 . The respective density enhancement in front of the targets is moderate (up to a factor of about three relative to the upstream density). These scenarios are also in basic agreement with B2/EIRENE code predictions. At n e >1.5 x 10 20 m -3 detachment seems to develop. Improvements are expected from additional coils controlling the field line pitch inside the islands, and from optimized targets which will better focus recycling neutrals into the islands. Both are in preparation. (orig.)

  1. Current profile control and magnetohydrodynamic stability in Tore Supra discharges with edge-plasma control by the ergodic divertor

    International Nuclear Information System (INIS)

    Zabiego, M.; Friant, C.; Ghendrih, P.; Becoulet, M.; Bucalossi, J.; Saint-Laurent, F.

    1999-01-01

    Although ergodic divertors are primarily designed to control particle and heat fluxes at the plasma edge, they also happen to affect the MHD stability of tokamak discharges. On Tore Supra, the ergodic divertor has long been known to stabilize the m/n=2/1 tearing mode induced, for instance, by edge radiation and detachment processes, thus allowing safe high-current and high-density operations. More recently, though, in discharges where ergodic divertor operations were optimised relative to the control of the edge-plasma (i.e., with large divertor perturbation), a detrimental increase in the disruptiveness has been observed. The action that the ergodic divertor has on the MHD activity is interpreted in terms of a redistribution of the current profile. The latter results from a large increase in the edge resistivity, primarily induced by the degradation of the electron energy confinement in the ergodic layer. The possibility that a transport barrier develops in the vicinity of the separatrix strongly affects the considered modelling. (authors)

  2. The effect of feedback-controlled divertor nitrogen seeding on the boundary plasma and power exhaust channel width in Alcator C-Mod

    Science.gov (United States)

    Labombard, B.; Brunner, D.; Kuang, A. Q.; McCarthy, W.; Terry, J. L.

    2017-10-01

    The scrape-off layer (SOL) power channel width, λq, is projected to be 0.5 mm in power reactors, based on multi-machine measurements of divertor target heat fluxes in H-mode at low levels of divertor dissipation. An important question is: does λq change with the level of divertor dissipation? We report results in which feedback controlled nitrogen seeding in the divertor was used to systematically vary divertor dissipation in a series of otherwise identical L-mode plasmas at three plasma currents: 0.55, 0.8 and 1.1 MA. Outer midplane profiles were recorded with a scanning Mirror Langmuir Probe; divertor plasma conditions were monitored with `rail' Langmuir probe and surface thermocouple arrays. Despite an order of magnitude reduction in divertor target heat fluxes (q// 400 MW m-2 to 40 MW m-2) and corresponding change in divertor regime from sheath-limited through high-recycling to near-detached, the upstream electron temperature profile is found to remain unchanged or to become slightly steeper in the near SOL and to drop significantly in the far SOL. Thus heat in the SOL appears to take advantage of this impurity radiation `heat sink' in the divertor by preferentially draining via the narrow (and perhaps an increasingly narrow) λq of the near SOL. Supported by USDoE award DE-FC02-99ER54512.

  3. Plasma characteristics of the end-cell of the GAMMA 10 tandem mirror for the divertor simulation experiment

    International Nuclear Information System (INIS)

    Nakashima, Y.; Sakamoto, M.; Yoshikawa, M.; Takeda, H.; Ichimura, K.; Hosoi, K.; Hirata, M.; Ichimura, M.; Ikezoe, R.; Imai, T.; Kariya, T.; Katanuma, I.; Kohagura, J.; Minami, R.; Numakura, T.; Oki, K.; Ueda, H.; Asakura, Nobuyuki; Furuta, T.; Hatayama, A.; Toma, M.; Hirooka, Y.; Masuzaki, S.; Sagara, A.; Shoji, M.; Kado, S.; Matsuura, H.; Nagata, S.; Nishino, N.; Ohno, N.; Tonegawa, A.; Ueda, Y.

    2012-11-01

    In this paper, detailed characteristics and controllability of plasmas emitted from the end-cell of the GAMMA 10 tandem mirror are described from the viewpoint of divertor simulation studies. The energy analysis of ion flux by using end-loss ion energy analyzer (ELIEA) proved that the obtained high ion temperature (100 - 400 eV) was comparable to SOL plasma parameters in toroidal devices and was controlled by changing the ICRF power. Parallel ion temperature T i∥ determined from the probe and calorimeter shows a linear relationship with the ICRF power in the central-cell and agrees with the results of ELIEA. Additional ICRF heating revealed a significant enhancement of particle flux, which indicated an effectiveness of additional plasma heating in adjacent cells toward the improvement of the performance. Superimposing the ECH pulse of 380 kW, 5 ms attained the maximum heat-flux more than 10 MW/m 2 on axis. This value comes up to the heat-load of the divertor plate of ITER, which gives a clear prospect of generating the required heat density for divertor studies by building up heating systems to the end-mirror cell. Initial results of plasma irradiation experiment and construction of new divertor module are also described. (author)

  4. 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

  5. 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

  6. 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)

  7. 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.

  8. Modeling of combined effects of divertor closure and advanced magnetic configuration on detachment in DIII-D by SOLPS

    Science.gov (United States)

    Si, H.; Guo, H. Y.; Covele, B.; Leonard, A. W.; Watkins, J. G.; Thomas, D.; Ding, R.

    2018-05-01

    One of the major challenges facing the design and operation of next-step high-power steady-state fusion devices is to develop a divertor solution for handling power exhaust, while ensuring acceptable divertor target plate erosion, which necessitates access to divertor detachment at relative low main plasma densities compatible with current drive and high plasma confinement. Detailed modeling with SOLPS is carried out to examine the effect of divertor closure on detachment with the normal single null divertor (SD) configuration, as well as one of the advanced divertor configurations, such as x-divertor (XD) respectively. The SOLPS modeling for a high confinement plasma in DIII-D finds that increasing divertor closure with SD reduces the upstream separatrix density at the onset of detachment from 1.18× {{10}19} {{m}-3} to 0.88× {{10}19} {{m}-3} . Moreover, coupling the divertor closure with XD further promotes the onset of divertor detachment at a still lower upstream separatrix density, down to the value of 0.67× {{10}19} {{m}-3} , thus, showing that divertor closure and advanced magnetic configuration can work synergistically to facilitate divertor detachment.

  9. The ITER Divertor Cassette Project meeting

    International Nuclear Information System (INIS)

    Akiba, M.; Tivey, R.

    2000-01-01

    The Divertor Cassette Project topical meeting took place on April 5-7, 2000 at the JAERI Naka site in Japan. The meeting focused on the progress made by the three parties under task agreements on the development of carbon-fibre composite and tungsten armored high flux plasma-facing components

  10. Innovative divertor concepts for LHD

    International Nuclear Information System (INIS)

    Ohyabu, N.; Komori, A.; Akaishi, K.

    1994-07-01

    We are developing various innovative divertor concepts which improve the LHD plasma performance. These are two divertor magnetic geometries (helical and local island divertors), three operational scenarios (radiative cooling in the high density, cold boundary, confinement improvement by generating high temperature divertor plasma and simultaneous achievement of radiative cooling and H-mode like confinement improvement) and technological development of new efficient hydrogen pumping schemes. (author)

  11. 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)

  12. Progress towards RF heated steady-state plasma operations on LHD by employing ICRF heating methods and improved divertor plates

    International Nuclear Information System (INIS)

    Kumazawa, R.; Mutoh, T.; Saito, K.

    2008-10-01

    A long pulse plasma discharge experiment was carried out using RF heating power in the Large Helical Device (LHD), a currentless magnetic confining system. Progress in long pulse operation is summarized since the 10th experimental campaign (2006). A scaling relation of the plasma duration time to the applied RF power has been derived from the experimental data so far collected. It indicates that there exists a critical divertor temperature and consequently a critical RF heating power P RFcrit =0.65 MW. The area on the graph of the duration time versus the RF heating power was extended over the scaling relation by replacing divertor plates with new ones with better heat conductivity. The cause of the plasma collapse at the end of the long pulse operation was found to be the penetration of metal impurities. Many thin flakes consisting of heavy metals and graphite in stratified layers were found on the divertor plates and it was thought that they were the cause of impurity metals penetrating into the plasma. In a simulation involving injecting a graphite-coated Fe pellet to the plasma it was found that 230 Eμm in the diameter of the Fe pellet sphere was the critical size which led the plasma to collapse. A mode-conversion heating method was examined in place of the minority ICRF heating which has been employed in almost all the long-pulse plasma discharges. It was found that this method was much better from the viewpoint of achieving uniformity of the plasma heat load to the divertors. It is expected that P RFcrit will be increased by using the mode-conversion heating method. (author)

  13. 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

  14. Local magnetic divertor for control of the plasma--limiter interaction in a tokamak

    International Nuclear Information System (INIS)

    Zweben, S.J.; Liewer, P.C.; Gould, R.W.

    1984-01-01

    An experiment is described in which plasma flow to a tokamak limiter is controlled through the use of a local toroidal divertor coil mounted inside the limiter itself. This coil produces a local perturbed field B/sub C/ approximately equal to the local unperturbed toroidal field B/sub T/approx. =3 kG, such that when B/sub C/ adds to B/sub T/ the field lines move into the limiter and the local plasma flow to it increases by a factor as great as 1.6, and when B/sub C/ subtracts from B/sub T/ the field lines move away from the limiter and the local plasma flow to it decreases by as much as a factor of 4. A simple theoretical model is used to interpret these results. Since these changes occur without significantly affecting global plasma confinement, such a control scheme may be useful for optimizing the performance of pumped limiters

  15. Characterization of the island divertor plasma of W7-AS stellarator in the deeply detached state with volume recombination

    International Nuclear Information System (INIS)

    Ramasubramanian, N.; Koenig, R.; Feng, Y.; Giannone, L.; Grigull, P.; Klinger, T.; McCormick, K.; Thomsen, H.; Wenzel, U.

    2004-01-01

    In the high-density H-mode of the Stellarator Wendelstein 7-AS, the plasma detaches from the island divertor targets when the line-averaged density exceeds a critical value. This quasi-stationary detachment is found to be partial and shows edge-localized, poloidally asymmetric radiation. The spectroscopic characteristics of the deeply detached plasma are reported, including evidence for volume recombination. The detached plasma radiates up to 90% of the absorbed power with larger contributions from the locations close to magnetic X-points outside the divertor region. The spectral analysis of the Balmer series indicate very high densities and low temperatures at the detached regions. The results of the spectral analysis underline the importance of three-dimensional modelling. An initial comparison is made with the latest results from EMC3-EIRENE modelling. (author)

  16. 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.

  17. 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

  18. 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

  19. Status of National Spherical Torus Experiment Liquid Lithium Divertor

    Science.gov (United States)

    Kugel, H. W.; Viola, M.; Ellis, R.; Bell, M.; Gerhardt, S.; Kaita, R.; Kallman, J.; Majeski, R.; Mansfield, D.; Roquemore, A. L.; Schneider, H.; Timberlake, J.; Zakharov, L.; Nygren, R. E.; Allain, J. P.; Maingi, R.; Soukhanovskii, V.

    2009-11-01

    Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. The next step in this work is the 2009 installation of a Liquid Lithium Divertor (LLD). The 20 cm wide LLD located on the lower outer divertor, consists of four, 80 degree sections; each section is separated by a row of graphite diagnostic tiles. The temperature controlled LLD structure consists of a 0.01cm layer of vacuum flame-sprayed, 50 percent porous molybdenum, on top of 0.02 cm, 316-SS brazed to a 1.9 cm Cu base. The physics design of the LLD encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

  20. Plans of LHD divertor experiment

    International Nuclear Information System (INIS)

    Ohyabu, Nobuyoshi; Komori, Akio; Sagara, Akio; Noda, Nobuaki; Motojima, Osamu

    1996-01-01

    Scenarios of the LHD divertor experiment are presented. In the LHD divertor experimental program, various innovative divertor concepts and technologies, developed during its design phase will be utilized to improve the plasma performance. Two divertor operational scenarios (confinement improvement by generating high temperature divertor plasma and simultaneous achievement of radiative cooling and H-mode-like confinement improvement) are among them. Local island divertor geometry has also been proposed. This new divertor has been successfully tested in the CHS device and is planned to be installed in the LHD device. In addition, technological development of new efficient hydrogen pumping schemes (carbon sheet pump and membrane pump) are being pursued for enhancement of the divertor control capability. (author)

  1. The symmetric quartic map for trajectories of magnetic field lines in elongated divertor tokamak plasmas

    Science.gov (United States)

    Jones, Morgin; Wadi, Hasina; Ali, Halima; Punjabi, Alkesh

    2009-04-01

    The coordinates of the area-preserving map equations for integration of magnetic field line trajectories in divertor tokamaks can be any coordinates for which a transformation to (ψt,θ,φ) coordinates exists [A. Punjabi, H. Ali, T. Evans, and A. Boozer, Phys. Lett. A 364, 140 (2007)]. ψt is toroidal magnetic flux, θ is poloidal angle, and φ is toroidal angle. This freedom is exploited to construct the symmetric quartic map such that the only parameter that determines magnetic geometry is the elongation of the separatrix surface. The poloidal flux inside the separatrix, the safety factor as a function of normalized minor radius, and the magnetic perturbation from the symplectic discretization are all held constant, and only the elongation is κ varied. The width of stochastic layer, the area, and the fractal dimension of the magnetic footprint and the average radial diffusion coefficient of magnetic field lines from the stochastic layer; and how these quantities scale with κ is calculated. The symmetric quartic map gives the correct scalings which are consistent with the scalings of coordinates with κ. The effects of m =1, n =±1 internal perturbation with the amplitude that is expected to occur in tokamaks are calculated by adding a term [H. Ali, A. Punjabi, A. H. Boozer, and T. Evans, Phys. Plasmas 11, 1908 (2004)] to the symmetric quartic map. In this case, the width of stochastic layer scales as 0.35 power of κ. The area of the footprint is roughly constant. The average radial diffusion coefficient of field lines near the X-point scales linearly with κ. The low mn perturbation changes the quasisymmetric structure of the footprint, and reorganizes it into a single, large scale, asymmetric structure. The symmetric quartic map is combined with the dipole map [A. Punjabi, H. Ali, and A. H. Boozer, Phys. Plasmas 10, 3992 (2003)] to calculate the effects of magnetic perturbation from a current carrying coil. The coil position and coil current coil are

  2. The symmetric quartic map for trajectories of magnetic field lines in elongated divertor tokamak plasmas

    International Nuclear Information System (INIS)

    Jones, Morgin; Wadi, Hasina; Ali, Halima; Punjabi, Alkesh

    2009-01-01

    The coordinates of the area-preserving map equations for integration of magnetic field line trajectories in divertor tokamaks can be any coordinates for which a transformation to (ψ t ,θ,φ) coordinates exists [A. Punjabi, H. Ali, T. Evans, and A. Boozer, Phys. Lett. A 364, 140 (2007)]. ψ t is toroidal magnetic flux, θ is poloidal angle, and φ is toroidal angle. This freedom is exploited to construct the symmetric quartic map such that the only parameter that determines magnetic geometry is the elongation of the separatrix surface. The poloidal flux inside the separatrix, the safety factor as a function of normalized minor radius, and the magnetic perturbation from the symplectic discretization are all held constant, and only the elongation is κ varied. The width of stochastic layer, the area, and the fractal dimension of the magnetic footprint and the average radial diffusion coefficient of magnetic field lines from the stochastic layer; and how these quantities scale with κ is calculated. The symmetric quartic map gives the correct scalings which are consistent with the scalings of coordinates with κ. The effects of m=1, n=±1 internal perturbation with the amplitude that is expected to occur in tokamaks are calculated by adding a term [H. Ali, A. Punjabi, A. H. Boozer, and T. Evans, Phys. Plasmas 11, 1908 (2004)] to the symmetric quartic map. In this case, the width of stochastic layer scales as 0.35 power of κ. The area of the footprint is roughly constant. The average radial diffusion coefficient of field lines near the X-point scales linearly with κ. The low mn perturbation changes the quasisymmetric structure of the footprint, and reorganizes it into a single, large scale, asymmetric structure. The symmetric quartic map is combined with the dipole map [A. Punjabi, H. Ali, and A. H. Boozer, Phys. Plasmas 10, 3992 (2003)] to calculate the effects of magnetic perturbation from a current carrying coil. The coil position and coil current coil are

  3. Innovations in the LHD divertor program

    International Nuclear Information System (INIS)

    Ohyabu, N.; Komori, A.; Noda, N.; Morisaki, T.; Sagara, A.; Suzuki, H.; Watanabe, T.; Motojima, O.; Takase, H.

    1995-01-01

    Various innovative divertor concepts have been developed to improve the LHD plasma performance. They are two divertor magnetic geometries (helical divertor configurations with and without n/m=1/1 island) and two operational scenarios (confinement improvement by generating high temperature divertor plasma and simultaneous achievement of radiative cooling and H-mode-like confinement improvement). In addition, technological development of new efficient hydrogen pumping schemes are being pursued for enhancing the divertor control capability. 16 refs., 4 figs

  4. 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.

  5. Transport in the plasma edge specific connection to the wall in the Tore Supra ergodic divertor experiments

    International Nuclear Information System (INIS)

    Grosman, A.; Ghendrih, P.; DeMichelis, C.; Monier-Garbet, P.; Vallet, J.C.; Capes, H.; Chatelier, M.; Geraud, A.; Goniche, M.; Grisolia, C.; Guilhem, D.; Harris, G.; Hess, W.; Nguyen, F.; Poutchy, L.; Samain, A.

    1992-01-01

    The ergodic divertor experiments in TORE SUPRA can be analysed along two main lines. The first one refers to the change of the heat and particle transport in the ergodized zone. This is especially true for the electron heat transport which is enhanced in the edge layer. But other distinctive features give evidence of the importance of the parallel connexion length between the plasma edge and the wall. The field lines, which are stochastic in the major part of the perturbed layer (10-15 cm) are such that, in the outermost layer (3 cm), the connexion topology is regular. This has obvious effects on the particle and power deposition, but also on the plasma parameters, and consequently influences the particle recycling and impurity shielding processes. The TORE SUPRA ergodic divertor experiments are reviewed in this framework

  6. Modelling of Plasma Response to Resonant Magnetic Perturbations and its Influence on Divertor Strike Points

    Energy Technology Data Exchange (ETDEWEB)

    Cahyna, P.; Peterka, M.; Panek, R., E-mail: cahyna@ipp.cas.cz [Institute of Plasma Physics AS CR, Prague (Czech Republic); Liu, Y.; Kirk, A.; Harrison, J.; Thornton, A.; Chapman, I. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon (United Kingdom); Nardon, E. [Association Euratom/CEA, CEA Cadarache, St. Paul-lez-Durance (France); Schmitz, O. [Forschung Zentrum Juelich, Juelich (Germany)

    2012-09-15

    Full text: Resonant magnetic perturbations (RMPs) for edge localized mode (ELM) mitigation in tokamaks can be modified by the plasma response and indeed strong screening of the applied perturbation is in some cases predicted by simulations. In this contribution we investigate what effect would such screening have on the spiralling patterns (footprints) which may appear at the divertor when RMPs are applied. We use two theoretical tools for investigation of the impact of plasma response on footprints: a simple model of the assumed screening currents, which can be used to translate the screening predicted by MHD codes in a simplified geometry into the real geometry, and the MHD code MARS-F. The former consistently predicts that footprints are significantly reduced when complete screening of the resonant perturbation modes (like it is the case in ideal MHD) is assumed. This result is supported by the result of MARS-F in ideal mode for the case of the MAST tokamak. To predict observed patterns of fluxes it is necessary to take into account the deformation of the scrape-off layer, and for this we developed an approximative method based on the Melnikov integral. If the screening of perturbations indeed reduces the footprints, it would provide us with an important tool to evaluate the amount of screening in experiments, as the footprints can be easily observed. We thus present a comparison between predictions and experimental data, especially for the MAST tokamak, where a significant amount of data has been collected. (author)

  7. Imitation of deuterium plasma interaction with the surface of carbon materials in gaseous divertor conditions

    Energy Technology Data Exchange (ETDEWEB)

    Korshunov, S.N. E-mail: sinet@nfi.kiae.ru; Guseva, M.I.; Gureev, V.M.; Danelyan, L.S.; Khripunov, B.I.; Kolbasov, B.N.; Kulikauskas, V.S.; Litnovsky, A.M.; Martynenko, Yu.V.; Petrov, V.B.; Zatekin, V.V

    2003-03-01

    The experiments on simulation of gas divertor conditions were done in the LENTA facility under interaction of a plasma flow with neutral gas. The samples of carbon materials were exposed in a steady-state deuterium plasma (ion energy 5 eV, ion flux 5x10{sup 21} m{sup -2} s{sup -1}, fluence 10{sup 26} m{sup -2}) at 1470 K (MPG-8) and at 1320 K (SEP NB31). Heavy deuterocarbon molecules (C{sub 2}D{sub 2}, C{sub 2}D{sub 4}, C{sub 2}D{sub 6}) were observed in mass spectra of the discharge. This fact and high erosion yields show the presence of chemical erosion. Deuterium accumulation in carbon materials was studied by elastic recoil detection analysis. The integral deuterium content is 6x10{sup 18} m{sup -2} in SEP NB31 and 1.95x10{sup 19} m{sup -2} in MPG-8. The profiles of C and Mo atom distributions in deposited layer on Mo collector is 'X'-like. Carbon atoms distribution in deposited layer on Si is uniform. The integral deuterium content in co-deposited layers is 1.4x10{sup 21} m{sup -2} on Si and 4.8x10{sup 20} m{sup -2} on Mo. A globular structure of co-deposited layer on Mo collector was found.

  8. Be I and Be II spectroscopy in divertor plasma relevant conditions

    Science.gov (United States)

    Nishijima, D.; Doerner, R. P.; Seraydarian, R. P.

    2013-07-01

    Intensity ratios of various Be I and Be II lines measured in Be-seeded D and He plasmas in the PISCES-B linear divertor plasma simulator are compared with the corresponding ratios of the photon emissivity coefficient, PEC, calculated by ADAS. Agreement of measured intensity ratios with calculated PEC ratios is satisfactory within a factor of ˜2 for both Be I and Be II. It is proposed that a Be I line ratio of 234.8 nm/265.0 nm and a Be II line ratio of 467.3 nm/313.1 nm can be used to estimate the electron temperature, while a 265.0 nm/332.1 nm Be I line ratio is sensitive to the electron density. Further, S/XB values of a Be I line at 457.3 nm were experimentally determined from a ratio of the sputtered Be flux to the emission intensity. Measured values are systematically lower than calculated ADAS values, which may be explained by the increased sputtering yield of redeposited Be atoms.

  9. Be I and Be II spectroscopy in divertor plasma relevant conditions

    Energy Technology Data Exchange (ETDEWEB)

    Nishijima, D., E-mail: dnishijima@ferp.ucsd.edu [Center for Energy Research, University of California at San Diego, 9500 Gilman Dr., La Jolla, CA 92093-0417 (United States); Doerner, R.P.; Seraydarian, R.P. [Center for Energy Research, University of California at San Diego, 9500 Gilman Dr., La Jolla, CA 92093-0417 (United States)

    2013-07-15

    Intensity ratios of various Be I and Be II lines measured in Be-seeded D and He plasmas in the PISCES-B linear divertor plasma simulator are compared with the corresponding ratios of the photon emissivity coefficient, PEC, calculated by ADAS. Agreement of measured intensity ratios with calculated PEC ratios is satisfactory within a factor of ∼2 for both Be I and Be II. It is proposed that a Be I line ratio of 234.8 nm/265.0 nm and a Be II line ratio of 467.3 nm/313.1 nm can be used to estimate the electron temperature, while a 265.0 nm/332.1 nm Be I line ratio is sensitive to the electron density. Further, S/XB values of a Be I line at 457.3 nm were experimentally determined from a ratio of the sputtered Be flux to the emission intensity. Measured values are systematically lower than calculated ADAS values, which may be explained by the increased sputtering yield of redeposited Be atoms.

  10. 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.

  11. Workshop on Molecule Assisted Recombination and Other Processes in Fusion Divertor Plasmas, September 8-9, 2000

    International Nuclear Information System (INIS)

    Janev, R.K.; Schultz, D.R.

    2000-01-01

    A brief proceedings of the two-day Workshop on Molecule Assisted Recombination and Other Processes in Fusion Divertor Plasmas, organized by the ORNL Controlled Fusion Atomic Data Center on September 8-9, 2000, is presented. The conclusions and recommendations of the workshop regarding the topics discussed and the collaboration of the U.S. fusion research and atomic physics communities are also summarized

  12. 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.)

  13. Radiation transport effects in divertor plasmas generated during a tokamak reactor disruption

    International Nuclear Information System (INIS)

    Peterson, R.R.; MacFarlane, J.J.; Wang, P.

    1994-01-01

    Vaporization of material from tokamak divertors during disruptions is a critical issue for tokamak reactors from ITER to commercial power plants. Radiation transport from the vaporized material onto the remaining divertor surface plays an important role in the total mass loss to the divertor. Radiation transport in such a vapor is very difficult to calculate in full detail, and this paper quantifies the sensitivity of the divertor mass loss to uncertainties in the radiation transport. Specifically, the paper presents the results of computer simulations of the vaporization of a graphite coated divertor during a tokamak disruption with ITER CDA parameters. The results show that a factor of 100 change in the radiation conductivity changes the mass loss by more than a factor of two

  14. Models for poloidal divertors

    International Nuclear Information System (INIS)

    Post, D.E.; Heifetz, D.; Petravic, M.

    1982-07-01

    Recent progress in models for poloidal divertors has both helped to explain current divertor experiments and contributed significantly to design efforts for future large tokamak (INTOR, etc.) divertor systems. These models range in sophistication from zero-dimensional treatments and dimensional analysis to two-dimensional models for plasma and neutral particle transport which include a wide variety of atomic and molecular processes as well as detailed treatments of the plasma-wall interaction. This paper presents a brief review of some of these models, describing the physics and approximations involved in each model. We discuss the wide variety of physics necessary for a comprehensive description of poloidal divertors. To illustrate the progress in models for poloidal divertors, we discuss some of our recent work as typical examples of the kinds of calculations being done

  15. Models for poloidal divertors

    Energy Technology Data Exchange (ETDEWEB)

    Post, D.E.; Heifetz, D.; Petravic, M.

    1982-07-01

    Recent progress in models for poloidal divertors has both helped to explain current divertor experiments and contributed significantly to design efforts for future large tokamak (INTOR, etc.) divertor systems. These models range in sophistication from zero-dimensional treatments and dimensional analysis to two-dimensional models for plasma and neutral particle transport which include a wide variety of atomic and molecular processes as well as detailed treatments of the plasma-wall interaction. This paper presents a brief review of some of these models, describing the physics and approximations involved in each model. We discuss the wide variety of physics necessary for a comprehensive description of poloidal divertors. To illustrate the progress in models for poloidal divertors, we discuss some of our recent work as typical examples of the kinds of calculations being done.

  16. 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.)

  17. 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.

  18. 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

  19. Limiter and divertor systems - conceptual and mechanical design for Aditya Tokamak upgrade

    International Nuclear Information System (INIS)

    Patel, Kaushal; Rathod, Kulav; Jadeja, Kumarpalsinh A.

    2015-01-01

    Existing Aditya tokamak with limiter configuration is being upgraded into a machine to have both the limiter and divertor configurations. Necessary modifications have been carried out to accommodate divertor coils by replacing the old vacuum vessel with a new circular section vacuum vessel. The upgraded Aditya tokamak will have different set of limiters and divertors, such as Safety limiter, Toroidal Inner limiter, outer limiter of smaller toroidal extent, Upper and lower divertor plates. The limiter and divertor locations inside the Aditya tokamak upgrade are decided based on the numerical simulation of the plasma equilibrium profiles. Initially graphite will be used as plasma facing material (PFM) in all the limiter and divertor plates. The dimensions of the limiter and divertor tiles are decided based on their installation inside the vacuum vessel as well as on the total plasma heat loads (∼ 1 MW) falling on them. Depending upon the heat loads; the thickness of graphite tiles for limiter and divertor plates is estimated. Shaped graphite tiles will be fixed on specially designed support structures made out of SS-304L inside the torus shaped vacuum vessel. In this paper mechanical structural design of limiter and divertor of Aditya Upgrade Tokamak is presented. (author)

  20. 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.)

  1. 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.)

  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. 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

  5. Experimental study of the recombination of a drifting low temperature plasma in the divertor simulator Mistral-B

    International Nuclear Information System (INIS)

    Brault, C.; Escarguel, A.; Koubiti, M.; Stamm, R.; Pierre, Th.; Quotb, K.; Guyomarc'h, D.

    2004-01-01

    In a new divertor simulator, an ultra-cold (T e 18 m -3 . The collector is segmented into two plates and a transverse electric field is applied through a potential difference between the plates. The Lorentz force induces the ejection of a very-low temperature plasma jet in the limiter shadow. The characteristic convection time and decay lengths have been obtained with an ultra-fast camera. The study of the atomic physics of the recombining plasma allows to understand the measured decay time and to explain the emission spectra. (authors)

  6. 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.

  7. 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

  8. Snowflake Divertor Configuration in NSTX

    International Nuclear Information System (INIS)

    Soukhanovskii, V.A.; Ahn, Joonwook; Bell, R.E.; Gates, D.A.; Gerhardt, S.; Kaita, R.; Kolemen, E.; Kugel, H.W.; LeBlanc, B.; Maingi, Rajesh; Maqueda, R.J.; McLean, Adam G.; Menard, J.E.; Mueller, D.; Paul, S.F.; Raman, R.; Roquemore, L.; Ryutov, D.D.; Scott, H.A.

    2011-01-01

    Steady-state handling of divertor heat flux is a critical issue for present and future conventional and spherical tokamaks with compact high power density divertors. A novel 'snowflake' divertor (SFD) configuration that takes advantage of magnetic properties of a second-order poloidal null has been predicted to have a larger plasma-wetted area and a larger divertor volume, in comparison with a standard first-order poloidal X-point divertor configuration. The SFD was obtained in 0.8 MA, 4-6 MW NBI-heated H-mode discharges in NSTX using two divertor magnetic coils. The SFD led to a partial detachment of the outer strike point even in low-collisionality scrape-off layer plasma obtained with lithium coatings in NSTX. Significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with good core confinement and MHD properties.

  9. 'Snowflake' divertor configuration in NSTX

    International Nuclear Information System (INIS)

    Soukhanovskii, V.A.; Ahn, J.-W.; Bell, R.E.; Gates, D.A.; Gerhardt, S.; Kaita, R.; Kolemen, E.; Kugel, H.W.; LeBlanc, B.P.; Maingi, R.; Maqueda, R.; McLean, A.; Menard, J.E.; Mueller, D.M.; Paul, S.F.; Raman, R.; Roquemore, A.L.; Ryutov, D.D.; Scott, H.A.

    2011-01-01

    Steady-state handling of divertor heat flux is a critical issue for present and future conventional and spherical tokamaks with compact high power density divertors. A novel 'snowflake' divertor (SFD) configuration that takes advantage of magnetic properties of a second-order poloidal null has been predicted to have a larger plasma-wetted area and a larger divertor volume, in comparison with a standard first-order poloidal X-point divertor configuration. The SFD was obtained in 0.8 MA, 4-6 MW NBI-heated H-mode discharges in NSTX using two divertor magnetic coils. The SFD led to a partial detachment of the outer strike point even in low-collisionality scrape-off layer plasma obtained with lithium coatings in NSTX. Significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with good core confinement and MHD properties.

  10. "Snowflake" divertor configuration in NSTX

    Science.gov (United States)

    Soukhanovskii, V. A.; Ahn, J.-W.; Bell, R. E.; Gates, D. A.; Gerhardt, S.; Kaita, R.; Kolemen, E.; Kugel, H. W.; Leblanc, B. P.; Maingi, R.; Maqueda, R.; McLean, A.; Menard, J. E.; Mueller, D. M.; Paul, S. F.; Raman, R.; Roquemore, A. L.; Ryutov, D. D.; Scott, H. A.

    2011-08-01

    Steady-state handling of divertor heat flux is a critical issue for present and future conventional and spherical tokamaks with compact high power density divertors. A novel "snowflake" divertor (SFD) configuration that takes advantage of magnetic properties of a second-order poloidal null has been predicted to have a larger plasma-wetted area and a larger divertor volume, in comparison with a standard first-order poloidal X-point divertor configuration. The SFD was obtained in 0.8 MA, 4-6 MW NBI-heated H-mode discharges in NSTX using two divertor magnetic coils. The SFD led to a partial detachment of the outer strike point even in low-collisionality scrape-off layer plasma obtained with lithium coatings in NSTX. Significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with good core confinement and MHD properties.

  11. Operation method for thermonuclear device and divertor for it

    International Nuclear Information System (INIS)

    Kotake, Michiko; Yoshioka, Ken; Fukumoto, Hideshi; Okazaki, Takashi; Kinoshita, Shigemi; Takeuchi, Kazuhiro.

    1992-01-01

    Divertor plates are disposed subsequently along with circumferential direction of a vacuum vessel in a region where magnetic fluxed generated from the divertor coils are injected toward a container wall. Each of the divertor plates is moved in a state that the injection position of the magnetic fluxes enter to the vacuum vessel is kept constant. Alternatively, each of the divertor plates is inclined at an angle facing the injection direction of plasma particle fluxes, or it is inclined so that the angle between the injection surface and the magnetic fluxes makes an acute angle. Since each of the divertor coils is moved in the state of keeping the injection position of the magnetic fluxes during firing of plasmas, in other words, with on change of the current of the divertor coils, the position of the magnetic fluxed is kept at a predetermined condition. Accordingly, charged particles are prevented from concentrating locally without causing eddy current in the coils and the vacuum vessel, which can contribute to the reduction of the wear of the divertor plates. (N.H.)

  12. Estimation of peak heat flux onto the targets for CFETR with extended divertor leg

    International Nuclear Information System (INIS)

    Zhang, Chuanjia; Chen, Bin; Xing, Zhe; Wu, Haosheng; Mao, Shifeng; Luo, Zhengping; Peng, Xuebing; Ye, Minyou

    2016-01-01

    Highlights: • A hypothetical geometry is assumed to extend the outer divertor leg in CFETR. • Density scan SOLPS simulation is done to study the peak heat flux onto target. • Attached–detached regime transition in out divertor occurs at lower puffing rate. • Unexpected delay of attached–detached regime transition occurs in inner divertor. - Abstract: China Fusion Engineering Test Reactor (CFETR) is now in conceptual design phase. CFETR is proposed as a good complement to ITER for demonstrating of fusion energy. Divertor is a crucial component which faces the plasmas and handles huge heat power for CFETR and future fusion reactor. To explore an effective way for heat exhaust, various methods to reduce the heat flux to divertor target should be considered for CFETR. In this work, the effect of extended out divertor leg on the peak heat flux is studied. The magnetic configuration of the long leg divertor is obtained by EFIT and Tokamak Simulation Code (TSC), while a hypothetical geometry is assumed to extend the out divertor leg as long as possible inside vacuum vessel. A SOLPS simulation is performed to study peak heat flux of the long leg divertor for CFETR. D 2 gas puffing is used and increasing of the puffing rate means increase of plasma density. Both peak heat flux onto inner and outer targets are below 10 MW/m 2 is achieved. A comparison between the peak heat flux between long leg and conventional divertor shows that an attached–detached regime transition of out divertor occurs at lower gas puffing gas puffing rate for long leg divertor. While for the inner divertor, even the configuration is almost the same, the situation is opposite.

  13. Estimation of peak heat flux onto the targets for CFETR with extended divertor leg

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Chuanjia; Chen, Bin [School of Nuclear Science and Technology, University of Science and Technology of China, 96 Jinzhai Road, Hefei, Anhui 230026 (China); Xing, Zhe [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Wu, Haosheng [School of Nuclear Science and Technology, University of Science and Technology of China, 96 Jinzhai Road, Hefei, Anhui 230026 (China); Mao, Shifeng, E-mail: sfmao@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, 96 Jinzhai Road, Hefei, Anhui 230026 (China); Luo, Zhengping; Peng, Xuebing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Ye, Minyou [School of Nuclear Science and Technology, University of Science and Technology of China, 96 Jinzhai Road, Hefei, Anhui 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

    2016-11-01

    Highlights: • A hypothetical geometry is assumed to extend the outer divertor leg in CFETR. • Density scan SOLPS simulation is done to study the peak heat flux onto target. • Attached–detached regime transition in out divertor occurs at lower puffing rate. • Unexpected delay of attached–detached regime transition occurs in inner divertor. - Abstract: China Fusion Engineering Test Reactor (CFETR) is now in conceptual design phase. CFETR is proposed as a good complement to ITER for demonstrating of fusion energy. Divertor is a crucial component which faces the plasmas and handles huge heat power for CFETR and future fusion reactor. To explore an effective way for heat exhaust, various methods to reduce the heat flux to divertor target should be considered for CFETR. In this work, the effect of extended out divertor leg on the peak heat flux is studied. The magnetic configuration of the long leg divertor is obtained by EFIT and Tokamak Simulation Code (TSC), while a hypothetical geometry is assumed to extend the out divertor leg as long as possible inside vacuum vessel. A SOLPS simulation is performed to study peak heat flux of the long leg divertor for CFETR. D{sub 2} gas puffing is used and increasing of the puffing rate means increase of plasma density. Both peak heat flux onto inner and outer targets are below 10 MW/m{sup 2} is achieved. A comparison between the peak heat flux between long leg and conventional divertor shows that an attached–detached regime transition of out divertor occurs at lower gas puffing gas puffing rate for long leg divertor. While for the inner divertor, even the configuration is almost the same, the situation is opposite.

  14. Energy and particle transport in the radiative divertor plasmas of DIII-D

    International Nuclear Information System (INIS)

    Leonard, A.W.; Allen, S.L.; Brooks, N.H.

    1997-06-01

    It has been argued that divertor energy transport dominated by parallel electron thermal conduction, or q parallel = -kT 5/2 2 dT e /ds parallel, leads to severe localization of the intense radiating region and ultimately limits the fraction of energy flux that can be radiated before striking the divertor target. This is due to the strong T 5/2 e dependence of electron heat conduction which results in very short spatial scales of the T e gradient at high power densities and low temperatures where deuterium and impurities radiate most effectively. However, we have greatly exceeded this constraint on DIII-D with deuterium gas puffing which reduces the peak heat flux to the divertor plate a factor of 5 while distributing the divertor radiation over a long length

  15. Change of the Magnetic-Field Topology by an Ergodic Divertor and the Effect on the Plasma Structure and Transport

    International Nuclear Information System (INIS)

    Jakubowski, M.W.; Schmitz, O.; Abdullaev, S.S.; Brezinsek, S.; Finken, K.H.; Kraemer-Flecken, A.; Lehnen, M.; Samm, U.; Unterberg, B.; Wolf, R.C.; Spatschek, K.H.

    2006-01-01

    The magnetic-field perturbation produced by the dynamic ergodic divertor in TEXTOR changes the topology of the magnetic field in the plasma edge, creating an open chaotic system. The perturbation spectrum contains only a few dominant harmonics and therefore it can be described by an analytical model. The modeling is performed in the vacuum approximation without assuming a backreaction of the plasma and does not rely on any experimentally obtained parameters. It is shown that this vacuum approximation predicts in many details the experimentally observed plasma structure. Several experiments have been performed to prove that the plasma edge behavior is defined mostly by the magnetic topology of the perturbed volume. The change in the transport can be explained with the knowledge of only the magnetic structures; i.e., the ergodic pattern dominates the plasma properties

  16. T-12 divertor experiment

    Energy Technology Data Exchange (ETDEWEB)

    Bortnikov, A V; Brevnov, N N; Gerasimov, S N; Zhukovskii, V G; Kuznetsov, N V; Naftulin, S M; Pergament, V I; Khimchenko, L N [Gosudarstvennyj Komitet po Ispol' zovaniyu Atomnoj Ehnergii SSSR, Moscow. Inst. Atomnoj Ehnergii

    1981-01-01

    In designing tokamak devices and reactors, in the last few years, the use of elongated-cross-section plasma discharges has been proposed to improve the economic and physical parameters. Application of a quadrupole poloidal magnetic field necessary for sustaining the elongated discharge cross-section serves, in this case, to create the magnetic configuration of an axisymmetric poloidal divertor. To-day, the creation of such a combination, including an elongated plasma cross-section and a divertor and using the outer poloidal magnetic field coils, seems to be the most reasonable approach, from the point of view of design and technology. Such a divertor was produced and studied at the T-12 tokamak. A stable equilibrium configuration of a finger-ring tokamak with a divertor has been produced by superposing the magnetic fields of the plasma current, the external quadrupole coils and the copper shell currents; the reactor blanket can fulfil the function of the latter. It is shown that both a symmetric magnetic configuration with two divertors and a droplet configuration with a single divertor may be realized by controlling the plasma column position with respect to the equatorial plane. The stability of the plasma column against vertical displacement depends on this position and the distance between the separatrix points. Vertical instability stabilization has been observed. The divertor layer efficiently screens the plasma from the impurity influx from the wall and unloads the wall from particle and energy fluxes. The results obtained from the tokamak T-12 experiment have demonstrated the capability of a system with outer poloidal field coils and a copper shell providing an elongated-cross-section plasma column with poloidal divertors.

  17. 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

  18. PREFACE: Light element atom, molecule and radical behaviour in the divertor and edge plasma regions

    Science.gov (United States)

    Braams, Bastiaan J.; Chung, Hyun-Kung

    2015-01-01

    This volume of Journal of Physics: Conference Series contains contributions by participants in an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on "Light element atom, molecule and radical behaviour in the divertor and edge plasma regions" (in magnetic fusion devices). Light elements are the dominant impurity species in fusion experiments and in the near-wall plasma they occur as atoms or ions and also as hydrides and other molecules and molecular ions. Hydrogen (H or D, and T in a reactor) is the dominant species in fusion experiments, but all light elements He - O and Ne are of interest for various reasons. Helium is a product of the D+T fusion reaction and is introduced in experiments for transport studies. Lithium is used for wall coating and also as a beam diagnostic material. Beryllium is foreseen as a wall material for the ITER experiment and is used on the Joint European Torus (JET) experiment. Boron may be used as a coating material for the vessel walls. Carbon (graphite or carbon-fiber composite) is often used as the target material for wall regions subject to high heat load. Nitrogen may be used as a buffer gas for edge plasma cooling. Oxygen is a common impurity in experiments due to residual water vapor. Finally, neon is another choice as a buffer gas. Data for collisional and radiative processes involving these species are important for plasma modelling and for diagnostics. The participants in the CRP met 3 times over the years 2009-2013 for a research coordination meeting. Reports and presentation materials for these meetings are available through the web page on coordinated research projects of the (IAEA) Atomic and Molecular Data Unit [1]. Some of the numerical data generated in the course of the CRP is available through the ALADDIN database [2]. The IAEA takes the opportunity to thank the participants in the CRP for their dedicated efforts in the course of the CRP and for their contributions to this volume. The IAEA

  19. 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)

  20. 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.

  1. 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.)

  2. The ITER divertor cassette project

    International Nuclear Information System (INIS)

    Ulrickson, M.; Tivey, R.; Akiba, M.

    2001-01-01

    The divertor ''Large Project'' was conceived with the aim of demonstrating the feasibility of meeting the lifetime requirements by employing the candidate armor materials of beryllium, tungsten (W) and carbon-fiber-composite (CFC). At the start, there existed only limited experience with constructing water-cooled high heat flux armored components for tokamaks. To this was added the complication posed by the need to use a silver-free joining technique that avoids the transmutation of n-irradiated silver to cadmium. The research project involving the four Home Teams (HTs) has focused on the design, development, manufacture and testing of full-scale Plasma Facing Components (PFCs) suitable for ITER. The task addressed all the issues facing ITER divertor design, such as providing adequate armor erosion lifetime, meeting the required armor-heat sink joint lifetime and heat sink fatigue life, sustaining thermal-hydraulic and electromechanical loads, and seeking to identify the most cost-effective manufacturing options. This paper will report the results of the divertor large project. (author)

  3. The ITER divertor cassette project

    International Nuclear Information System (INIS)

    Ulrickson, M.; Tivey, R.; Akiba, M.

    1999-01-01

    The divertor 'Large Project' was conceived with the aim of demonstrating the feasibility of meeting the lifetime requirements by employing the candidate armor materials of beryllium, tungsten (W) and carbon-fiber-composite (CFC). At the start, there existed only limited experience with constructing water-cooled high heat flux armored components for tokamaks. To this was added the complication posed by the need to use a silver-free joining technique that avoids the transmutation of n-irradiated silver to cadmium. The research project involving the four Home Teams (HTs) has focused on the design, development, manufacture and testing of full-scale Plasma Facing Components (PFCs) suitable for ITER. The task addressed all the issues facing ITER divertor design, such as providing adequate armor erosion lifetime, meeting the required armor-heat sink joint lifetime and heat sink fatigue life, sustaining thermal-hydraulic and electromechanical loads, and seeking to identify the most cost-effective manufacturing options. This paper will report the results of the divertor large project. (author)

  4. Progress towards modeling tokamak boundary plasma turbulence and understanding its role in setting divertor heat flux widths

    Science.gov (United States)

    Chen, B.; Xu, X. Q.; Xia, T. Y.; Li, N. M.; Porkolab, M.; Edlund, E.; LaBombard, B.; Terry, J.; Hughes, J. W.; Ye, M. Y.; Wan, Y. X.

    2018-05-01

    The heat flux distributions on divertor targets in H-mode plasmas are serious concerns for future devices. We seek to simulate the tokamak boundary plasma turbulence and heat transport in the edge localized mode-suppressed regimes. The improved BOUT++ model shows that not only Ip but also the radial electric field Er plays an important role on the turbulence behavior and sets the heat flux width. Instead of calculating Er from the pressure gradient term (diamagnetic Er), it is calculated from the plasma transport equations with the sheath potential in the scrape-off layer and the plasma density and temperature profiles inside the separatrix from the experiment. The simulation results with the new Er model have better agreement with the experiment than using the diamagnetic Er model: (1) The electromagnetic turbulence in enhanced Dα H-mode shows the characteristics of quasi-coherent modes (QCMs) and broadband turbulence. The mode spectra are in agreement with the phase contrast imaging data and almost has no change in comparison to the cases which use the diamagnetic Er model; (2) the self-consistent boundary Er is needed for the turbulence simulations to get the consistent heat flux width with the experiment; (3) the frequencies of the QCMs are proportional to Er, while the divertor heat flux widths are inversely proportional to Er; and (4) the BOUT++ turbulence simulations yield a similar heat flux width to the experimental Eich scaling law and the prediction from the Goldston heuristic drift model.

  5. Numerical modeling and experimental simulation of vapor shield formation and divertor material erosion for ITER typical plasma disruptions

    International Nuclear Information System (INIS)

    Wuerz, H.; Arkhipov, N.I.; Bakhin, V.P.; Goel, B.; Hoebel, W.; Konkashbaev, I.; Landman, I.; Piazza, G.; Safronov, V.M.; Sherbakov, A.R.; Toporkov, D.A.; Zhitlukhin, A.M.

    1994-01-01

    The high divertor heat load during a tokamak plasma disruption results in sudden evaporation of a thin layer of divertor plate material, which acts as vapor shield and protects the target from further excessive evaporation. Formation and effectiveness of the vapor shield are theoretically modeled and experimentally investigated at the 2MK-200 facility under conditions simulating the thermal quench phase of ITER tokamak plasma disruptions. In the optical wavelength range C II, C III, C IV emission lines for graphite, Cu I, Cu II lines for copper and continuum radiation for tungsten samples are observed in the target plasma. The plasma expands along the magnetic field lines with velocities of (4±1)x10 6 cm/s for graphite and 10 5 cm/s for copper. Modeling was done with a radiation hydrodynamics code in one-dimensional planar geometry. The multifrequency radiation transport is treated in flux limited diffusion and in forward reverse transport approximation. In these first modeling studies the overall shielding efficiency for carbon and tungsten defined as ratio of the incident energy and the vaporization energy for power densities of 10 MW/cm 2 exceeds a factor of 30. The vapor shield is established within 2 μs, the power fraction to the target after 10 μs is below 3% and reaches in the stationary state after about 20 μs a value of around 1.5%. ((orig.))

  6. 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

  7. Investigation of detached recombining deuterium plasma and carbon chemical erosion in the toroidal divertor simulator NAGDIS-T

    International Nuclear Information System (INIS)

    Yada, K.; Matsui, N.; Ohno, N.; Kajita, S.; Takamura, S.; Takagi, M.

    2009-01-01

    Detached deuterium recombining plasma has been generated in the toroidal divertor simulator. The electron temperature (0.1-0.4 eV) and density (∼10 18 m -3 ) in the detached plasmas were evaluated with a spectroscopic method using a series of deuterium Balmer line emission from highly excited levels and the Stark broadening of D(2-12). We have investigated the role of volume plasma recombination through Electron-Ion Recombination (EIR) and Molecular Activated Recombination (MAR) processes. Moreover, the carbon erosion in the detached deuterium plasma has been studied with a weight loss method. It is found that deuterium neutrals generated by EIR process could have strong influence on the carbon chemical erosion.

  8. 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.

  9. 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.

  10. 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.)

  11. Fusion plasma theory Task II: ECRH and transport modeling in tandem mirrors and divertor physics. Final report, January 1-December 31, 1985

    International Nuclear Information System (INIS)

    Emmert, G.A.

    1985-07-01

    The research reported here focuses on: (1) the coupling of an ECRH ray tracing and absorption code to a tandem mirror transport code in order to self-consistently model the temporal and spatial evolution of the plasma, and (2) the further development of semi-analytical models for plasma flow in divertors and pumped limiters. 5 refs., 1 fig

  12. 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.

  13. Comparisons of physical and chemical sputtering in high density divertor plasmas with the Monte Carlo Impurity (MCI) transport model

    International Nuclear Information System (INIS)

    Evans, T.E.; Loh, Y.S.; West, W.P.; Finkenthal, D.F.

    1997-11-01

    The MCI transport model was used to compare chemical and physical sputtering for a DIII-D divertor plasma near detachment. With physical sputtering alone the integrated carbon influx was 8.4 x 10 19 neutral/s while physical plus chemical sputtering produced an integrated carbon influx of 1.7 x 10 21 neutrals/s. The average carbon concentration in the computational volume increased from 0.012% with only physical sputtering to 0.182% with both chemical and physical sputtering. This increase in the carbon inventory produced more radiated power which is in better agreement with experimental measurements

  14. 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

  15. Deuterium to helium plasma-wall change-over experiments in the JET MkII-gas box divertor

    International Nuclear Information System (INIS)

    Hillis, D.L.; Loarer, T.; Bucalossi, J.; Pospieszczyk, A.; Fundamenski, W.; Matthews, G.; Meigs, A.; Morgan, P.; Phillips, V.; Pitts, R.; Stamp, M.; Hellermann, M. von

    2003-01-01

    The deuterium and helium dynamics in the plasma and subdivertor regions of JET are compared during a sequence of similar ohmic and ICRH pulses where 100% He gas is injected into the JET vacuum vessel, whose graphite walls were previously saturated with deuterium. After the first six He fueled change-over discharges, only He plasma operation was performed. Following this investigation, the situation is reversed and the change-over from an initially saturated He wall is investigated when only D 2 plasma fuelling is used. The He concentration is measured in the subdivertor with a species selective Penning gauge. Comparison of the time dependence of the divertor concentrations with those at the edge and strike point shows significant differences during the first six discharges. This difference along with a global He particle balance is used to assess the status of the wall saturation over the initial 6-7 He change-over discharges

  16. 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.

  17. Particle and power deposition on divertor targets in EAST H-mode plasmas

    DEFF Research Database (Denmark)

    Wang, L.; Xu, G.S.; Guo, H.Y.

    2012-01-01

    ELMs were chosen for analysis in order to reduce the uncertainty resulting from the influence of fast electrons on Langmuir triple-probe evaluation during ELMs. The power deposition obtained from Langmuir triple probes was consistent with that from the divertor infra-red camera during an ELM...

  18. Design integration of liquid surface divertors

    International Nuclear Information System (INIS)

    Nygren, R.E.; Cowgill, D.F.; Ulrickson, M.A.; Nelson, B.E.; Fogarty, P.J.; Rognlien, T.D.; Rensink, M.E.; Hassanein, A.; Smolentsev, S.S.; Kotschenreuther, M.

    2004-01-01

    The US Enabling Technology Program in fusion is investigating the use of free flowing liquid surfaces facing the plasma. We have been studying the issues in integrating a liquid surface divertor into a configuration based upon an advanced tokamak, specifically the ARIES-RS configuration. The simplest form of such a divertor is to extend the flow of the liquid first wall into the divertor and thereby avoid introducing additional fluid streams. In this case, one can modify the flow above the divertor to enhance thermal mixing. For divertors with flowing liquid metals (or other electrically conductive fluids) MHD (magneto-hydrodynamics) effects are a major concern and can produce forces that redirect flow and suppress turbulence. An evaluation of Flibe (a molten salt) as a working fluid was done to assess a case in which the MHD forces could be largely neglected. Initial studies indicate that, for a tokamak with high power density, an integrated Flibe first wall and divertor does not seem workable. We have continued work with molten salts and replaced Flibe with Flinabe, a mixture of lithium, sodium and beryllium fluorides, that has some potential because of its lower melting temperature. Sn and Sn-Li have also been considered, and the initial evaluations on heat removal with minimal plasma contamination show promise, although the complicated 3D MHD flows cannot yet be fully modeled. Particle pumping in these design concepts is accomplished by conventional means (ports and pumps). However, trapping of hydrogen in these flowing liquids seems plausible and novel concepts for entrapping helium are also being studied

  19. Numerical studies on divertor experiments

    International Nuclear Information System (INIS)

    Ueda, N.; Itoh, K.; Itoh, S.-I.; Tanaka, M.; Hasegawa, M.; Shoji, T.; Sugihara, M.

    1988-04-01

    Numerical analysis on the divertor experiments such as JFT-2M tokamak is made by use of the two-dimensional time-dependent simulation code. The plasma in the scrape-off layer (SOL) and divertor region is solved for the given particle and heat sources from the main plasma, Γ p and Q T . Effect of the direction of the toroidal magnetic field is studied. It is found that the heat flux which is proportional to b vector x ∇T i has influences on the divertor plasmas, but has a small effect on the parameters on the midplane in the framework of the fluid model. Parameter survey on Γ p and Q T is made. The transient response of the SOL/divertor plasma to the sudden change of Γ p and Q T is studied. Time delay in the SOL and divertor region is calculated. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

  1. Temporal evolutions of electron temperature and density with edge localized mode in the JT-60U divertor plasma

    International Nuclear Information System (INIS)

    Nakano, T; Kubo, H; Asakura, N

    2010-01-01

    From the intensity ratios of the three He I lines measured at 20 kHz, the temporal evolutions of the electron temperature and density during and after the power and the particle flow into the divertor plasma caused by edge localized modes are determined. The electron temperature increases from 70 eV to 80 eV with increasing D α intensity. Then, at the peak of D α intensity, the electron temperature starts decreasing down to 60 eV. The electron density increases from 0.1 x 10 19 m -3 to 0.3 x 10 19 m -3 with increasing D α intensity, and then starts to decrease more gradually compared with the electron temperature after the peak of D α intensity. It is interpreted that the increase of the electron temperature is ascribed to the power and the particle flow into the divertor plasma, and that the decrease of the electron temperature and the increase of the electron density are ascribed to the ionization of the recycled neutrals, which consumes the electron energy and produces electrons.

  2. 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

  3. 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)

  4. Module of lithium divertor for KTM tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lyublinski, I., E-mail: yublinski@yandex.ru [FSUE ' Red Star' , Moscow (Russian Federation); Vertkov, A.; Evtikhin, V.; Balakirev, V.; Ionov, D.; Zharkov, M. [FSUE ' Red Star' , Moscow (Russian Federation); Tazhibayeva, I. [IAE NNC RK, Kurchatov (Kazakhstan); Mirnov, S. [TRINITI, Troitsk, Moscow Region (Russian Federation); Khomiakov, S.; Mitin, D. [OJSC Dollezhal Institute, Moscow (Russian Federation); Mazzitelli, G. [ENEA RC Frascati (Italy); Agostini, P. [ENEA RC Brasimone (Italy)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium-metal with low Z. Black-Right-Pointing-Pointer Capillary-porous system (CPS) - new material in which liquid lithium fill a solid matrix from porous material. Black-Right-Pointing-Pointer Lithium divertor module for KTM tokamak is under development. Black-Right-Pointing-Pointer Lithium filled tungsten felt is offered as the base plasma facing material of divertor. Black-Right-Pointing-Pointer Results of this project addresses to the progress in the field of fusion neutrons source and fusion energy source creation. - Abstract: Activity on projects of ITER and DEMO reactors has shown that solution of problems of divertor target plates and other plasma facing elements (PFEs) based on the solid plasma facing materials cause serious difficulties. Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium-metal with low Z. Application of lithium will allow to create a self-renewal and MHD stable liquid metal surface of the in-vessel devices possessing practically unlimited service life; to reduce power flux due to intensive re-irradiation on lithium atoms in plasma periphery that will essentially facilitate a problem of heat removal from PFE; to reduce Z{sub eff} of plasma to minimally possible level close to 1; to exclude tritium accumulation, that is provided with absence of dust products and an opportunity of the active control of the tritium contents in liquid lithium. Realization of these advantages is based on use of so-called lithium capillary-porous system (CPS) - new material in which liquid lithium fill a solid matrix from porous material. The progress in development of lithium technology and also activity in lithium experiments in the tokamaks TFTR, T-11M, T-10, FTU, NSTX, HT-7 and stellarator TJ II permits of solving the problems in development of

  5. Textor bundle divertor

    International Nuclear Information System (INIS)

    Yang, T.F.; Wan, A.; Gierszewski, P.; Rapperport, E.; Montgomery, D.B.

    1982-01-01

    This report presents a preliminary bundle divertor conceptual design for installation on the TEXTOR tokamak. An advanced cascade T-shaped coil configuration is used. This divertor design has the following important characteristics: (1) the current density in the conductor is less than 6 kAmp/cm 2 , and the maximum field is less than 6 Tesla; (2) the divertor can be operated at steady-state either for copper or superconducting conductors; (3) the power consumption is about 7 MW for a normal conductor; (4) the divertor can be inserted into the existing geometry of TEXTOR; (5) the ripple on axis is only 0.3% and the mirror ratio is 2 to 4; (6) the stagnation axis is concave toward the plasma, therefore q/sub D/ is smaller, the acceptance angle is larger, and the efficiency may be better than the conventional circular coil design

  6. TEXTOR bundle divertor

    International Nuclear Information System (INIS)

    Yang, T.F.; Wan, A.; Gierszewski, P.; Rapperport, E.; Montgomery, D.B.

    1982-01-01

    This report presents a preliminary bundle divertor conceptual design for installation on the TEXTOR tokamak. An advanced cascade T-shaped coil configuration is used. This divertor design has the following important characteristics: (1) the current density in the conductor is less than 6 kAmp/cm 2 , and the maximum field is less than 6 Tesla; (2) the divertor can be operated at steady-state either for copper or superconducting conductors; (3) the power consumption is about 7 MW for a normal conductor; (4) the divertor can be inserted into the existing geometry of TEXTOR; (5) the ripple on axis is only 0.3% and the mirror ratio is 2 to 4; (6) the stagnation axis is concave toward the plasma, therefore q/sub D/ is smaller, the acceptance angle is larger, and the efficiency may be better than the conventional circular coil design

  7. NSTX Tangential Divertor Camera

    International Nuclear Information System (INIS)

    Roquemore, A.L.; Ted Biewer; Johnson, D.; Zweben, S.J.; Nobuhiro Nishino; Soukhanovskii, V.A.

    2004-01-01

    Strong magnetic field shear around the divertor x-point is numerically predicted to lead to strong spatial asymmetries in turbulence driven particle fluxes. To visualize the turbulence and associated impurity line emission near the lower x-point region, a new tangential observation port has been recently installed on NSTX. A reentrant sapphire window with a moveable in-vessel mirror images the divertor region from the center stack out to R 80 cm and views the x-point for most plasma configurations. A coherent fiber optic bundle transmits the image through a remotely selected filter to a fast camera, for example a 40500 frames/sec Photron CCD camera. A gas puffer located in the lower inboard divertor will localize the turbulence in the region near the x-point. Edge fluid and turbulent codes UEDGE and BOUT will be used to interpret impurity and deuterium emission fluctuation measurements in the divertor

  8. Effect of the resonant magnetic perturbation on the plasma parameters in COMPASS tokamak’s divertor region

    Science.gov (United States)

    Dimitrova, M.; Cahyna, P.; Peterka, M.; Hasan, E.; Popov, Tsv K.; Ivanova, P.; Vasileva, E.; Panek, R.; Cavalier, J.; Seidl, J.; Markovic, T.; Havlicek, J.; Dejarnac, R.; Weinzettl, V.; Hacek, P.; Tomes, M.; the COMPASS Team; the EUROfusion MST1 Team

    2018-02-01

    The resonant magnetic perturbation (RMP) has proven to be a useful way to suppress edge-localized modes that under certain conditions can damage the device by the large power fluxes carried from the bulk plasma to the wall. The effect of RMP on the L-mode plasma parameters in the divertor region of the COMPASS tokamak was studied using the array of 39 Langmuir probes embedded into the divertor target. The current-voltage (IV) probe characteristics were processed by the first-derivative probe technique to obtain the plasma potential and the electron energy distribution function (EEDF) which was approximated by a bi-Maxwellian EEDF with a low-energy (4-6 eV) fraction and a high-energy (11-35 eV) one, the both factions having similar electron density. Clear splitting was observed during the RMP pulse in the low-field-side scrape-off-layer profiles of the floating potential U fl and the ion saturation current density J sat; these two quantities were obtained both by direct continuous measurement and by evaluation of the IV characteristics of probes with swept bias. The negative peaks of U fl induced by RMP spatially overlaps with the local minima of J sat (and n e) rather than with its local maxima which is partly caused by the spatial variation of the plasma potential and partly by the changed shape of the EEDF. The effective temperature of the whole EEDF is not correlated with the negative peaks of U fl, and the profile of the parallel power flux density shows secondary maxima due to RMP which mimic those of J sat.

  9. 2D statistical analysis of Non-Diffusive transport under attached and detached plasma conditions of the linear divertor simulator

    International Nuclear Information System (INIS)

    Tanaka, H.; Ohno, N.; Tsuji, Y.; Kajita, S.

    2010-01-01

    We have analyzed the 2D convective motion of coherent structures, which is associated with plasma blobs, under attached and detached plasma conditions of a linear divertor simulator, NAGDIS-II. Data analysis of probes and a fast-imaging camera by spatio-temporal correlation with three decomposition and proper orthogonal decomposition (POD) was carried out to determine the basic properties of coherent structures detached from a bulk plasma column. Under the attached plasma condition, the spatio-temporal correlation with three decomposition based on the probe measurement showed that two types of coherent structures with different sizes detached from the bulk plasma and the azimuthally localized structure radially propagated faster than the larger structure. Under the detached plasma condition, movies taken by the fast-imaging camera clearly showed the dynamics of a 2D spiral structure at peripheral regions of the bulk plasma; this dynamics caused the broadening of the plasma profile. The POD method was used for the data processing of the movies to obtain low-dimensional mode shapes. It was found that the m=1 and m=2 ring-shaped coherent structures were dominant. Comparison between the POD analysis of both the movie and the probe data suggested that the coherent structure could be detached from the bulk plasma mainly associated with the m=2 fluctuation. This phenomena could play an important role in the reduction of the particle and heat flux as well as the plasma recombination processes in plasma detachment (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  10. 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.

  11. 3D plasma fluid simulations in divertor tokamaks. Final technical report, 1993--1995

    International Nuclear Information System (INIS)

    Strauss, H.R.

    1995-08-01

    The main accomplishment of this grant was the development of a finite element time dependent magnetofluid code, FEMHD. The code is nonlinear and three dimensional. In the poloidal plane, the elemental cells of the mesh are triangles, which offer both simplicity and adaptability. In the third, toroidal, direction, there is an option of a standard staggered finite difference mesh, or Fourier transforms. The FEMHD code runs on several platforms, including Crays, UNIX workstations, and a parallel version runs on an IBM SP1. Several problems have been considered with the unstructured mesh FEMHD code. They are (1) MHD simulations in divertor tokamaks; (2) simulations of ELM-like ballooning modes in divertor tokamaks; and (3) reconnection and singular MHD equilibria

  12. Thermomechanical simulation of WEST actively cooled upper divertor

    International Nuclear Information System (INIS)

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

    2016-01-01

    The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support to the ITER tungsten divertor strategy. The WEST project aims to test ITER-like W monoblock Plasma Facing Units (PFU). This ITER-like divertor will be tested under long plasma discharge up to 1000 s, with high heat flux density up to 20 MW/m 2 . This paper presents the results of ANSYS thermal-structural simulations of the WEST upper divertor. The upper divertor is made of twelve 30° sectors, each one composed of 38 PFU. The PFUs are actively cooled CuCrZr heat sinks and the incidence surface is coated with a thin tungsten layer. The fixing system is made of pins engaged in slotted holes. Besides, the fixing system of the sector assembly is the same as WEST lower divertor, so one upper divertor sector can be used indifferently in upper or Lower position during transitional operation phases in WEST. The total surface of the upper divertor is 8 m 2 , and it has to be able to extract up to 4 MW in steady-state, with peak heat flux values up to 8 MW/m 2 . The fixing system was designed to handle structural loads such as forces and torques resulting from halo and eddy current, respectively, especially during disruptions and Vertical Displacement Event (VDE). The torque resulting from eddy current is first calculated thanks to an internal CEA ANSYS APDL routine. Then the ANSYS structural and thermal-structural simulations of the PFU are presented, and its design is validated thanks to A-level RCC-MRx criteria. Finally, the most conservative load case is determined in order to validate the design of the pins and the support structure.

  13. Thermomechanical simulation of WEST actively cooled upper divertor

    Energy Technology Data Exchange (ETDEWEB)

    Batal, T., E-mail: tristan.batal@cea.fr; Richou, M.; Guilhem, D.; Firdaouss, M.; Larroque, S.; Ferlay, F.; Missirlian, M.; Bucalossi, J.

    2016-11-15

    The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support to the ITER tungsten divertor strategy. The WEST project aims to test ITER-like W monoblock Plasma Facing Units (PFU). This ITER-like divertor will be tested under long plasma discharge up to 1000 s, with high heat flux density up to 20 MW/m{sup 2}. This paper presents the results of ANSYS thermal-structural simulations of the WEST upper divertor. The upper divertor is made of twelve 30° sectors, each one composed of 38 PFU. The PFUs are actively cooled CuCrZr heat sinks and the incidence surface is coated with a thin tungsten layer. The fixing system is made of pins engaged in slotted holes. Besides, the fixing system of the sector assembly is the same as WEST lower divertor, so one upper divertor sector can be used indifferently in upper or Lower position during transitional operation phases in WEST. The total surface of the upper divertor is 8 m{sup 2}, and it has to be able to extract up to 4 MW in steady-state, with peak heat flux values up to 8 MW/m{sup 2}. The fixing system was designed to handle structural loads such as forces and torques resulting from halo and eddy current, respectively, especially during disruptions and Vertical Displacement Event (VDE). The torque resulting from eddy current is first calculated thanks to an internal CEA ANSYS APDL routine. Then the ANSYS structural and thermal-structural simulations of the PFU are presented, and its design is validated thanks to A-level RCC-MRx criteria. Finally, the most conservative load case is determined in order to validate the design of the pins and the support structure.

  14. 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

  15. 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.

  16. 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.

  17. 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.

  18. Scrape-off layer and divertor theory meeting: Proceedings

    International Nuclear Information System (INIS)

    1994-03-01

    This report contains viewgraphs on the following topics: fluid modelling of neutrals in the SOL and divertor; instabilities of gas-fueled divertors: theory and adaptive simulations; stability of ionization fronts of gaseous divertor plasmas; monte carlo calculation of heat transport; reduced charge model for edge impurity flows; thermally collapsed solutions for gaseous/radiative divertors; adaptive grid methods in transport simulation; advanced numerical solution algorithms applied to the multispecies edge plasma equations; two-dimensional edge plasma simulation using the multigrid method; neutral behavior and the effects of neutral-neutral and neutral-ion elastic scattering in the ITER gaseous divertor; particle throughput in the TPX divertor; marfes in tokamaks; a comparative study of the limiter and divertor edge plasmas in TEXT-U; issues of toroidal tokamak-type divertor simulators; ASDEX upgrade; the ITER divertor; the DIII-D divertor program and TPX divertor; DEGAS 2: a transmission/escape probabilities model for neutral particle transport: comparison with DEGAS 2; a collisional radiative model of hydrogen for high recycling divertors; comparison of fluid and non- fluid neutral models in B2.5; DIII-D radiative divertor simulations; 3-D fluid simulations of turbulence from conducting wall mode; turbulence and drifts in SOL plasmas; recent results for 1 1/2-D ITER gas target divertor modelling; evaluation of pumping and fueling in coupled core, SOL, and divertor chamber calculations; and ITER gas target divertors: comparison of volume recombination and large radial transport scenarios using DEGAS

  19. Divertor characterization experiments

    International Nuclear Information System (INIS)

    Porter, G.D.; Allen, S.; Fenstermacher, M.; Hill, D.; Brown, M.; Jong, R.A.; Rognlien, T.; Rensink, M.; Smith, G.; Stambaugh, R.; Mahdavi, M.A.; Leonard, A.; West, P., Evans, T.

    1996-01-01

    Recent DIII-D experiments with enhanced Scrape-off Layer (SOL) diagnostics permit detailed characterization of the SOL and divertor plasma under various operating conditions. We observe two distinct plasma modes: attached and detached divertor plasmas. Detached plasmas are characterized by plate temperatures of only 1 to 2 eV. Simulation of detached plasmas using the UEDGE code indicate that volume recombination and charge exchange play an important role in achieving detachment. When the power delivered to the plate is reduced by enhanced radiation to the point that recycled neutrals can no longer be efficiently ionized, the plate temperature drops from around 10 eV to 1-2 eV. The low temperature region extends further off the plate as the power continues to be reduced, and charge exchange processes remove momentum, reducing the plasma flow. Volume recombination becomes important when the plasma flow is reduced sufficiently to permit recombination to compete with flow to the plate

  20. Study of the heavy impurity influx into the plasma using laser fluorescence spectroscopy in the TO-2 tokamak with toroidal divertor

    International Nuclear Information System (INIS)

    Vukolov, K.Yu.; Shvindt, N.N.

    1992-01-01

    Measurement cycle for determination of iron atom absolute concentrations was carried out in divertor and diaphragm modes of laser fluorescence spectroscopy. The conclusion is made on effective wall shielding by divertor layer as compared to material diaphragm. The basic result of the work consists in creating and testing on the tokamak TO-2 of multichannel diagnostic complex for remote measurement of atom (ion) absolute concentrations of metallic impurities in the near-wall plasma with high spatial and time resolution through laser fluorescence spectroscopy method intended for studies at the Tokamak-15 facility

  1. 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

  2. Towards the procurement of the ITER divertor

    International Nuclear Information System (INIS)

    Merola, M.; Tivey, R.; Martin, A.; Pick, M.

    2006-01-01

    The procurement of the ITER divertor is planned to start in 2009. On the basis of the present common understanding of the sharing of the ITER components, the Japanese Participating Team (JAPT) will supply the outer vertical target, the Russian Federation (RF) PT the dome liner and will perform the high heat flux testing, the EU PT will supply the inner vertical targets and the cassette bodies, including final assembly of the divertor plasma-facing components (PFCs). The manufacturing of the PFCs of the ITER divertor represents a challenging endeavor due to the high technologies which are involved, and due to the unprecedented series production. To mitigate the associated risks, special arrangements need to be put in place prior to and during procurement to ensure quality and to keep to the time schedule. Before procurement can start, an ITER review of the qualification and production capability of each candidate PT is planned. Well in advance of the assumed start of the procurement, each PT which would like to contribute to the divertor PFC procurement, should first demonstrate its technical qualification to carry out the procurement with the required quality, and in an efficient and timely manner. Appropriate precautions, like subdivision of the procurement into stages, are also to be adopted during the procurement phase to mitigate the consequences of possible unexpected manufacturing problems. In preparation for writing the procurement specification for the vertical targets, the topic of setting acceptance criteria is also being addressed. This activity has the objective of defining workable acceptance criteria for the PFC armour joints. A complete set of analyses is also in progress to assess the latest design modifications against the design requirements. This task includes neutronic, shielding, thermo-mechanical and electromagnetic analyses. More than half of the ITER plasma parameters that must be measured and the related diagnostics are located in the

  3. 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.)

  4. 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.

  5. 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.

  6. 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.

  7. 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)

  8. First results from the dynamic ergodic divertor at TEXTOR

    International Nuclear Information System (INIS)

    Lehnen, M.; Abdullaev, S.S.; Biel, W.; Brezinsek, S.; Finken, K.H.; Harting, D.; Hellermann, M. von; Jakubowski, M.; Jaspers, R.; Kobayashi, M.; Koslowski, H.R.; Kraemer-Flecken, A.; Matsunaga, G.; Pospieszczyk, A.; Reiter, D.; Van Rompuy, T.; Samm, U.; Schmitz, O.; Sergienko, G.; Unterberg, B.; Wolf, R.; Zimmermann, O.

    2005-01-01

    Experimental results from the dynamic ergodic divertor (DED) at TEXTOR are given, describing the complex structure of the edge plasma and the properties of the divertor as well as its influence on the plasma rotation

  9. Spatial radiation profiles in the ASDEX Upgrade divertor for detached plasmas

    International Nuclear Information System (INIS)

    Wenzel, U.; Thoma, A.; Dux, R.; Fuchs, C.; Herrmann, A.; Hirsch, S.; Kallenbach, A.; Kastelewicz, H.; Laux, M.; Mast, F.; Napiontek, B.

    1997-01-01

    In this paper we describe impurity line emission measurements in the divertor of ASDEX Upgrade during high power neutral beam heated discharges. We focus on detached conditions where the dominating part of the radiation comes from the X-point region. Spatially resolved line emission in the VUV and visible spectral region of the intrinsic carbon and additionally puffed impurities (neon and nitrogen) is presented. A simple interpretation of the line emission profiles is given and they are also compared to the results of bolometry. (orig.)

  10. Modelling of profile control with LH wave injection in the HL-2A single-null divertor plasma

    International Nuclear Information System (INIS)

    Gao Qingdi; Yuan Baoshan; Li Fangzhu; Wang, Aike; Budny, R.V.

    2005-01-01

    In the HL-2A tokamak a single-null divertor configuration has been established. The separatrix of the single-null diverted plasma was identified with a filament model, and the determined striking area on the target plate is in agreement with the measurements of electric probe array. Higher power LH wave (1.5MW) is injected to the diverted plasma with a nearly symmetric spectrum. Dominant electron heating and current profile control are investigated with numerical simulation. Plasma heating by electron Landau interaction results in operation scenarios of preferentially dominant electron heating. Due to the off-axis driven current, an optimized q-profile is formed, and an enhanced confinement regime with steep electron temperature gradient is produced. The clear decrease of the electron thermal conductivity in the LH power deposition region shows that an electron-ITB is developed. When higher LH power injects into the target plasma that is heated by NBI (0.5MW), the ion temperature has a large increment in addition to the high increase of electron temperature. The temperature profiles indicate that an enhanced core confinement is established with both ion-ITB and electron-ITB developed. (author)

  11. The trace ion module for the Monte Carlo code Eirene, a unified approach to plasma chemistry in the ITER divertor

    International Nuclear Information System (INIS)

    Seebacher, J.; Reiter, D.; Borner, P.

    2007-01-01

    Modelling of kinetic transport effects in magnetic fusion devices is of great importance for understanding the physical processes in both the core and and the scrape off layer (SOL) plasma. For SOL simulation the EIRENE code is a well established tool for modelling of neutral, impurities and radiation transport. Recently a new trace ion transport module (tim), has been developed and incorporated into EIRENE. The tim essentially consists of two parts: 1) A trajectory integrator tracing the deterministic motion of a guiding centre particle in general 3D electric and magnetic fields. 2) A stochastic representation of the Fokker Planck collision operator in suitable guiding centre coordinates treating Coulomb collisions with the plasma background species. The TIM enables integrated SOL simulation packages such as B2-EIRENE, EDGE2D-EIRENE (2D) or EMC3-EIRENE (3D) to treat the physical and chemical processes near the divertor targets and in the bulk of the SOL in greater detail than before, and in particular on a kinetic rather than a fluid level. One of the physics applications is the formation and transport of hydrocarbon molecules and ions in the divertor in tokamaks, where the tritium co deposition via hydrocarbons remains a serious issue for next generation fusion devices like ITER. Real tokamak modelling scenarios will be discussed with the code packages B2-EIRENE (2D) and EMC3-EIRENE (3D). A brief overview of the theoretical basis of the tim will be given including code verification studies of the basic physics properties. Applications to hydrocarbon transport studies in TEXTOR and ITER, comparing present (fluid) approximations in edge modelling with the new extended kinetic model, will be presented. (Author)

  12. Divertor design and its integration into the ITER-FEAT machine

    International Nuclear Information System (INIS)

    Janeschitz, G.; Antipenkov, A.; Federici, G.; Ibbott, C.; Kukushkin, A.; Ladd, P.; Martin, E.; Tivey, R.

    2001-01-01

    The physics of the edge and divertor plasma is strongly coupled with the divertor and the fuel cycle design. Due to the limited space available the design as well as the remote maintenance approach for the ITER divertor are highly optimized to allow maximum space for the divertor plasma. Several auxiliary systems (e.g. in vessel viewing, glow discharge electrodes...) as well as a part of the pumping and fuelling system have to be integrated together with the divertor into the lower level of the ITER machine. Two main options exist for the choice of the plasma-facing material in the divertor, i.e. W and CFC. Based on already existing R and D results one can be optimistic that the material choice will be mainly based on physics considerations and material issues (e.g. C-T co-deposition). The requirements for the ITER fuel cycle arise from plasma physics as well as from the envisaged operation scenarios. Due to the complex dynamic relationship of the fuel cycle subsystems among themselves and with the plasma, codes are employed for their optimization. This paper elaborates these interacting issues and gives the latest design status. (author)

  13. 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

    More and more 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 CO 2 -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 meterials have to be developed and qualified which are able to withstand the extreme environments in a commercial thermonuclear power reactor. The plasma facing materials (PEMs) 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. Materials for plasma facing components have to fulfill a number of requirements. First of all the materials have to be plasma compatible, i.e. they should exhibit a low atomic number to avoid radiative losses whenever atoms from the wall material will be ionized in the plasma. In addition, the materials must have a high melting point, a high thermal conductivity, and adequate mechanical properties. To select the most suitable material candidates, a comprehensive data base is required which includes all thermo-physical and mechanical properties. In present-day and next step devices the resulting thermal steady state heat loads to the first wall remain below 1 MWm -2 , meanwhile 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

  14. 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

    More and more 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 CO{sub 2}-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 meterials have to be developed and qualified which are able to withstand the extreme environments in a commercial thermonuclear power reactor. The plasma facing materials (PEMs) 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. Materials for plasma facing components have to fulfill a number of requirements. First of all the materials have to be plasma compatible, i.e. they should exhibit a low atomic number to avoid radiative losses whenever atoms from the wall material will be ionized in the plasma. In addition, the materials must have a high melting point, a high thermal conductivity, and adequate mechanical properties. To select the most suitable material candidates, a comprehensive data base is required which includes all thermo-physical and mechanical properties. In present-day and next step devices the resulting thermal steady state heat loads to the first wall remain below 1 MWm{sup -2}, meanwhile 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{sup -2} for next step tokamaks such as ITER or DEMO. These requirements are responsible for high demands on the selection of qualified PFMs

  15. Advanced divertor experiments on DIII-D

    International Nuclear Information System (INIS)

    Schaffer, M.J.; Mahdavi, M.A.; Osborne, T.; Petrie, T.W.; Stambaugh, R.D.; Buchenauer, D.; Hill, D.N.; Klepper, C.C.

    1991-01-01

    The poloidal divertor is presently favored for next-step, high-power tokamaks. The DIII-D Advanced Divertor Program (ADP) aims to gain increased control over the divertor plasma and tokamak boundary conditions. This paper reports experiments done in the first phase of the ADP. The DIII-D lower divertor was modified by the addition of a toroidally symmetric, graphite-armoured, water-cooled divertor-biasing ring electrode at the entrance to a gas plenum. (In the past DIII-D operated with an open divertor.) The plenum will eventually contain a He cryogenic loop for active divertor pumping. The separatrix 'strike' position is controlled by the lower poloidal field shaping coils and can be varied smoothly from the ring electrode upper surface to the divertor floor far from the entrance aperture. External power, at up to 550 V and 8 kA separately, has been applied to the electrode to date. (author) 5 refs., 5 figs

  16. TCV divertor upgrade for alternative magnetic configurations

    Directory of Open Access Journals (Sweden)

    H. Reimerdes

    2017-08-01

    Full Text Available The Swiss Plasma Center (SPC is planning a divertor upgrade for the TCV tokamak. The upgrade aims at extending the research of conventional and alternative divertor configurations to operational scenarios and divertor regimes of greater relevance for a fusion reactor. The main elements of the upgrade are the installation of an in-vessel structure to form a divertor chamber of variable closure and enhanced diagnostic capabilities, an increase of the pumping capability of the divertor chamber and the addition of new divertor poloidal field coils. The project follows a staged approach and is carried out in parallel with an upgrade of the TCV heating system. First calculations using the EMC3-Eirene code indicate that realistic baffles together with the planned heating upgrade will allow for a significantly higher compression of neutral particles in the divertor, which is a prerequisite to test the power dissipation potential of various divertor configurations.

  17. THERMAL HYDRAULIC ANALYSIS OF FIRE DIVERTOR

    International Nuclear Information System (INIS)

    C.B. bAXI; M.A. ULRICKSON; D.E. DRIMEYER; P. HEITZENROEDER

    2000-01-01

    The Fusion Ignition Research Experiment (FIRE) is being designed as a next step in the US magnetic fusion program. The FIRE tokamak has a major radius of 2 m, a minor radius of 0.525 m, and liquid nitrogen cooled copper coils. The aim is to produce a pulse length of 20 s with a plasma current of 6.6 MA and with alpha dominated heating. The outer divertor and baffle of FIRE are water cooled. The worst thermal condition for the outer divertor and baffle is the baseline D-T operating mode (10 T, 6.6 MA, 20 s) with a plasma exhaust power of 67 MW and a peak heat flux of 20 MW/m 2 . A swirl tape (ST) heat transfer enhancement method is used in the outer divertor cooling channels to increase the heat transfer coefficient and the critical heat flux (CHF). The plasma-facing surface consists of tungsten brush. The finite element (FE) analysis shows that for an inlet water temperature of 30 C, inlet pressure of 1.5 MPa and a flow velocity of 10 m/s, the incident critical heat flux is greater than 30 MW/m 2 . The peak copper temperature is 490 C, peak tungsten temperature is 1560 C, and the pressure drop is less than 0.5 MPa. All these results fulfill the design requirements

  18. 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.

  19. 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.

  20. 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.)

  1. Plasma spraying process of disperse carbides for spraying and facing

    International Nuclear Information System (INIS)

    Blinkov, I.V.; Vishnevetskaya, I.A.; Kostyukovich, T.G.; Ostapovich, A.O.

    1989-01-01

    A possibility to metallize carbides in plasma of impulsing capacitor discharge is considered. Powders granulation occurs during plasma spraying process, ceramic core being completely capped. X-ray phase and chemical analyses of coatings did not show considerable changes of carbon content in carbides before and after plasma processing. This distinguishes the process of carbides metallization in impulsing plasma from the similar processing in arc and high-frequency plasma generator. Use of powder composites produced in the impulsing capacitor discharge, for plasma spraying and laser facing permits 2-3 times increasing wear resistance of the surface layer as against the coatings produced from mechanical powders mixtures

  2. 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.

  3. The ITER divertor concept

    International Nuclear Information System (INIS)

    Janeschitz, G.; Borrass, K.; Federici, G.; Igitkhanov, Y.; Kukushkin, A.; Pacher, H.D.; Pacher, G.W.; Sugihara, M.

    1995-01-01

    The ITER divertor must exhaust most of the alpha particle power and the He ash at acceptable erosion rates. The high recycling regime of the ITER-CDA for present parameters would yield high power loads and erosion rates on conventional targets. Improvement by radiation in the SOL at constant pressure is limited in principle. To permit a higher radiation fraction, the plasma pressure along the field must be reduced by more than a factor 10, reducing also the target ion flux. This pressure reduction can be obtained by strong plasma-neutral interaction below the X-point. Under these conditions T e in the divertor can be reduced to <5 eV along a flame like ionisation front by impurity radiation and CX losses. Downstream of the front, neutrals undergo more CX or i-n collisions than ionisation events, resulting in significant momentum loss via neutrals to the divertor chamber wall. The pressure reduction by this mechanism depends on the along-field length for neutral-plasma interaction, the parallel power flux, the neutral density, the ratio of neutral-neutral collision length to the plasma-wall distance and on the Mach number of ions and neutrals. A supersonic transition in the main plasma-neutral interaction region, expected to occur near the ionisation front, would be beneficial for momentum removal. The momentum transfer fraction to the side walls is calculated: low Knudsen number is beneficial. The impact of the different physics effects on the chosen geometry and on the ITER divertor design and the lifetime of the various divertor components are discussed. ((orig.))

  4. Fast plasma shutdown by killer pellet injection in JT-60U with reduced heat flux on the divertor plate and avoiding runaway electron generation

    International Nuclear Information System (INIS)

    Yoshino, R.; Kondoh, T.; Neyatani, Y.; Itami, K.; Kawano, Y.; Isei, N.

    1997-01-01

    A killer pellet is an impurity pellet that is injected into a tokamak plasma in order to terminate a discharge without causing serious damage to the tokamak machine. In JT-60U neon ice pellets have been injected into OH and NB heated plasmas and fast plasma shutdowns have been demonstrated without large vertical displacement. The heat pulse on the divertor plate has been greatly reduced by killer pellet injections (KPI), but a low-power heat flux tail with a long time duration is observed. The total energy on the divertor plate increases with longer heat flux tail, so it has been reduced by shortening the tail. Runaway electron (RE) generation has been observed just after KPI and/or in the later phase of the plasma current quench. However, RE generation has been avoided when large magnetic perturbations are excited. These experimental results clearly show that KPI is a credible fast shutdown method avoiding large vertical displacement, reducing heat flux on the divertor plate, and avoiding (or minimizing) RE generation. (Author)

  5. 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.)

  6. 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

  7. Experimental study of the topological aspect of the ergodic divertor in Tore-supra tokamak; Etude experimentale des aspects topologiques du divertor ergodique de Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Costanzo, L

    2001-10-01

    The control of power deposition onto plasma facing components in tokamaks is a determining factor for future thermonuclear fusion reactors. Plasma surface interaction can be performed using limiters or divertors. The ergodic divertor installed on Tore Supra is an atypical example of a magnetic divertor. It consists in applying a magnetic perturbation which establishes a particular topology of the plasma in contact with the wall (edge plasma). We carried out dedicated experiments in order to study parallel heat flux which strike the divertor neutralizers. This quantitative and qualitative analysis of heat flux as a function of experimental conditions allows to determine the profiles of power deposition along the neutralizers. The influence of plasma electron density, additional heating, impurities and injected gas was established. An experimental study of the sheath heat transmission factor {gamma} was carried out by correlating measurements made with Langmuir probes and infrared imaging. This study gave rise to a major conclusion: for ohmic discharges with deuterium injection and most of the time with helium, it was experimentally confirmed that {gamma}=7 in agreement with classical sheath theory. However, an increase of this factor with additional power has been shown. Detached plasma, which is an attractive regime in order to reduce the power deposition, requires an optimized control. A new measurement of the detachment onset has been developed. It is based on the variation of heat flux onto the plates derived from infrared measurements. A detachment cartography with the determination of a new 2D 'IR' Degree of Detachment was carried out allowing to locate the zone where the detachment starts. We can apply this concept both to other tokamaks such as JET and ITER. A comparison between the axisymmetric divertor and the ergodic divertor is also presented concerning the power deposition in the two configurations. Low heat flux with the ergodic divertor is a

  8. 3. IAEA research co-ordination meeting on atomic and plasma-wall interaction data for fusion reactor divertor modeling. Summary report

    International Nuclear Information System (INIS)

    Janev, R.K.

    1999-04-01

    A brief description of the proceedings and the conclusions of the 3rd Research Co-ordination Meeting on 'Atomic and Plasma-Wall Interaction Data for Fusion Reactor Divertor Modeling', held on March 8-9, 1999, at the IAEA Headquarters in Vienna, Austria, is provided. The reports on the activities within the individual projects pertinent to the IAEA Co-ordinated Research program with the same title are given as appendix to the present report. (author)

  9. Light element atom, molecule and radical behaviour in the divertor and edge plasma regions. Summary report of the 1. research coordination meeting

    International Nuclear Information System (INIS)

    Braams, B.J.

    2010-01-01

    The first research coordination meeting of the Coordinated Research Project (CRP) on Light Element Atom, Molecule and Radical Behaviour in the Divertor and Edge Plasma Regions was held 18-20 November 2009 at IAEA headquarters, bringing together experts representing 14 institutions. Participants summarized their recent and ongoing work pertinent to the research project. The specific objectives of the CRP and a detailed work plan were formulated. The discussions, conclusions and recommendations of the meeting are summarized in this report. (author)

  10. CIT divertor conceptual design

    International Nuclear Information System (INIS)

    Wesley, J.C.; Sevier, D.L.

    1988-06-01

    A conceptual design of the divertor target assembly for the 1.75-m CIT baseline device has been developed. The divertor target assembly consists of four toroidal arrays of pyrolytic graphite plates that cover the inside surface of the ends of the vacuum vessel in the locations where the magnetic separatrices of the plasma intersect the vessel wall. During the course of the plasma discharge, the currents on the poloidal field coils that establish the plasma equilibrium are varied to sweep the separatrix strike locations across the divertor targets. This spreads the plasma heat loading over sufficient area to keep the peak target surface temperature within allowable limits. The required magnetic sweep (/+-/5 cm for the inside strike location and /+-/12 cm for the outside strike location) can be affected by programming either the external poloidal strike location) can be effected by programming either the external poloidal field (PF) coils or the internal PF control coils plus the external PF solenoid coils (PF1 and PF2). The ensuing variations in the elongation and triangularity of the plasma are modest, and fall within the ranges of plasma elongation and triangularity specified in the CIT General Requirements Document. 17 figs., 13 tabs

  11. 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.

  12. 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.)

  13. A new scaling for divertor detachment

    Science.gov (United States)

    Goldston, R. J.; Reinke, M. L.; Schwartz, J. A.

    2017-05-01

    The ITER design, and future reactor designs, depend on divertor ‘detachment,’ whether partial, pronounced or complete, to limit heat flux to plasma-facing components and to limit surface erosion due to sputtering. It would be valuable to have a measure of the difficulty of achieving detachment as a function of machine parameters, such as input power, magnetic field, major radius, etc. Frequently the parallel heat flux, estimated typically as proportional to P sep/R or P sep B/R, is used as a proxy for this difficulty. Here we argue that impurity cooling is dependent on the upstream density, which itself must be limited by a Greenwald-like scaling. Taking this into account self-consistently, we find the impurity fraction required for detachment scales dominantly as power divided by poloidal magnetic field. The absence of any explicit scaling with machine size is concerning, as P sep surely must increase greatly for an economic fusion system, while increases in the poloidal field strength are limited by coil technology and plasma physics. This result should be challenged by comparison with 2D divertor codes and with measurements on existing experiments. Nonetheless, it suggests that higher magnetic field, stronger shaping, double-null operation, ‘advanced’ divertor configurations, as well as alternate means to handle heat flux such as metallic liquid and/or vapor targets merit greater attention.

  14. 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

  15. Three-dimensional simulation of H-mode plasmas with localized divertor impurity injection on Alcator C-Mod using the edge transport code EMC3-EIRENE

    International Nuclear Information System (INIS)

    Lore, J. D.; Reinke, M. L.; Lipschultz, B.; Brunner, D.; LaBombard, B.; Terry, J.; Pitts, R. A.; Feng, Y.

    2015-01-01

    Experiments in Alcator C-Mod to assess the level of toroidal asymmetry in divertor conditions resulting from poloidally and toroidally localized extrinsic impurity gas seeding show a weak toroidal peaking (∼1.1) in divertor electron temperatures for high-power enhanced D-alpha H-mode plasmas. This is in contrast to similar experiments in Ohmically heated L-mode plasmas, which showed a clear toroidal modulation in the divertor electron temperature. Modeling of these experiments using the 3D edge transport code EMC3-EIRENE [Y. Feng et al., J. Nucl. Mater. 241, 930 (1997)] qualitatively reproduces these trends, and indicates that the different response in the simulations is due to the ionization location of the injected nitrogen. Low electron temperatures in the private flux region (PFR) in L-mode result in a PFR plasma that is nearly transparent to neutral nitrogen, while in H-mode the impurities are ionized in close proximity to the injection location, with this latter case yielding a largely axisymmetric radiation pattern in the scrape-off-layer. The consequences for the ITER gas injection system are discussed. Quantitative agreement with the experiment is lacking in some areas, suggesting potential areas for improving the physics model in EMC3-EIRENE

  16. The influence of plasma-surface interaction on the performance of tungsten at the ITER divertor vertical targets

    Science.gov (United States)

    De Temmerman, G.; Hirai, T.; Pitts, R. A.

    2018-04-01

    The tungsten (W) material in the high heat flux regions of the ITER divertor will be exposed to high fluxes of low-energy particles (e.g. H, D, T, He, Ne and/or N). Combined with long-pulse operations, this implies fluences well in excess of the highest values reached in today’s tokamak experiments. Shaping of the individual monoblock top surface and tilting of the vertical targets for leading-edge protection lead to an increased surface heat flux, and thus increased surface temperature and a reduced margin to remain below the temperature at which recrystallization and grain growth begin. Significant morphology changes are known to occur on W after exposure to high fluences of low-energy particles, be it H or He. An analysis of the formation conditions of these morphology changes is made in relation to the conditions expected at the vertical targets during different phases of operations. It is concluded that both H and He-related effects can occur in ITER. In particular, the case of He-induced nanostructure (also known as ‘fuzz’) is reviewed. Fuzz formation appears possible over a limited region of the outer vertical target, the inner target being generally a net Be deposition area. A simple analysis of the fuzz growth rate including the effect of edge-localized modes (ELMs) and the reduced thermal conductivity of fuzz shows that the fuzz thickness is likely to be limited by the occurrence of annealing during ELM-induced thermal excursions. Not only the morphology, but the material mechanical and thermal properties can be modified by plasma exposure. A review of the existing literature is made, but the existing data are insufficient to conclude quantitatively on the importance and extent of these effects for ITER. As a consequence of the high surface temperatures in ITER, W recrystallization is an important effect to consider, since it leads to a decrease in material strength. An approach is proposed here to develop an operational budget for the W material, i

  17. A Lithium Vapor Box Divertor Similarity Experiment

    Science.gov (United States)

    Cohen, Robert A.; Emdee, Eric D.; Goldston, Robert J.; Jaworski, Michael A.; Schwartz, Jacob A.

    2017-10-01

    A lithium vapor box divertor offers an alternate means of managing the extreme power density of divertor plasmas by leveraging gaseous lithium to volumetrically extract power. The vapor box divertor is a baffled slot with liquid lithium coated walls held at temperatures which increase toward the divertor floor. The resulting vapor pressure differential drives gaseous lithium from hotter chambers into cooler ones, where the lithium condenses and returns. A similarity experiment was devised to investigate the advantages offered by a vapor box divertor design. We discuss the design, construction, and early findings of the vapor box divertor experiment including vapor can construction, power transfer calculations, joint integrity tests, and thermocouple data logging. Heat redistribution of an incident plasma-based heat flux from a typical linear plasma device is also presented. This work supported by DOE Contract No. DE-AC02-09CH11466 and The Princeton Environmental Institute.

  18. Analysis of particle transport in a gas target divertor

    Energy Technology Data Exchange (ETDEWEB)

    Ohtsu, Shigeki; Tanaka, Satoru [Tokyo Univ. (Japan). Faculty of Engineering

    1996-10-01

    2-dimensional modelling of divertor plasma was performed with three types of the divertor geometry configuration. Pumping is effective to reduce neutral recycling to core region in the configuration without baffle. In baffle configuration, a good shielding of neutrals in the divertor region can be achieved. The dome configuration reduces plasma density near the null region and flow shear near the separatrix. (author)

  19. 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.

  20. 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

  1. Operating windows of pebble divertor

    International Nuclear Information System (INIS)

    Matsuhiro, K.; Isobe, M.; Ohtsuka, Y.; Ueda, Y.; Nishikawa, M.

    2001-01-01

    A marked feature of the pebble divertor is an effect by use of functional multi-layer coated pebble, which consists of a surface plasma facing layer, an intermediate tritium permeation barrier layer, and a kernel for heat removal. The dimensions, structure and the irradiation conditions of pebbles are the important issues for the development of the pebble divertor. From the view point of resistance of the induced thermal stress, the pebble is taken as small as possible in size. On the other hand, from the view point of the pumping performance, the suitable irradiation temperature range of the surface layer of pebble was estimated from the experiments and the numerical analysis. The pumping process enhanced by dynamic retention is available to extend the higher allowable irradiation temperature range from 900K to 1100K. As taking the temperature rise limitation due to pumping effect and the fractural strength due to the induced thermal stress limitation, it was found that the diameter of the pebble is possible to be 1-2 mm in about 20 MW/m 2 for the SiC kernel and 2-3 mm in less than 30 MW/m 2 for the graphite kernel. (author)

  2. Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake

    International Nuclear Information System (INIS)

    Kotschenreuther, Mike; Valanju, Prashant; Covele, Brent; Mahajan, Swadesh

    2013-01-01

    Advanced divertors are magnetic geometries where a second X-point is added in the divertor region to address the serious challenges of burning plasma power exhaust. Invoking physical arguments, numerical work, and detailed model magnetic field analysis, we investigate the magnetic field structure of advanced divertors in the physically relevant region for power exhaust—the scrape-off layer. A primary result of our analysis is the emergence of a physical “metric,” the Divertor Index DI, which quantifies the flux expansion increase as one goes from the main X-point to the strike point. It clearly separates three geometries with distinct consequences for divertor physics—the Standard Divertor (DI = 1), and two advanced geometries—the X-Divertor (XD, DI > 1) and the Snowflake (DI < 1). The XD, therefore, cannot be classified as one variant of the Snowflake. By this measure, recent National Spherical Torus Experiment and DIIID experiments are X-Divertors, not Snowflakes

  3. Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake

    Energy Technology Data Exchange (ETDEWEB)

    Kotschenreuther, Mike; Valanju, Prashant; Covele, Brent; Mahajan, Swadesh [Institute for Fusion Studies, The University of Texas at Austin, Austin, Texas 78712 (United States)

    2013-10-15

    Advanced divertors are magnetic geometries where a second X-point is added in the divertor region to address the serious challenges of burning plasma power exhaust. Invoking physical arguments, numerical work, and detailed model magnetic field analysis, we investigate the magnetic field structure of advanced divertors in the physically relevant region for power exhaust—the scrape-off layer. A primary result of our analysis is the emergence of a physical “metric,” the Divertor Index DI, which quantifies the flux expansion increase as one goes from the main X-point to the strike point. It clearly separates three geometries with distinct consequences for divertor physics—the Standard Divertor (DI = 1), and two advanced geometries—the X-Divertor (XD, DI > 1) and the Snowflake (DI < 1). The XD, therefore, cannot be classified as one variant of the Snowflake. By this measure, recent National Spherical Torus Experiment and DIIID experiments are X-Divertors, not Snowflakes.

  4. Design and analysis of the DII-D radiative divertor water-cooled structures

    International Nuclear Information System (INIS)

    Hollerbach, M.A.; Smith, J.P.; Baxi, C.B.; Bozek; Chin, E.; Phelps, R.D.; Redler, K.M.; Reis, E.E.

    1995-10-01

    The Radiative Divertor is a major modification to the divertor of DIII-D and is being designed and fabricated for installation in late 1996. The Radiative Divertor Program (RDP) will enhance the dissipative processes in the edge and divertor plasmas to reduce the heat flux and plasma erosion at the divertor target. This approach will have major implications for the heat removal methods used in future devices. The divertor is of slot-type configuration designed to minimize the flow of sputtered and injected impurities back to the core plasma. The new divertor will be composed of toroidally continuous, Inconel 625 water-cooled rings of sandwich construction with an internal water channel, incorporating seam welding to provide the water-to-vacuum seal as well as structural integrity. The divertor structure is designed to withstand electromagnetic loads as a result of halo currents and induced toroidal currents. It also accommodates the thermal differences experienced during the 400 degrees C bake used on DIII-D. A low Z plasma-facing surface is provided by mechanically attached graphite tiles. Water flow through the rings will inertially cool these tiles which will be subjected to 38 MW, 10 second pulses. Current schedules call for detailed design in 1996 with installation completed in March 1997. A full size prototype, one-quarter of one ring, is being built to validate manufacturing techniques, machining, roll-forming, and seam welding. The experience and knowledge gained through the fabrication of the prototype is discussed. The design of the electrically isolated (5 kV) vacuum-to-air water feedthroughs supplying the water-cooled rings is also discussed

  5. Design and analysis of the DIII-D radiative divertor water-cooled structures

    International Nuclear Information System (INIS)

    Hollerbach, M.A.; Smith, J.P.; Baxi, C.B.; Bozek, A.S.; Chin, E.; Phelps, R.D.; Redler, K.M.; Reis, E.E.

    1995-01-01

    The Radiative Divertor is a major modification to the divertor of DIII-D and is being designed and fabricated for installation in late 1996. The Radiative Divertor Program (RDP) will enhance the dissipative processes in the edge and divertor plasmas to reduce the heat flux and plasma erosion at the divertor target. This approach will have major implications for the heat removal methods used in future devices. The divertor is of slot-type configuration designed to minimize the flow of sputtered and injected impurities back to the core plasma. The new divertor will be composed of toroidally continuous, Inconel 625 water-cooled rings of sandwich construction with an internal water channel, incorporating seam welding to provide the water-to-vacuum seal as well as structural integrity. The divertor structure is designed to withstand electro-magnetic loads as a result of halo currents and induced toroidal currents. It also accommodates the thermal differences experienced during the 400 C bake used on DIII-D. A low Z plasma-facing surface is provided by mechanically attached graphite tiles. Water flow through the rings will inertially cool these tiles which will be subjected to 38 MW, 10 second pulses. Current schedules call for detailed design in 1996 with installation completed in March 1997. A full size prototype, one-quarter of one ring, is being built to validate manufacturing techniques, machining, roll-forming, and seam welding. The experience and knowledge gained through the fabrication of the prototype is discussed. The design of the electrically isolated (5 kV) vacuum-to-air water feedthroughs supplying the water-cooled rings is also discussed

  6. Advanced divertor experiments on DIII-D

    International Nuclear Information System (INIS)

    Schaffer, M.J.; Mahdavi, M.A.; Osborne, T.; Petrie, T.W.; Stambaugh, R.D.; Buchenauer, D.; Hill, D.N.; Klepper, C.C.

    1991-04-01

    The poloidal divertor is presently favored for next-step, high-power tokamaks. The DIII-D Advanced Divertor Program (ADP) aims to gain increased control over the divertor plasma and tokamak boundary conditions. This paper reports experiments done in the first phase of the ADP. The DIII-D lower divertor was modified by the addition of a toroidally symmetric, graphite-armoured, water-cooled divertor-biasing ring electrode at the entrance to a gas plenum. The plenum will eventually contain a He cryogenic loop for active divertor pumping. The separatrix ''strike'' position is controlled by the lower poloidal field shaping coils and can be varied smoothly from the ring electrode upper surface to the divertor floor far from the entrance aperture. External power, at up to 550 V and 8 kA separately, has been applied to the electrode to date. 5 refs., 5 figs

  7. Mechanical design issues associated with mounting, maintenance, and handling of an ITER divertor

    International Nuclear Information System (INIS)

    Goranson, D.L.; Fogarty, D.J.; Jones, G.H.

    1992-01-01

    Several designs that address plasma-facing plate configurations and thermal-hydraulic design issues have been developed for the ITER divertor. Design criteria growing out of physics requirements, physical constraints, and remote handling requirements impose severe mechanical requirements on the support structure and its attachments. These pose a challenge to the mechanical design of a divertor, which must be addressed before a functional divertor is practical that is, one that can be remotely handled, aligned, and maintained; that functions reliably under thermal loading and disruptions; and that gives the required life in the nuclear environment predicted for ITER. This paper discusses the design criteria for the divertor mounting structure and identifies the mechanical design issues that need to be addressed

  8. Engineering design of a radiative divertor for DIII-D

    International Nuclear Information System (INIS)

    Smith, J.P.; Baxi, C.B.; Bozek, A.S.

    1995-10-01

    A new divertor configuration is being developed for the DIII-D tokamak. This divertor will operate in the radiative mode. Experiments and modeling form the basis for the new design. The Radiative Divertor reduces the heat flux on the divertor plates by dispersing the power with radiation in the divertor region. In addition, the Radiative Divertor structure will allow density control in plasma shapes required for advanced tokamak operation. The divertor structure allows for operation in either double-null or single-null plasma configurations. Four independently controlled divertor cryopumps will enable pumping at either the inboard (upper and lower) or the outboard (upper and lower) divertor plates. An upgrade to the DIII-D cryogenic system is part of this project. The increased capabilities of the cryogenic system will allow delivery of liquid helium and nitrogen to the three new cryopumps. The Radiative Divertor design is very flexible, and will allow physics studies of the effects of slot width and length. Radiative Divertor diagnostics are being designed in parallel to provide comprehensive measurements for diagnosing the divertor. The Radiative divertor installation is scheduled for late 1996. Engineering experience gained in the DIII-D Advanced Divertor program form a foundation for the design work on the Radiative Divertor

  9. Mechanical Design of the NSTX Liquid Lithium Divertor

    Energy Technology Data Exchange (ETDEWEB)

    R. Ellis, R. Kaita, H. Kugel, G. Paluzzi, M. Viola and R. Nygren

    2009-02-19

    The Liquid Lithium Divertor (LLD) on NSTX will be the first test of a fully-toroidal liquid lithium divertor in a high-power magnetic confinement device. It will replace part of the lower outboard divertor between a specified inside and outside radius, and ultimately provide a lithium surface exposed to the plasma with enough depth to absorb a significant particle flux. There are numerous technical challenges involved in the design. The lithium layer must be as thin as possible, and maintained at a temperature between 200 and 400 degrees Celsius to minimize lithium evaporation. This requirement leads to the use of a thick copper substrate, with a thin stainless steel layer bonded to the plasma-facing surface. A porous molybdenum layer is then plasma-sprayed onto the stainless steel, to provide a coating that facilitates full wetting of the surface by the liquid lithium. Other challenges include the design of a robust, vacuumcompatible heating and cooling system for the LLD. Replacement graphite tiles that provided the proper interface between the existing outer divertor and the LLD also had to be designed, as well as accommodation for special LLD diagnostics. This paper describes the mechanical design of the LLD, and presents analyses showing the performance limits of the LLD.

  10. Mechanical Design of the NSTX Liquid Lithium Divertor

    International Nuclear Information System (INIS)

    Ellis, R.; Kaita, R.; Kugel, H.; Paluzzi, G.; Viola, M.; Nygren, R.

    2009-01-01

    The Liquid Lithium Divertor (LLD) on NSTX will be the first test of a fully-toroidal liquid lithium divertor in a high-power magnetic confinement device. It will replace part of the lower outboard divertor between a specified inside and outside radius, and ultimately provide a lithium surface exposed to the plasma with enough depth to absorb a significant particle flux. There are numerous technical challenges involved in the design. The lithium layer must be as thin as possible, and maintained at a temperature between 200 and 400 degrees Celsius to minimize lithium evaporation. This requirement leads to the use of a thick copper substrate, with a thin stainless steel layer bonded to the plasma-facing surface. A porous molybdenum layer is then plasma-sprayed onto the stainless steel, to provide a coating that facilitates full wetting of the surface by the liquid lithium. Other challenges include the design of a robust, vacuum compatible heating and cooling system for the LLD. Replacement graphite tiles that provided the proper interface between the existing outer divertor and the LLD also had to be designed, as well as accommodation for special LLD diagnostics. This paper describes the mechanical design of the LLD, and presents analyses showing the performance limits of the LLD.

  11. Surface roughness effects on plasma near a divertor plate and local impact angle

    Directory of Open Access Journals (Sweden)

    Wanpeng Hu

    2017-08-01

    Full Text Available The impact of rough surface topography on the electric potential and electric field is generally neglected due to the small scale of surface roughness compared to the width of the plasma sheath. However, the distributions of the electric potential and field on rough surfaces are expected to influence the characteristics of edge plasma and the local impact angle. The distributions of plasma sheath and local impact angle on rough surfaces are investigated by a two dimension-in-space and three dimension-in-velocity (2d3v Particle-In-Cell (PIC code. The influences of the plasma temperature andsurface morphology on the plasma sheath, local impact angle and resulting physical sputtering yield on rough surfaces are investigated.

  12. Lifetime evaluation of first wall and divertor plate by crack analyses during plasma disruptions

    International Nuclear Information System (INIS)

    Ohmori, Junji; Kobayashi, Takeshi; Yamada, Masao; Iida, Hiromasa

    1988-05-01

    The first wall and divertor armor in fusion devices are subjected to high heat and particle fluxes. In particular, disruption heating is an intense thermal shock which may cause melting or vaporization of the armor surfaces. The behavior of the armor materials is one of the major factors limiting the lifetime of these components. Generally the surface temperature of armor due to disruption gets so high that the surface may become cracked. However, even if only the surface of the armor is cracked, the function of the armor will not be lost as long as the damage is limited to within a small depth of the surface. In this study, the lifetime of the armor is evaluated by two stages: crack initiation life and crack propagation life which are related to the fatigue life and the energy release rate, respectively. Materials are graphite and C/C composite (carbon fiber reinforced carbon composite) for the first wall, and tungsten for the dinertor. For disruption conditions of Fusion Experimental Reactor, the fatigue life and the energy release rates are calculated by thermal, and stress analyses. Results show that crack initiation is expected after only a few disruptions, and the energy release rate as a function of the crack length comes up to the maximum value at a small crack length, and decreases with increasing of the crack length. This decreasing means that a crack propagation rate reduces. An unstable fracture does not occur if the maximum energy release rate does not exceed the critical energy release rate which can be obtained from the fracture toughness. (author)

  13. 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.)

  14. Connexion topology and sol physics induced by the ergodic divertor in Tore Supra

    International Nuclear Information System (INIS)

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

    1990-01-01

    The diffuse connexion induced by the ergodic divertor in Tore Supra leads to heat load patterns on the plasma facing components. The topology of those patterns is analysed using field line tracing and analytical derivation of effective heat transport coefficients. Control of the heat load via the magnetic configuration is investigated

  15. Behavior of plasma facing surface in the large helical device

    International Nuclear Information System (INIS)

    Hino, T.; Nobuta, Y.; Sagara, A.

    2002-01-01

    Material probes have been installed at the inner walls along poloidal direction in LHD from the first experimental campaign. After each campaign, the impurity deposition and the gas retention have been examined to clarify the plasma surface interaction and the degree of wall cleaning. In the 2nd campaign, the entire wall was considerably cleaned by helium glow discharge conditionings. For the 3rd and 4th campaigns, graphite tiles were installed at entire divertor strike region, and then the wall condition significantly changed compared to the case of stainless steel wall. The erosion of graphite took place during the main discharges and the eroded carbon deposited on the entire wall. In particular, the deposition thickness was large at the wall far from the plasma. Since the entire wall was well carbonized, amount of retained discharge gas such as H and He became large. In particular, the helium retention was large at the position close to the anodes used for helium glow discharge cleanings. One characteristics of the LHD wall is a large retention of helium gas since the wall temperature is limited below 368 K. In order to reduce the recycling of discharge gas, the wall heating before the experimental campaign and the surface heating between the main discharge shots are planned. (author)

  16. Behavior of plasma facing surface in the large helical device

    Energy Technology Data Exchange (ETDEWEB)

    Hino, T.; Nobuta, Y. [Hokkaido Univ., Dept. of Nuclear Engineering, Sapporo, Hokkaido (Japan); Sagara, A. [National Inst. for Fusion Science, Toki, Gifu (Japan)] [and others

    2002-11-01

    Material probes have been installed at the inner walls along poloidal direction in LHD from the first experimental campaign. After each campaign, the impurity deposition and the gas retention have been examined to clarify the plasma surface interaction and the degree of wall cleaning. In the 2nd campaign, the entire wall was considerably cleaned by helium glow discharge conditionings. For the 3rd and 4th campaigns, graphite tiles were installed at entire divertor strike region, and then the wall condition significantly changed compared to the case of stainless steel wall. The erosion of graphite took place during the main discharges and the eroded carbon deposited on the entire wall. In particular, the deposition thickness was large at the wall far from the plasma. Since the entire wall was well carbonized, amount of retained discharge gas such as H and He became large. In particular, the helium retention was large at the position close to the anodes used for helium glow discharge cleanings. One characteristics of the LHD wall is a large retention of helium gas since the wall temperature is limited below 368 K. In order to reduce the recycling of discharge gas, the wall heating before the experimental campaign and the surface heating between the main discharge shots are planned. (author)

  17. Behavior of plasma facing surfaces in the large helical device

    International Nuclear Information System (INIS)

    Hino, T.; Nobuta, Y.; Sagara, A.

    2003-01-01

    Material probes have been installed at the inner walls along the poloidal direction in LHD from the first experimental campaign. After each campaign, the impurity deposition and the gas retention have been examined to clarify the plasma surface interaction and the degree of wall cleaning. In the 2nd campaign, the entire wall was thoroughly cleaned by helium glow discharge conditioning. For the 3rd and 4th campaigns, graphite tiles were installed over the entire divertor strike region, and then the wall condition was significantly changed compared to the case of a stainless steel wall. Graphite erosion took place during the main discharges and the eroded carbon was deposited on the entire wall. In particular, the deposition thickness was large at the wall far from the plasma. Since the entire wall was well carbonized, the amount of retained discharge gases such as H and He became large. In particular, the helium retention was large at the position close to the anodes used for helium glow discharge cleanings. One characteristic of the LHD wall is a large retention of helium gas since the wall temperature is limited to below 368 K. In order to reduce the recycling of discharge gas, wall heating before the experimental campaign and surface heating between the main discharge shots are planned. (author)

  18. Behavior of plasma facing surface in the large helical device

    International Nuclear Information System (INIS)

    Hino, T.; Nobuta, Y.; Sagara, A.

    2002-10-01

    Material probes have been installed at the inner walls along poloidal direction in LHD from the first experimental campaign. After each the campaign, the impurity deposition and the gas retention have been examined to clarify the plasma surface interaction and the degree of wall cleaning. In the 2nd campaign, the entire wall was considerably cleaned by helium glow discharge conditionings. For the 3rd and 4th campaigns, graphite tiles were installed at entire divertor strike region, and then the wall condition significantly changed compared to the case of stainless steel wall. The erosion of graphite took place during the main discharges and the eroded carbon deposited on the entire wall. In particular, the deposition thickness was large at the wall far from the plasma. Since the entire wall was well carbonized, amount of retained discharge gas such as H and He became large. In particular, the helium retention was large at the position close to the anodes used for helium glow discharge cleanings. One characteristics of the LHD wall is a large retention of helium gas since the wall temperature is limited below 368 K. In order to reduce the recycling of discharge gas, the wall heating before the experimental campaign and the surface heating between the main discharge shots are planned. (author)

  19. Model for screening of resonant magnetic perturbations by plasma in a realistic tokamak geometry and its impact on divertor strike points

    Czech Academy of Sciences Publication Activity Database

    Cahyna, Pavel; Nardon, E.

    2011-01-01

    Roč. 415, č. 1 (2011), S927-S931 ISSN 0022-3115. [International Conference on Plasma-Surface Interactions in Controlled Fusion Device/19th./. San Diego, 24.05.2010-28.05.2010] R&D Projects: GA MŠk 7G09042; GA MŠk LA08048 Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamaks * ELM control * resonant magnetic perturbations * divertor Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.052, year: 2011 http://dx.doi.org/10.1016/j.jnucmat.2011.01.117

  20. 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.

  1. VUV Spectroscopy in DIII-D Divertor

    International Nuclear Information System (INIS)

    Alkesh Punjabi; Nelson Jalufka

    2004-01-01

    The research carried out on this grant was motivated by the high power emission from the CIV doublet at 155 nm in the DIII-D divertor and to study the characteristics of the radiative divertor. The radiative divertor is designed to reduce the heat load to the target plates of the divertor by reducing the energy in the divertor plasma using upstream scrape-off-layer (SOL) radiation. In some cases, particularly in Partially Detached Divertor (PDD) operations, this emission accounts for more than 50% of the total radiation from the divertor. In PDD operation, produced by neutral gas injection, the particle flow to the target plate and the divertor temperature are significantly reduced. A father motivation was to study the CIV emission distribution in the lower, open divertor and the upper baffled divertor. Two Vacuum Ultra Violet Tangential viewing Television cameras (VUV TTV) were constructed and installed in the upper, baffled and the lower, open divertor. The images recorded by these cameras were then inverted to produce two-dimensional distributions of CIV in the poloidal plane. Results obtained in the project are summarized in this report

  2. 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

  3. Snowflake divertor configuration studies in National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V. A.; McLean, A. G.; Rognlien, T. D.; Ryutov, D. D.; Umansky, M. V. [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States); Bell, R. E.; Diallo, A.; Gerhardt, S.; Kaye, S.; Kolemen, E.; LeBlanc, B. P.; Menard, J. E.; Paul, S. F.; Podesta, M.; Roquemore, A. L.; Scotti, F.; Battaglia, D.; Bell, M. G.; Gates, D. A.; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); and others

    2012-08-15

    Experimental results from NSTX indicate that the snowflake divertor (D. Ryutov, Phys. Plasmas 14, 064502 (2007)) may be a viable solution for outstanding tokamak plasma-material interface issues. Steady-state handling of divertor heat flux and divertor plate erosion remains to be critical issues for ITER and future concept devices based on conventional and spherical tokamak geometry with high power density divertors. Experiments conducted in 4-6 MW NBI-heated H-mode plasmas in NSTX demonstrated that the snowflake divertor is compatible with high-confinement core plasma operation, while being very effective in steady-state divertor heat flux mitigation and impurity reduction. A steady-state snowflake divertor was obtained in recent NSTX experiments for up to 600 ms using three divertor magnetic coils. The high magnetic flux expansion region of the scrape-off layer (SOL) spanning up to 50% of the SOL width {lambda}{sub q} was partially detached in the snowflake divertor. In the detached zone, the heat flux profile flattened and decreased to 0.5-1 MW/m{sup 2} (from 4-7 MW/m{sup 2} in the standard divertor) indicative of radiative heating. An up to 50% increase in divertor, P{sub rad} in the snowflake divertor was accompanied by broadening of the intrinsic C III and C IV radiation zones, and a nearly order of magnitude increase in divertor high-n Balmer line emission indicative of volumetric recombination onset. Magnetic reconstructions showed that the x-point connection length, divertor plasma-wetted area and divertor volume, all critical parameters for geometric reduction of deposited heat flux, and increased volumetric divertor losses were significantly increased in the snowflake divertor, as expected from theory.

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

    Science.gov (United States)

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

    2013-11-01

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

  5. Evolution of transiently melt damaged tungsten under ITER-relevant divertor plasma heat loading

    Energy Technology Data Exchange (ETDEWEB)

    Bardin, S., E-mail: s.bardin@differ.nl [FOM Institute DIFFER – Dutch Institute For Fundamental Energy Research, Ass EURATOM-FOM, Trilateral Euregio Cluster, Nieuwegein (Netherlands); Morgan, T.W. [FOM Institute DIFFER – Dutch Institute For Fundamental Energy Research, Ass EURATOM-FOM, Trilateral Euregio Cluster, Nieuwegein (Netherlands); Glad, X. [Université de Lorraine, Institut Jean Lamour, Vandoeuvre-les-Nancy (France); Pitts, R.A. [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); De Temmerman, G. [FOM Institute DIFFER – Dutch Institute For Fundamental Energy Research, Ass EURATOM-FOM, Trilateral Euregio Cluster, Nieuwegein (Netherlands); ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

    2015-08-15

    A high-repetition-rate ELM simulation system was used at both the Pilot-PSI and Magnum-PSI linear plasma devices to investigate the nature of W damage under multiple shallow melt events and the subsequent surface evolution under ITER relevant plasma fluence and high ELM number. First, repetitive shallow melting of two W monoblocks separated by a 0.5 mm gap was obtained by combined pulsed/steady-state hydrogen plasma loading at normal incidence in the Pilot-PSI device. Surface modifications including melting, cracking and strong net-reshaping of the surface are obtained. During the second step, the pre-damaged W sample was exposed to a high flux plasma regime in the Magnum-PSI device with a grazing angle of 35°. SEM analysis indicates no measurable change to the surface state after the exposure in Magnum-PSI. An increase in transient-induced temperature rise of 40% is however observed, indicating a degradation of thermal properties over time.

  6. Studies on divertor effects by means of the Doublet-III high-temperature plasma device

    International Nuclear Information System (INIS)

    Shimada, Michiya

    1982-12-01

    The diverter action on impurity removal, helium ash compression and radiative cooling was studied in Doublet-3, placing emphasis on the applicability to reacting plasma grade devices such as Intor. The following principal results were obtained with a single-null poloidal diverter without the diverter chamber and the diverter throat (referred to as ''open diverter''), and the diverter coils being installed outside the vacuum chamber. The diverter reduced metallic impurities in the central plasma volume, carbon influx and radiation loss, and changed a typically peaked radiation power profile to a hollow profile. In helium-seeded diverter discharge, helium gas pressure near the diverter rose with the increase of main plasma density, and the pressure was high enough to demonstrate the possibility of helium ash exhaust in a diverted tokamak. The radiation power in the diverter volume significantly increased with the increasing main plasma density to as much as 50 % of the input ohmic power. The remote radiation cooling reduced the thermal load on the diverter plate, and the electron temperature near the diverter plate was cooled down. The source of this remote radiative cooling power was the mixture of line radiation of hydrogen neutral and oxygen. (Kako, I.)

  7. The influence of electric fields and neutral particles on the plasma sheath at ITER divertor conditions

    NARCIS (Netherlands)

    Shumack, A.E.

    2011-01-01

    The purpose of this thesis is to support the optimization of the ‘exhaust-pipe’, or so-called ‘divertor’, of the nuclear fusion experiment ITER, a large international fusion reactor now under construction in the south of France. We focus particularly on two ‘tools’ for optimization of the plasma

  8. 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)

  9. 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.

  10. Plasma-surface interactions under high heat and particle fluxes

    NARCIS (Netherlands)

    De Temmerman, G.; Bystrov, K.; Liu, F.; Liu, W.; Morgan, T.; Tanyeli, I.; van den Berg, M.; Xu, H.; Zielinski, J.

    2013-01-01

    The plasma-surface interactions expected in the divertor of a future fusion reactor are characterized by extreme heat and particle fluxes interacting with the plasma-facing surfaces. Powerful linear plasma generators are used to reproduce the expected plasma conditions and allow plasma-surface

  11. 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.

  12. Overview of the divertor design and its integration into RTO/RC-ITER

    International Nuclear Information System (INIS)

    Janeschitz, G.; Tivey, R.; Antipenkov, A.; Barabash, V.; Chiocchio, S.; Federici, G.; Heidl, H.; Ibbott, C.; Martin, E.

    2000-01-01

    channel between the plasma facing components (PFC) and the cassette body (P. Ladd, C. Ibbott, G. Janeschitz, E. Martin, Design of the RTO/RC-ITER primary pumping system, this conference). This gives a somewhat better conductance from the private flux region to the pumping ports than in the previous design. Diagnostic access in the divertor now also uses the cut-outs provided for pumping instead of the space below the cassettes. Developments, in particular in the area of the PFCs, aimed at reducing the cost of the divertor are reported in C. Ibbott, A. Antipenkov, S. Chiocchio, G. Federici, H. Heidl, G. Janeschitz, E. Martin, R. Tivey, Design issues and cost implications of RTO/RC-ITER divertor, this conference

  13. 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.)

  14. 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.)

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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

  1. 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.

  2. 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)

  3. 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

  4. 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.

  5. 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.

  6. 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.

  7. Neutral particle retention in the JET MK I divertor

    International Nuclear Information System (INIS)

    Ehrenberg, J.K.; Campbell, D.J.; Harbour, P.J.; Horton, L.D.; Loarte, A.; McCormick, G.K.; Monk, R.D.; Saibene, G.R.; Simonini, R.; Taroni, A.; Stamp, M.F.

    1997-01-01

    Retention of neutral deuterium and nitrogen in the JET MK I divertor has been investigated. Results show that ohmic plasma detachment reduces deuterium retention, that the magnetic divertor configuration has some influence on the achievable deuterium retention, and that nitrogen in nitrogen-seeded steady state detached H-mode discharges accumulates in the divertor. (orig.)

  8. Thermal effects of runaway electrons in an armoured divertor

    International Nuclear Information System (INIS)

    Stad, R.C.L. van der.

    1993-12-01

    This report describes the results of a numerical thermal analysis of the heat deposition of runaway electrons accompanying plasma disruptions in a armoured divertor. The divertor concepts studied are carbon on molybdenum and beryllium on copper. The conclusion is that the runaway electrons can cause melting of the armour as well as melting of the structure and can damage the divertor severely. (orig.)

  9. Moving Divertor Plates in a Tokamak

    International Nuclear Information System (INIS)

    Zweben, S.J.; Zhang, H.

    2009-01-01

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions

  10. Moving Divertor Plates in a Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S.J. Zweben, H. Zhang

    2009-02-12

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions.

  11. Development of divertor remote maintenance system

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Nobukazu; Oka, Kiyoshi; Akou, Kentaro; Takiguchi, Yuji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    The ITER divertor is categorized as a scheduled maintenance component because of extreme heat and particle loads it is exposed to by plasma. It is also highly activated by 14 MeV neutrons. Reliable remote handling equipment and tools are required for divertor maintenance under intense gamma radiation. To facilitate remote maintenance, the divertor is segmented into 60 cassettes, and each cassette weighing about 25 tons and maintained and replaced through four maintenance ports each 90 degrees. Divertor cassettes must be transported toroidally and radially for replacement through maintenance ports. Remote handling involving cassette movers and carriers for toroidal and radial transport has been developed. Under the ITER R and D program, technology critical to divertor cassette maintenance is being developed jointly by Japan, E.U., and U.S. home teams. This paper summarizes divertor remote maintenance design and the status of technology development by the Japan Home Team. (author)

  12. Development of divertor remote maintenance system

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Oka, Kiyoshi; Akou, Kentaro; Takiguchi, Yuji

    1998-01-01

    The ITER divertor is categorized as a scheduled maintenance component because of extreme heat and particle loads it is exposed to by plasma. It is also highly activated by 14 MeV neutrons. Reliable remote handling equipment and tools are required for divertor maintenance under intense gamma radiation. To facilitate remote maintenance, the divertor is segmented into 60 cassettes, and each cassette weighing about 25 tons and maintained and replaced through four maintenance ports each 90 degrees. Divertor cassettes must be transported toroidally and radially for replacement through maintenance ports. Remote handling involving cassette movers and carriers for toroidal and radial transport has been developed. Under the ITER R and D program, technology critical to divertor cassette maintenance is being developed jointly by Japan, E.U., and U.S. home teams. This paper summarizes divertor remote maintenance design and the status of technology development by the Japan Home Team. (author)

  13. Engineering conceptual design of CFETR divertor

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Xuebing, E-mail: pengxb@ipp.cas.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei Anhui (China); Ye, Minyou [School of Nuclear Science and Technology, University of Science and Technology of China, Jinzhai Road 96, 230026 Hefei Anhui (China); Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei Anhui (China); Song, Yuntao [Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei Anhui (China); School of Nuclear Science and Technology, University of Science and Technology of China, Jinzhai Road 96, 230026 Hefei Anhui (China); Mao, Xin [Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei Anhui (China); Chen, Peiming; Qian, Xinyuan [School of Nuclear Science and Technology, University of Science and Technology of China, Jinzhai Road 96, 230026 Hefei Anhui (China)

    2015-10-15

    Highlights: • Three divertor structures for two plasma configurations, ITER-like and snowflake. • Property of enlarging wet area for all three divertors is analyzed. • The divertor accommodating with both the plasma configurations is unfeasible. • Divertor cooling system is developed. - Abstract: The China Fusion Engineering Test Reactor (CFETR), which is in conceptual design phase, aims at producing fusion power of 50–200 MW with tritium breeding ratio of ∼1.2 and duty cycle time of 0.3–0.5. Its designed main parameters are major/minor radii of 5.7 m/1.6 m and plasma current of 10 MA. Although the fusion power is lower than the one of ITER, the relative smaller machine dimensions and planed much higher auxiliary heating power of 100–140 MW make that the power exhausting for the CFETR divertor is a very critical issue. To solve this issue, the divertor should be better designed with advanced physical operation mode, advanced configuration/geometry or high efficient cooling structure. In the paper, much effort was put on the divertor configuration and geometry. With designed magnet system, three divertor configurations can be realized, ITER-like, snowflake and super-X. However, considering structural design feasibility and remote handling compatibility, only the first two configurations were selected for the first step of engineering design. Three divertors were designed. They have different first wall geometries to accommodate with different plasma configurations, one for the ITER-like, one for the snowflake and the third one for both the configurations. All three divertors employ the same cassette body as the support and the cooling water manifold for the first wall. This feature simplifies the interface of the divertor to other components in the vacuum vessel. Besides, the cooling structure and the remote maintenance concept are also introduced in the paper.

  14. Non-local transport in a tokamak plasma divertor with recycling

    International Nuclear Information System (INIS)

    Abou-Assaleh, Z.; Petravic, M.; Vesey, R.

    1993-01-01

    The plasma transport, particle and energy fluxes, near the diverter plate with high recycling has been modeled by using an electron kinetic code (Fokker-Planck International) in conjunction with a two-fluid ambipolar code. We include the effects of ionization and excitation of the hydrogen atoms. The electron energy distribution calculated from the kinetic code shows a large deviation from Maxwellian especially near the plate. This deviation from Maxwellian is due to the non-local transport of the suprathermal electrons from the SOL, and due also to the absorption of the fast electrons by the target plate. The heat flux near the plate is shown to be nonlocal, in that it is not determined uniquely by the local plasma parameters. Therefore the classical transport coefficients in the fluid model must be modified by including a nonlocal effect to produce the kinetic results. The kinetic calculation is compared with those of the fluid code with different values of the electron heat flux limiter factor (f). To reduce the computer load, the initial condition we used corresponds to the equilibrium solution already found with the fluid code with f=0.2. The fluid and Fokker-Planck codes are relaxed until all transients associated with electron dynamics have disappeared. In section 2, we present the kinetic code. The fluid code is presented in section 3. The boundary conditions used in these simulations are given in section 4. Finally the results and conclusion of these simulations are presented in section 5

  15. Engineering design of a Radiative Divertor for DIII-D

    International Nuclear Information System (INIS)

    Smith, J.P.; Allen, S.L.; Anderson, P.M.; Baxi, C.B.; Chin, E.; Fenstermacher, M.E.; Hill, D.N.; Hollerbach, M.A.; Hyatt, A.W.; Junge, R.; Mahdavi, M.A.; Porter, G.D.; Redler, K.; Reis, E.E.; Schaffer, M.J.; Sevier, D.L.; Stambaugh, R.D.

    1995-01-01

    A new divertor called the Radiative Divertor is presently being designed for the DIII-D tokamak. Input from tokamak experiments and modeling form the basis for the new design. The Radiative Divertor is intended to reduce the heat flux on the divertor plates by dispersing the power with radiation. Gas puffing experiments in the current open divertor have shown a reduction of the divertor heat flux with either deuterium or impurity puffing. However, either the plasma density (D 2 ) or the core Z eff (impurities) increases in these experiments. The radiative divertor uses a slot structure to isolate the divertor plasma region from the area surrounding the core plasma. Modeling has shown that the Radiative Divertor hardware will provide better baffling and particle control and thereby minimize the effect of the gas puffing in the divertor region on the plasma core. In addition, the Radiative Divertor structure will allow density control in plasma shapes with high triangularity (>0.8) required for advanced tokamak operation. The divertor structure allows for operation in either double or single-null plasma configurations. Four independently controlled divertor cryopumps will enable pumping at either the inboard (upper and lower) or the outboard (upper and lower) divertor plates. Biasing is an integral part of the design and is based on experience at the Tokamak de Varennes (TdeV) and DIII-D. Boron nitride tiles electrically insulate the inner and outer strike points and a low current electrode is used to apply a radial electric field to the scrape-off layer. TdeV has shown that biasing can provide particle and impurity control. The design is extremely flexible, and will allow physics studies of the effect of slot width and height. This is extremely important, as the amount of chamber volume needed for the divertor in future machines such as International Thermonuclear Experiment Reactor (ITER) and Tokamak Physics Experiment (TPX) must be determined. (orig./WL)

  16. Towards the development of workable acceptance criteria for the divertor CFC monoblock armour

    Energy Technology Data Exchange (ETDEWEB)

    D' Agata, E. [ITER International Team, ITER Joint Work Site, Boltzmannstr. 2, D-85748 Garching (Germany)]. E-mail: dagatae@itereu.de; Tivey, R. [ITER International Team, ITER Joint Work Site, Boltzmannstr. 2, D-85748 Garching (Germany)

    2005-11-15

    The plasma-facing components (PFCs) of the ITER divertor will be subjected to high heat flux (HHF). Carbon-fibre composite (CFC) is selected as the armour for the region of highest heat flux where the scrape-off layer of the plasma intercepts the vertical targets (VT). Failure of the armour to heat sink joints will compromise the performance of the divertor and could ultimately result in its failure and the shut down of the ITER machine. There are tens of thousands of CFCs to CuCrZr joints. The aim of the PFC design is to ensure that the divertor can continue to function even with the failure of a few joints. In preparation for writing the procurement specification for the ITER vertical target PFCs, a programme of work is underway with the objective of defining workable acceptance criteria for the PFC armour joints.

  17. Towards the development of workable acceptance criteria for the divertor CFC monoblock armour

    International Nuclear Information System (INIS)

    D'Agata, E.; Tivey, R.

    2005-01-01

    The plasma-facing components (PFCs) of the ITER divertor will be subjected to high heat flux (HHF). Carbon-fibre composite (CFC) is selected as the armour for the region of highest heat flux where the scrape-off layer of the plasma intercepts the vertical targets (VT). Failure of the armour to heat sink joints will compromise the performance of the divertor and could ultimately result in its failure and the shut down of the ITER machine. There are tens of thousands of CFCs to CuCrZr joints. The aim of the PFC design is to ensure that the divertor can continue to function even with the failure of a few joints. In preparation for writing the procurement specification for the ITER vertical target PFCs, a programme of work is underway with the objective of defining workable acceptance criteria for the PFC armour joints

  18. 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.

  19. 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.))

  20. 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.

  1. Divertor tungsten tiles erosion in the region of the castellated gaps

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  2. Divertor design through shape optimization

    International Nuclear Information System (INIS)

    Dekeyser, W.; Baelmans, M.; Reiter, D.

    2012-01-01

    Due to the conflicting requirements, complex physical processes and large number of design variables, divertor design for next step fusion reactors is a challenging problem, often relying on large numbers of computationally expensive numerical simulations. In this paper, we attempt to partially automate the design process by solving an appropriate shape optimization problem. Design requirements are incorporated in a cost functional which measures the performance of a certain design. By means of changes in the divertor shape, which in turn lead to changes in the plasma state, this cost functional can be minimized. Using advanced adjoint methods, optimal solutions are computed very efficiently. The approach is illustrated by designing divertor targets for optimal power load spreading, using a simplified edge plasma model (copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  3. Modelling of DEMO core plasma consistent with SOL/divertor simulations for long-pulse scenarios with impurity seeding

    International Nuclear Information System (INIS)

    Pacher, G.W.; Pacher, H.D.; Janeschitz, G.; Kukushkin, A.S.; Kotov, V.; Reiter, D.

    2007-01-01

    The integrated core-pedestal-SOL model is applied to the simulation of a typical DEMO operation. Impurity seeding is used to reduce the power load on the divertor to acceptable levels. The influence on long-pulse operation of impurity seeding with various impurities is investigated. DEMO operation at acceptable peak power loads and long-pulse lengths is demonstrated

  4. 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

  5. Divertor conceptual designs for a fusion power plant

    International Nuclear Information System (INIS)

    Norajitra, P.; Ihli, T.; Janeschitz, G.; Abdel-Khalik, S.; Mazul, I.; Malang, S.

    2007-01-01

    The development of a divertor concept for post-ITER fusion power plants is deemed to be an urgent task to meet the EU Fast Track scenario. Developing a divertor is particularly challenging due to the wide range of requirements to be met including the high incident peak heat flux, the blanket design with which the divertor has to be integrated, sputtering erosion of the plasma-facing material caused by the incident a particles, radiation effects on the properties of structural materials, and efficient recovery and conversion of the divertor thermal power (∝15% of the total fusion thermal power) by maximizing the coolant operating temperature while minimizing the pumping power. In the course of the EU PPCS, three near-term (A, B and AB) and two advanced power plant models (C, D) were investigated. Model A utilizes a water-cooled lead-lithium (WCLL) blanket and a water-cooled divertor with a peak heat flux of 15 MW/m 2 . Model B uses a He-cooled ceramics/beryllium pebble bed (HCPB) blanket and a He-cooled divertor concept (10 MW/m 2 ). Model AB uses a He-cooled lithium-lead (HCLL) blanket and a He-cooled divertor concept (10 MW/m 2 ). Model C is based on a dual-coolant (DC) blanket (lead/lithium self-cooled bulk and He-cooled structures) and a He-cooled divertor (10 MW/m 2 ). Model D employs a self-cooled lead/lithium (SCLL) blanket and lead-lithiumcooled divertor (5 MW/m 2 ). The values in parenthesis correspond to the maximum peak heat fluxes required. It can be noted that the helium-cooled divertor is used in most of the EU plant models; it has also been proposed for the US ARIES-CS reactor study. Since 2002, it has been investigated extensively in Europe under the PPCS with the goal of reaching a maximum heat flux of at least 10 MW/m2. Work has covered many areas including conceptual design, analysis, material and fabrication issues, and experiments. Generally, the helium-cooled divertor is considered to be a suitable solution for fusion power plants, as it

  6. 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

  7. 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

  8. 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)

  9. Diagnostic setup for investigation of plasma wall interactions at Wendelstein 7-X

    International Nuclear Information System (INIS)

    Neubauer, Olaf; Biel, Wolfgang; Czymek, Guntram; Denner, Peter; Effenberg, Florian; Krämer-Flecken, Andreas; Liang, Yunfeng; Marchuk, Oleksandr; Offermanns, Guido; Rack, Michael; Samm, Ulrich; Schmitz, Oliver; Schweer, Bernd; Terra, Alexis

    2015-01-01

    Graphical abstract: - Highlights: • We are investigating plasma wall interactions at Wendelstein 7-X stellarator. • Steady state operation and island divertor are unique. • We are developing diagnostics for divertor plasma and plasma facing surfaces. • A multi-purpose fast manipulator allows for exposure of probes and samples. • Versatile endoscopes allow for local divertor spectroscopy from IR to UV. - Abstract: Wendelstein 7-X being the most advanced stellarator is currently prepared for commissioning at Greifswald. Forschungszentrum Jülich is preparing a research programme in the field of plasma wall interactions (PWI) by developing a dedicated set of diagnostic systems. The specific interest at Wendelstein 7-X is to understand PWI processes in presence of a 3D plasma boundary of an island divertor. Furthermore, for the first time steady state plasma at high density and low temperature in the divertor region will be available. Since PWI only could be understood in conjunction with the edge plasma properties the aim of the setup is to observe both the edge plasma as well as surface processes. For optimum combination of different diagnostic methods the edge diagnostic systems are aligned toroidally along one out of five magnetic islands. Main systems are a multipurpose fast probe manipulator, two gas boxes in opposite divertor modules together with two endoscopes each observing the divertor regions, a poloidal correlation reflectometer, a dispersion interferometer in the divertor, and VUV and X-ray spectroscopy in the plasma core. The concept of the diagnostic setup is presented in this paper.

  10. Diagnostic setup for investigation of plasma wall interactions at Wendelstein 7-X

    Energy Technology Data Exchange (ETDEWEB)

    Neubauer, Olaf, E-mail: o.neubauer@fz-juelich.de [Institute of Energy and Climate Research, Forschungszentrum Jülich GmbH, D-52425 Jülich (Germany); Biel, Wolfgang; Czymek, Guntram; Denner, Peter [Institute of Energy and Climate Research, Forschungszentrum Jülich GmbH, D-52425 Jülich (Germany); Effenberg, Florian [University Wisconsin–Madison, Madison, WI (United States); Krämer-Flecken, Andreas; Liang, Yunfeng; Marchuk, Oleksandr; Offermanns, Guido; Rack, Michael; Samm, Ulrich [Institute of Energy and Climate Research, Forschungszentrum Jülich GmbH, D-52425 Jülich (Germany); Schmitz, Oliver [University Wisconsin–Madison, Madison, WI (United States); Schweer, Bernd [Laboratoire de Physique des Plasmas – Laboratorium voor Plasmafysica, ERM/KMS, 1000 Brussels (Belgium); Terra, Alexis [Institute of Energy and Climate Research, Forschungszentrum Jülich GmbH, D-52425 Jülich (Germany)

    2015-10-15

    Graphical abstract: - Highlights: • We are investigating plasma wall interactions at Wendelstein 7-X stellarator. • Steady state operation and island divertor are unique. • We are developing diagnostics for divertor plasma and plasma facing surfaces. • A multi-purpose fast manipulator allows for exposure of probes and samples. • Versatile endoscopes allow for local divertor spectroscopy from IR to UV. - Abstract: Wendelstein 7-X being the most advanced stellarator is currently prepared for commissioning at Greifswald. Forschungszentrum Jülich is preparing a research programme in the field of plasma wall interactions (PWI) by developing a dedicated set of diagnostic systems. The specific interest at Wendelstein 7-X is to understand PWI processes in presence of a 3D plasma boundary of an island divertor. Furthermore, for the first time steady state plasma at high density and low temperature in the divertor region will be available. Since PWI only could be understood in conjunction with the edge plasma properties the aim of the setup is to observe both the edge plasma as well as surface processes. For optimum combination of different diagnostic methods the edge diagnostic systems are aligned toroidally along one out of five magnetic islands. Main systems are a multipurpose fast probe manipulator, two gas boxes in opposite divertor modules together with two endoscopes each observing the divertor regions, a poloidal correlation reflectometer, a dispersion interferometer in the divertor, and VUV and X-ray spectroscopy in the plasma core. The concept of the diagnostic setup is presented in this paper.

  11. The investigation of structure, chemical composition, hydrogen isotope trapping and release processes in deposition layers on surfaces exposed to DIII-D divertor plasma

    International Nuclear Information System (INIS)

    Buzhinskij, O.I.; Opimach, I.V.; Barsuk, V.A.; Arkhipov, I.I.; Whyte, D.; Wampler, W.R.

    1998-05-01

    The exposure of ATG graphite sample to DIII-D divertor plasma was provided by the DiMES (Divertor Material Evaluation System) mechanism. The graphite sample arranged to receive the parallel heat flux on a small region of the surface was exposed to 600ms of outer strike point plasma. The sample was constructed to collect the eroded material directed downward into a trapping zone onto s Si disk collector. The average heat flux onto the graphite sample during the exposure was about 200W/cm 2 , and the parallel heat flux was about 10 KW/cm 2 . After the exposure the graphite sample and Si collector disk were analyzed using SEM, NRA, RBS, Auger spectroscopy. IR and Raman spectroscopy. The thermal desorption was studied also. The deposited coating on graphite sample is amorphous carbon layer. Just upstream of the high heat flux zone the redeposition layer has a globular structure. The deposition layer on Si disk is composed also from carbon but has a diamond-like structure. The areal density of C and D in the deposited layer on Si disk varied in poloidal and toroidal directions. The maximum D/C areal density ratio is about 0.23, maximum carbon density is about 3.8 x 10 18 cm -2 , maximum D area density is about 3 x 10 17 cm 2 . The thermal desorption spectrum had a peak at 1,250K

  12. 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.

  13. 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.

  14. 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)

  15. Snowflake divertor configuration studies for NSTX-Upgrade

    International Nuclear Information System (INIS)

    Soukhanovskii, V.A.

    2011-01-01

    Snowflake divertor experiments in NSTX provide basis for PMI development toward NSTX-Upgrade. Snowflake configuration formation was followed by radiative detachment. Significant reduction of steady-state divertor heat flux observed in snowflake divertor. Impulsive heat loads due to Type I ELMs are partially mitigated in snowflake divertor. Magnetic control of snowflake divertor configuration is being developed. Plasma material interface development is critical for NSTX-U success. Four divertor coils should enable flexibility in boundary shaping and control in NSTX-U. Snowflake divertor experiments in NSTX provide good basis for PMI development in NSTX-Upgrade. FY 2009-2010 snowflake divertor experiments in NSTX: (1) Helped understand control of magnetic properties; (2) Core H-mode confinement unchanged; (3) Core and edge carbon concentration reduced; and (4) Divertor heat flux significantly reduced - (a) Steady-state reduction due to geometry and radiative detachment, (b) Encouraging results for transient heat flux handling, (c) Combined with impurity-seeded radiative divertor. Outlook for snowflake divertor in NSTX-Upgrade: (1) 2D fluid modeling of snowflake divertor properties scaling - (a) Edge and divertor transport, radiation, detachment threshold, (b) Compatibility with cryo-pump and lithium conditioning; (2) Magnetic control development; and (3) PFC development - PFC alignment and PFC material choice.

  16. 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.

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

    Directory of Open Access Journals (Sweden)

    Kuroda Yusuke

    2013-11-01

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

  18. Long-term erosion and re-deposition of carbon in the divertor region of JT-60U

    International Nuclear Information System (INIS)

    Gotoh, Y.; Tanabe, T.; Ishimoto, Y.; Masaki, K.; Arai, T.; Kubo, H.; Tsuzuki, K.; Miya, N.

    2006-01-01

    Erosion and redeposition profiles of carbon tiles used in the W-shaped divertor of JT-60U with all carbon plasma facing wall are studied. The inner divertor is mostly covered by carbon redeposited layers, while the outer divertor mostly eroded. In the dome region, the erosion dominates on the inner dome-wing, while redeposited on the outer dome-wing. The redeposited layers on the outer dome-wing show very clear columnar structures indicating local carbon transport form the outer divertor to the outer dome-wing. The weight gain by the redeposition extrapolated to the whole divertor area is 0.55 kg. Since the extrapolated total erosion is about 0.33 kg, the remaining 0.22 kg must be originated from the main chamber erosion. Significant amount of the redeposition is caused locally by multiple processes of erosion, ionization and prompt redeposition toward inboard direction owing to gyration along magnetic filed line. This inboard transport is one of the reasons for small redeposition on the plasma shadowed area of the W shaped divertor of JT-60U with pumping slots placed at the bottoms side

  19. Application of the radiating divertor approach to innovative tokamak divertor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Petrie, T.W., E-mail: petrie@fusion.gat.com [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Allen, S.L.; Fenstermacher, M.E. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Groebner, R.J. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Holcomb, C.T. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Kolemen, E. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); La Haye, R.J. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Lasnier, C.J. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Leonard, A.W.; Luce, T.C. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Maingi, R. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); Moyer, R.A. [University of California San Diego, 9500 Gilman Dr., La Jolla, CA 92093-0417 (United States); Solomon, W.M. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Turco, F. [Columbia University, 2960 Broadway, New York, NY 10027 (United States); Watkins, J.G. [Sandia National Laboratory, PO Box 5800, Albuquerque, NM 87185 (United States)

    2015-08-15

    We survey the results of recent DIII-D experiments that tested the effectiveness of three innovative tokamak divertor concepts in reducing divertor heat flux while still maintaining acceptable energy confinement under neon/deuterium-based radiating divertor (RD) conditions: (1) magnetically unbalanced high performance double-null divertor (DND) plasmas, (2) high performance double-null “Snowflake” (SF-DN) plasmas, and (3) single-null H-mode plasmas having different isolation from their divertor targets. In general, all three concepts adapt well to RD conditions, achieving significant reduction in divertor heat flux (q{sub ⊥p}) and maintaining high performance metrics, e.g., 50–70% reduction in peak divertor heat flux for DND and SF-DN plasmas that are characterized by β{sub N} ≅ 3.0 and H{sub 98(y,2)} ≈ 1.35. It is also demonstrated that q{sub ⊥p} could be reduced ≈50% by extending the parallel connection length (L{sub ||-XPT}) in the scrape-off layer between the X-point and divertor targets over a variety of the RD and non-RD environments tested.

  20. Deposition of deuterium and metals on divertor tiles in the DIII--D tokamak

    International Nuclear Information System (INIS)

    Walsh, D.S.; Doyle, B.L.; Jackson, G.L.

    1992-01-01

    Hydrogen recycling and impurity influx are important issues in obtaining high confinement discharges in the DIII--D tokamak. To reduce metallic impurities in DIII--D, 40% of the wall area, including the highest heat flux zones, have been covered with graphite tiles. However, erosion, redeposition, and hydrogen retention in the tiles, as well as metal transport from the remaining Inconel walls, can lead to enhanced recycling and impurity influx. Hydrogen and metal retention in divertor floor tiles have been measured using external ion beam analysis techniques following four campaigns where tiles were exposed to several thousand tokamak discharges. The areal density of deuterium retained following exposure to tokamak plasmas was measured with external nuclear reaction analysis. External proton-induced x-ray emission analysis was used to measure the areal densities of metallic impurities deposited upon the divertor tiles either by sputtering of metallic components during discharges or as contamination during tile fabrication. Measurements for both deuterium and metallic impurities were taken on both the tile surfaces which face the operating plasma and the surfaces on the sides of the tiles which form the small gaps separating each of the tiles in the divertor. The highest areal densities of both deuterium (from 2 to 8 x 10 18 atoms/cm 2 ) and metals (from 0.2 to 1 x 10 18 atoms/cm 2 ) were found on the plasma-facing surface near the inner strike point region of each set of divertor tiles. Significant deposits, extending as far as 1 cm from the plasma-facing surface and containing up to 40% of the total divertor deposition, were also observed on the gap-forming surfaces of the tiles

  1. Deposition of deuterium and metals on divertor tiles in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Walsh, D.S.; Doyle, B.L.; Jackson, G.L.

    1991-01-01

    Hydrogen recycling and impurity influx are important issues in obtaining high confinement discharges in the D3-D tokamak. To reduce metallic impurities in D3-D, 40% of the wall area, including the highest heat flux zones, have been covered with graphite tiles. However erosion, redeposition and hydrogen retention in the tiles, as well as metal transport from the remaining Inconel walls can lead to enhanced recycling and impurity influx. Hydrogen and metal retention in divertor floor tiles have been measured using external ion beam analysis techniques following four campaigns where tiles were exposed to several thousand tokamak discharges. The areal density of deuterium retained following exposure to tokamak plasmas was measured with external nuclear reaction analysis. External proton-induced x-ray emission analysis was used to measure the areal densities of metallic impurities deposited upon the divertor tiles either by sputtering of metallic components during discharges or as contamination during tile fabrication. Measurements for both deuterium and metallic impurities were taken on both the tile surfaces which face the operating plasma and the surfaces on the side of the tiles which form the small gaps separating each of the tiles in the divertor. The highest areal densities of both deuterium and metals were found on the plasma-facing surface near the inner strike point region of each set of divertor tiles. Significant deposits, extending as fast a 1 cm from the plasma-facing and containing up to forty percent of the total divertor deposition, were also observed on the gap-forming surfaces of the tiles

  2. Experimental investigation of heat transport and divertor loads of fusion plasmas in all metal ASDEX upgrade and JET

    International Nuclear Information System (INIS)

    Sieglin, Bernhard A.

    2014-01-01

    This work presents divertor heat load studies conducted at two of the largest tokamaks currently in operation, ASDEX Upgrade and the Joint European Torus (JET). A commonly agreed empirical scaling for the power fall-off length in H-mode obtained in carbon devices is validated in JET with the ILW. Bohm and Gyro-Bohm like models are identified as possible candidates describing the divertor broadening. Quantities for the assessment of the thermal load induced by transient heat loads are defined. JET with the ILW exhibits an on average longer ELM duration as compared to the carbon wall. For identical pedestal conditions the ELM durations in both cases are found to be the same within error bars. The energy fluency is found to depend mainly on the pedestal pressure with a weak dependence on the relative loss in stored energy. This is noteworthy since the current extrapolation to ITER assumes a linear dependence on the relative ELM size.

  3. Small angle slot divertor concept for long pulse advanced tokamaks

    Science.gov (United States)

    Guo, H. Y.; Sang, C. F.; Stangeby, P. C.; Lao, L. L.; Taylor, T. S.; Thomas, D. M.

    2017-04-01

    SOLPS-EIRENE edge code analysis shows that a gas-tight slot divertor geometry with a small-angle (glancing-incidence) target, named the small angle slot (SAS) divertor, can achieve cold, dissipative/detached divertor conditions at relatively low values of plasma density at the outside midplane separatrix. SAS exhibits the following key features: (1) strong enhancement of the buildup of neutral density in a localized region near the plasma strike point on the divertor target; (2) spreading of the cooling front across the divertor target with the slot gradually flaring out from the strike point, thus effectively reducing both heat flux and erosion on the entire divertor target surface. Such a divertor may potentially provide a power and particle handling solution for long pulse advanced tokamaks.

  4. Design analysis of the ITER divertor

    International Nuclear Information System (INIS)

    Samuelli, G.; Marin, A.; Roccella, M.; Lucca, F.; Merola, M.; Riccardi, B.; Petrizzi, L.; Villari, R.

    2007-01-01

    The divertor is one of the most challenging components of the ITER machine. Its function is to reduce the impurity in the plasma and consists essentially of two parts: the plasma facing components (PFCs) and a massive support structure called the cassette body (CB). Considerable R and D effort (developed by EFDA CSU GARCHING and the ITER International Team together with the EU Associations and the EU Industries) has been spent in designing divertor components capable of withstanding the expected electromagnetic (EM) loads and to take into account the latest ITER design conditions. In support of such efforts extensive and very detailed Neutronic, Thermal, EM and Structural analyses have been performed. A summary of the analyses performed will be presented. One of the main result is a typical exercise of integration between the different kind of analyses and the importance of keeping the consistency between the different assumptions and simplifications. The models used for the numerical analyses include a detailed geometrical description of the CB, the inlet, outlet hydraulic manifolds, the CB to vacuum vessel locking system and three configurations of the PFU. The effect of electrical bridging, both in poloidal and toroidal direction, of the PFU castellation, due to a possible melting at the W mono-block or tiles, occurring during the plasma disruptions, has been analyzed. For all these configurations 2 VDE scenarios including the effect of the Toroidal Field Variation and the HaloCurrent with the related out of plane induced EM forces have been extensively analyzed and a detailed poloidal and radial distribution of the nuclear heating has been used for the neutronic flux on the divertor components. The aim of this activity is to produce a comprehensive design and assessment of the ITER divertor via: -The estimation of the neutronic heat deposition and shielding capability; -The calculation of the related thermal and mechanical effects and the comparison of the

  5. Design analysis of the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Samuelli, G.; Marin, A.; Roccella, M.; Lucca, F. [L.T. Calcoli SaS, Merate (Lecco) (Italy); Merola, M. [ITER Team, Cadarache (France); Riccardi, B. [EFDA CSU Garching (Germany); Petrizzi, L.; Villari, R. [CRE ENEA sulla Fusione Frascati, Roma (Italy)

    2007-07-01

    The divertor is one of the most challenging components of the ITER machine. Its function is to reduce the impurity in the plasma and consists essentially of two parts: the plasma facing components (PFCs) and a massive support structure called the cassette body (CB). Considerable R and D effort (developed by EFDA CSU GARCHING and the ITER International Team together with the EU Associations and the EU Industries) has been spent in designing divertor components capable of withstanding the expected electromagnetic (EM) loads and to take into account the latest ITER design conditions. In support of such efforts extensive and very detailed Neutronic, Thermal, EM and Structural analyses have been performed. A summary of the analyses performed will be presented. One of the main result is a typical exercise of integration between the different kind of analyses and the importance of keeping the consistency between the different assumptions and simplifications. The models used for the numerical analyses include a detailed geometrical description of the CB, the inlet, outlet hydraulic manifolds, the CB to vacuum vessel locking system and three configurations of the PFU. The effect of electrical bridging, both in poloidal and toroidal direction, of the PFU castellation, due to a possible melting at the W mono-block or tiles, occurring during the plasma disruptions, has been analyzed. For all these configurations 2 VDE scenarios including the effect of the Toroidal Field Variation and the HaloCurrent with the related out of plane induced EM forces have been extensively analyzed and a detailed poloidal and radial distribution of the nuclear heating has been used for the neutronic flux on the divertor components. The aim of this activity is to produce a comprehensive design and assessment of the ITER divertor via: -The estimation of the neutronic heat deposition and shielding capability; -The calculation of the related thermal and mechanical effects and the comparison of the

  6. 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)

  7. Two-dimensional numerical study of ELMs-induced erosion of tungsten divertor target tiles with different edge shapes

    International Nuclear Information System (INIS)

    Huang, Yan; Sun, Jizhong; Hu, Wanpeng; Sang, Chaofeng; Wang, Dezhen

    2016-01-01

    Highlights: • Thermal performance of three edge-shaped divertor tiles was assessed numerically. • All the divertor tiles exposed to type-I ELMs like ITER's will melt. • The rounded edge tile thermally performs the best in all tiles of interest. • The incident energy flux density was evaluated with structural effects considered. - Abstract: Thermal performance of the divertor tile with different edge shapes was assessed numerically along the poloidal direction by a two-dimensional heat conduction model with considering the geometrical effects of castellated divertor tiles on the properties of its adjacent plasma. The energy flux density distribution arriving at the castellated divertor tile surface was evaluated by a two-dimension-in-space and three-dimension-in-velocity particle-in-cell plus Monte Carlo Collisions code and then the obtained energy flux distribution was used as input for the heat conduction model. The simulation results showed that the divertor tiles with any edge shape of interest (rectangular edge, slanted edge, and rounded edge) would melt, especially, in the edge surface region of facing plasma poloidally under typical heat flux density of a transient event of type-I ELMs for ITER, deposition energy of 1 MJ/m"2 in a duration of 600 μs. In comparison with uniform energy deposition, the vaporizing erosion was reduced greatly but the melting erosion was aggravated noticeably in the edge area of plasma facing diveror tile. Of three studied edge shapes, the simulation results indicated that the divertor plate with rounded edge was the most resistant to the thermal erosion.

  8. Design and test program of a simplified divertor dummy coil structure for the WEST project

    Energy Technology Data Exchange (ETDEWEB)

    Doceul, L., E-mail: louis.doceul@cea.fr [CEA, IRFM, Saint-Paul-Lez-Durance Cedex F-13108 (France); Bucalossi, J.; Dougnac, H.; Ferlay, F.; Gargiulo, L.; Keller, D.; Larroque, S.; Lipa, M.; Pilia, A. [CEA, IRFM, Saint-Paul-Lez-Durance Cedex F-13108 (France); Portafaix, C. [ITER Organization, Route de Vinon-sur-Verdon 13115, St. Paul-lez-Durance (France); Saille, A. [CEA, IRFM, Saint-Paul-Lez-Durance Cedex F-13108 (France); Salami, M. [AVANTIS Engineering Groupe, ZI de l’Aiguille 46100, Figeac (France); Samaille, F.; Soler, B.; Thouvenin, D.; Verger, J.M.; Zago, B. [CEA, IRFM, Saint-Paul-Lez-Durance Cedex F-13108 (France)

    2013-12-15

    Highlights: • The mechanical design and integration of the divertor structure has been performed. • The design of the casing and the winding-pack has been finalized. • The coil assembly process has been validated. • The realization of a coil mock-up scale one is in progress. -- Abstract: In order to fully validate actively cooled tungsten plasma facing components (industrial fabrication, operation with long plasma duration), the implementation of a tungsten axisymmetric divertor structure in the tokamak Tore-Supra is studied. With this major upgrade, so-called WEST (Tungsten Environment in Steady state Tokamak), Tore-Supra will be able to address the problematic of long plasma discharges with a metallic divertor target. To do so, it is planned to install two symmetric divertor coils inside the vacuum vessel. This assembly, called divertor structure, is made up of two stainless steel casings containing a copper winding pack cooled by a pressurized hot water circuit (up to 180 °C, 4 MPa) and is designed to perform steady state plasma operation (up to 1000 s). The divertor structure will be a complex assembly ring of 4 m diameter representing a total weight of around 20 tons. The technical challenge of this component will be the implementation of angular sectors inside the vacuum vessel environment (TIG welding of the coil casing, induction brazing and electrical insulation of the copper winding). Moreover, this complex assembly must sustain harsh environmental conditions in terms of ultra high vacuum conditions, electromagnetical loads and electrical isolation (13 kV ground voltage) under high temperature. In order to fully validate the assembly and the performance of this complex component, the production of a scale one dummy coil is in progress. The paper will illustrate, the technical developments performed in order to finalize the design for the call for tender for fabrication. The progress and the first results of the simplified dummy coils will be also

  9. Design and test program of a simplified divertor dummy coil structure for the WEST project

    International Nuclear Information System (INIS)

    Doceul, L.; Bucalossi, J.; Dougnac, H.; Ferlay, F.; Gargiulo, L.; Keller, D.; Larroque, S.; Lipa, M.; Pilia, A.; Portafaix, C.; Saille, A.; Salami, M.; Samaille, F.; Soler, B.; Thouvenin, D.; Verger, J.M.; Zago, B.

    2013-01-01

    Highlights: • The mechanical design and integration of the divertor structure has been performed. • The design of the casing and the winding-pack has been finalized. • The coil assembly process has been validated. • The realization of a coil mock-up scale one is in progress. -- Abstract: In order to fully validate actively cooled tungsten plasma facing components (industrial fabrication, operation with long plasma duration), the implementation of a tungsten axisymmetric divertor structure in the tokamak Tore-Supra is studied. With this major upgrade, so-called WEST (Tungsten Environment in Steady state Tokamak), Tore-Supra will be able to address the problematic of long plasma discharges with a metallic divertor target. To do so, it is planned to install two symmetric divertor coils inside the vacuum vessel. This assembly, called divertor structure, is made up of two stainless steel casings containing a copper winding pack cooled by a pressurized hot water circuit (up to 180 °C, 4 MPa) and is designed to perform steady state plasma operation (up to 1000 s). The divertor structure will be a complex assembly ring of 4 m diameter representing a total weight of around 20 tons. The technical challenge of this component will be the implementation of angular sectors inside the vacuum vessel environment (TIG welding of the coil casing, induction brazing and electrical insulation of the copper winding). Moreover, this complex assembly must sustain harsh environmental conditions in terms of ultra high vacuum conditions, electromagnetical loads and electrical isolation (13 kV ground voltage) under high temperature. In order to fully validate the assembly and the performance of this complex component, the production of a scale one dummy coil is in progress. The paper will illustrate, the technical developments performed in order to finalize the design for the call for tender for fabrication. The progress and the first results of the simplified dummy coils will be also

  10. Rapidly Moving Divertor Plates In A Tokamak

    International Nuclear Information System (INIS)

    Zweben, S.

    2011-01-01

    It may be possible to replace conventional actively cooled tokamak divertor plates with a set of rapidly moving, passively cooled divertor plates on rails. These plates would absorb the plasma heat flux with their thermal inertia for ∼10-30 sec, and would then be removed from the vessel for processing. When outside the tokamak, these plates could be cooled, cleaned, recoated, inspected, and then returned to the vessel in an automated loop. This scheme could provide nearoptimal divertor surfaces at all times, and avoid the need to stop machine operation for repair of damaged or eroded plates. We describe various possible divertor plate designs and access geometries, and discuss an initial design for a movable and removable divertor module for NSTX-U.

  11. Reactor application of an improved bundle divertor

    International Nuclear Information System (INIS)

    Yang, T.F.; Ruck, G.W.; Lee, A.Y.; Smeltzer, G.; Prevenslik, T.

    1978-11-01

    A Bundle Divertor was chosen as the impurity control and plasma exhaust system for the beam driven Demonstration Tokamak Hybrid Reactor - DTHR. In the context of a preconceptual design study of the reactor and associated facility a bundle divertor concept was developed and integrated into the reactor system. The overall system was found feasible and scalable for reactors with intermediate torodial field strengths on axis. The important design characteristics are: the overall average current density of the divertor coils is 0.73 kA for each tesla of toroidal field on axis; the divertor windings are made from super-conducting cables supported by steel structures and are designed to be maintainable; the particle collection assembly and auxiliary cryosorption vacuum pump are dual systems designed such that they can be reactivated alterntively to allow for continuous reactor operation; and the power requirement for energizing and operating the divertor is about 5 MW

  12. Development of conductively cooled first wall armor and actively cooled divertor structure for ITER/FER

    International Nuclear Information System (INIS)

    Ioki, K.; Yamada, M.; Sakata, S.; Okada, K.; Toyoda, M.; Shimizu, K.; Tsujimura, S.; Iimura, M.; Akiba, M.; Araki, M.; Seki, M.

    1991-01-01

    Based on the design requirements for the plasma facing components in ITER/FER, we have performed design studies on the conductively cooled first wall armor and the divertor plate with sliding supports. The full-scale armor tiles were fabricated for heat load tests, and good thermal performances were obtained in heat load tests of 0.2-0.4 MW/m 2 . It is shown by the thermomechanical analysis on the divertor plate that thermal stresses and bending deformation are reduced significantly by using the sliding supports. The divertor test module with the sliding supports has been fabricated to investigate its fabricability and to verify the functions of the sliding supports during a high heat load of about 10 MW/m 2 . (orig.)

  13. A large divertor manipulator for ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Herrmann, Albrecht, E-mail: albrecht.herrmann@ipp.mpg.de; Jaksic, Nikola; Leitenstern, Peter; Greuner, Henri; Krieger, Karl; Marné, Pascal de; Oberkofler, Martin; Rohde, Volker; Schall, Gerd

    2015-10-15

    Highlights: • A large divertor manipulator for ASDEX Upgrade is developed and tested. • It allows replacing a relevant part of the divertor by dedicated targets and probes. • Modified solid standard targets. • Electrical and mechanical probes for dedicated investigations. • Test of actively cooled component. - Abstract: In 2013 a new bulk tungsten divertor, Div-III, was installed in ASDEX Upgrade (AUG). During the concept and design phase of Div-III the option of adaptable divertor instrumentation and divertor modification as contribution for divertor investigations in preparation of ITER was given a high priority. To gain flexibility for the test of divertor modifications without affecting the operational space of AUG, the large divertor manipulator, DIM-II, was designed and installed. DIM-II allows to retract 2 out of 128 outer divertor target tiles including the water cooled support structure into a target exchange box and to replace these targets without breaking the vacuum of the AUG vessel. DIM-II is based on a carriage-rail system with a driving rod pushing a front-end with the target module into the divertor position for plasma operation. Three types of front-ends are foreseen for physics investigations: (i) modified standard targets clamped to the standard cooling structure, (ii) dedicated front-ends making use of the whole available volume of about 230 × 160 × 80 mm{sup 3} and (iii) actively cooled/heated targets for cooling water temperatures up to 230 °C. This paper presents the DIM-II design including the FEM calculations for the modified divertor support structure and the front-end options, as well as the test procedure and operation mode.

  14. 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

  15. 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.

  16. Comparison of wall/divertor deuterium retention and plasma fueling requirements on the DIII-D, TdeV, and ASDEX-upgrade tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Maingi, R. [Oak Ridge Associated Universities, TN (United States); Terreault, B. [Inst. National de la Recherche Scientifique, Varennes, Quebec (Canada); Haas, G. [Max Planck Inst. fuer Plasmaphysik, Garching (Germany)] [and others

    1996-06-01

    The authors present a comparison of the wall deuterium retention and plasma fueling requirements of three diverted tokamaks, DIII-D, TdeV, and ASDEX-Upgrade, with different fractions of graphite coverage of stainless steel or Inconel outer walls and different heating modes. Data from particle balance experiments on each tokamak demonstrate well-defined differences in wall retention of deuterium gas, even though all three tokamaks have complete graphite coverage of divertor components and all three are routinely boronized. This paper compares the evolution of the change in wall loading and net fueling efficiency for gas during dedicated experiments without Helium Glow Discharge Cleaning on the DIII-D and TdeV tokamaks. On the DIII-D tokamak, it was demonstrated that the wall loading could be increased by > 1,250 Torr-1 (equivalent to 150 {times} plasma particle content) plasma inventories resulting in an increase in fueling efficiency from 0.08 to 0.25, whereas the wall loading on the TdeV tokamak could only be increased by < 35 Torr-{ell} (equivalent to 50{times} plasma particle content) plasma inventories at a maximum fueling efficiency {approximately} 1. Data from the ASDEX-Upgrade tokamak suggests qualitative behavior of wall retention and fueling efficiency similar to DIII-D.

  17. Tritium distribution on plasma facing graphite tiles of JT-60U

    International Nuclear Information System (INIS)

    Tanabe, T.; Sugiyama, K.; Masaki, K.; Gotoh, Y.; Tobita, K.; Miya, N.

    2003-01-01

    Tritium distributions on the graphite divertor tiles, the dome units and the baffle plates of JT-60U were successfully measured. Poloidally, the highest tritium level was found at the dome top tiles and the outer baffle plates, where the plasma did not hit directly. On the other hand, although the toroidal tritium profiles on each tiles appeared uniform, detailed profiles in full toroidal direction clearly showed a periodic variation corresponding to the position of the magnetic field coils, indicating the ripple loss of high energy tritons as suggested by the OFMC code. Finally, the temperature increase owing to the plasma heat load was found to release the once retained tritium. (author)

  18. 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.

  19. 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

  20. 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.

  1. 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...

  2. 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

  3. Engineering studies for the installation of an axi-symmetric metallic divertor in Tore Supra

    International Nuclear Information System (INIS)

    Doceul, L.; Portafaix, C.; Bucalossi, J.; Saille, A.; Bertrand, B.; Lipa, M.; Missirlian, M.; Jiolat, G.; Samaille, F.; Soler, B.

    2011-01-01

    Tore Supra (TS) has been designed to operate using technologies that allow long plasma operation (a few minutes), by means of superconducting magnets and actively-cooled high heat flux plasma facing components (PFCs). Actively cooled tungsten PFC will be used in the baffle area of the first ITER divertor. In order to validate such a technology fully (industrial manufacturing, operation with long plasma duration), the implementation of a tungsten axi-symmetric divertor in the tokamak Tore Supra has been studied . With this second major upgrade, Tore Supra should be able to address the problematic of long plasma discharges with a metallic divertor. The proposed divertor is made up of two stainless steel casings containing a copper coil winding located at the top and bottom area of the vacuum vessel. These casings are firmly maintained by connection beams and protected by PFC. This paper describes the mechanical design of this major component and its integration in TS, the associated electromagnetic and thermomechanical analysis, the manufacturing issues and finally the integration of ITER representative PFCs.

  4. 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

  5. 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

  6. The lithium vapor box divertor

    International Nuclear Information System (INIS)

    Goldston, R J; Schwartz, J; Myers, R

    2016-01-01

    It has long been recognized that volumetric dissipation of the plasma heat flux from a fusion power system is preferable to its localized impingement on a material surface. Volumetric dissipation mitigates both the anticipated very high heat flux and intense particle-induced damage due to sputtering. Recent projections to a tokamak demonstration power plant suggest an immense upstream parallel heat flux, of order 20 GW m −2 , implying that fully detached operation may be a requirement for the success of fusion power. Building on pioneering work on the use of lithium by Nagayama et al and by Ono et al as well as earlier work on the gas box divertor by Watkins and Rebut, we present here a concept for a lithium vapor box divertor, in which lithium vapor extracts momentum and energy from a fusion-power-plant divertor plasma, using fully volumetric processes. At the high powers and pressures that are projected this requires a high density of lithium vapor, which must be isolated from the main plasma in order to avoid lithium build-up on the chamber walls or in the plasma. Isolation is achieved through a powerful multi-box differential pumping scheme available only for condensable vapors. The preliminary box-wise calculations are encouraging, but much more work is required to demonstrate the practical viability of this scheme, taking into account at least 2D plasma and vapor flows within and between the vapor boxes and out of the vapor boxes to the main plasma. (paper)

  7. Numerical exploration of non-axisymmetric divertor closure in the small angle slot (SAS) divertor at DIII-D

    Science.gov (United States)

    Frerichs, H.; Schmitz, O.; Covele, B.; Feng, Y.; Guo, H. Y.; Hill, D.

    2018-05-01

    Numerical simulations of toroidal asymmetries in a tightly baffled small angle slot (SAS) divertor on the DIII-D tokamak show that toroidal asymmetries in divertor closure result in (non-axisymmetric) local onset of detachment within a density window of 10-15% on top of the nominal threshold separatrix density. The SAS divertor is explored at DIII-D for improving access to cold, dissipative/detached divertor conditions. The narrow width of the slot divertor coupled with a small magnetic field line-to-target angle facilitates the buildup of neutral density, thereby increasing radiative and neutrals-related (atoms and molecules) losses in the divertor. Small changes in the strike point location can be expected to have a large impact on divertor conditions. The combination of misaligned slot structure and non-axisymmetric perturbations to the magnetic field configuration causes the strike point to move along the divertor target plate, possibly leaving the divertor slot at some locations. The latter extreme case essentially introduces an opening in the divertor slot from where recycling neutrals can easily escape, and thereby degrade the performance of the slot divertor. Such a strike point dislocation is approximated by a finite gap in the divertor baffle for which 3D edge plasma and neutral gas simulations are performed with the EMC3-EIRENE code.

  8. 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

  9. 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

  10. 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)

  11. Design of DIII-D advanced divertor

    International Nuclear Information System (INIS)

    Smith, J.P.; Baxi, C.B.; Reis, E.; Schaffer, M.; Thruston, G.

    1989-01-01

    The Advanced Divertor is a modification being designed for the plasma chamber of the DIII-D tokamak in order to optimize the divertor configuration and allow a broader range of experiments to be carried out. The Advanced Divertor will enable two classes of physics experiments to be run in DIII-D: Divertor biasing and Divertor baffing. The Advanced Divertor has two principal components: ( 1) a toroidally symmetric baffle; and (2) a continuous ring electrode. The tokamak can be run in baffle, bias, or standard DIII-D divertor modes by accurate positioning of the outer divertor strike point through the use of the DIII-D plasma control system. The baffle will contain approximately 50,000 l/s pumping for particle removal in the outer bottom corner of the vacuum vessel. The strike point will be positioned at the entrance aperture for the baffle mode. The aperture geometry is designed to facilitate a large particle influx plus a high probability that backstreaming particles will be reionized and redirected to the aperture. Where the baffling plates meet, gas sealing is required to prevent recycling of neutrals back into the plasma. The electrode is a continuous water-cooled ring, armored with graphite. The ring is electrically isolated from the vessel wall and is biasable to 1 kV and 20 kA. The outer leg of the divertor will be positioned on the graphite covered ring during biasing experiments. The supports for the ring are radially flexible to handle the differential thermal growth between the ring and the vessel wall but stiff in the vertical direction to restrain the ring against large disruption forces. The coolant and electrical feeds are designed in a similar manner. 2 refs., 4 figs

  12. Design of DIII-D Advanced Divertor

    International Nuclear Information System (INIS)

    Smith, J.P.; Baxi, C.B.; Reis, E.; Schaffer, M.; Thurston, G.

    1989-11-01

    The Advanced Divertor is a modification being designed for the plasma chamber of the DIII-D tokamak in order to optimize the divertor configuration and allow a broader range of experiments to be carried out. The Advanced Divertor will enable two classes of physics experiments to be run in DIII-D: Divertor biasing and Divertor baffling. The Advanced Divertor has two principal components: a toroidally symmetric baffle; and a continuous ring electrode. The tokamak can be run in baffle, bias, or standard DIII-D divertor modes by accurate positioning of the outer divertor strike point through the use of the DIII-D plasma control system. The baffle will contain approximately 50,000 l/s pumping for particle removal in the outer bottom corner of the vacuum vessel. The strike point will be positioned at the entrance aperture for the baffle mode. The aperture geometry is designed to facilitate a large particle influx plus a high probability that backstreaming particles will be reionized and redirected to the aperture. Where the baffling plates meet, gas sealing is required to prevent recycling of neutrals back into the plasma. The electrode is a continuous water-cooled ring, armored with graphite. The ring is electrically isolated from the vessel wall and is biasable to 1 kV and 20 kA. The outer leg of the divertor will be positioned on the graphite covered ring during biasing experiments. The supports for the ring are radially flexible to handle the differential thermal growth between the ring and the vessel wall but stiff in the vertical direction to restrain the ring against large disruption forces. The coolant and electrical feeds are designed in a similar manner. All the feeds are supported from and maintain a 5 kV isolation to the vessel wall. 2 refs., 4 figs

  13. Fabrication and installation of the DIII-D radiative divertor structures

    International Nuclear Information System (INIS)

    Hollerbach, M.A.; Smith, J.P.

    1997-11-01

    Phase 1A of the Radiative Divertor Program (RDP) is now installed in the DIII-D tokamak located at General Atomics. This hardware was added to enhance both the Divertor and Advanced Tokamak research elements of the DIII-D program. This installation consists of a divertor baffle enveloping a cryocondensation pump at the upper outer divertor target of DIII-D. The divertor baffle consists of two toroidally continuous Inconel 625 water-cooled rings and a toroidal array of discontinuous radiatively-cooled plates. The water-cooled rings are each comprised of four quadrants, mechanically formed, chem.-milled, and resistance and TIG welded Inconel 625 panels. The supports attaching the panels to the vessel wall are designed to accommodate the differential thermal expansion between the rings and vessel during bake and to react the electromagnetic loads induced during disruptions. They are made from either Inconel 625 or Inconel 718 depending on the stress levels predicted in Finite Element Analysis. Gas seals are designed to limit the leakage from the baffle chamber back to the core plasma to 2,500 ell/s and incorporate plasma sprayed alumina to minimize currents flowing through them. The bulk of the water-cooled ring fabrication was performed by a vendor, however, the final machining of penetrations in the conical ring for diagnostic access was performed in-house using a unique machining configuration. This configuration, and the machining of the diagnostic cutouts is described. Graphite tiles were machined from ATJ graphite to form a smooth plasma-facing surface. The installation of all divertor components required only four weeks

  14. Design and Test of Wendelstein 7-X Water-Cooled Divertor Scraper

    Energy Technology Data Exchange (ETDEWEB)

    Boscary, J. [Max-Planck-Institut fur Plasmaphysik, EURATOM Association, Garching, Germany; Greuner, Henri [Max Planck Institute for Plasma Physics, Garching, Germany; Ehrke, Gunnar [Max Planck Institute of Plasma Physics, Greifswald, Germany; Boeswirth, Bernd [Max Planck Institute for Plasma Physics, Garching, Germany; Wang, Zhongwei [Max Planck Institute for Plasma Physics, Garching, Germany; Clark, Emily [The University of Tennessee, Knoxville; Lumsdaine, Arnold [ORNL; Tretter, Jorg [Max Planck Institute for Plasma Physics, Garching, Germany; Junghanns, Patrick [Max Planck Institute for Plasma Physics, Garching, Germany; Stadler, Reinhold [Max Planck Institute for Plasma Physics, Garching, Germany; McGinnis, William Dean [ORNL; Lore, Jeremy D. [ORNL; Team, W7-X [Max-Planck-Institut fur Plasmaphysik, Griefswald, Germany

    2018-04-01

    Heat load calculations have indicated the possible overloading of the ends of the water-cooled divertor facing the pumping gap beyond their technological limit. The intention of the scraper is the interception of some of the plasma fluxes both upstream and downstream before they reach the divertor surface. The scraper is divided into six modules of four plasma facing components (PFCs); each module has four PFCs hydraulically connected in series by two water boxes (inlet and outlet). A full-scale prototype of one module has been manufactured. Development activities have been carried out to connect the water boxes to the cooling pipes of the PFCs by tungsten inert gas internal orbital welding. This prototype was successfully tested in the GLADIS facility with 17 MW/m2 for 500 cycles. The results of these activities have confirmed the possible technological basis for a fabrication of the water-cooled scraper.

  15. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-08-01

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

  17. An Asdex-type divertor for ITER

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1989-01-01

    An Asdex-type local divertor is proposed for ITER consisting of a copper poloidal field coil adjacent to the plasma. Estimates indicate that the power consumption is acceptable. Advantages would be a much reduced heat load not very sensitive to magnetic perturbations. A disadvantage is the finite lifetime under neutron bombardment that would require periodic replacement of the divertor coils in a reactor, but probably not in ITER because of its limited fluence. Another disadvantage would be poorer blanket coverage unless the divertor coil itself incorporates breeding material. 3 figs

  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. Some problems of brazing technology for the divertor plate manufacturing

    Science.gov (United States)

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

    1992-09-01

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

  20. Some problems of brazing technology for the divertor plate manufacturing

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  1. Some problems of brazing technology for the divertor plate manufacturing

    Energy Technology Data Exchange (ETDEWEB)

    Prokofiev, Yu.G.; Barabash, V.R.; Gervash, A.A. (D.V. Efremov Scientific Research Inst. of Electrophysical Apparatus, St. Petersburg (Russia)); Khorunov, V.F.; Maksimova, S.V. (E.O. Paton Inst. of Electronwelding, Kiev (Ukraine)); Vinokurov, V.F. (Central Scientific Research Inst. of Structural Materials ' Prometey' , St. Petersburg (Russia)); Fabritsiev, S.A.

    1992-09-01

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

  2. Intermittent Divertor Filaments in the National Spherical Torus Experiment and Their Relation to Midplane Blobs

    International Nuclear Information System (INIS)

    Maqueda, R.J.; Stotler, D.P.

    2010-01-01

    While intermittent filamentary structures, also known as blobs, are routinely seen in the low-field-side scrape-off layer of the National Spherical Torus Experiment (NSTX) (Ono et al 2000 Nucl. Fusion 40 557), fine structured filaments are also seen on the lower divertor target plates of NSTX. These filaments, not associated with edge localized modes, correspond to the interaction of the turbulent blobs seen near the midplane with the divertor plasma facing components. The fluctuation level of the neutral lithium light observed at the divertor, and the skewness and kurtosis of its probability distribution function, is similar to that of midplane blobs seen in D α ; e.g. increasing with increasing radii outside the outer strike point (OSP) (separatrix). In addition, their toroidal and radial movement agrees with the typical movement of midplane blobs. Furthermore, with the appropriate magnetic topology, i.e. mapping between the portion of the target plates being observed into the field of view of the midplane gas puff imaging diagnostic, very good correlation is observed between the blobs and the divertor filaments. The correlation between divertor plate filaments and midplane blobs is lost close to the OSP. This latter observation is consistent with the existence of 'magnetic shear disconnection' due to the lower X-point, as proposed by Cohen and Ryutov (1997 Nucl. Fusion 37 621).

  3. Non-destructive examination of the bonding interface in DEMO divertor fingers

    International Nuclear Information System (INIS)

    Richou, Marianne; Missirlian, Marc; Vignal, Nicolas; Cantone, Vincent; Hernandez, Caroline; Norajitra, Prachai; Spatafora, Luigi

    2013-01-01

    Highlights: • SATIR tests on DEMO divertor fingers (integrating or not He cooling system). • Millimeter size artificial defects were manufactured. • Detectability of millimeter size artificial defects was evaluated. • SATIR can detect defect in DEMO divertor fingers. • Simulations are well correlated to SATIR tests. -- Abstract: Plasma facing components (PFCs) with tungsten (W) armor materials for DEMO divertor require a high heat flux removal capability (at least 10 MW/m 2 in steady-state conditions). The reference divertor PFC concept is a finger with a tungsten tile as a protection and sacrificial layer brazed to a thimble made of tungsten alloy W – 1% La 2 O 3 (WL10). Defects may be located at the W thimble to W tile interface. As the number of fingers is considerable (>250,000), it is then a major issue to develop a reliable control procedure in order to control with a non-destructive examination the fabrication processes. The feasibility for detecting defect with infrared thermography SATIR test bed is presented. SATIR is based on the heat transient method and is used as an inspection tool in order to assess component heat transfer capability. SATIR tests were performed on fingers integrating or not the complex He cooling system (steel cartridge with jet holes). Millimeter size artificial defects were manufactured and their detectability was evaluated. Results of this study demonstrate that the SATIR method can be considered as a relevant non-destructive technique examination for the defect detection of DEMO divertor fingers

  4. Conceptual design studies for the European DEMO divertor: Rationale and first results

    Czech Academy of Sciences Publication Activity Database

    You, J.H.; Mazzone, F.; Visca, E.; Bachmann, Ch.; Autissier, E.; Barrett, T.; Cocilovo, V.; Crescenzi, F.; Domalapally, P.K.; Dongiovanni, D.; Entler, Slavomír; Federici, G.; Frosi, P.; Fursdon, M.; Greuner, H.; Hancock, D.; Marzullo, D.; McIntosh, S.; Müller, A.V.; Porfiri, M.T.; Ramogida, G.; Reiser, J.; Richou, M.; Rieth, M.; Rydzy, A.; Villari, R.; Widak, V.

    109-111, November (2016), s. 1598-1603 ISSN 0920-3796. [International Symposium on Fusion Nuclear Technology (ISFNT-12)/12./. Jeju, 14.09.2015-18.09.2015] EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : DEMO * Tokamak * Divertor * Plasma-facing component * Conceptual design * Eurofusiona Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379615303331

  5. Tritium permeation evaluation through vertical target of divertor based on recent tritium transport properties

    OpenAIRE

    中村 博文; 西 正孝

    2003-01-01

    Re-evaluation of tritium permeation through vertical target of divertor under the ITER operation condition was carried out using tritium transport properties in the candidate materials such as the diffusion coefficient and the trapping factors in tungsten for armor, and the surface recombination coefficient on copper for the heat sink obtained by authors' recent investigation (authors' data), which simulated the plasma-facing conditions of ITER. Evaluation with the data set of previous evalua...

  6. Integrated core-edge-divertor modeling studies

    International Nuclear Information System (INIS)

    Stacey, W.M.

    2001-01-01

    An integrated calculation model for simulating the interaction of physics phenomena taking place in the plasma core, in the plasma edge and in the SOL and divertor of tokamaks has been developed and applied to study such interactions. The model synthesises a combination of numerical calculations (1) the power and particle balances for the core plasma, using empirical confinement scaling laws and taking into account radiation losses (2), the particle, momentum and power balances in the SOL and divertor, taking into account the effects of radiation and recycling neutrals, (3) the transport of feeling and recycling neutrals, explicitly representing divertor and pumping geometry, and (4) edge pedestal gradient scale lengths and widths, evaluation of theoretical predictions (5) confinement degradation due to thermal instabilities in the edge pedestals, (6) detachment and divertor MARFE onset, (7) core MARFE onsets leading to a H-L transition, and (8) radiative collapse leading to a disruption and evaluation of empirical fits (9) power thresholds for the L-H and H-L transitions and (10) the width of the edge pedestals. The various components of the calculation model are coupled and must be iterated to a self-consistent convergence. The model was developed over several years for the purpose of interpreting various edge phenomena observed in DIII-D experiments and thereby, to some extent, has been benchmarked against experiment. Because the model treats the interactions of various phenomena in the core, edge and divertor, yet is computationally efficient, it lends itself to the investigation of the effects of different choices of various edge plasma operating conditions on overall divertor and core plasma performance. Studies of the effect of feeling location and rate, divertor geometry, plasma shape, pumping and over 'edge parameters' on core plasma properties (line average density, confinement, density limit, etc.) have been performed for DIII-D model problems. A

  7. Spectroscopic diagnostics for liquid lithium divertor studies on National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Soukhanovskii, V. A.; Roquemore, A. L.; Bell, R. E.; Kaita, R.; Kugel, H. W.

    2010-01-01

    The use of lithium-coated plasma facing components for plasma density control is studied in the National Spherical Torus Experiment (NSTX). A recently installed liquid lithium divertor (LLD) module has a porous molybdenum surface, separated by a stainless steel liner from a heated copper substrate. Lithium is deposited on the LLD from two evaporators. Two new spectroscopic diagnostics are installed to study the plasma surface interactions on the LLD: (1) A 20-element absolute extreme ultraviolet (AXUV) diode array with a 6 nm bandpass filter centered at 121.6 nm (the Lyman-α transition) for spatially resolved divertor recycling rate measurements in the highly reflective LLD environment, and (2) an ultraviolet-visible-near infrared R=0.67 m imaging Czerny-Turner spectrometer for spatially resolved divertor D I, Li I-II, C I-IV, Mo I, D 2 , LiD, CD emission and ion temperature on and around the LLD module. The use of photometrically calibrated measurements together with atomic physics factors enables studies of recycling and impurity particle fluxes as functions of LLD temperature, ion flux, and divertor geometry.

  8. Transient thermal hydraulic modeling and analysis of ITER divertor plate system

    International Nuclear Information System (INIS)

    El-Morshedy, Salah El-Din; Hassanein, Ahmed

    2009-01-01

    A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m 2 plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.

  9. Transient thermal hydraulic modeling and analysis of ITER divertor plate system

    Energy Technology Data Exchange (ETDEWEB)

    El-Morshedy, Salah El-Din [Argonne National Laboratory, Argonne, IL (United States); Atomic Energy Authority, Cairo (Egypt)], E-mail: selmorshedy@etrr2-aea.org.eg; Hassanein, Ahmed [Purdue University, West Lafayette, IN (United States)], E-mail: hassanein@purdue.edu

    2009-12-15

    A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m{sup 2} plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.

  10. Influence of stray light for divertor spectroscopy in ITER

    International Nuclear Information System (INIS)

    Kajita, Shin; Veshchev, Evgeny; Lisgo, Steve; Barnsley, Robin; Morgan, Philip; Walsh, Michael; Ogawa, Hiroaki; Sugie, Tatsuo; Itami, Kiyoshi

    2015-01-01

    The influence of stray light in the divertor spectroscopy system in ITER is quantitatively investigated using a ray tracing simulation. Simulation results show that the stray light is negligible at positions in the divertor where the plasma emission is strong. However, it is also shown that the stray light can be significantly greater than the real signal if the plasma intensity is low. Deuterium and beryllium emissions are used for the assessment; for beryllium cases in particular, since the emission profile may be non-uniform in the divertor region, the influence of stray light can be non-negligible at some positions, e.g., above the divertor dome

  11. Operational Characteristics of Liquid Lithium Divertor in NSTX

    Science.gov (United States)

    Kaita, R.; Kugel, H.; Abrams, T.; Bell, M. G.; Bell, R. E.; Gerhardt, S.; Jaworski, M. A.; Kallman, J.; Leblanc, B.; Mansfield, D.; Mueller, D.; Paul, S.; Roquemore, A. L.; Scotti, F.; Skinner, C. H.; Timberlake, J.; Zakharov, L.; Maingi, R.; Nygren, R.; Raman, R.; Sabbagh, S.; Soukhanovskii, V.

    2010-11-01

    Lithium coatings on plasma-facing components (PFC's) have resulted in improved plasma performance on NSTX in deuterium H-mode plasmas with neutral beam heating.^ Salient results included improved electron confinement and ELM suppression. In CDX-U, the use of lithium-coated PFC's and a large-area liquid lithium limiter resulted in a six-fold increase in global energy confinement time. A Liquid Lithium Divertor (LLD) has been installed in NSTX for the 2010 run campaign. The LLD PFC consists of a thin film of lithium on a temperature-controlled substrate to keep the lithium liquefied between shots, and handle heat loads during plasmas. This capability was demonstrated when the LLD withstood a strike point on its surface during discharges with up to 4 MW of neutral beam heating.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bazylev, B.N. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany)]. E-mail: bazylev@ihm.fzk.de; Janeschitz, G. [Forschungszentrum Karlsruhe, Fusion, P.O. Box 3640, 76021 Karlsruhe (Germany); Landman, I.S. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany); Loarte, A. [EFDA-CSU, Max-Planck-Institut fuer Plasmaphysik, D-85748 Garching (Germany); Pestchanyi, S.E. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2007-06-15

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

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

    Science.gov (United States)

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

    2007-06-01

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  15. Plasma Surface interaction in Controlled fusion devices

    International Nuclear Information System (INIS)

    1990-05-01

    The subjects presented in the 9th conference on plasma surface interaction in controlled fusion devices were: the modifications of power scrape-off-length and power deposition during various configurations in Tore Supra plasmas; the effects observed in ergodic divertor experiments in Tore-Supra; the diffuse connexion induced by the ergodic divertor and the topology of the heat load patterns on the plasma facing components in Tore-Supra; the study of the influence of air exposure on graphite implanted by low energy high density deuterium plasma

  16. 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)

  17. 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

  18. Characteristics of the Secondary Divertor on DIII-D

    Science.gov (United States)

    Watkins, J. G.; Lasnier, C. J.; Leonard, A. W.; Evans, T. E.; Pitts, R.; Stangeby, P. C.; Boedo, J. A.; Moyer, R. A.; Rudakov, D. L.

    2009-11-01

    In order to address a concern that the ITER secondary divertor strike plates may be insufficiently robust to handle the incident pulses of particles and energy from ELMs, we performed dedicated studies of the secondary divertor plasma and scrape-off layer (SOL). Detailed measurements of the ELM energy and particle deposition footprint on the secondary divertor target plates were made with a fast IR camera and Langmuir probes and SOL profile and transport measurements were made with reciprocating probes. The secondary divertor and SOL conditions depended on changes in the magnetic balance and the core plasma density. Larger density resulted in smaller ELMs and the magnetic balance affected how many ELM particles coupled to the secondary SOL and divertor. Particularly striking are the images from a new fast IR camera that resolve ELM heat pulses and show spiral patterns with multiple peaks during ELMs in the secondary divertor.

  19. 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-SR