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Sample records for iter divertor plasma

  1. 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

  2. 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

  3. 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.

  4. 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

  5. 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.

  6. 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.))

  7. 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

  8. 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)

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.)

  15. 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

  16. 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.))

  17. 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

  18. 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

  19. 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.)

  20. 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

  1. 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.

  2. 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

  3. 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)

  4. 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)

  5. 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

  6. 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

  7. Divertor development for ITER

    International Nuclear Information System (INIS)

    Janeschitz, G.; Ando, T.; Antipenkov, A.; Barabash, V.; Chiocchio, S.; Federici, G.; Ibbott, C.; Jakeman, R.; Matera, R.; Martin, E.; Parker, R.; Tivey, R.; Pacher, H.D.

    1998-01-01

    The requirements for the ITER divertor design, i.e. power and He ash exhaust, neutral leakage control, lifetime, disruption load resistance and exchange by remote handling, are described in this paper. These requirements and the physics requirements for detached and semi-attached operation result in the vertical target configuration. This is realised by a concept incorporating 60 cassettes carrying the high heat flux components. The armour choice for these components is CFC monoblock in the strike zone near at the lower part of the vertical target, and a W brush elsewhere. Cooling is by swirl tubes or hypervapotrons depending on the component. The status of the heat sink and joining technology R and D is given. Finally, the resulting design of the high heat flux components is presented. (orig.)

  8. 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.

  9. 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.

  10. 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.

  11. 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

  12. 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.

  13. 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.

  14. Structural analysis of the ITER Divertor toroidal rails

    Energy Technology Data Exchange (ETDEWEB)

    Viganò, F., E-mail: Fabio.Vigano@LTCalcoli.it [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy); Escourbiac, F.; Gicquel, S.; Komarov, V. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Lucca, F. [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy); Merola, M. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Ngnitewe, R. [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy)

    2013-10-15

    The Divertor is one of the most technically challenging components of the ITER machine, which has the main function of extracting the power conducted in the scrape-off layer while maintaining the plasma purity. There are 54 Divertor cassettes installed in the vacuum vessel (VV). Each cassette body (CB) is fastened on the inner and outer concentric Divertor toroidal rails. The comprehensive assessment (in accordance with the Structural Design Criteria for ITER In-vessel Components: ITER SDC-IC) of the Divertor toroidal rails has been performed during design activity based on performing of thermal and stress analyses at operating conditions of neutron stage of ITER operation. This paper outlines the engineering aspects of the ITER Divertor toroidal rails and focuses on some critical regions of the present design highlighted by the performed structural assessment. The structural assessment has been performed with help of using Finite Element (FE) Abaqus code and based on criteria given by ITER SDC-IC.

  15. 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.

  16. 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.

  17. 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.))

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  19. Thermal effects of divertor sweeping in ITER

    International Nuclear Information System (INIS)

    Wesley, J.C.

    1992-01-01

    In this paper, thermal effects of magnetically sweeping the separatrix strike point on the outer divertor target of the International Thermonuclear Fusion Reactor (ITER) are calculated. For the 0. 2 Hz x ± 12 cm sweep scenario proposed for ITER operations, the thermal capability of a generic target design is found to be slightly inadequate (by ∼ 5%) to accommodate the full degree of plasma scrape-off peaking postulated as a design basis. The principal problem identified is that the 5 s sweep period is long relative to the 1. 4 s thermal time constant of the divertor target. An increase of the sweep frequency to ∼ 1 Hz is suggested: this increase would provide a power handling margin of ∼ 25% relative to present operational criteria

  20. 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

  1. 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

  2. The ITER divertor cassette project meeting

    International Nuclear Information System (INIS)

    Merola, M.; Riccardi, B.; Tivey, R.

    1999-01-01

    The Divertor Cassette Project topical meeting was held on May 26-28, 1999 at the ENEA Brasimone Research Centre in Camugnano (Bologna), Italy. Specialists from all the four Parties and the JCT participated in the meeting. It was concluded that the Divertor Cassette Project has significantly contributed to solving a large part of the critical issues of the ITER divertor design

  3. 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 .

  4. 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

  5. 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)

  6. 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.)

  7. 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.

  8. 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

  9. 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

  10. Divertor cassette movers prototypes for ITER

    International Nuclear Information System (INIS)

    Bogusch, E.; Batz, R.; Bieber, O.; Gottfried, R.; Cerdan, G.

    1998-01-01

    Following competitive tendering, in October 1996 Siemens was contracted by the European Commission to design and supply an assembly of four Divertor Cassette Movers Prototypes including the control and command systems for the movers proper. The assembly consisting of one Cassette Toroidal Mover (CTM), one Radial Mover Tractor (TRC), one Second Cassette Carrier (SCC), and one Radial Cassette Carrier (RCC) represents key components of the Divertor Test Platform at Brasimone, one of the seven large R+D projects for ITER. By detailed design, high-precision manufacturing and testing of these devices, Siemens contributed to the verification of an important task within the European R and D program towards ITER construction. Replacement of the divertor cassettes is a scheduled maintenance operation throughout the life of ITER. The successful fabrication and testing of the Divertor Cassette Movers Prototypes is all important milestone to verify this delicate operation. (authors)

  11. Latest status of manufacturing activity of ITER divertor and engineering issues on tungsten divertor

    International Nuclear Information System (INIS)

    Suzuki, Satoshi

    2011-01-01

    Divertors for ITER are now in construction. In the present chapter, the specification and the latest status of manufacturing of ITER divertors are presented. In addition, issues in the development of divertors for the fusion demo reactor are given on the basis of experiences on the ITER divertor development. (J.P.N.)

  12. Aberrations in preliminary design of ITER divertor impurity influx monitor

    Energy Technology Data Exchange (ETDEWEB)

    Kitazawa, Sin-iti, E-mail: kitazawa.siniti@jaea.go.jp [Naka Fusion Institute, Japan Atomic Energy Agency, JAEA, Naka 311-0193 (Japan); Ogawa, Hiroaki [Naka Fusion Institute, Japan Atomic Energy Agency, JAEA, Naka 311-0193 (Japan); Katsunuma, Atsushi; Kitazawa, Daisuke [Core Technology Center, Nikon Corporation, Yokohama 244-8533 (Japan); Ohmori, Keisuke [Customized Products Business Unit, Nikon Corporation, Mito 310-0843 (Japan)

    2015-12-15

    Highlights: • Divertor impurity influx monitor for ITER (DIM) is procured by JADA. • DIM is designed to observe light from nuclear fusion plasma directly. • DIM is under preliminary design phase. • The spot diagrams were suppressed within the core of receiving fiber. • The aberration of DIM is suppressed in the preliminary design. - Abstract: Divertor impurity influx monitor for ITER (DIM) is a diagnostic system that observes light from nuclear fusion plasma directly. This system is affected by various aberrations because it observes light from the fan-array chord near the divertor in the ultraviolet–near infrared wavelength range. The aberrations should be suppressed to the extent possible to observe the light with very high spatial resolution. In the preliminary design of DIM, spot diagrams were suppressed within the core of the receiving fiber's cross section, and the resulting spatial resolutions satisfied the design requirements.

  13. Aberrations in preliminary design of ITER divertor impurity influx monitor

    International Nuclear Information System (INIS)

    Kitazawa, Sin-iti; Ogawa, Hiroaki; Katsunuma, Atsushi; Kitazawa, Daisuke; Ohmori, Keisuke

    2015-01-01

    Highlights: • Divertor impurity influx monitor for ITER (DIM) is procured by JADA. • DIM is designed to observe light from nuclear fusion plasma directly. • DIM is under preliminary design phase. • The spot diagrams were suppressed within the core of receiving fiber. • The aberration of DIM is suppressed in the preliminary design. - Abstract: Divertor impurity influx monitor for ITER (DIM) is a diagnostic system that observes light from nuclear fusion plasma directly. This system is affected by various aberrations because it observes light from the fan-array chord near the divertor in the ultraviolet–near infrared wavelength range. The aberrations should be suppressed to the extent possible to observe the light with very high spatial resolution. In the preliminary design of DIM, spot diagrams were suppressed within the core of the receiving fiber's cross section, and the resulting spatial resolutions satisfied the design requirements.

  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. Physics conclusions in support of ITER W divertor monoblock shaping

    Czech Academy of Sciences Publication Activity Database

    Pitts, R.A.; Bardin, S.; Bazylev, B.; van den Berg, M.A.; Bunting, P.; Carpentier-Chouchana, S.; Coenen, J.W.; Corre, Y.; Dejarnac, Renaud; Escourbiac, F.; Gaspar, J.; Gunn, J. P.; Hirai, T.; Hong, S.-H.; Horáček, Jan; Iglesias, D.; Komm, Michael; Krieger, K.; Lasnier, C.; Matthews, G.F.; Morgan, T.W.; Panayotis, S.; Pestchanyi, S.; Podolník, Aleš; Nygren, R.E.; Rudakov, D.L.; De Temmerman, G.; Vondráček, Petr; Watkins, J.G.

    2017-01-01

    Roč. 12, August (2017), s. 60-74 ISSN 2352-1791. [International Conference on Plasma Surface Interactions 2016, PSI2016 /22./. Roma, 30.05.2016-03.06.2016] Institutional support: RVO:61389021 Keywords : ITER * Tungsten * Divertor * Shaping * Melting * MEMOS Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) http://www.sciencedirect.com/science/ article /pii/S2352179116302885

  16. Divertor armour issues: lifetime, safety and influence on ITER performance

    International Nuclear Information System (INIS)

    Pestchanyi, S.

    2009-01-01

    Comprehensive simulations of the ITER divertor armour vaporization and brittle destruction under ELMs of different sizes have revealed that the erosion rate of CFC armour is intolerable for an industrial reactor, but it can be considerably reduced by the armour fibre structure optimization. The ITER core contamination with carbon is tolerable for medium size ELMs, but large type I ELM can run the confinement into the disruption. Erosion of tungsten, an alternative armour material, under ELMs influence is satisfactory, but the danger of the core plasma contamination with tungsten is still not enough understood and potentially it could be very dangerous. Vaporization of tungsten, its cracking and dust production during ELMs are rather urgent issues to be investigated for proper choice of the divertor armour material for ITER. However, the erosion rate under action of the disruptive heat loads is tolerable for both armour materials assuming few hundred disruptions falls out during ITER lifetime

  17. Surface heat loads on the ITER divertor vertical targets

    Czech Academy of Sciences Publication Activity Database

    Gunn, J. P.; Carpentier-Chouchana, S.; Escourbiac, F.; Hirai, T.; Panayotis, S.; Pitts, R.A.; Corre, Y.; Dejarnac, Renaud; Firdaouss, M.; Kočan, M.; Komm, Michael; Kukushkin, A.; Languille, P.; Missirlian, M.; Zhao, W.; Zhong, G.

    2017-01-01

    Roč. 57, č. 4 (2017), č. článku 046025. ISSN 0029-5515 Institutional support: RVO:61389021 Keywords : ITER * divertor * ELM heat load * inter-ELM heat load * tungsten 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/10.1088/1741-4326/aa5e2a

  18. Surface mechanical attrition treatment of tungsten and its behavior under low energy deuterium plasma implantation relevant to ITER divertor conditions

    Energy Technology Data Exchange (ETDEWEB)

    Xu, H.Y.; Yuan, Y.; Fu, B.Q.; Godfrey, A.; Liu, W. [Tsinghua Univ.. Lab. of Advanced Materials, Beijing (China); Zhang, Y.B. [Technical Univ. og Denmark. DTU Risoe Campus, Roskilde (Denmark); Tao, N.R. [Chinese Academy of Sciences, Shenyang (China)

    2012-11-01

    In the light of a foreseen application for tungsten (W) as an ITER divertor material samples have been plastically deformed by a surface mechanical attrition treatment (SMAT) and by cold rolling. The resistance to blister formation by low energy deuterium implantation in these samples has been examined, with the result that the structure is significantly improved as the structural scale is reduced to the nanometer range in the SMAT sample. The distribution of blisters in this sample is however bimodal, due to the formation of several very large blisters, which are heterogeneously distributed. The observations suggest that process optimization must be a next step in the development with a view to the application of plastically deformed W in a fusion reactor. (Author)

  19. 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

  20. 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.

  1. 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

  2. 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.

  3. 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)

  4. 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

  5. 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

  6. Particle-in-cell simulations of the plasma interaction with poloidal gaps in the ITER divertor outer vertical target.

    Czech Academy of Sciences Publication Activity Database

    Komm, Michael; Gunn, J. P.; Dejarnac, Renaud; Pánek, Radomír; Pitts, R.A.; Podolník, Aleš

    2017-01-01

    Roč. 57, č. 12 (2017), č. článku 126047. ISSN 0029-5515 R&D Projects: GA ČR(CZ) GA16-14228S; GA MŠk(CZ) 8D15001 Grant - others:Ga MŠk(CZ) LM2015070 Institutional support: RVO:61389021 Keywords : tokamak * plasma * ITER * particle-in-cell * heat loads * monoblock 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/10.1088/1741-4326/aa8a9a/meta

  7. Simulation of divertor targets shielding during transients in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Pestchanyi, Sergey, E-mail: serguei.pestchanyi@kit.edu [KIT, Hermann-von-Helmholtz-Platz 1, Eggenstein-Leopoldshafen (Germany); Pitts, Richard; Lehnen, Michael [ITER Organization,Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2016-11-01

    Highlights: • We simulated plasma shielding effect during disruption in ITER using the TOKES code. • It has been found that vaporization is unavoidable under action of ITER transients, but plasma shielding drastically reduces the divertor target damage: the melt pool and the vaporization region widths reduced 10–15 times. • A simplified 1D model describing the melt pool depth and the shielded heat flux to the divertor targets have been developed. • The results of the TOKES simulations have been compared with the analytic model when the model is valid. - Abstract: Direct extrapolation of the disruptive heat flux on ITER conditions predicts severe melting and vaporization of the divertor targets causing their intolerable damage. However, tungsten vaporized from the target at initial stage of the disruption can create plasma shield in front of the target, which effectively protects the target surface from the rest of the heat flux. Estimation of this shielding efficiency has been performed using the TOKES code. The shielding effect under ITER conditions is found to be very strong: the maximal depth of the melt layer reduced 4 times, the melt layer width—more than 10 times and vaporization region shrinks 10–15 times due to shielding for unmitigated disruption of 350 MJ discharge. The simulation results show complex, 2D plasma dynamics of the shield under ITER conditions. However, a simplified analytic model, valid for rough estimation of the maximum value for the shielded flux to the target and for the melt depth at the target surface has been developed.

  8. 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)

  9. Application of powerful quasi-steady-state plasma accelerators for simulation of ITER transient heat loads on divertor surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Tereshin, V I [Institute of Plasma Physics of the NSC KIPT, Kharkov 61108 (Ukraine); Bandura, A N [Institute of Plasma Physics of the NSC KIPT, Kharkov 61108 (Ukraine); Byrka, O V [Institute of Plasma Physics of the NSC KIPT, Kharkov 61108 (Ukraine); Chebotarev, V V [Institute of Plasma Physics of the NSC KIPT, Kharkov 61108 (Ukraine); Garkusha, I E [Institute of Plasma Physics of the NSC KIPT, Kharkov 61108 (Ukraine); Landman, I [Forschungszentrum Karlsruhe, IHM, Karlsruhe 76021 (Germany); Makhlaj, V A [Institute of Plasma Physics of the NSC KIPT, Kharkov 61108 (Ukraine); Neklyudov, I M [Institute of Plasma Physics of the NSC KIPT, Kharkov 61108 (Ukraine); Solyakov, D G [Institute of Plasma Physics of the NSC KIPT, Kharkov 61108 (Ukraine); Tsarenko, A V [Institute of Plasma Physics of the NSC KIPT, Kharkov 61108 (Ukraine)

    2007-05-15

    The paper presents the investigations of high power plasma interaction with material surfaces under conditions simulating the ITER disruptions and type I ELMs. Different materials were exposed to plasma with repetitive pulses of 250 {mu}s duration, the ion energy of up to 0.6 keV, and the heat loads varying in the 0.5-25 MJ m{sup -2} range. The plasma energy transfer to the material surface versus impact load has been analysed. The fraction of plasma energy that is absorbed by the target surface is rapidly decreased with the achievement of the evaporation onset for exposed targets. The distributions of evaporated material in front of the target surface and the thickness of the shielding layer are found to be strongly dependent on the target atomic mass. The surface analysis of tungsten targets exposed to quasi-steady-state plasma accelerators plasma streams is presented together with measurements of the melting onset load and evaporation threshold, and also of erosion patterns with increasing heat load and the number of plasma pulses.

  10. Is Carbon a Realistic Choice for ITER's Divertor?

    International Nuclear Information System (INIS)

    Skinner, C.H.; Federici, G.

    2005-01-01

    Tritium retention by co-deposition with carbon on the divertor target plate is predicted to limit ITER's DT burning plasma operations (e.g. to about 100 pulses for the worst conditions) before the in-vessel tritium inventory limit, currently set at 350 g, is reached. At this point, ITER will only be able to continue its burning plasma program if technology is available that is capable of rapidly removing large quantities of tritium from the vessel with over 90% efficiency. The removal rate required is four orders of magnitude faster than that demonstrated in current tokamaks. Eighteen years after the observation of co-deposition on JET and TFTR, such technology is nowhere in sight. The inexorable conclusion is that either a major initiative in tritium removal should be funded or that research priorities for ITER should focus on metal alternatives

  11. 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.

  12. Development of the armoring technique for ITER Divertor Dome

    Energy Technology Data Exchange (ETDEWEB)

    Litunovsky, Nikolay, E-mail: nlitunovsky@sintez.niiefa.spb.su [D.V. Efremov Reseasch Institute, 3, Doroga na Metallostroy, Saint Petersburg (Russian Federation); Alekseenko, Evgeny; Makhankov, Alexey; Mazul, Igor [D.V. Efremov Reseasch Institute, 3, Doroga na Metallostroy, Saint Petersburg (Russian Federation)

    2011-10-15

    This paper describes the current status of the technique for armoring of Plasma Facing Units (PFUs) of the ITER Divertor Dome with flat tungsten tiles planned for application at the procurement stage. Application of high-temperature vacuum brazing for armoring of High Heat Flux (HHF) plasma facing components was traditionally developed at the Efremov Institute and successfully tried out at the ITER R and D stage by manufacturing and HHF testing of a number of W- and Be-armored mock-ups . Nevertheless, the so-called 'fast brazing' technique successfully applied in the past was abandoned at the stage of manufacturing of the Dome Qualification Prototypes (Dome QPs), as it failed to retain the mechanical properties of CuCrZr heat sink of the substrate. Another problem was a substantially increased number of armoring tiles brazed onto one substrate. Severe ITER requirements for the joints quality have forced us to refuse from production of W/Cu joints by brazing in favor of casting. These modifications have allowed us to produce ITER Divertor Dome QPs with high-quality tungsten armor, which then passed successfully the HHF testing. Further preparation to the procurement stage is in progress.

  13. Behavior of divertor and first wall armour materials at plasma heat fluxes relevant to ITER ELMs and disruptions

    Directory of Open Access Journals (Sweden)

    D.V. Kovalenko

    2017-08-01

    Full Text Available The paper presents the main results of numerous experiments carried out over the past 10 years at QSPA-T and QSPA-Be plasma guns in support of ITER. Special targets made of pure W, W-1%La2O3 and two types of Be (TGP-56FW and S65-C were tested under the series of repeated plasma stream and photonic flux impact. Maximum heat load on the target surface was up to 2.5MJ/m2 in the case of plasma testing and was equal to 0.5MJ/m2 in the case of photonic flux testing. Pulse waveform was rectangular with tpulse= 0.5ms. It was found that the main erosion mechanisms of W and Be under plasma stream impact are the melt layer movement, the ejection of droplets and the cracks formation. As a result of repeated photonic fluxes a regular, “corrugated” structure are eventually formed on the Be target surface. Study of erosion products of W formed under plasma stream impact on the W target has shown that the D/W atomic ratio in the deposited W films during pulsed events may be the same or even higher than that for stationary processes.

  14. 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)

  15. 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.

  16. Fluid simulation of beryllium transport in the ITER gaseous divertor

    International Nuclear Information System (INIS)

    Knoll, D.A.; Campbell, R.B.; McHugh, P.R.

    1994-01-01

    The transport of either intrinsic or injected impurities will play a crucial role in the energy loss mechanisms in the ITER gaseous/cold plasma target divertor. Both 1-D and 2-D multi-charge state fluid codes are used to model the transport of beryllium in the ITER SOL. Our major conclusion is that in order to model the containment of impurities, the background flow field must be known in detail. Comparing 1-D and 2-D solutions, hydrogen flow reversal plays an important role in the entrainment process. Further, the flow of particles from the core plasma also has a strong impact on the resultant entrainment of the impurities in both 1-D and 2-D. It is imperative that those components of poloidal velocity due to E x B and diamagnetic drifts be included in the models. (orig.)

  17. The WEST project: Current status of the ITER-like tungsten divertor

    International Nuclear Information System (INIS)

    Missirlian, M.; Bucalossi, J.; Corre, Y.; Ferlay, F.; Firdaouss, M.; Garin, P.; Grosman, A.; Guilhem, D.; Gunn, J.; Languille, P.; Lipa, M.; Richou, M.; Tsitrone, E.

    2014-01-01

    Highlights: • We presented the ITER-like W components occurred for the WEST divertor. • The main features including key elements of the design were detailed. • The main results of studies investigating the integration constraints or issues were reported. • The WEST ITER-like divertor design reached a mature stage to enable the launching of the procurement phase. - Abstract: The WEST (W – for tungsten – Environment in Steady-state Tokamak) project is an upgrade of Tore Supra from a limiter based tokamak with carbon PFCs into an X-point divertor tokamak with full-tungsten armour while keeping its long discharge capability. The WEST project will primarily offer the key capability of testing for the first time the ITER technology in real plasma environment. In particular, the main divertor (i.e. the lower divertor) of the WEST project will be based on actively cooled tungsten monoblock components and will follow as closely as possible the design and the assembling technology, foreseen for the ITER divertor units. The current design of WEST ITER-like tungsten divertor has now reached a mature stage following the 2013 WEST Final Design Review. This paper presents the key elements of the design, reports the technological requirements and reviews the main design and integration issues

  18. The WEST project: Current status of the ITER-like tungsten divertor

    Energy Technology Data Exchange (ETDEWEB)

    Missirlian, M., E-mail: marc.missirlian@cea.fr; Bucalossi, J.; Corre, Y.; Ferlay, F.; Firdaouss, M.; Garin, P.; Grosman, A.; Guilhem, D.; Gunn, J.; Languille, P.; Lipa, M.; Richou, M.; Tsitrone, E.

    2014-10-15

    Highlights: • We presented the ITER-like W components occurred for the WEST divertor. • The main features including key elements of the design were detailed. • The main results of studies investigating the integration constraints or issues were reported. • The WEST ITER-like divertor design reached a mature stage to enable the launching of the procurement phase. - Abstract: The WEST (W – for tungsten – Environment in Steady-state Tokamak) project is an upgrade of Tore Supra from a limiter based tokamak with carbon PFCs into an X-point divertor tokamak with full-tungsten armour while keeping its long discharge capability. The WEST project will primarily offer the key capability of testing for the first time the ITER technology in real plasma environment. In particular, the main divertor (i.e. the lower divertor) of the WEST project will be based on actively cooled tungsten monoblock components and will follow as closely as possible the design and the assembling technology, foreseen for the ITER divertor units. The current design of WEST ITER-like tungsten divertor has now reached a mature stage following the 2013 WEST Final Design Review. This paper presents the key elements of the design, reports the technological requirements and reviews the main design and integration issues.

  19. 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   ×  \

  20. 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)

  1. 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

  2. Design of ITER divertor VUV spectrometer and prototype test at KSTAR tokamak

    Science.gov (United States)

    Seon, Changrae; Hong, Joohwan; Song, Inwoo; Jang, Juhyeok; Lee, Hyeonyong; An, Younghwa; Kim, Bosung; Jeon, Taemin; Park, Jaesun; Choe, Wonho; Lee, Hyeongon; Pak, Sunil; Cheon, MunSeong; Choi, Jihyeon; Kim, Hyeonseok; Biel, Wolfgang; Bernascolle, Philippe; Barnsley, Robin; O'Mullane, Martin

    2017-12-01

    Design and development of the ITER divertor VUV spectrometer have been performed from the year 1998, and it is planned to be installed in the year 2027. Currently, the design of the ITER divertor VUV spectrometer is in the phase of detail design. It is optimized for monitoring of chord-integrated VUV signals from divertor plasmas, chosen to contain representative lines emission from the tungsten as the divertor material, and other impurities. Impurity emission from overall divertor plasmas is collimated through the relay optics onto the entrance slit of a VUV spectrometer with working wavelength range of 14.6-32 nm. To validate the design of the ITER divertor VUV spectrometer, two sets of VUV spectrometers have been developed and tested at KSTAR tokamak. One set of spectrometer without the field mirror employs a survey spectrometer with the wavelength ranging from 14.6 nm to 32 nm, and it provides the same optical specification as the spectrometer part of the ITER divertor VUV spectrometer system. The other spectrometer with the wavelength range of 5-25 nm consists of a commercial spectrometer with a concave grating, and the relay mirrors with the same geometry as the relay mirrors of the ITER divertor VUV spectrometer. From test of these prototypes, alignment method using backward laser illumination could be verified. To validate the feasibility of tungsten emission measurement, furthermore, the tungsten powder was injected in KSTAR plasmas, and the preliminary result could be obtained successfully with regard to the evaluation of photon throughput. Contribution to the Topical Issue "Atomic and Molecular Data and their Applications", edited by Gordon W.F. Drake, Jung-Sik Yoon, Daiji Kato, Grzegorz Karwasz.

  3. Estimation of the contribution of gaps to tritium retention in the divertor of ITER

    Czech Academy of Sciences Publication Activity Database

    Matveev, D.; Kirschner, A.; Schmid, K.; Litnovsky, A.; Borodin, D.; Komm, Michael; Van Oost, G.; Samm, U.

    -, T159 (2014), 014063-014063 ISSN 0031-8949 Institutional support: RVO:61389021 Keywords : plasma * tokamak * tritium retention * ITER * castellated surfaces * gaps * divertor * impurity deposition Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.126, year: 2014 http://iopscience.iop.org/1402-4896/2014/T159/014063/

  4. ITER divertor, design issues and research and development

    International Nuclear Information System (INIS)

    Tivey, R.; Ando, T.; Antipenkov, A.; Barabash, V.; Chiocchio, S.; Federici, G.; Ibbott, C.; Jakeman, R.; Janeschitz, G.; Raffray, R.; Mazul, I.; Pacher, H.; Ulrickson, M.; Vieider, G.

    1999-01-01

    Over the period of the ITER Engineering Design Activity (EDA) the results from physics experiments, modelling, engineering analyses and R and D, have been brought together to provide a design for an ITER divertor. The design satisfies all necessary requirements for steady state and transient heat flux, nuclear shielding, pumping, tritium inventory, impurity control, armour lifetime, electromagnetic loads, diagnostics, and remote maintenance. The design consists of 60 cassettes each comprising a cassette body onto which the plasma facing components (PFCs) are mounted. Each cassette is supported by toroidal rails which are attached to the vacuum vessel. For the PFCs the final armour choice is carbon-fibre-composite (CfC) for the strike point regions and tungsten in all remaining areas. R and D has demonstrated that CfC monoblocks can routinely withstand heat loads up to 20 MW m -2 10 MW m -2 . Analysis and experiment show that a CfC armour thickness of ∝20 mm will provide sufficient lifetime for at least 1000 full power pulses. The thickness of the cassette body is sufficient to shield the vacuum vessel, so that, if necessary, rewelding is possible, and also provides sufficient stiffness against electromagnetically generated loads. The cassette design provides efficient and proven remote maintenance which should allow exchange of a complete divertor within ∝6 months. (orig.)

  5. ITER divertor, design issues and research and development

    Energy Technology Data Exchange (ETDEWEB)

    Tivey, R.; Ando, T.; Antipenkov, A.; Barabash, V.; Chiocchio, S.; Federici, G.; Ibbott, C.; Jakeman, R.; Janeschitz, G.; Raffray, R. [ITER Joint Central Team, Garching (Germany). Joint Central Work Site; Akiba, M. [Japan Atomic Energy Research Institute, Naka-machi, Ibaraki-ken (Japan); Mazul, I. [Efremov Institute, St Petersburg (Russian Federation); Pacher, H. [NET Team, Boltzmannstr. 2, D-85748, Garching (Germany); Ulrickson, M. [Sandia National Laboratories, Albuquerque, NM (United States); Vieider, G. [NET Team, Boltzmannstr. 2, D-85748, Garching (Germany)

    1999-11-01

    Over the period of the ITER Engineering Design Activity (EDA) the results from physics experiments, modelling, engineering analyses and R and D, have been brought together to provide a design for an ITER divertor. The design satisfies all necessary requirements for steady state and transient heat flux, nuclear shielding, pumping, tritium inventory, impurity control, armour lifetime, electromagnetic loads, diagnostics, and remote maintenance. The design consists of 60 cassettes each comprising a cassette body onto which the plasma facing components (PFCs) are mounted. Each cassette is supported by toroidal rails which are attached to the vacuum vessel. For the PFCs the final armour choice is carbon-fibre-composite (CfC) for the strike point regions and tungsten in all remaining areas. R and D has demonstrated that CfC monoblocks can routinely withstand heat loads up to 20 MW m{sup -2}10 MW m{sup -2}. Analysis and experiment show that a CfC armour thickness of {proportional_to}20 mm will provide sufficient lifetime for at least 1000 full power pulses. The thickness of the cassette body is sufficient to shield the vacuum vessel, so that, if necessary, rewelding is possible, and also provides sufficient stiffness against electromagnetically generated loads. The cassette design provides efficient and proven remote maintenance which should allow exchange of a complete divertor within {proportional_to}6 months. (orig.)

  6. 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.

  7. An exploration of advanced X-divertor scenarios on ITER

    Science.gov (United States)

    Covele, B.; Valanju, P.; Kotschenreuther, M.; Mahajan, S.

    2014-07-01

    It is found that the X-divertor (XD) configuration (Kotschenreuther et al 2004 Proc. 20th Int. Conf. on Fusion Energy (Vilamoura, Portugal, 2004) (Vienna: IAEA) CD-ROM file [IC/P6-43] www-naweb.iaea.org/napc/physics/fec/fec2004/datasets/index.html, Kotschenreuther et al 2006 Proc. 21st Int. Conf. on Fusion Energy 2006 (Chengdu, China, 2006) (Vienna: IAEA), CD-ROM file [IC/P7-12] www-naweb.iaea.org/napc/physics/FEC/FEC2006/html/index.htm, Kotschenreuther et al 2007 Phys. Plasmas 14 072502) can be made with the conventional poloidal field (PF) coil set on ITER (Tomabechi et al and Team 1991 Nucl. Fusion 31 1135), where all PF coils are outside the TF coils. Starting from the standard divertor, a sequence of desirable XD configurations are possible where the PF currents are below the present maximum design limits on ITER, and where the baseline divertor cassette is used. This opens the possibility that the XD could be tested and used to assist in high-power operation on ITER, but some further issues need examination. Note that the increased major radius of the super-X-divertor (Kotschenreuther et al 2007 Bull. Am. Phys. Soc. 53 11, Valanju et al 2009 Phys. Plasmas 16 5, Kotschenreuther et al 2010 Nucl. Fusion 50 035003, Valanju et al 2010 Fusion Eng. Des. 85 46) is not a feature of the XD geometry. In addition, we present an XD configuration for K-DEMO (Kim et al 2013 Fusion Eng. Des. 88 123) to demonstrate that it is also possible to attain the XD configuration in advanced tokamak reactors with all PF coils outside the TF coils. The results given here for the XD are far more encouraging than recent calculations by Lackner and Zohm (2012 Fusion Sci. Technol. 63 43) for the Snowflake (Ryutov 2007 Phys. Plasmas 14 064502, Ryutov et al 2008 Phys. Plasmas 15 092501), where the required high PF currents represent a major technological challenge. The magnetic field structure in the outboard divertor SOL (Kotschenreuther 2013 Phys. Plasmas 20 102507) in the recently created

  8. 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.

  9. Physics conclusions in support of ITER W divertor monoblock shaping

    Directory of Open Access Journals (Sweden)

    R.A. Pitts

    2017-08-01

    Full Text Available The key remaining physics design issue for the ITER tungsten (W divertor is the question of monoblock (MB front surface shaping in the high heat flux target areas of the actively cooled targets. Engineering tolerance specifications impose a challenging maximum radial step between toroidally adjacent MBs of 0.3mm. Assuming optical projection of the parallel heat loads, magnetic shadowing of these edges is required if quasi-steady state melting is to be avoided under certain conditions during burning plasma operation and transiently during edge localized mode (ELM or disruption induced power loading. An experiment on JET in 2013 designed to investigate the consequences of transient W edge melting on ITER, found significant deficits in the edge power loads expected on the basis of simple geometric arguments, throwing doubt on the understanding of edge loading at glancing field line angles. As a result, a coordinated multi-experiment and simulation effort was initiated via the International Tokamak Physics Activity (ITPA and through ITER contracts, aimed at improving the physics basis supporting a MB shaping decision from the point of view both of edge power loading and melt dynamics. This paper reports on the outcome of this activity, concluding first that the geometrical approximation for leading edge power loading on radially misaligned poloidal leading edges is indeed valid. On this basis, the behaviour of shaped and unshaped monoblock surfaces under stationary and transient loads, with and without melting, is compared in order to examine the consequences of melting, or power overload in context of the benefit, or not, of shaping. The paper concludes that MB top surface shaping is recommended to shadow poloidal gap edges in the high heat flux areas of the ITER divertor targets.

  10. 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

  11. 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

    The design of the divertor and its integration into the reduced technical objectives/reduced cost-international thermonuclear energy reactor (RTO/RC-ITER) is based on the experience gained from the 1998 design of international thermonuclear energy reactor (ITER) and on the research and development performed throughout the engineering design activities (EDA). This paper gives an overview of the layout and functional design of the RTO/RC-ITER divertor, including the integration into the machine and the remote replacement of the divertor cassettes. Design guidelines are presented which have allowed quick preparation of divertor layouts suitable for further study using the B2-EIRENE edge plasma code. As in the 1998 design, the divertor is segmented into cassettes, and the segmentation, which is three per sector, is driven by access through the divertor level ports. Maintaining this access and avoiding interference with poloidal field coils means that the divertor level ports need to be inclined (7 deg.). This opens up the possibility of incorporating inboard and outboard baffles into the divertor cassettes. The cassettes are transported in-vessel by making use of the toroidal rails onto which the cassettes are finally clamped in position. Significant reduction of the space available between the X-point and the vacuum vessel results in re-positioning of the toroidal rails in order to retain sufficient depth for the inner and outer divertor legs. This, in turn, requires some changes to the remote handling (RH) concept. Remote handling (RH) is now based on using a cantilevered articulated gripper during the radial movement of the cassettes inside the RH ports. However, the principle to use a cassette toroidal mover (CTM) for in vessel handling is unchanged, hence maintaining the validity of previous EDA research and development. The space previously left below the cassettes for RH was also used for pumping. Elimination of this space has led to re-siting of the pumping

  12. Effects of ELMs on ITER divertor armour materials

    Energy Technology Data Exchange (ETDEWEB)

    Zhitlukhin, A. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation)]. E-mail: zhitlukh@triniti.ru; Klimov, N. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Landman, I. [Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe (Germany); Linke, J. [Forschungszentrum Juelich, EURATOM-Association, Juelich (Germany)]. E-mail: j.linke@fz-juelich.de; Loarte, A. [EFDA, Boltzmannstr. 2, 85748 Garching (Germany); Merola, M. [EFDA, Boltzmannstr. 2, 85748 Garching (Germany); Podkovyrov, V. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Federici, G. [ITER JWS Garching, Boltzmannstr. 2, 85748 Garching (Germany); Bazylev, B. [Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe (Germany); Pestchanyi, S. [Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe (Germany); Safronov, V. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Hirai, T. [Forschungszentrum Juelich, EURATOM-Association, Juelich (Germany); Maynashev, V. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Levashov, V. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Muzichenko, A. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation)

    2007-06-15

    This paper is concerned with investigation of an erosion of the ITER-like divertor plasma facing components under plasma heat loads expected during the Type I ELMs in ITER. These experiments were carried out on plasma accelerator QSPA at the SRC RF TRINITI under EU/RF collaboration. Targets were exposed by series repeated plasma pulses with heat loads in a range of 0.5-1.5 MJ/m{sup 2} and pulse duration 0.5 ms. Erosion of CFC macrobrushes was determined mainly by sublimation of PAN-fibres that was less than 2.5 {mu}m per pulse. The CFC erosion was negligible at the energy density less than 0.5 MJ/m{sup 2} and was increased to the average value 0.3 {mu}m per pulse at 1.5 MJ/m{sup 2}. The pure tungsten macrobrushes erosion was small in the energy range of 0.5-1.3 MJ/m{sup 2}. The sharp growth of tungsten erosion and the intense droplet ejection were observed at the energy density of 1.5 MJ/m{sup 2}.

  13. Effects of ELMs on ITER divertor armour materials

    Science.gov (United States)

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

    2007-06-01

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

  14. Advances in optical thermometry for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Lott, F. [CEA, IRFM, F-13108 St Paul lez Durance (France)], E-mail: fraser.lott@gmail.com; Netchaieff, A. [Laboratoire National de Metrologie et d' Essais (LNE), ZA de Trappes-Elancourt, 29 avenue Roger Hennequin, 78197 TRAPPES Cedex (France); Escourbiac, F. [CEA, IRFM, F-13108 St Paul lez Durance (France); Jouvelot, J.-L.; Constans, S. [AREVA NP, Centre Technique-FE200, Porte Magenta BP 181, 71205 Le Creusot (France); Hernandez, D. [Procedes, Materiaux et Energie Solaire (PROMES), Centre National de la Recherche Scientifique (CNRS), B.P. 5, 66125 Font-Romeu Cedex (France)

    2010-01-15

    Thermography will be an important diagnostic on the ITER tokamak, but the inclusion of reflective materials such as tungsten in the design for ITER's first wall and divertor region presents problems for optical temperature measurement. The ongoing testing of ITER plasma facing components (PFCs) provides an excellent opportunity to resolve such problems. This has focused on the variation of PFC emissivity with temperature and time, as well as environmental influence on thermography. The sensitivity of these systems to ambient temperature, due primarily to modification of the transmission of the optical path, has been established and minimised. The accuracy of the system is then sufficient to measure the variation of emissivity in heated material samples, by comparing its front-face luminance measured with an infrared camera to the temperature given by an implanted thermocouple. Measurements on both tungsten and carbon fibre composite are in broad agreement with theory, and thus give the material's function of emissivity with temperature at the start of its life. To determine its evolution, a bicolour pyroreflectometer was then installed. This uses two lasers to measure the reflectivity in addition to the luminance at two wavelengths, and thus the true temperature can be calculated. This was validated against the instrumented sample, then used along with the camera to observe an ITER mock-up during {approx}50,000 s of 5 MW/m{sup 2} testing. Emissivity was seen to vary little in the 500 deg. C region. Higher temperature tests are ongoing.

  15. 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

  16. Design study of ITER-like divertor target for DEMO

    International Nuclear Information System (INIS)

    Crescenzi, Fabio; Bachmann, C.; Richou, M.; Roccella, S.; Visca, E.; You, J.-H.

    2015-01-01

    Highlights: • ‘DEMO’ is a near-term Power Plant Conceptual Study (PPCS). • The ITER-like design concept represents a promising solution also for DEMO plasma facing units. • The optimization of PFUs aims to enhance the thermo-mechanical behaviour of the component. • The optimized geometry was evaluated by ITER SDC-IC criteria and in terms of low cycle fatigue (LCF). - Abstract: A near-term water-cooled target solution has to be evaluated together with the required technologies and its power exhaust limit under ‘DEMO’ conditions. The ITER-like design concept based on the mono-block technology using W as armour material and the CuCrZr-IG as structural material with an interlayer of pure copper represents a promising solution also for DEMO. This work reports the design study of an “optimized” ITER-like Water Cooled Divertor able to withstand a heat flux of 10 MW m"−"2, as requested for DEMO operating conditions. The optimization of plasma facing unit (PFU) aims to enhance the thermo-mechanical behaviour of the component by varying some geometrical parameters (monoblock size, interlayer thickness and, tube diameter and thickness). The optimization was performed by means of the multi-variable optimization algorithms using the FEM code ANSYS. The coolant hydraulic conditions (inlet pressure, temperature and velocity) were fixed for simplicity. This study is based on elastic analysis and 3 dimensional modelling. The resulting optimized geometry was evaluated on the basis of the ITER SDC-IC criteria and in terms of low cycle fatigue (LCF). The margin to the critical heat flux (CHF) was also estimated. Further design study (taking into account the effect of neutron radiation on the material properties) together with mock-up fabrication and high-heat-flux (HHF) tests are foreseen in next work programmes.

  17. Design study of ITER-like divertor target for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Crescenzi, Fabio, E-mail: fabio.crescenzi@enea.it [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Bachmann, C. [EFDA, Power Plant Physics and Technology, Boltzmannstraße 2, 85748 Garching (Germany); Richou, M. [CEA, IRFM, F-13108 Saint Paul Lez Durance (France); Roccella, S.; Visca, E. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); You, J.-H. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany)

    2015-10-15

    Highlights: • ‘DEMO’ is a near-term Power Plant Conceptual Study (PPCS). • The ITER-like design concept represents a promising solution also for DEMO plasma facing units. • The optimization of PFUs aims to enhance the thermo-mechanical behaviour of the component. • The optimized geometry was evaluated by ITER SDC-IC criteria and in terms of low cycle fatigue (LCF). - Abstract: A near-term water-cooled target solution has to be evaluated together with the required technologies and its power exhaust limit under ‘DEMO’ conditions. The ITER-like design concept based on the mono-block technology using W as armour material and the CuCrZr-IG as structural material with an interlayer of pure copper represents a promising solution also for DEMO. This work reports the design study of an “optimized” ITER-like Water Cooled Divertor able to withstand a heat flux of 10 MW m{sup −2}, as requested for DEMO operating conditions. The optimization of plasma facing unit (PFU) aims to enhance the thermo-mechanical behaviour of the component by varying some geometrical parameters (monoblock size, interlayer thickness and, tube diameter and thickness). The optimization was performed by means of the multi-variable optimization algorithms using the FEM code ANSYS. The coolant hydraulic conditions (inlet pressure, temperature and velocity) were fixed for simplicity. This study is based on elastic analysis and 3 dimensional modelling. The resulting optimized geometry was evaluated on the basis of the ITER SDC-IC criteria and in terms of low cycle fatigue (LCF). The margin to the critical heat flux (CHF) was also estimated. Further design study (taking into account the effect of neutron radiation on the material properties) together with mock-up fabrication and high-heat-flux (HHF) tests are foreseen in next work programmes.

  18. 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.)

  19. Dissociative recombination and electron-impact de-excitation in CH photon emission under ITER divertor-relevant plasma conditions

    NARCIS (Netherlands)

    van Swaaij, G. A.; Bystrov, K.; Borodin, D.; Kirschner, A.; van der Vegt, L. B.; van Rooij, G. J.; De Temmerman, G.; W. J. Goedheer,

    2012-01-01

    For understanding carbon erosion and redeposition in nuclear fusion devices, it is important to understand the transport and chemical break-up of hydrocarbon molecules in edge plasmas, often diagnosed by emission of the CH A(2)Delta-X-2 Pi Gero band around 430 nm. The CH A-level can be excited

  20. Assessment of radiation maps during activated divertor moving in the ITER building

    International Nuclear Information System (INIS)

    Ying Dongchuan; Zeng Qin; Qiu Yuefeng; Dang Tongqiang; Wu Yican; Loughlin, Michael

    2011-01-01

    As the main interface components between plasma and vacuum vessel, the divertor is foreseen to be removed to the hot cell for refurbishment during the 20 years of ITER operation. During this process, the activated divertor will cause a large increase of radiation in the ITER building. 3D analysis has been performed to assess the radiation maps throughout the ITER building for assisting the shielding design for personnel and sensitive equipment. The activation of the divertor has been determined by coupled neutron transport and inventory calculations, radiation maps have been obtained from gamma transport calculations. The neutron and gamma transport calculations have been performed by MCNP5 code with FENDL2.1library. The inventory calculations have been performed by FISPACT2007 code with EAF-2007 library. The results of these 3D decay gamma radiation maps are presented by pictures in this paper, including the biological dose maps and maps of heat deposition in electronic equipment.

  1. R and D activities on manufacturing plasma-facing unit for prototype of ITER divertor outer target in JADA

    International Nuclear Information System (INIS)

    Ezato, Koichiro; Suzuki, Satoshi; Seki, Yohji; Nishi, Hiroshi; Mohri, Kensuke; Enoeda, Mikio

    2012-01-01

    Japan Domestic Agency (JADA) carried out R and Ds activities to improve joining CFC monoblocks onto a CuCrZr cooling tube in PFUs to boost the success rate of joint and to confirm load carrying capability of the monoblock attachments to Steel Support Structure (SSS) against tensile force simulating electromagnetic load to pull PFUs from SSS. In joining the CFC monoblocks to the cooling tube, JADA has adopted brazing by using noble-metal-free filler with the following improvements; (1) metalizing joint surface of CFC using Ti-coating with accurate thickness controlling, (2) Changing buffer layer material from soft pure copper to Cu–W alloy. By using the present improved joint, JADA has manufactured three mock-ups with 5 CFC monoblocks and tested against repetitive high heat loads more than 20 MW/m 2 . All of CFC monoblocks of each mockup can survive the high heat loads throughout 1000 cycles with no degradation of heat removal capability. Regarding the load carrying capability of monoblock attachments to SSS, tensile experiments were carried out using the same geometries of CFC and tungsten monoblocks in PFUs and the results show that both geometries and joints meet the ITER requirements, that is, 3 kN and 8 kN, respectively.

  2. R and D activities on manufacturing plasma-facing unit for prototype of ITER divertor outer target in JADA

    Energy Technology Data Exchange (ETDEWEB)

    Ezato, Koichiro, E-mail: ezatok.koichiro@jaea.go.jp [Naka Fusion Institute, Japan Atomic Energy Agency, JAEA, Naka, Ibaraki (Japan); Suzuki, Satoshi; Seki, Yohji; Nishi, Hiroshi; Mohri, Kensuke; Enoeda, Mikio [Naka Fusion Institute, Japan Atomic Energy Agency, JAEA, Naka, Ibaraki (Japan)

    2012-08-15

    Japan Domestic Agency (JADA) carried out R and Ds activities to improve joining CFC monoblocks onto a CuCrZr cooling tube in PFUs to boost the success rate of joint and to confirm load carrying capability of the monoblock attachments to Steel Support Structure (SSS) against tensile force simulating electromagnetic load to pull PFUs from SSS. In joining the CFC monoblocks to the cooling tube, JADA has adopted brazing by using noble-metal-free filler with the following improvements; (1) metalizing joint surface of CFC using Ti-coating with accurate thickness controlling, (2) Changing buffer layer material from soft pure copper to Cu-W alloy. By using the present improved joint, JADA has manufactured three mock-ups with 5 CFC monoblocks and tested against repetitive high heat loads more than 20 MW/m{sup 2}. All of CFC monoblocks of each mockup can survive the high heat loads throughout 1000 cycles with no degradation of heat removal capability. Regarding the load carrying capability of monoblock attachments to SSS, tensile experiments were carried out using the same geometries of CFC and tungsten monoblocks in PFUs and the results show that both geometries and joints meet the ITER requirements, that is, 3 kN and 8 kN, respectively.

  3. 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

  4. 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.

  5. 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.

  6. Ion orbit modelling of ELM heat loads on ITER divertor vertical targets.

    Czech Academy of Sciences Publication Activity Database

    Gunn, J. P.; Carpentier-Chouchana, S.; Dejarnac, Renaud; Escourbiac, F.; Hirai, T.; Komm, Michael; Kukushkin, A.; Panayotis, S.; Pitts, R.A.

    2017-01-01

    Roč. 12, August (2017), s. 75-83 ISSN 2352-1791. [International Conference on Plasma Surface Interactions 2016, PSI2016 /22./. Roma, 30.05.2016-03.06.2016] Institutional support: RVO:61389021 Keywords : ITER * Divertor * ELM heat loads Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) http://www.sciencedirect.com/science/article/pii/S2352179116302745

  7. 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.

  8. 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.

  9. Design of a diagnostic residual gas analyzer for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Klepper, C.C., E-mail: kleppercc@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6169 (United States); Biewer, T.M.; Graves, V.B. [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6169 (United States); Andrew, P. [ITER Organisation, Route de Vinon-sur-Verdon, 13067 St. Paul-lez-Durance (France); Lukens, P.C. [US ITER Project Office, 1055 Commerce Park Dr #1, Oak Ridge, TN 37830 (United States); Marcus, C. [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6169 (United States); Shimada, M., E-mail: shimada.michiya@jaea.go.jp [ITER Organisation, Route de Vinon-sur-Verdon, 13067 St. Paul-lez-Durance (France); Hughes, S.; Boussier, B. [ITER Organisation, Route de Vinon-sur-Verdon, 13067 St. Paul-lez-Durance (France); Johnson, D.W. [US ITER Diagnostics Office, Princeton Plasma Physics Laboratory, Princeton, NJ 08540 (United States); Gardner, W.L. [US ITER Project Office, 1055 Commerce Park Dr #1, Oak Ridge, TN 37830 (United States); Hillis, D.L. [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6169 (United States); Vayakis, G.; Walsh, M. [ITER Organisation, Route de Vinon-sur-Verdon, 13067 St. Paul-lez-Durance (France)

    2015-10-15

    Highlights: • The divertor DRGA for ITER will measure neutral gas composition in the pumping ducts during plasma. • System must respond in timescales relevant to compositional changes in the divertor plasma. • It is shown that times can vary from 1 to 6 s for fuel (H2, D2, T2) up to 50 s for He (fusion reaction ash). • It is shown that present design delivers ∼ 1 s response even via an 8m long sampling pipe sampling. • Response time validated with VacTran{sup ®} over anticipated the 0.1–10 Pa pressure range in the ducts. - Abstract: One of the ITER diagnostics having reached an advanced design stage is a diagnostic RGA for the divertor, i.e. residual gas analysis system for the ITER divertor, which is intended to sample the divertor pumping duct region during the plasma pulse and to have a response time compatible with plasma particle and impurity lifetimes in the divertor region. Main emphasis is placed on helium (He) concentration in the ducts, as well as the relative concentration between the hydrogen isotopes (mainly in the form of diatomic molecules of H, D, and T). Measurement of the concentration of radiative gases, such as neon (Ne) and nitrogen (N{sub 2}), is also intended. Numerical modeling of the gas flow from the sampled region to the cluster of analysis sensors, through a long (∼8 m long, ∼110 mm diameter) sampling pipe originating from a pressure reducing orifice, confirm that the desired response time (∼1 s for He or D{sub 2}) is achieved with the present design.

  10. Erosion of ITER divertor armour and contamination of sol after transient events erosion products

    International Nuclear Information System (INIS)

    Bazylev, B.N.; Landman, I.S.; Pestchanyi, S.E.

    2005-01-01

    Plasma impact to the divertor expected in the tokamak ITER during ELMs or disruptions can result in a significant surface damage to CFC- and tungsten armours (brittle destruction and melting respectively) as well as in contamination of SOL by evaporated impurities. Numerical investigations for tungsten and CFC targets provide important details of the material erosion process. The simulations carried out in FZK on the material damage, carbon plasma expansion and the radiation fluxes from the carbon impurity are surveyed

  11. 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.

  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. Failure mode analysis of preliminary design of ITER divertor impurity monitor

    International Nuclear Information System (INIS)

    Kitazawa, Sin-iti; Ogawa, Hiroaki

    2016-01-01

    Highlights: • Divertor impurity influx monitor for ITER (DIM) is procured by JADA. • DIM is designed to observe light from nuclear fusion plasma directly. • DIM is under preliminary design phase. • Failure mode of DIM was prepared for RAMI analysis. • RAMI analysis on DIM was performed to reduce technical risks. - Abstract: The objective of the divertor impurity influx monitor (DIM) for ITER is to measure the parameters of impurities and hydrogen isotopes (tritium, deuterium, and hydrogen) in divertor plasma using visible and UV spectroscopic techniques in the 200–1000 nm wavelength range. In ITER, special provisions are required to ensure accuracy and full functionality of the diagnostic components under harsh conditions (high temperature, high magnetic field, high vacuum condition, and high radiation field). Japan Domestic Agency is preparing the preliminary design of the ITER DIM system, which will be installed in the upper, equatorial and lower ports. The optical and mechanical designs of the DIM are conducted to fit ITER’s requirements. The optical and mechanical designs meet the requirements of spatial resolution. Some auxiliary systems were examined via prototyping. The preliminary design of the ITER DIM system was evaluated by RAMI analysis. The availability of the designed system is adequately high to satisfy the project requirements. However, some equipment does not have certain designs, and this may cause potential technical risks. The preliminary design should be modified to reduce technical risks and to prepare the final design.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-09-15

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

  17. 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.))

  18. Development and optimisation of tungsten armour geometry for ITER divertor

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  19. Acceptance criteria for the ITER divertor vertical target

    International Nuclear Information System (INIS)

    Fouquet, S.; Schlosser, J.; Merola, M.; Durocher, A.; Escourbiac, F.; Grosman, A.; Missirlian, M.; Portafaix, C.

    2006-01-01

    In the frame of the toroidal pump limiter fabrication for Tore Supra, CEA developed a large experience of infrared test for acceptance of high heat flux components armoured with carbon fibre composite flat tiles. The test is based on a thermal transient induced by an alternative hot/cold water flow in the heat sink structure. The tile surface temperature transients are compared with those of a reference element, the maximum difference for each tile leading to a value called ΔT ref m ax . This method is proposed for the commissioning of plasma facing components for the ITER divertor vertical target. This paper describes the determination of the best acceptance criteria for the 'monoblock' geometry of the carbon part. First, it has been shown that the location and the extension of the defects could reliably be determined by monitoring both top and lateral surfaces during the test. Second, it was possible to fix an acceptance method based on ΔT ref m ax . Samples with calibrated defects are now under fabrication to validate the results

  20. 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.

  1. 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.)

  2. 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)

  3. 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.)

  4. Engineering challenges and development of the ITER Blanket System and Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Merola, Mario, E-mail: mario.merola@iter.org; Escourbiac, Frederic; Raffray, Alphonse Rene; Chappuis, Philippe; Hirai, Takeshi; Gicquel, Stefan

    2015-10-15

    The ITER Blanket System and the Divertor are the main components which directly face the plasma. Being the first physical barrier to the plasma, they have very demanding design requirements, which include accommodating: (1) surface heat flux and neutronic volumetric heating, (2) electromagnetic loads, (3) nuclear shielding function, (4) capability of being assembled and remote-handled, (5) interfaces with other in-vessel components, and (6) high heat flux technologies and complex welded structures in the design. The main functions of the Blanket System have been substantially expanded and it has now also to provide limiting surfaces that define the plasma boundary during startup and shutdown. As regards the Divertor, the ITER Council decided in November 2013 to start the ITER operation with a full-tungsten armour in order to minimize costs and already gain operational experience with tungsten during the non-active phase of the machine. This paper gives an overview of the design and technology qualification of the Blanket System and the Divertor.

  5. Remote operational trials with the ITER FDR divertor handling equipment

    International Nuclear Information System (INIS)

    Irving, M.; Baldi, L.; Benamati, G.; Galbiati, L.; Giacomelli, S.; Lorenzelli, L.; Micciche, G.; Muro, L.; Polverari, A.; Palmer, J.; Martin, E.

    2003-01-01

    The ITER divertor test platform (DTP) located at ENEA's Research Centre in Brasimone, Italy is a full-scale mock-up of a 72 deg. arc of the ITER 1998 vessel divertor region--the result of a major initiative over the period 1996-2000. Since the implementation of this facility, the design of the ITER vessel--and therefore much of the remote maintenance equipment--has changed substantially. However, the nature and principles of the remote handling equipment are still very similar, and hence many valuable lessons can yet be learned from the existing equipment for the future. In particular, true remote handling tests of the major maintenance subsystems were seen as an important step in determining their suitability for ITER. This paper describes and documents a series of three, discrete, remote-handling trials carried out using most of the major DTP subsystems, and presents an overview of the conclusions and suggestions for future development of ITER cassette remote handling equipment

  6. Design issues and cost implications of RTO/RC-ITER divertor

    International Nuclear Information System (INIS)

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

    2000-01-01

    This paper reports on the conceptual divertor design developed for the reduced technical objectives/reduced cost-international thermonuclear experimental reactor (RTO/RC-ITER). The cost drivers are discussed and a number of cost-reducing measures identified. Scaled costs, based on industrial estimates of the 1998 ITER design (Technical Basis for the ITER Final Design Report, Cost Review and Safety Analysis (FDR). ITER EDA Documentation Series No. 16. IAEA, Vienna, 1998), give for the RTO/RC-ITER ∼60% of the FDR costs. Plasma facing components (PFCs) account for 75% of the total divertor costs. Hence, PFC design simplifications are outlined in the paper showing the possibility of achieving a cost reduction of 50%. The design proposals, outlined in the paper, focus on minimising the number of sub-components and simplifying the manufacturing cycle. These changes contribute to improved reliability based on a more robust coolant design layout. The reduced space allocated to the divertor (G. Janeschitz, A. Antipenkov, V. Barabash, S. Chiocchio, G. Federici, C. Ibbott, E. Martin, R. Tivey, Overview of the Divertor Design and its Integration into RTO/RC-ITER, this conference) requires changes to the design that minimise the cassette body thickness, relocate the cassette attachments and revise the remote handling philosophy. Results of supporting electro-magnetic, neutron shielding, thermo-hydraulic and pumping conductance analyses are reported, qualifying the cassette design. A reduction in the coolant inlet temperature to 100-120 deg. C is discussed in terms of thermal-hydraulic performance and fatigue life of the heat sink. Finally, an R and D plan sets out the work needed: (1) to develop the cost saving measures of the new design; and (2) to demonstrate the reliability of the chosen technologies

  7. Neutron activation behavior of NET/ITER divertor structural materials

    International Nuclear Information System (INIS)

    Smid, I.; Weimann, G.; Kny, E.; Kneringer, G.; Reheis, N.

    1995-01-01

    The post-activation behavior of the materials carbon, TZM (99.3 % Mo) and Mo.41Re, as well as of high temperature brazes suitable for their joining after irradiation with 14 MeV neutrons has been evaluated. The activity, dose rate and energy generation after exposure to an ignited fusion plasma is presented for various time steps after shutdown. The impact of the activity and the afterheat production on the handling and storage conditions of retired divertor components is simulated, the required protection for maintenance is discussed. Further the temperature of stored divertor elements after a full time operation in NET was calculated. No major afterheat production will occur and thus no special cooling is to be provided after approximately one month. Taking into account convection and radiation the equilibrium temperature of vertically stored environment/aircooled divertor elements is predicted to be approximately 100 degree C. (author)

  8. Development of a full-size divertor cassette prototype for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Ulrickson, M.A. [Sandia National Labs., Albuquerque, NM (United States); Vieider, G.; Pacher, H.D. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany). NET Design Team] [and others

    1996-10-01

    Production of a full-size divertor cassette involves eight major components. All of the components are mounted on the cassette body. Inner divertor channel components for the vertical target design are being provided by the Japan Home Team. Outer divertor channel components for the vertical target design are being provided by the European and United States Home Teams. Gas box liners are being provided by the Russian Home Team. The full-size components manufactured by the four parties will be shipped to the US Home Team for assembly into a full size divertor cassette. The techniques for assembly and maintenance of the cassette will be demonstrated during this process. The assembled cassette will be tested for proper flow distribution and proof of the filling and draining procedures. The testing will include vacuum leak tightness at full temperature and pressure, cyclic heating to 150 {degrees}C, verification of dimensional accuracy of the assembled components, and application of thermal gradients to measure dimensional stability. The development of the divertor for the International Thermonuclear Experimental Reactor (ITER) depends on successful R&D efforts on materials, joining, and plasma materials interactions. Results of the development program are presented. The scale-up of the processes developed in the basic research and development tasks is accomplished by producing and high-heat-flux testing medium and full-scale mock- ups. The design of the mock-ups is discussed.

  9. Development of a full-size divertor cassette prototype for ITER

    International Nuclear Information System (INIS)

    Ulrickson, M.A.; Vieider, G.; Pacher, H.D.

    1996-01-01

    Production of a full-size divertor cassette involves eight major components. All of the components are mounted on the cassette body. Inner divertor channel components for the vertical target design are being provided by the Japan Home Team. Outer divertor channel components for the vertical target design are being provided by the European and United States Home Teams. Gas box liners are being provided by the Russian Home Team. The full-size components manufactured by the four parties will be shipped to the US Home Team for assembly into a full size divertor cassette. The techniques for assembly and maintenance of the cassette will be demonstrated during this process. The assembled cassette will be tested for proper flow distribution and proof of the filling and draining procedures. The testing will include vacuum leak tightness at full temperature and pressure, cyclic heating to 150 degrees C, verification of dimensional accuracy of the assembled components, and application of thermal gradients to measure dimensional stability. The development of the divertor for the International Thermonuclear Experimental Reactor (ITER) depends on successful R ampersand D efforts on materials, joining, and plasma materials interactions. Results of the development program are presented. The scale-up of the processes developed in the basic research and development tasks is accomplished by producing and high-heat-flux testing medium and full-scale mock- ups. The design of the mock-ups is discussed

  10. Engineering and design aspects related to the development of the ITER divertor

    International Nuclear Information System (INIS)

    Dietz, J.; Chiocchio, S.; Antipenkov, A.

    1994-01-01

    Most of the divertor concepts proposed for the Next Step devices relied on the exhaust of the SOL power to target plates which intersect the magnetic field fines. The resulting highly peaked thermal load, together with the concentrated fluxes of energetic particles, posed severe design constraints and ultimately led to unacceptably short target lifetime. The ITER high density gas target divertor concept is based on transferring the nominal power perpendicular to the magnetic field lines from the plasma edge onto large surfaces and on dissipating the particles' energy through atomic and molecular mechanisms. While the basic ideas for this approach have been motivated by recent results in present tokamaks, a full assessment of this concept still requires extensive experimental and modelling work. The paper describes the engineering and design aspects involving the development of the ITER divertor and shows how the physics assumptions translate into engineering requirements, and how the additional existing constraints (such as the limited space, neutron load, electromagnetic effects, compatibility with other components, remote maintainability) have been taken into account for the design definition. The concept developed takes advantage of the spatial separation of the several physics phenomena anticipated to take place in the divertor, thus relaxing the needs to accommodate in the same region opposing requirements

  11. 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.

  12. 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.

  13. 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)

  14. Plasma-safety assessment model and safety analyses of ITER

    International Nuclear Information System (INIS)

    Honda, T.; Okazaki, T.; Bartels, H.-H.; Uckan, N.A.; Sugihara, M.; Seki, Y.

    2001-01-01

    A plasma-safety assessment model has been provided on the basis of the plasma physics database of the International Thermonuclear Experimental Reactor (ITER) to analyze events including plasma behavior. The model was implemented in a safety analysis code (SAFALY), which consists of a 0-D dynamic plasma model and a 1-D thermal behavior model of the in-vessel components. Unusual plasma events of ITER, e.g., overfueling, were calculated using the code and plasma burning is found to be self-bounded by operation limits or passively shut down due to impurity ingress from overheated divertor targets. Sudden transition of divertor plasma might lead to failure of the divertor target because of a sharp increase of the heat flux. However, the effects of the aggravating failure can be safely handled by the confinement boundaries. (author)

  15. 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)

  16. 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)

  17. 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)

  18. 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

  19. 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.

  20. Experimental test campaign on an ITER divertor mock-up

    International Nuclear Information System (INIS)

    Dell'Orco, G.; Malavasi, A.; Merola, M.; Polazzi, G.; Simoncini, M.; Zito, D.

    2002-01-01

    In 1998, in the frame of the European R and D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body (CB), designed with some mechanical and hydraulic simplifications with respect to the reference body and its actively cooled Dummy Armour Prototype (DAP). This DAP consists of a Vertical Target (VT), a Wing (WI) and a Dump Target (DT), manufactured by European industries, which are integrated to the Gas Box Liner (GBL) supplied by the Russian Federation ITER Home Team. In 1999, in parallel with the manufacturing activity, the ITER European Home Team decided to assign to ENEA a Task for checking the component integration and performing the thermal-hydraulic and thermal mechanical testing of the DAP and CB. In 1999-2000, ENEA performed the experimental campaign at Brasimone Labs. The present work presents the experimental results of the component integration and the thermal-hydraulic and thermo-mechanical fatigue tests

  1. Experimental test campaign on an ITER divertor mock-up

    Energy Technology Data Exchange (ETDEWEB)

    Dell' Orco, G. E-mail: giovanni.dellorco@brasimone.enea.it; Malavasi, A.; Merola, M.; Polazzi, G.; Simoncini, M.; Zito, D

    2002-11-01

    In 1998, in the frame of the European R and D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body (CB), designed with some mechanical and hydraulic simplifications with respect to the reference body and its actively cooled Dummy Armour Prototype (DAP). This DAP consists of a Vertical Target (VT), a Wing (WI) and a Dump Target (DT), manufactured by European industries, which are integrated to the Gas Box Liner (GBL) supplied by the Russian Federation ITER Home Team. In 1999, in parallel with the manufacturing activity, the ITER European Home Team decided to assign to ENEA a Task for checking the component integration and performing the thermal-hydraulic and thermal mechanical testing of the DAP and CB. In 1999-2000, ENEA performed the experimental campaign at Brasimone Labs. The present work presents the experimental results of the component integration and the thermal-hydraulic and thermo-mechanical fatigue tests.

  2. High heat flux tests of mock-ups for ITER divertor application

    International Nuclear Information System (INIS)

    Giniatulin, R.; Gervash, A.; Komarov, V.L.; Makhankov, A.; Mazul, I.; Litunovsky, N.; Yablokov, N.

    1998-01-01

    One of the most difficult tasks in fusion reactor development is the designing, fabrication and high heat flux testing of actively cooled plasma facing components (PFCs). At present, for the ITER divertor project it is necessary to design and test components by using mock-ups which reflect the real design and fabrication technology. The cause of failure of the PFCs is likely to be through thermo-cycling of the surface with heat loads in the range 1-15 MW m -2 . Beryllium, tungsten and graphite are considered as the most suitable armour materials for the ITER divertor application. This work presents the results of the tests carried out with divertor mock-ups clad with beryllium and tungsten armour materials. The tests were carried out in an electron beam facility. The results of high heat flux screening tests and thermo-cycling tests in the heat load range 1-9 MW m -2 are presented along with the results of metallographic analysis carried out after the tests. (orig.)

  3. Hydrogen embrittlement considerations in niobium-base alloys for application in the ITER divertor

    International Nuclear Information System (INIS)

    Peterson, D.T.; Hull, A.B.; Loomis, B.A.

    1991-01-01

    The ITER divertor will be subjected to hydrogen from aqueous corrosion by the coolant and by transfer from the plasma. Global hydrogen concentrations are one factor in assessing hydrogen embrittlement but local concentrations affected by source fluxes and thermotransport in thermal gradients are more important considerations. Global hydrogen concentrations is some corrosion- tested alloys will be presented and interpreted. The degradation of mechanical properties of Nb-base alloys due to hydrogen is a complex function of temperature, hydrogen concentration, stresses and alloy composition. The known tendencies for embrittlement and hydride formation in Nb alloys are reviewed

  4. 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.)

  5. 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

  6. Critical heat flux acoustic detection: Methods and application to ITER divertor vertical target monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Courtois, X., E-mail: xavier.courtois@cea.fr [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Escourbiac, F. [ITER Organization, Route de Vinon sur Verdon, F-13115 Saint-Paul-Lez-Durance (France); Richou, M.; Cantone, V. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Constans, S. [AREVA-NP, Le Creusot (France)

    2013-10-15

    Actively cooled plasma facing components (PFCs) have to exhaust high heat fluxes from plasma radiation and plasma–wall interaction. Critical heat flux (CHF) event may occur in the cooling channel due to unexpected heat loading or operational conditions, and has to be detected as soon as possible. Therefore it is essential to develop means of monitoring based on precursory signals providing an early detection of this destructive phenomenon, in order to be able to stop operation before irremediable damages appear. Capabilities of CHF early detection based on acoustic techniques on PFC mock-ups cooled by pressurised water were already demonstrated. This paper addresses the problem of the detection in case of flow rate reduction and of flow dilution resulting from multiple plasma facing units (PFU) which are hydraulically connected in parallel, which is the case of ITER divertor. An experimental study is launched on a dedicated mock-up submitted to heat loads up to the CHF. It shows that the measurement of the acoustic waves, generated by the cooling phenomena, allows the CHF detection in conditions similar to that of the ITER divertor, with a reasonable number of sensors. The paper describes the mock-ups and the tests sequences, and comments the results.

  7. Estimation of the contribution of gaps to tritium retention in the divertor of ITER

    International Nuclear Information System (INIS)

    Matveev, D; Kirschner, A; Litnovsky, A; Borodin, D; Samm, U; Schmid, K; Komm, M; Van Oost, G

    2014-01-01

    An estimation of the contribution of gaps to beryllium deposition and resulting tritium retention in the divertor of ITER is presented. Deposition of beryllium layers in gaps of the full tungsten divertor is simulated with the 3D-GAPS code. For gaps aligned along the poloidal direction, non-shaped and shaped solutions are compared. Plasma and impurity ion fluxes from Schmid (2008 Nucl. Fusion 48 105004) are used as input. Ion penetration into gaps is considered to be geometrical along magnetic field lines. The effect of realistic ion penetration into gaps is discussed. In total, gaps in the divertor are estimated to contribute about 0.3 mgT s −1 to the overall tritium retention dominated by toroidal gaps, which are not shaped. This amount corresponds to about 7800 ITER discharges up to the safety limit of 1 kg in-vessel tritium; excluding, however, tritium release during wall baking and retention at plasma-wetted and remote areas. (paper)

  8. A high-recycle divertor for ITER [International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Werley, K.A.; Bathke, C.G.

    1988-01-01

    A coupled one-dimensional (axial/radial) edge-plasma model (SOLAR) has been used to investigate tradeoffs between collector-plate and edge-plasma conditions in a doublenull, open, high-recycle divertor (HRD) for a preliminary International Thermonuclear Experimental Reactor (ITER) design. A steady-state HRD produces in attractive high-density edge plasma (5 /times/ 10 19 m/sup /minus/3/) with sufficiently low plasma temperature (10-20eV) at a tungsten plat that the sheath-accelerated ions are below sputtering threshold energies. Manageable plate heat fluxes (3-6 MW/m 2 ) are achieved by positioning the plate poloidal cross section at a minimum angle of 15-30/degree/ with respect to flux surfaces. 6 refs., 9 figs

  9. 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)

  10. Manufacturing and testing of reference samples for the definition of acceptance criteria for the ITER divertor

    International Nuclear Information System (INIS)

    Visca, Eliseo; Cacciotti, E.; Libera, S.; Mancini, A.; Pizzuto, A.; Roccella, S.; Riccardi, B.; Escourbiac, F.; Sanguinetti, G.P.

    2010-01-01

    The most critical part of a high heat flux (HHF) plasma facing component (PFC) is the armour to heat sink joint. An experimental study was launched by EFDA in order to define the acceptance criteria to be used for the procurements of the ITER Divertor PFCs. ENEA is involved in the European International Thermonuclear Experimental Reactor (ITER) R and D activities and together with Ansaldo Ricerche S.p.A. has manufactured several PFCs mock-ups using the Hot Radial Pressing and Pre-Brazed Casting technologies. According to the technical specifications issued by EFDA, ENEA and Ansaldo have collaborated to manufacture half of the samples with calibrated artificial defects required for this experimental study. After manufacturing, the samples were examined by ultrasonic and SATIR non-destructive examination (NDE) methods in order to confirm the size and position of the artificial defects. In particular, it was concluded that defects are detectable with these NDE techniques and they finally gave indication about the threshold of propagation during high heat flux experiments relevant with heat fluxes expected in ITER Divertor. This paper reports the manufacturing procedure used to obtain the required calibrated artificial defects in the CFC and W armoured samples as well as the NDE results and the thermal high heat flux results.

  11. Manufacturing and testing of reference samples for the definition of acceptance criteria for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Visca, Eliseo, E-mail: visca@frascati.enea.i [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Cacciotti, E.; Libera, S.; Mancini, A.; Pizzuto, A.; Roccella, S. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Riccardi, B., E-mail: Bruno.Riccardi@f4e.europa.e [Fusion For Energy, Barcelona (Spain); Escourbiac, F., E-mail: frederic.escourbiac@iter.or [ITER Organization, Cadarache (France); Sanguinetti, G.P., E-mail: gianpaolo.sanguinetti@aen.ansaldo.i [Ansaldo Energia S.p.A., Genova (Italy)

    2010-12-15

    The most critical part of a high heat flux (HHF) plasma facing component (PFC) is the armour to heat sink joint. An experimental study was launched by EFDA in order to define the acceptance criteria to be used for the procurements of the ITER Divertor PFCs. ENEA is involved in the European International Thermonuclear Experimental Reactor (ITER) R and D activities and together with Ansaldo Ricerche S.p.A. has manufactured several PFCs mock-ups using the Hot Radial Pressing and Pre-Brazed Casting technologies. According to the technical specifications issued by EFDA, ENEA and Ansaldo have collaborated to manufacture half of the samples with calibrated artificial defects required for this experimental study. After manufacturing, the samples were examined by ultrasonic and SATIR non-destructive examination (NDE) methods in order to confirm the size and position of the artificial defects. In particular, it was concluded that defects are detectable with these NDE techniques and they finally gave indication about the threshold of propagation during high heat flux experiments relevant with heat fluxes expected in ITER Divertor. This paper reports the manufacturing procedure used to obtain the required calibrated artificial defects in the CFC and W armoured samples as well as the NDE results and the thermal high heat flux results.

  12. Novel aspects of plasma control in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Humphreys, D.; Jackson, G.; Walker, M.; Welander, A. [General Atomics P.O. Box 85608, San Diego, California 92186-5608 (United States); Ambrosino, G.; Pironti, A. [CREATE/University of Naples Federico II, Napoli (Italy); Vries, P. de; Kim, S. H.; Snipes, J.; Winter, A.; Zabeo, L. [ITER Organization, St. Paul Lez durance Cedex (France); Felici, F. [Eindhoven University of Technology, Eindhoven (Netherlands); Kallenbach, A.; Raupp, G.; Treutterer, W. [Max-Planck Institut für Plasmaphysik, Garching (Germany); Kolemen, E. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Lister, J.; Sauter, O. [Centre de Recherches en Physique des Plasmas, Ecole Polytechnique Federale de Lausanne, Lausanne (Switzerland); Moreau, D. [CEA, IRFM, 13108 St. Paul-lez Durance (France); Schuster, E. [Lehigh University, Bethlehem, Pennsylvania (United States)

    2015-02-15

    ITER plasma control design solutions and performance requirements are strongly driven by its nuclear mission, aggressive commissioning constraints, and limited number of operational discharges. In addition, high plasma energy content, heat fluxes, neutron fluxes, and very long pulse operation place novel demands on control performance in many areas ranging from plasma boundary and divertor regulation to plasma kinetics and stability control. Both commissioning and experimental operations schedules provide limited time for tuning of control algorithms relative to operating devices. Although many aspects of the control solutions required by ITER have been well-demonstrated in present devices and even designed satisfactorily for ITER application, many elements unique to ITER including various crucial integration issues are presently under development. We describe selected novel aspects of plasma control in ITER, identifying unique parts of the control problem and highlighting some key areas of research remaining. Novel control areas described include control physics understanding (e.g., current profile regulation, tearing mode (TM) suppression), control mathematics (e.g., algorithmic and simulation approaches to high confidence robust performance), and integration solutions (e.g., methods for management of highly subscribed control resources). We identify unique aspects of the ITER TM suppression scheme, which will pulse gyrotrons to drive current within a magnetic island, and turn the drive off following suppression in order to minimize use of auxiliary power and maximize fusion gain. The potential role of active current profile control and approaches to design in ITER are discussed. Issues and approaches to fault handling algorithms are described, along with novel aspects of actuator sharing in ITER.

  13. 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

  14. New achievements of the Divertor Test Platform programme for the ITER divertor remote maintenance R and D

    International Nuclear Information System (INIS)

    Damiani, C.; Baldi, L.; Galbiati, L.; Irving, M.; Lorenzelli, L.; Micciche, G.; Muro, L.; Nucci, S.; Varocchi, G.; Poggianti, A.; Fermani, G.; Maisonnier, D.; Palmer, J.; Martin, E.; Friconneau, J.P.; Gravez, P.; Takeda, N.

    2001-01-01

    The divertor assembly for the ITER fusion reactor consists of a number of rail mounted cassettes (54 now in ITER FEAT) located in the bottom region of the vacuum vessel. These cassettes shall be removed/installed remotely during the life of the reactor by means of specific devices. To demonstrate and optimise the feasibility of the in-vessel maintenance process the Divertor Test Platform (DTP) has been established at the ENEA Research Centre in Brasimone, Italy, as a major part of the large ITER R and D project L7. A first set of tests has been already carried out and reported during 1998, when the basic feasibility of the divertor replacement was demonstrated. In the present period (January 1999-July 2000), new activities, including both site tests and other 'external' R and D works, have been carried out in order to refine and improve the ITER divertor maintenance scenario. These include the study of abnormal maintenance operations and of possible handling equipment failure and its consequences; the procurement and testing of new sub-systems (e.g. a force reflection manipulator arm), and the development of remote handling techniques including a virtual reality system. Following a short description of the DTP, this paper reports on the new results and achievements, draws the relevant conclusions, and finally discusses future activities

  15. Materials issues in the design of the ITER first wall, blanket, and divertor

    International Nuclear Information System (INIS)

    Mattas, R.F.; Smith, D.L.; Wu, C.H.; Shatalov, G.

    1992-01-01

    During the ITER conceptual design study, a property data base was assembled, the key issues were identified, and a comprehensive R ampersand D plan was formulated to resolve these issues. The desired properties of candidate ITER divertor, first wall, and blanket materials are briefly reviewed, and the major materials issues are presented. Estimates of the influence of materials properties on the performance limits of the first wall, blanket, and divertor are presented

  16. Integrated simulations of H-mode operation in ITER including core fuelling, divertor detachment and ELM control

    Science.gov (United States)

    Polevoi, A. R.; Loarte, A.; Dux, R.; Eich, T.; Fable, E.; Coster, D.; Maruyama, S.; Medvedev, S. Yu.; Köchl, F.; Zhogolev, V. E.

    2018-05-01

    ELM mitigation to avoid melting of the tungsten (W) divertor is one of the main factors affecting plasma fuelling and detachment control at full current for high Q operation in ITER. Here we derive the ITER operational space, where ELM mitigation to avoid melting of the W divertor monoblocks top surface is not required and appropriate control of W sources and radiation in the main plasma can be ensured through ELM control by pellet pacing. We apply the experimental scaling that relates the maximum ELM energy density deposited at the divertor with the pedestal parameters and this eliminates the uncertainty related with the ELM wetted area for energy deposition at the divertor and enables the definition of the ITER operating space through global plasma parameters. Our evaluation is thus based on this empirical scaling for ELM power loads together with the scaling for the pedestal pressure limit based on predictions from stability codes. In particular, our analysis has revealed that for the pedestal pressure predicted by the EPED1  +  SOLPS scaling, ELM mitigation to avoid melting of the W divertor monoblocks top surface may not be required for 2.65 T H-modes with normalized pedestal densities (to the Greenwald limit) larger than 0.5 to a level of current of 6.5–7.5 MA, which depends on assumptions on the divertor power flux during ELMs and between ELMs that expand the range of experimental uncertainties. The pellet and gas fuelling requirements compatible with control of plasma detachment, core plasma tungsten accumulation and H-mode operation (including post-ELM W transient radiation) have been assessed by 1.5D transport simulations for a range of assumptions regarding W re-deposition at the divertor including the most conservative assumption of zero prompt re-deposition. With such conservative assumptions, the post-ELM W transient radiation imposes a very stringent limit on ELM energy losses and the associated minimum required ELM frequency. Depending on

  17. High heat flux testing of EU tungsten monoblock mock-ups for the ITER divertor

    International Nuclear Information System (INIS)

    Gavila, P.; Riccardi, B.; Pintsuk, G.; Ritz, G.; Kuznetsov, V.; Durocher, A.

    2015-01-01

    Highlights: • All the tested items sustained the ITER Full W divertor qualification program requirements. This confirms that the technology for the manufacturing of the first set of the ITER Divertor is available in Europe. • The surface roughening and local melting of the W surface under high heat flux was proven to be significantly reduced for an armour thickness lower or equal to 6 mm. • However, this campaign highlighted some specific areas of improvement to be implemented ideally before the upcoming ITER Divertor IVT serial production. • The issue of the self-castellation of the W monoblocks, which typically appears after a few tenths of cycles at 20 MW/m"2, is critical because it generates some uncontrolled defects at the amour to heat sink joints. Besides, they create a gap which exposure is almost perpendicular to the magnetic field lines and which might lead to local W melting in the strike point region. • This campaign also evidenced that the minimum IO requirements on the CuCrZr ductility could be revised to avoid the occurrence of rather early fatigue failures. Although the W material characterization program has been set up by the IO, the strategy on the CuCrZr still needs to be defined. - Abstract: With the aim to assess the option to start the ITER operation with a full tungsten divertor, an R&D program was launched in order to evaluate the performances of tungsten (W) armoured plasma facing components (PFCs) under high heat flux. The F4E program consisted in the manufacturing and high heat flux (HHF) testing of W monoblock mock-ups and medium scale prototypes up to 20 MW/m"2. During the test campaign, 26 W mock-ups and two medium scale prototypes manufactured by Plansee SE (Austria) and by Ansaldo Nucleare (Italy) have been tested at the FE200 (AREVA, Le Creusot, France) and ITER Divertor Test Facility (IDTF) (Efremov Institute Saint Petersburg, Russian Federation) electron beam test facilities. The high heat flux (HHF) testing program

  18. High heat flux testing of EU tungsten monoblock mock-ups for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Gavila, P., E-mail: pierre.gavila@f4e.europa.eu [Fusion for Energy, 08019 Barcelona (Spain); Riccardi, B. [Fusion for Energy, 08019 Barcelona (Spain); Pintsuk, G. [Forschungszentrum Juelich, 52425 Juelich (Germany); Ritz, G. [AREVA NP, Centre Technique France, 71205 Le Creusot (France); Kuznetsov, V. [JCS “Efremov Institute”, Doroga na Metallostroy 3, Metallostroy, Saint-Petersburg 196641 (Russian Federation); Durocher, A. [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 Saint Paul-lez-Durance (France)

    2015-10-15

    Highlights: • All the tested items sustained the ITER Full W divertor qualification program requirements. This confirms that the technology for the manufacturing of the first set of the ITER Divertor is available in Europe. • The surface roughening and local melting of the W surface under high heat flux was proven to be significantly reduced for an armour thickness lower or equal to 6 mm. • However, this campaign highlighted some specific areas of improvement to be implemented ideally before the upcoming ITER Divertor IVT serial production. • The issue of the self-castellation of the W monoblocks, which typically appears after a few tenths of cycles at 20 MW/m{sup 2}, is critical because it generates some uncontrolled defects at the amour to heat sink joints. Besides, they create a gap which exposure is almost perpendicular to the magnetic field lines and which might lead to local W melting in the strike point region. • This campaign also evidenced that the minimum IO requirements on the CuCrZr ductility could be revised to avoid the occurrence of rather early fatigue failures. Although the W material characterization program has been set up by the IO, the strategy on the CuCrZr still needs to be defined. - Abstract: With the aim to assess the option to start the ITER operation with a full tungsten divertor, an R&D program was launched in order to evaluate the performances of tungsten (W) armoured plasma facing components (PFCs) under high heat flux. The F4E program consisted in the manufacturing and high heat flux (HHF) testing of W monoblock mock-ups and medium scale prototypes up to 20 MW/m{sup 2}. During the test campaign, 26 W mock-ups and two medium scale prototypes manufactured by Plansee SE (Austria) and by Ansaldo Nucleare (Italy) have been tested at the FE200 (AREVA, Le Creusot, France) and ITER Divertor Test Facility (IDTF) (Efremov Institute Saint Petersburg, Russian Federation) electron beam test facilities. The high heat flux (HHF) testing

  19. International Thermonuclear Experimental Reactor (ITER) divertor plate performance and lifetime considerations

    International Nuclear Information System (INIS)

    Mattas, R.F.

    1990-03-01

    The ITER divertor plate performance during the technology phase of operation has been analyzed. High-Z materials, such as tungsten and tantalum, have been considered as plasma side materials, and refractory metal alloys, Ta-10W, TZM, Nb-1Zr, and V-15Cr-5Ti, plus copper alloys have been considered as the structural materials. The fatigue lifetime have been predicted for structural plates and for duplex plates with the plasma side material bonded to the structure. The results indicate that refractory alloys have a comparable or improved performance to copper alloys. Peak allowable heat fluxes for these analyses are in the range of 15--20 MW/m 2 for 2 mm thick structural plates and 7--11 MW/m 2 for 4 mm thick duplex plates. 4 refs., 55 figs., 6 tabs

  20. Using the Tritium Plasma Experiment to evaluate ITER PFC safety

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Anderl, R.A.; Bartlit, J.R.; Causey, R.A.; Haines, J.R.

    1993-01-01

    The Tritium Plasma Experiment was assembled at Sandia National Labs., Livermore and is being moved to the Tritium Systems Test Assembly facility at Los Alamos National Lab. to investigate interactions between dense plasmas at low energies and plasma-facing component materials. This apparatus has the unique capability of replicating plasma conditions in a tokamak divertor with particle flux densities of 2 x 10 23 ions/m 2 .s and a plasma temperature of about 15 eV using a plasma that includes tritium. An experimental program has been initiated using the Tritium Plasma Experiment to examine safety issues related to tritium in plasma-facing components, particularly the ITER divertor. Those issues include tritium retention and release characteristics, tritium permeation rates and transient times to coolant streams, surface modification and erosion by the plasma, the effects of thermal loads and cycling, and particulate production. An industrial consortium led by McDonnell Douglas will design and fabricate the test fixtures

  1. Streaming through the gaps around divertor pipings in ITER

    International Nuclear Information System (INIS)

    Sato, Satoshi; Seki, Yasushi; Takatsu, Hideyuki; Mori, Seiji; Zimin, S.; Maki, Koichi; Kuroda, Toshimasa.

    1993-03-01

    Neutron and gamma ray streaming through the annular gap around divertor piping in International Thermonuclear Experimental Reactor (ITER) was investigated. A stepwise gap is proposed near the midpoint of the annular gap in order to reduce the dose rate at the upper port. The optimal step position and width to satisfy the design limit of dose rates were examined. From these studies, the following results were obtained. (1) In case of the straight annular 1 cm wide gap around cooling pipes through the 3 m thick shield, dose rate at the upper port in a day after shutdown is about 4 orders larger than the reference value of 25 μSv/h (2.5 mrem/h) for the biological shielding design. But by providing a step structure with the offset ratio of 2.2 times of the gap width at the midpoint of the shield, the dose rate can be evaluated as low as 1/20 of the biological shielding value 2.5 μSv/h (0.25 mrem/h) including a safety factor of 10 for the reference value. It satisfies the requirement of the shielding design. (2) The optimal step position to minimize the dose rate at the upper port is the midpoint of the shield. (3) The dose rates are not further more reduced even if the offset width is set more than twice of the gap width, and the offset width of twice the gap width is recommended. (author)

  2. Manufacturing and testing of ITER divertor gas box liners

    International Nuclear Information System (INIS)

    Mazul, I.; Giniatulin, R.; Komarov, V.L.; Krylov, V.; Kuzmin, Ye.; Makhankov, A.; Odintsov, V.; Zhuk, A.

    1998-01-01

    Among a variety of R and D works performed by different ITER parties there are seven large projects which deal with the development, manufacturing and testing of most important complex reactor components. One of the projects is directed to produce a prototype of divertor cassette. In according with integration plan two full size liners with dummy armour are manufactured by RF Home Team. Except for liners with dummy armors the large - scale mock-up with real armour have to be manufactured in order to demonstrate the semi-industrial possibilities for joining of Be and W to CuCrZr heat - sink structure. The design of this liners, technological approaches to their manufacturing are presented. The description of brazing facility and joining technology which use a fast ohmic heating by 15 kA current is made. A mock-up of 800 mm in length and 90 mm in width was armored by 18 Be tiles (44 x 44 mm 2 in plane, 10 mm - thick) and 16 W-Cu tiles (44 x 44 mm 2 in plane, 3 mm - thick W). The preliminary results of high heat flux testing of the armored mock-ups are also presented. (author)

  3. The ITER divertor cassette. Steady state characterisation and draining and drying transient hydraulic analyses

    International Nuclear Information System (INIS)

    Pietro Alessandro Di Maio; Valerio Tomarchio; Giuseppe Vella; Irene Zammuto; Giovanni Dell'Orco

    2005-01-01

    Full text of publication follows: The divertor is one of the most challenging components of the next step ITER nuclear fusion reactor. It is aimed at controlling the characteristics of boundary plasma, reducing the impurities in the plasma and sustaining the heat and particle fluxes arising from it, during normal and transient operations as well as during disruption events. The ITER divertor consists of 54 cassettes, each one mainly composed of three Plasma-Facing Components (PFCs), namely the inner vertical target, the outer vertical target and the dome-liner, actively cooled by subcooled pressurized water. Each PFC consists in a number of plasma facing units, cooled in parallel and assembled onto a supporting structure. The water maximum total flow rate, for the whole divertor, should be 1000 kg/s, with 100-150 deg. C inlet/outlet temperatures, 4.2 MPa inlet pressure and a maximum pressure drop of 1.4 MPa. The PFCs are cooled in series, with a maximum water velocity in the channel of 11 m/s, whilst the water coolant is routed via the cassette body. Due to the extremely high heat loads expected onto the PFCs (up to 20 MW/m 2 over 20 s), the hydraulic design of the divertor is particularly demanding. It shall ensure that the foreseen flow rate actually reaches each plasma-facing unit to ensure an adequate cooling and to prevent any risk of Critical Heat Flux (CHF). Sufficient margin ( > 40 %) to avoid the reaching of a CHR limit on the PFCs could be obtained by using hypervapotron design inside the flat channels and swirl flow turbulence tape promoters inside the vertical target cooling tubes. Furthermore the overall pressure drop and flow rate shall be within the specified design limit to avoid an unduly high pumping power. Another important issue is the definition of a proper procedure to drain the coolant and dry the divertor components prior to the maintenance operations as well as to refill them with water after maintenance, ensuring a complete elimination of

  4. Power flux in the ITER divertor tile gaps during ELMs

    Czech Academy of Sciences Publication Activity Database

    Dejarnac, Renaud; Komm, Michael; Gunn, J. P.; Pánek, Radomír

    390-391, - (2009), s. 818-821 ISSN 0022-3115. [International Conference on Plasma-Surface Interactions in Controlled Fusion Devices/18th./. Toledo, 26.05.2008-30.05.2008] R&D Projects: GA AV ČR KJB100430602 Institutional research plan: CEZ:AV0Z20430508 Keywords : Edge modeling * Ion-surface interactions * ITER * Sheaths Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.933, year: 2009

  5. 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.

  6. Estimation of carbon fibre composites as ITER divertor armour

    Science.gov (United States)

    Pestchanyi, S.; Safronov, V.; Landman, I.

    2004-08-01

    Exposure of the carbon fibre composites (CFC) NB31 and NS31 by multiple plasma pulses has been performed at the plasma guns MK-200UG and QSPA. Numerical simulation for the same CFCs under ITER type I ELM typical heat load has been carried out using the code PEGASUS-3D. Comparative analysis of the numerical and experimental results allowed understanding the erosion mechanism of CFC based on the simulation results. A modification of CFC structure has been proposed in order to decrease the armour erosion rate.

  7. Estimation of carbon fibre composites as ITER divertor armour

    International Nuclear Information System (INIS)

    Pestchanyi, S.; Safronov, V.; Landman, I.

    2004-01-01

    Exposure of the carbon fibre composites (CFC) NB31 and NS31 by multiple plasma pulses has been performed at the plasma guns MK-200UG and QSPA. Numerical simulation for the same CFCs under ITER type I ELM typical heat load has been carried out using the code PEGASUS-3D. Comparative analysis of the numerical and experimental results allowed understanding the erosion mechanism of CFC based on the simulation results. A modification of CFC structure has been proposed in order to decrease the armour erosion rate

  8. A program to evaluate the erosion on the CFC tiles of the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    D' Agata, E. [ITER International Team, ITER Joint Work Site, Boltzmannstr 2, 85748 Garching (Germany)], E-mail: elio.dagata@iter.org; Ogorodnikova, O.V. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint-Paul-Lez-Durance (France); Tivey, R. [ITER International Team, ITER Joint Work Site, Boltzmannstr 2, 85748 Garching (Germany); Lowry, C.; Schlosser, J. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint-Paul-Lez-Durance (France)

    2007-10-15

    The plasma-facing surfaces of the ITER divertor are armoured with tungsten in the upper part of the inner and outer vertical targets, and carbon fibre composite (CFC) in the lower part, the region where the scrape-off layer intercepts the divertor. The CFC in the form of a monoblock in the vertical target is the most loaded part of the plasma-facing surfaces, and hence it is subjected to high erosion and has a significant risk of failure. A program has been developed with the aim of understanding the impact on the erosion lifetime due to a combination of two main effects: the material property variations (particularly pronounced in CFC) and the presence of joining defects. The software allows the evolution of the surface profile of the armour to be predicted and the margin on critical heat flux at the heat-sink-to-coolant interface to be estimated for a range of postulated defects, from start-of-life through to end-of-life of the component. In assessing erosion, the code takes account of geometry and sublimation, and physical and chemical erosion of the CFC armour. The incident angle (a glancing angle of a few degrees) of the particle and heat flux onto the target is taken into account. The program has been validated by comparison with analytical approximations very well validated against experimental data. The code has been developed in the APDL language to operate inside a commercial and certificated finite element program such as ANSYS.

  9. A program to evaluate the erosion on the CFC tiles of the ITER divertor

    International Nuclear Information System (INIS)

    D'Agata, E.; Ogorodnikova, O.V.; Tivey, R.; Lowry, C.; Schlosser, J.

    2007-01-01

    The plasma-facing surfaces of the ITER divertor are armoured with tungsten in the upper part of the inner and outer vertical targets, and carbon fibre composite (CFC) in the lower part, the region where the scrape-off layer intercepts the divertor. The CFC in the form of a monoblock in the vertical target is the most loaded part of the plasma-facing surfaces, and hence it is subjected to high erosion and has a significant risk of failure. A program has been developed with the aim of understanding the impact on the erosion lifetime due to a combination of two main effects: the material property variations (particularly pronounced in CFC) and the presence of joining defects. The software allows the evolution of the surface profile of the armour to be predicted and the margin on critical heat flux at the heat-sink-to-coolant interface to be estimated for a range of postulated defects, from start-of-life through to end-of-life of the component. In assessing erosion, the code takes account of geometry and sublimation, and physical and chemical erosion of the CFC armour. The incident angle (a glancing angle of a few degrees) of the particle and heat flux onto the target is taken into account. The program has been validated by comparison with analytical approximations very well validated against experimental data. The code has been developed in the APDL language to operate inside a commercial and certificated finite element program such as ANSYS

  10. 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.

  11. 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

  12. Using the Tritium Plasma Experiment to evaluate ITER PFC safety

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Anderl, R.A.; Bartlit, J.R.; Causey, R.A.; Haines, J.R.

    1993-01-01

    The Tritium Plasma Experiment was assembled at Sandia National Laboratories, Livermore to investigate interactions between dense plasmas at low energies and plasma-facing component materials. This apparatus has the unique capability of replicating plasma conditions in a tokamak divertor with particle flux densities of 2 x 10 19 ions/cm 2 · s and a plasma temperature of about 15 eV using a plasma that includes tritium. With the closure of the Tritium Research Laboratory at Livermore, the experiment was moved to the Tritium Systems Test Assembly facility at Los Alamos National Laboratory. An experimental program has been initiated there using the Tritium Plasma Experiment to examine safety issues related to tritium in plasma-facing components, particularly the ITER divertor. Those issues include tritium retention and release characteristics, tritium permeation rates and transient times to coolant streams, surface modification and erosion by the plasma, the effects of thermal loads and cycling, and particulate production. A considerable lack of data exists in these areas for many of the materials, especially beryllium, being considered for use in ITER. Not only will basic material behavior with respect to safety issues in the divertor environment be examined, but innovative techniques for optimizing performance with respect to tritium safety by material modification and process control will be investigated. Supplementary experiments will be carried out at the Idaho National Engineering Laboratory and Sandia National Laboratory to expand and clarify results obtained on the Tritium Plasma Experiment

  13. Testing of high heat flux components manufactured by ENEA for ITER divertor

    International Nuclear Information System (INIS)

    Visca, Eliseo; Escourbiac, F.; Libera, S.; Mancini, A.; Mazzone, G.; Merola, M.; Pizzuto, A.

    2009-01-01

    ENEA is involved in the International Thermonuclear Experimental Reactor (ITER) R and D activities and in particular in the manufacturing of high heat flux plasma-facing components, such as the divertor targets. During the last years ENEA has manufactured actively cooled mock-ups by using different technologies, namely brazing, diffusion bonding and HIPping. A new manufacturing process that combines two main techniques PBC (Pre-Brazed Casting) and the HRP (Hot Radial Pressing) has been set up and widely tested. A full monoblock medium scale vertical target, having a straight CFC armoured part and a curved W armoured part, was manufactured using this process. The ultrasonic method was used for the non-destructive examinations performed during the manufacturing of the component, from the monoblock preparation up to the final mock-up assembling. The component was also examined by thermography on SATIR facility (CEA, France), afterwards it was thermal fatigue tested at FE200 (200 kW electron beam facility, CEA/AREVA France). The successful results of the thermal fatigue testing performed according the ITER requirements (10 MW/m 2 , 3000 cycles of 10 s on both CFC and W part, then 20/15 MW/m 2 , 2000 cycles of 10 s on CFC/W part, respectively) have confirmed that the developed process can be considerate a candidate for the manufacturing of monoblock divertor components. Furthermore, a 35-MW/m 2 Critical Heat Flux was measured at relevant thermal-hydraulics conditions at the end of the testing campaign. This paper reports the manufacturing route, the thermal fatigue testing results, the pre and post non-destructive examination and the destructive examination performed on the ITER vertical target medium scale mock-up. These activities were performed in the frame of EFDA contracts (04-1218 with CEA, 93-851 JN with AREVA and 03-1054 with ENEA).

  14. Numerical simulations for ITER divertor armour erosion and SOL contamination due to disruptions and ELMs

    International Nuclear Information System (INIS)

    Landman, I.S.; Pestchanyi, S.E.; Bazylev, B.N.

    2005-01-01

    The divertor armour materials for ITER are going to be tungsten (as brushe or plates) and CFC. Disruptive loads with the heat deposition Q up to 30 MJ/m 2 on the time scale τ of 3 ms or operation with ELMs at repetitive loads of Q ∼ 3 MJ/m 2 and τ ∼ 0.3 ms cause enhanced armour erosion and produce contamination of SOL. Recent numerical investigations of erosion mechanisms with the anisotropic thermomechanics code PEGASUS-3D and the surface melt motion code MEMOS-1.5D as well as hot hydrogen plasma dynamics, heat loads at the armour surface and backward propagation of material plasma in SOL with the radiation-magnetohydrodynamics code FOREV-2D are survived. For CFC targets, the local overheating model is explained and numerically demonstrated. For the tungsten targets the numerical analysis of melt motion erosion of W-brushe and bulk tungsten targets on the base of MEMOS-1.5D calculations is developed and accompanied by numerical results. For validation of the codes at the regimes relevant to ITER disruptions and ELMs, the simulation results are compared with available experiments carried out at plasma guns, electron beam test facilities and the tokamak JET. (author)

  15. ATHENA calculation model for the ITER-FEAT divertor cooling system. Final report with updates

    International Nuclear Information System (INIS)

    Eriksson, John; Sjoeberg, A.; Sponton, L.L.

    2001-05-01

    An ATHENA model of the ITER-FEAT divertor cooling system has been developed for the purpose of calculating and evaluating consequences of different thermal-hydraulic accidents as specified in the Accident Analysis Specifications for the ITER-FEAT Generic Site Safety Report. The model is able to assess situations for a variety of conceivable operational transients from small flow disturbances to more critical conditions such as total blackout caused by a loss of offsite and emergency power. The main objective for analyzing this type of scenarios is to determine margins against jeopardizing the integrity of the divertor cooling system components and pipings. The model of the divertor primary heat transport system encompasses the divertor cassettes, the port limiter systems, the pressurizer, the heat exchanger and all feed and return pipes of these components. The development was pursued according to practices and procedures outlined in the ATHENA code manuals using available modelling components such as volumes, junctions, heat structures and process controls

  16. 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.)

  17. R(and)D on full tungsten divertor and beryllium wall for JET ITER-like Wall Project

    International Nuclear Information System (INIS)

    Hirai, T.; Maier, H.; Rubel, M.

    2006-01-01

    The ITER-like Wall Project was initiated at JET, with the goal of testing the reference material combination chosen for ITER: beryllium (Be) in the main chamber (wall and limiters) and tungsten (W) in the divertor. The major aims are to study the tritium retention, material mixing, melt layer behavior and to optimize plasma operation scenarios with a full metal wall. The project requires major design and engineering efforts in R(and)D: (i) bulk W tile, (ii) W coatings on carbon fibre composites (CFC) (iii) Be coatings on Inconel, (iv) Be marker tiles. For the W divertor, two R(and)D tasks were initiated: (1) development of a conceptual design for a bulk W tile as the main outer divertor target plate, and (2) W coating selection from 14 different samples produced by various techniques for the other divertor plates and neutral beam shine. The bulk W tile must withstand power loads of 7 MW/m 2 for 10 s. JET divertor plates are not actively cooled, therefore, heat capacity of the tiles is an important design parameter. In addition to power handling, mechanical structural stability under electromagnetic forces and compatibility with remote handling are the key requirements in the design. The design has been completed. The test-tile survived 100 pulses at 7 MW/m 2 for 10 s in the electron beam facility, JUDITH. The W coatings with different thickness, thin ( 2 and 200 pulses at 10 MW/m 2 for 5 s. In all tested samples cracks developed perpendicularly to the fiber bundles in CFC because of contraction of the coating in the cooling phase. Coatings were also exposed to 1000 ELM-like loading pulses. The thin coatings showed fatigue leading to delamination, whereas for thick coatings better resistance in ELM-like loading. As a result of R(and)D a full W divertor was decided: bulk metal at the outer divertor and W coating at other areas. Be-related R(and)D activities are in two areas. Production of 8-9 μm layers on inner wall cladding Inconel tiles ensures the full coating of

  18. 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

  19. 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

  20. Effects of ELMs and disruptions on ITER divertor armour materials

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  1. Effects of ELMs and disruptions on ITER divertor armour materials

    Science.gov (United States)

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

    2005-03-01

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

  2. 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.)

  3. Plasma control concepts for ITER

    International Nuclear Information System (INIS)

    Lister, J.B.; Nieswand, C.

    1997-01-01

    This overview paper skims over a wide range of issues related to the control of ITER plasmas. Although operation of the ITER project will require extensive developmental work to achieve the degree of control required, there is no indication that any of the identified problems will present overwhelming difficulties compared with the operation of present tokamaks. However, the precision of control required and the degree of automation of the final ITER plasma control system will present a challenge which is somewhat greater than for present tokamaks. In order to operate ITER optimally, integrated use of a large amount of diagnostic information will be necessary, evaluated and interpreted automatically. This will challenge both the diagnostics themselves and their supporting interpretation codes. The intervening years will provide us with the opportunity to implement and evaluate most of the new features required for ITER on existing tokamaks, with the exception of the control of an ignited plasma. (author) 7 figs., 7 refs

  4. A program to Evaluate the Erosion on the CFC tiles of the ITER Divertor

    International Nuclear Information System (INIS)

    DAgata, E.; Tivey, R.; Ogorodnikova, O.; Lowry, Ch.; Schlosser, J.

    2006-01-01

    The plasma-facing surfaces of the ITER divertor are armoured with tungsten in the upper part of the inner and outer vertical targets and carbon-fibre composite (CFC) in the lower part, the region where the scrape-off layer intercepts the divertor. The CFC in the form of a monoblock in the vertical target is the most loaded part of the plasma-facing surfaces, and hence it is subjected to high erosion and has a significant risk of failure. A program has been developed with the aim of understanding the impact on the erosion lifetime and on the probability of a critical heat flux event in the heat sink of a combination of two main effects: the material property variations (particularly pronounced in CFC) and the presence of joining defects. The software allows the evolution of the surface profile of the armour to be predicted and the margin on critical heat flux at the heat-sink-to-coolant interface to be estimated for a range of postulated defects, for start-of-life through to end-of-life of the component. In assessing erosion, the code takes account of geometry and sublimation, and physical and chemical erosion of the CFC armour. The code allows the computation of the effect of normal and off-normal (ELMs, etc.) operation. The incident angle (a glancing angle of a few degrees) of the particle and heat flux onto the target is taken into account. The program has been validated by comparison with analytical approximations and experimental data. The code has been developed in APDL language to operate inside a commercial and certificate finite element program such as ANSYS. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

  6. Status of the ITER full-tungsten divertor shaping and heat load distribution analysis

    International Nuclear Information System (INIS)

    Carpentier-Chouchana, S; Hirai, T; Escourbiac, F; Durocher, A; Fedosov, A; Ferrand, L; Kocan, M; Kukushkin, A S; Jokinen, T; Komarov, V; Lehnen, M; Merola, M; Mitteau, R; Pitts, R A; Sugihara, M; Firdaouss, M; Stangeby, P C

    2014-01-01

    In September 2011, the ITER Organization (IO) proposed to begin operation with a full-tungsten (W) armoured divertor, with the objective of taking a decision on the final target material (carbon fibre composite or W) by the end of 2013. This period of 2 years would enable the development of a full-W divertor design compatible with nuclear operations, the investigation of further several physics R and D aspects associated with the use of W targets and the completion of technology qualification. Beginning with a brief overview of the reference heat load specifications which have been defined for the full-W engineering activity, this paper will report on the current status of the ITER divertor shaping and will summarize the results of related three-dimensional heat load distribution analysis performed as part of the design validation. (paper)

  7. Thermal–hydraulic analysis of a candidate design for ITER divertor neutron flux monitor (DNFM)

    Energy Technology Data Exchange (ETDEWEB)

    Tanchuk, Victor, E-mail: Victor.Tanchuk@sintez.niiefa.spb.su [Scientific Technical Center SINTEZ, D.V. Efremov Institute, 196641 St. Petersburg (Russian Federation); Alexandrov, Evgeny [Institution “Project Center ITER”, 1, Akademika Kurchatova sq., 123182 Moscow (Russian Federation); Batyunin, Alexander; Kashchuk, Yuri [State Research Center of Russian Federation Troitsk Institute for Innovation and Fusion Research, ul. Pushkovykh, vladenie 12, 142190 Troitsk, Moscow Region (Russian Federation); Korban, Svetlana; Lyublin, Boris [Scientific Technical Center SINTEZ, D.V. Efremov Institute, 196641 St. Petersburg (Russian Federation); Obudovsky, Sergey [State Research Center of Russian Federation Troitsk Institute for Innovation and Fusion Research, ul. Pushkovykh, vladenie 12, 142190 Troitsk, Moscow Region (Russian Federation); Senik, Konstantin [Scientific Technical Center SINTEZ, D.V. Efremov Institute, 196641 St. Petersburg (Russian Federation)

    2013-10-15

    The key role in direct measurement of the ITER fusion power is assigned to the neutron diagnostic system for measurement of total neutron flux of the D–D and D–T fusion reaction with the help of a neutron flux monitor located under the divertor dome. High plasma heat loads in this position implies stringent requirements for the detector design and its cooling system to ensure the required temperature operation regime of the neutron detector. The paper describes the neutron flux monitor design developed in close collaboration with IO ITER diagnostic division. Two numerical models (hydraulic and thermal) built up to simulate the water flow in the cooling system and the temperature state of detector components are also presented and discussed. The numerical investigations carried out on the developed models have shown that only good thermal contact between the shell of the detector blocks and water-cooled casing of the monitor (fit, brazing) will provide the required temperature operation regimes of the most temperature-sensitive IFC electrodes. The obtained high temperature of the detector supports makes necessary an auxiliary direct cooling of the supports or their redesign so as to provide their higher thermal conductivity.

  8. Thermal–hydraulic analysis of a candidate design for ITER divertor neutron flux monitor (DNFM)

    International Nuclear Information System (INIS)

    Tanchuk, Victor; Alexandrov, Evgeny; Batyunin, Alexander; Kashchuk, Yuri; Korban, Svetlana; Lyublin, Boris; Obudovsky, Sergey; Senik, Konstantin

    2013-01-01

    The key role in direct measurement of the ITER fusion power is assigned to the neutron diagnostic system for measurement of total neutron flux of the D–D and D–T fusion reaction with the help of a neutron flux monitor located under the divertor dome. High plasma heat loads in this position implies stringent requirements for the detector design and its cooling system to ensure the required temperature operation regime of the neutron detector. The paper describes the neutron flux monitor design developed in close collaboration with IO ITER diagnostic division. Two numerical models (hydraulic and thermal) built up to simulate the water flow in the cooling system and the temperature state of detector components are also presented and discussed. The numerical investigations carried out on the developed models have shown that only good thermal contact between the shell of the detector blocks and water-cooled casing of the monitor (fit, brazing) will provide the required temperature operation regimes of the most temperature-sensitive IFC electrodes. The obtained high temperature of the detector supports makes necessary an auxiliary direct cooling of the supports or their redesign so as to provide their higher thermal conductivity

  9. Design and material selection for ITER first wall/blanket, divertor and vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Gohar, Y.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Lousteau, D.; Onozuka, M.; Parker, R.; Sannazzaro, G.; Tivey, R. [ITER JCT, Garching (Germany)

    1998-10-01

    Design and R and D have progressed on the ITER vacuum vessel, shielding and breeding blankets, and the divertor. The principal materials have been selected and the fabrication methods selected for most of the components based on design and R and D results. The resulting design changes are discussed for each system. (orig.) 11 refs.

  10. Design and material selection for ITER first wall/blanket, divertor and vacuum vessel

    Science.gov (United States)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Gohar, Y.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Lousteau, D.; Onozuka, M.; Parker, R.; Sannazzaro, G.; Tivey, R.

    1998-10-01

    Design and R&D have progressed on the ITER vacuum vessel, shielding and breeding blankets, and the divertor. The principal materials have been selected and the fabrication methods selected for most of the components based on design and R&D results. The resulting design changes are discussed for each system.

  11. 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

  12. 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

  13. Engineering analyses of ITER divertor diagnostic rack design

    Energy Technology Data Exchange (ETDEWEB)

    Modestov, Victor S., E-mail: modestov@compmechlab.com [St Petersburg State Polytechnical University, 195251 St Petersburg, 29 Polytechnicheskaya (Russian Federation); Nemov, Alexander S.; Borovkov, Aleksey I.; Buslakov, Igor V.; Lukin, Aleksey V. [St Petersburg State Polytechnical University, 195251 St Petersburg, 29 Polytechnicheskaya (Russian Federation); Kochergin, Mikhail M.; Mukhin, Eugene E.; Litvinov, Andrey E.; Koval, Alexandr N. [Ioffe Physico-Technical Institute, 194021 St Petersburg, 26 Polytechnicheskaya (Russian Federation); Andrew, Philip [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: • The approach developed early has been used for the assessment of new design of DTS racks and neutron shield units. • Results of most critical EM and seismic analyses indicate that introduced changes significantly improved the system behaviour under these loads. • However further research is required to finalize the design and check it upon meeting all structural, thermal, seismic, EM and fatigue requirements. -- Abstract: The divertor port racks used as a support structure of the divertor Thomson scattering equipment has been carefully analyzed to be consistent with electromagnetic and seismic loads. It follows from the foregoing simulations that namely these analyses demonstrate critical challenges associated with the structure design. Based on the results of the reference structure [2] a modified design of the diagnostic racks is proposed and updated simulation results are given. The results signify a significant improvement over the previous reference layout and the design will be continued towards finalization.

  14. 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)

  15. 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)

  16. Assessment of erosion of the ITER divertor targets during type I ELMs

    Science.gov (United States)

    Federici, G.; Loarte, A.; Strohmayer, G.

    2003-09-01

    This paper presents the results of a preliminary assessment conducted to estimate the thermal response and erosion lifetime of the ITER divertor targets clad either with carbon-fibre composite or tungsten during type I ELMs. The one-dimensional thermal/erosion model, used for the analyses, is briefly described. It includes all the key surface heat transfer processes such as evaporation, melting, and radiation, and their interaction with the bulk thermal response, and it is based on an implicit finite-difference scheme, which allows for temperature-dependent material properties. The cases analysed clarify the influence of several ELM parameters on the heat transfer and erosion processes at the target (i.e. characteristic plasma ELM energy loss from the pedestal, fraction of the energy reaching the divertor, broadening of the strike-points during ELMs, duration and waveform of the ELM heat load) and design/material parameters (i.e. inclination of the target, type and thickness of the armour material, and for tungsten only, fraction of the melt layer loss). Comparison is made between cases where all ELMs are characterized by the same fixed averaged parameters, and cases where instead the characteristic parameters of each ELM are evaluated in a random fashion by using a standard Monte Carlo technique, based on distributions of some of the variables of interest derived from experiments in today's machines. Although uncertainties rule out providing firm quantitative predictions, the results of this study are useful to illustrate trends. Based on the results, the implications on the design and operation are discussed and priorities are determined for the R&D needed to reduce the remaining uncertainties.

  17. Assessment of erosion of the ITER divertor targets during type I ELMs

    International Nuclear Information System (INIS)

    Federici, G; Loarte, A; Strohmayer, G

    2003-01-01

    This paper presents the results of a preliminary assessment conducted to estimate the thermal response and erosion lifetime of the ITER divertor targets clad either with carbon-fibre composite or tungsten during type I ELMs. The one-dimensional thermal/erosion model, used for the analyses, is briefly described. It includes all the key surface heat transfer processes such as evaporation, melting, and radiation, and their interaction with the bulk thermal response, and it is based on an implicit finite-difference scheme, which allows for temperature-dependent material properties. The cases analysed clarify the influence of several ELM parameters on the heat transfer and erosion processes at the target (i.e. characteristic plasma ELM energy loss from the pedestal, fraction of the energy reaching the divertor, broadening of the strike-points during ELMs, duration and waveform of the ELM heat load) and design/material parameters (i.e. inclination of the target, type and thickness of the armour material, and for tungsten only, fraction of the melt layer loss). Comparison is made between cases where all ELMs are characterized by the same fixed averaged parameters, and cases where instead the characteristic parameters of each ELM are evaluated in a random fashion by using a standard Monte Carlo technique, based on distributions of some of the variables of interest derived from experiments in today's machines. Although uncertainties rule out providing firm quantitative predictions, the results of this study are useful to illustrate trends. Based on the results, the implications on the design and operation are discussed and priorities are determined for the R and D needed to reduce the remaining uncertainties

  18. Magnetic Configuration Control of ITER Plasmas

    International Nuclear Information System (INIS)

    Albanese, R.; Artaserse, G.; Mattei, M.; Ambrosino, G.; Crisanti, F.; Tommasi, G. de; Fresa, R.; Portone, A.; Sartori, F.; Villone, F.

    2006-01-01

    The aim of this paper is to review the capability of the ITER Poloidal Field (PF) system of controlling the broad range of plasma configurations presently forecasted during ITER operation. The attention is focused on the axi-symmetric aspects of plasma magnetic configuration control since they pose the greatest challenges in terms of control power and they have the largest impact on machine capital cost. The paper is broadly divided in two main sections devoted, respectively, to open loop (feed-forward) and closed loop (feedback) control. In the first part of the study the PF system is assessed with respect to the initiation, ramp-up, sustained burn, ramp-down phases of the main plasma inductive scenario. The limiter-to-divertor configuration transition phase is considered in detail with the aim of assessing the PF capability to form an X-point at the lowest possible current and, therefore, to relax the thermal load on the limiter surfaces. Moreover, during the sustained burn it is important to control plasmas with a broad range of current density profiles. In the second part of the study the plasma vertical feedback control requirements are assessed in details, in particular for the high elongation configurations achievable during the early limiter-to-X point transition phase. Non-rigid plasma displacement models are used to assess the control system voltage and current requirements of different radial field control circuits obtained, for example, by connecting the outermost PF coils, some CS coils, coils sub-sections etc. At last, the main 3D effects of the vessel ports are modeled and their impact of vertical stabilization evaluated. (author)

  19. 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

  20. 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.

  1. 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.

  2. 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

  3. 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)

  4. 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)

  5. ITER-EDA physics design requirements and plasma performance assessments

    International Nuclear Information System (INIS)

    Uckan, N.A.; Galambos, J.; Wesley, J.; Boucher, D.; Perkins, F.; Post, D.; Putvinski, S.

    1996-01-01

    Physics design guidelines, plasma performance estimates, and sensitivity of performance to changes in physics assumptions are presented for the ITER-EDA Interim Design. The overall ITER device parameters have been derived from the performance goals using physics guidelines based on the physics R ampersand D results. The ITER-EDA design has a single-null divertor configuration (divertor at the bottom) with a nominal plasma current of 21 MA, magnetic field of 5.68 T, major and minor radius of 8.14 m and 2.8 m, and a plasma elongation (at the 95% flux surface) of ∼1.6 that produces a nominal fusion power of ∼1.5 GW for an ignited burn pulse length of ≥1000 s. The assessments have shown that ignition at 1.5 GW of fusion power can be sustained in ITER for 1000 s given present extrapolations of H-mode confinement (τ E = 0.85 x τ ITER93H ), helium exhaust (τ* He /τ E = 10), representative plasma impurities (n Be /n e = 2%), and beta limit [β N = β(%)/(I/aB) ≤ 2.5]. The provision of 100 MW of auxiliary power, necessary to access to H-mode during the approach to ignition, provides for the possibility of driven burn operations at Q = 15. This enables ITER to fulfill its mission of fusion power (∼ 1--1.5 GW) and fluence (∼1 MWa/m 2 ) goals if confinement, impurity levels, or operational (density, beta) limits prove to be less favorable than present projections. The power threshold for H-L transition, confinement uncertainties, and operational limits (Greenwald density limit and beta limit) are potential performance limiting issues. Improvement of the helium exhaust (τ* He /τ E ≤ 5) and potential operation in reverse-shear mode significantly improve ITER performance

  6. 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

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

    Science.gov (United States)

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

    2017-12-01

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

  8. 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

  9. 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

  10. 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.

  11. 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.)

  12. 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

  13. 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.

  14. 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.)

  15. 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.

  16. Divertor materials for ITER - Tungsten and carbon/carbon composite behavior under coupled ionic irradiation and high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Raunier, S.; Balat-Pichelin, M.; Sans, J.L.; Hernandez, D. [Laboratoire PROMES-CNRS, Laboratoire Procedes, Materiaux et Energie Solaire, 7 rue du Four Solaire, 66120 Font-Romeu Odeillo (France)

    2007-07-01

    Full text of publication follows: In the frame of the International Thermonuclear Experimental Reactor ITER, the physical-chemical characterization of plasma-facing components (divertor and structural materials) is essential because they are subjected to simultaneous high thermal and ionic fluxes. In this paper, an experimental and theoretical study of the physical-chemical behavior of carbon/carbon composite and tungsten (materials for ITER divertor) under extreme conditions is performed. The simulation of the interaction of hydrogen ions with the material, the theoretical study of physical erosion (TRIM and TRIDYN codes) and the chemical erosion (GEMINI code) are carried out. The conditions of nominal or accidental mode that can occur during the operation of the reactor (high temperature 1300 - 2500 K, high vacuum, H{sup +} ionic flux with different energies) are experimentally simulated. In this work, we have studied the material degradation, the mass loss kinetics, the characterization of the emitted neutral and charged species of heated and both heated and irradiated materials, and the determination of the thermo-radiative properties versus time. This study, done in collaboration with CEA Cadarache, is realized using the MEDIASE experimental device (Moyen d'Essai et de Diagnostic en Ambiance Solaire Extreme) located at the focus of the 1000 kW solar furnace of PROMES-CNRS laboratory in Odeillo. Material characterization pre- and post-processing is performed with classical techniques as SEM, XRD and XPS and also by measuring the BRDF (Bidirectional Reflectivity Diffusion Function). (authors)

  17. Divertor materials for ITER - Tungsten and carbon/carbon composite behavior under coupled ionic irradiation and high temperature

    International Nuclear Information System (INIS)

    Raunier, S.; Balat-Pichelin, M.; Sans, J.L.; Hernandez, D.

    2007-01-01

    Full text of publication follows: In the frame of the International Thermonuclear Experimental Reactor ITER, the physical-chemical characterization of plasma-facing components (divertor and structural materials) is essential because they are subjected to simultaneous high thermal and ionic fluxes. In this paper, an experimental and theoretical study of the physical-chemical behavior of carbon/carbon composite and tungsten (materials for ITER divertor) under extreme conditions is performed. The simulation of the interaction of hydrogen ions with the material, the theoretical study of physical erosion (TRIM and TRIDYN codes) and the chemical erosion (GEMINI code) are carried out. The conditions of nominal or accidental mode that can occur during the operation of the reactor (high temperature 1300 - 2500 K, high vacuum, H + ionic flux with different energies) are experimentally simulated. In this work, we have studied the material degradation, the mass loss kinetics, the characterization of the emitted neutral and charged species of heated and both heated and irradiated materials, and the determination of the thermo-radiative properties versus time. This study, done in collaboration with CEA Cadarache, is realized using the MEDIASE experimental device (Moyen d'Essai et de Diagnostic en Ambiance Solaire Extreme) located at the focus of the 1000 kW solar furnace of PROMES-CNRS laboratory in Odeillo. Material characterization pre- and post-processing is performed with classical techniques as SEM, XRD and XPS and also by measuring the BRDF (Bidirectional Reflectivity Diffusion Function). (authors)

  18. Feasibility of ''gas target'' mode of divertor operation in ITER

    International Nuclear Information System (INIS)

    Kukushkin, A.S.

    1994-01-01

    Power load upon the divertor target remains one of the most critical issues for a tokamak reactor. Simple estimates, confirmed by 2D modelling, together with some indications from tokamak experiments, showed that the profile of power flow gets narrower along with increase of the reactor power, because strong temperature dependence of the parallel heat conductance, χ parallel αΤ 5/2 , favours parallel heat transport in competition with the cross-field one. This leads to unacceptable peak loads and makes one to look for a means to spread the power more evenly across the magnetic field. The scope of the present paper is to show the results of the modelling studies and to discuss the physical and computational issues which are still missing or are insufficiently developed. I must apologize for partiality for my own calculations with the DDC83 code, but there are some reasons justifying this: they have been the first calculations on this issue, they seem to be the most extensive, and they are certainly the most familiar to me. (orig.)

  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. 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

  4. 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.)

  5. 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.)

  6. 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

  7. 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.

  8. Design, fabrication, and testing of a helium-cooled module for the ITER divertor

    International Nuclear Information System (INIS)

    Baxi, C.B.; Smith, J.P.; Youchison, D.

    1994-08-01

    The International Thermonuclear Reactor (ITER) will have a single-null divertor with total power flow of 200 MW and a peak heat flux of about 5 MW/m 2 . The reference coolant for the divertor is water. However, helium is a viable alternative and offers advantages from safety considerations, such as excellent radiation stability and chemical inertness. In order to prove the feasibility of helium cooling at ITER relevant heat flux conditions, General Atomics designed, fabricated, and tested a helium-cooled divertor module. The module was made from dispersion strengthened copper, with a heat flux surface 25 mm wide and 80 mm long, designed for twice the ITER divertor heat flux. Different techniques were examined to enhance the heat transfer, which in turn reduced the flow and pumping power required to cool the module. It was concluded that an extended surface was the most practical solution. An optimization study was performed to find the best extended surface parameters. The optimum extended surface geometry consisted of fins: 10 mm high, 0.4 mm thick with a 1 mm pitch. It was estimated to require a pumping power of 150 W to remove 20 kW of power. This is more than an order of magnitude reduction in pumping power requirement, compared to smooth surface. The module was fabricated by electric discharge machining (EDM) process. The testing was carried out at SNLA during August 1993. The testing confirmed the design calculations. The peak heat flux during the test was 10 MW/m 2 applied over a surface area of 20 cm 2 . The pumping power calculated from flow rate and pressure drop measurement was about 160 W, which was less than 1% of the power removed. It is planned to test the module to higher temperature limits and higher heat fluxes during coming months. As a result of this effort we conclude that helium cooling of the ITER divertor is feasible without requiring a very large helium pressure or a large pumping power

  9. 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

  10. 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)

  11. Fabrication of the wing and vertical target dummy armour prototypes of the ITER divertor

    International Nuclear Information System (INIS)

    Grattarola, M.; Bet, M.; Biagiotti, B.; Gandini, G.; Merola, M.; Ottonello, G.B.; Riccardi, B.; Vieider, G.; Zacchia, F.

    2000-01-01

    The dummy armour prototypes are identical to the reference components in terms of geometry, cooling circuit and material except for the armour material, which is replaced by an equivalent thickness of copper alloy. The main objectives of the dummy armour prototypes are the demonstration of the overall engineering concept of the Divertor, the integration in a 3 deg. cassette together with components manufactured by the other ITER Home Teams and the successive thermo-hydraulic tests on the whole Divertor module. This paper describes the realization of both the wing and the vertical target dummy armour prototypes focusing on the critical aspects of the fabrication and their impact on a further industrialization of the components

  12. Fabrication of the wing and vertical target dummy armour prototypes of the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Grattarola, M. E-mail: gratta@ari.ansaldo.it; Bet, M.; Biagiotti, B.; Gandini, G.; Merola, M.; Ottonello, G.B.; Riccardi, B.; Vieider, G.; Zacchia, F

    2000-11-01

    The dummy armour prototypes are identical to the reference components in terms of geometry, cooling circuit and material except for the armour material, which is replaced by an equivalent thickness of copper alloy. The main objectives of the dummy armour prototypes are the demonstration of the overall engineering concept of the Divertor, the integration in a 3 deg. cassette together with components manufactured by the other ITER Home Teams and the successive thermo-hydraulic tests on the whole Divertor module. This paper describes the realization of both the wing and the vertical target dummy armour prototypes focusing on the critical aspects of the fabrication and their impact on a further industrialization of the components.

  13. Experimental study of a RF plasma source with helicon configuration in the mix Ar/H_2. Application to the chemical etching of carbon materials surfaces in the framework of the plasma-wall interactions studies of ITER's divertor

    International Nuclear Information System (INIS)

    Bieber, T.

    2012-01-01

    The issue of the interaction wall-plasma is important in thermonuclear devices. The purpose of this work is to design a very low pressure atomic plasma source in order to study chemical etching of carbon surfaces in the same conditions as edge plasma in tokamaks. The experimental work has consisted in 2 stages: first, the characterisation of the new helicon configuration reactor developed for this research and secondly the atomic hydrogen source used for the chemical etching. The first chapter recalls what thermonuclear fusion is. The helicon configuration reactor as well as its diagnostics (optical emission spectroscopy, laser induced fluorescence - LIF, and Langmuir probe) are described in the second chapter. The third chapter deals with the different coupling modes (RF power and plasma) identified in pure argon plasmas and how they are obtained by setting experimental parameters such as injected RF power, magnetic fields or pressure. The fourth chapter is dedicated to the study of the difference in behavior between the electronic density and the relative density of metastable Ar"+ ions. The last chapter presents the results in terms of mass losses of the carbon material surfaces obtained with the atomic hydrogen source. (A.C.)

  14. Plasma regimes and research goals of JT-60SA towards ITER and DEMO

    International Nuclear Information System (INIS)

    Kamada, Y.; Ide, S.; Fujita, T.; Suzuki, T.; Matsunaga, G.; Yoshida, M.; Shinohara, K.; Urano, H.; Nakano, T.; Sakurai, S.; Kawashima, H.; Barabaschi, P.; Lackner, K.; Ishida, S.; Bolzonella, T.

    2011-01-01

    The JT-60SA device has been designed as a highly shaped large superconducting tokamak with a variety of plasma actuators (heating, current drive, momentum input, stability control coils, resonant magnetic perturbation coils, W-shaped divertor, fuelling, pumping, etc) in order to satisfy the central research needs for ITER and DEMO. In the ITER- and DEMO-relevant plasma parameter regimes and with DEMO-equivalent plasma shapes, JT-60SA quantifies the operation limits, plasma responses and operational margins in terms of MHD stability, plasma transport and confinement, high-energy particle behaviour, pedestal structures, scrape-off layer and divertor characteristics. By integrating advanced studies in these research fields, the project proceeds 'simultaneous and steady-state sustainment of the key performances required for DEMO' with integrated control scenario development applicable to the highly self-regulating burning high-β high bootstrap current fraction plasmas.

  15. Damage evaluation under thermal fatigue of a vertical target full scale component for the ITER divertor

    International Nuclear Information System (INIS)

    Missirlian, M.; Escourbiac, F.; Merola, M.; Durocher, A.; Bobin-Vastra, I.; Schedler, B.

    2007-01-01

    An extensive development programme has been carried out in the EU on high heat flux components within the ITER project. In this framework, a Full Scale Vertical Target (VTFS) prototype was manufactured with all the main features of the corresponding ITER divertor design. The fatigue cycling campaign on CFC and W armoured regions, proved the capability of such a component to meet the ITER requirements in terms of heat flux performances for the vertical target. This paper discusses thermographic examination and thermal fatigue testing results obtained on this component. The study includes thermal analysis, with a tentative proposal to evaluate with finite element approach the location/size of defects and the possible propagation during fatigue cycling

  16. Examination of high heat flux components for the ITER divertor after thermal fatigue testing

    International Nuclear Information System (INIS)

    Missirlian, M.; Escourbiac, F.; Schmidt, A.; Riccardi, B.; Bobin-Vastra, I.

    2011-01-01

    An extensive development programme has been carried out in the EU on high heat flux components within the ITER project. In this framework, a full-scale vertical target (VTFS) prototype was manufactured with all the main features of the corresponding ITER divertor design. The fatigue cycling campaign on CFC and W armoured regions, proved the capability of such a component to meet the ITER requirements in terms of heat flux performances for the vertical target. This paper discusses metallographic observations performed on both CFC and W part after this intensive thermal fatigue testing campaign for a better understanding of thermally induced mechanical stress within the component, especially close to the armour-heat sink interface.

  17. Examination of high heat flux components for the ITER divertor after thermal fatigue testing

    Energy Technology Data Exchange (ETDEWEB)

    Missirlian, M., E-mail: marc.missirlian@cea.fr [CEA, IRFM, F-13108 Saint Paul lez Durance (France); Escourbiac, F., E-mail: frederic.escourbiac@cea.fr [CEA, IRFM, F-13108 Saint Paul lez Durance (France); Schmidt, A., E-mail: a.schmidt@fz-juelich.de [Forschungszentrum Juelich, IFE-2 (Germany); Riccardi, B., E-mail: Bruno.Riccardi@f4e.europa.eu [Fusion For Energy, E-08019 Barcelona (Spain); Bobin-Vastra, I., E-mail: isabelle.bobinvastra@areva.com [AREVA-NP, 71200 Le Creusot (France)

    2011-10-01

    An extensive development programme has been carried out in the EU on high heat flux components within the ITER project. In this framework, a full-scale vertical target (VTFS) prototype was manufactured with all the main features of the corresponding ITER divertor design. The fatigue cycling campaign on CFC and W armoured regions, proved the capability of such a component to meet the ITER requirements in terms of heat flux performances for the vertical target. This paper discusses metallographic observations performed on both CFC and W part after this intensive thermal fatigue testing campaign for a better understanding of thermally induced mechanical stress within the component, especially close to the armour-heat sink interface.

  18. 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

  19. Analyses of microstructure, composition and retention of hydrogen isotopes in divertor tiles of JET with the ITER-like wall

    Science.gov (United States)

    Masuzaki, S.; Tokitani, M.; Otsuka, T.; Oya, Y.; Hatano, Y.; Miyamoto, M.; Sakamoto, R.; Ashikawa, N.; Sakurada, S.; Uemura, Y.; Azuma, K.; Yumizuru, K.; Oyaizu, M.; Suzuki, T.; Kurotaki, H.; Hamaguchi, D.; Isobe, K.; Asakura, N.; Widdowson, A.; Heinola, K.; Jachmich, S.; Rubel, M.; contributors, JET

    2017-12-01

    Results of the comprehensive surface analyses of divertor tiles and dusts retrieved from JET after the first ITER-like wall campaign (2011-2012) are presented. The samples cored from the divertor tiles were analyzed. Numerous nano-size bubble-like structures were observed in the deposition layer on the apron of the inner divertor tile, and a beryllium dust with the same structures were found in the matter collected from the inner divertor after the campaign. This suggests that the nano-size bubble-like structures can make the deposition layer to become brittle and may lead to cracking followed by dust generation. X-ray photoelectron spectroscopy analyses of chemical states of species in the deposition layers identified the formation of beryllium-tungsten intermetallic compounds on an inner vertical tile. Different tritium retention profiles along the divertor tiles were observed at the top surfaces and at deeper regions of the tiles by using the imaging plate technique.

  20. Critical heat flux performance of hypervapotrons proposed for use in the ITER divertor vertical target

    International Nuclear Information System (INIS)

    Youchison, D.L.; Marshall, T.D.; McDonald, J.M.; Lutz, T.J.; Watson, R.D.; Driemeyer, D.E.; Kubik, D.L.; Slattery, K.T.; Hellwig, T.H.

    1997-09-01

    Task T-222 of the International Thermonuclear Experimental Reactor (ITER) program addresses the manufacturing and testing of permanent components for use in the ITER divertor. Thermalhydraulic and critical heat flux performance of the heat sinks proposed for use in the divertor vertical target are part of subtask T-222.4. As part of this effort, two single channel, medium scale, bare copper alloy, hypervapotron mockups were designed, fabricated, and tested using the EB-1200 electron beam system. The objectives of the effort were to develop the design and manufacturing procedures required for construction of robust high heat flux (HHF) components, verify thermalhydraulic, thermomechanical and critical heat flux (CHF) performance under ITER relevant conditions, and perform analyses of HHF data to identify design guidelines and failure criteria and possibly modify any applicable CHF correlations. The design, fabrication, and finite element modeling of two types of hypervapotrons are described; a common version already in use at the Joint European Torus (JET) and a new attached fin design. HHF test data on the attached fin hypervapotron will be used to compare the CHF performance under uniform heating profiles on long heated lengths with that of localized, highly peaked, off nominal profiles

  1. Critical heat flux performance of hypervapotrons proposed for use in the ITER divertor vertical target

    Energy Technology Data Exchange (ETDEWEB)

    Youchison, D.L.; Marshall, T.D.; McDonald, J.M.; Lutz, T.J.; Watson, R.D. [Sandia National Labs., Albuquerque, NM (United States); Driemeyer, D.E. Kubik, D.L.; Slattery, K.T.; Hellwig, T.H. [McDonnell Douglas Aerospace, St. Louis, MO (United States)

    1997-09-01

    Task T-222 of the International Thermonuclear Experimental Reactor (ITER) program addresses the manufacturing and testing of permanent components for use in the ITER divertor. Thermalhydraulic and critical heat flux performance of the heat sinks proposed for use in the divertor vertical target are part of subtask T-222.4. As part of this effort, two single channel, medium scale, bare copper alloy, hypervapotron mockups were designed, fabricated, and tested using the EB-1200 electron beam system. The objectives of the effort were to develop the design and manufacturing procedures required for construction of robust high heat flux (HHF) components, verify thermalhydraulic, thermomechanical and critical heat flux (CHF) performance under ITER relevant conditions, and perform analyses of HHF data to identify design guidelines and failure criteria and possibly modify any applicable CHF correlations. The design, fabrication, and finite element modeling of two types of hypervapotrons are described; a common version already in use at the Joint European Torus (JET) and a new attached fin design. HHF test data on the attached fin hypervapotron will be used to compare the CHF performance under uniform heating profiles on long heated lengths with that of localized, highly peaked, off nominal profiles.

  2. Low cycle fatigue behavior of ITER-like divertor target under DEMO-relevant operation conditions

    Energy Technology Data Exchange (ETDEWEB)

    Li, Muyuan; Werner, Ewald [Lehrstuhl für Werkstoffkunde und Werkstoffmechanik, Technische Universität München, Boltzmannstr. 15, 85748 Garching (Germany); You, Jeong-Ha, E-mail: you@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany)

    2015-01-15

    Highlights: • LCF behavior of the cooling tube and the interlayer of an ITER-like divertor target is studied. • For the cooling tube, LCF failure will not be an issue under an HHF load of up to 18 MW/m{sup 2}. • Plastic strain in the interlayer is concentrated at the free surface edge of the bond interface. • The predicted LCF lifetime of the interlayer may not meet the design requirement. - Abstract: In this work the low cycle fatigue (LCF) behavior of the copper alloy cooling tube and the copper interlayer of an ITER-like divertor target is reported for nine different combinations of loading and cooling conditions relevant to DEMO divertor operation. The LCF lifetime is presented as a function of loading and cooling conditions considered here by means of cyclic plasticity simulation and using LCF data of materials relevant for ITER. The numerical predictions indicate, that fatigue failure will not be an issue for the copper alloy tube under a high heat flux (HHF) load of up to 18 MW/m{sup 2} as long as it preserves its initial strength. In contrast, the copper interlayer exhibits significant plastic dissipation at the free surface edge of the bond interface adjacent to the cooling tube, where the LCF lifetime is predicted to be below 3000 load cycles for HHF loads higher than 15 MW/m{sup 2}. Most of the bulk region of the copper interlayer away from the free surface edge does not experience severe plastic fatigue and hence does not pose any critical concern as the LCF lifetime is predicted to be at least 7000 load cycles. LCF lifetime decreases as HHF load is increased or coolant temperature is decreased.

  3. 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.)

  4. 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)

  5. 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

  6. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

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

  8. Design and fabrication methods of FW/blanket, divertor and vacuum vessel for ITER

    International Nuclear Information System (INIS)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Ibbott, C.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Tivey, R.; Utin, Y.; Yamada, M.

    2000-01-01

    Design has progressed on the vacuum vessel, FW/blanket and Divertor for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design [K. Ioki et al., J. Nucl. Mater. 258-263 (1998) 74]. Design and fabrication methods of the components have been improved to achieve ∼50% reduction of the construction cost. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R and D performed so far during the Engineering Design Activities (EDAs) are still applicable. Further cost reduction methods are also being investigated and additional R and D is being performed

  9. Ex-vessel break in ITER divertor cooling loop analysis with the ECART code

    CERN Document Server

    Cambi, G; Parozzi, F; Porfiri, MT

    2003-01-01

    A hypothetical double-ended pipe rupture in the ex-vessel section of the International Thermonuclear Experimental Reactor (ITER) divertor primary heat transfer system during pulse operation has been assessed using the nuclear source term ECART code. That code was originally designed and validated for traditional nuclear power plant safety analyses, and has been internationally recognized as a relevant nuclear source term codes for nuclear fission plants. It permits the simulation of chemical reactions and transport of radioactive gases and aerosols under two-phase flow transients in generic flow systems, using a built-in thermal-hydraulic model. A comparison with the results given in ITER Generic Site Safety Report, obtained using a thermal-hydraulic system code (ATHENA), a containment code (INTRA) and an aerosol transportation code (NAUA), in a sequential way, is also presented and discussed.

  10. Design and fabrication methods of FW/blanket, divertor and vacuum vessel for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K. E-mail: iokik@itereu.deiokik@ipp.mpg.de; Barabash, V.; Cardella, A.; Elio, F.; Ibbott, C.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Tivey, R.; Utin, Y.; Yamada, M

    2000-12-01

    Design has progressed on the vacuum vessel, FW/blanket and Divertor for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design [K. Ioki et al., J. Nucl. Mater. 258-263 (1998) 74]. Design and fabrication methods of the components have been improved to achieve {approx}50% reduction of the construction cost. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R and D performed so far during the Engineering Design Activities (EDAs) are still applicable. Further cost reduction methods are also being investigated and additional R and D is being performed.

  11. Design and fabrication methods of FW/blanket, divertor and vacuum vessel for ITER

    Science.gov (United States)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Ibbott, C.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Tivey, R.; Utin, Y.; Yamada, M.

    2000-12-01

    Design has progressed on the vacuum vessel, FW/blanket and Divertor for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design [K. Ioki et al., J. Nucl. Mater. 258-263 (1998) 74]. Design and fabrication methods of the components have been improved to achieve ˜50% reduction of the construction cost. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R&D performed so far during the Engineering Design Activities (EDAs) are still applicable. Further cost reduction methods are also being investigated and additional R&D is being performed.

  12. 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

  13. Critical heat flux analysis and R and D for the design of the ITER divertor

    International Nuclear Information System (INIS)

    Raffray, A.R.; Chiocchio, S.; Merola, M.; Tivey, R.; Vieider, G.; Schlosser, J.; Driemeyer, D.; Escourbiac, F.; Grigoriev, S.; Youchison, D.

    1999-01-01

    The vertical target and dump target of the ITER divertor have to be designed for high heat fluxes (up to 20 MW/m 2 over ∼10 s). Accommodation of such high heat fluxes gives rise to several issues, including the critical heat flux (CHF) margin which is a key requirement influencing the choice of cooling channel geometry and coolant conditions. An R and D programme was evolved to address the overall CHF issue and to help focus the design. It involved participation of the four ITER home teams and has been very successful in substantially expanding the CHF data base for one-sided heating and in providing more accurate experimental measurements of pressure drop (and derived correlations) for these geometries. This paper describes the major R and D results and the design analysis performed in converging on a choice of reference configuration and parameters which resulted in a CHF margin of ∼1.4 or more for all divertor components. (orig.)

  14. 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

  15. 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

  16. Research and development of remote maintenance equipment for ITER divertor maintenance

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka

    2005-02-01

    To facilitate easy remote maintainability, the ITER divertor is divided into 60 cassettes, which are transported in the toroidal and radial directions for replacement through maintenance ports located every 90 degrees using the divertor remote maintenance equipment such as in- and ex-vessel transporters. The cassette of 25 tons has to be transported and installed in the vacuum vessel with a positioning accuracy less than 2 mm in the limited space of the vacuum vessel and maintenance port under the intense gamma radiation field. Based on these requirements, the following design and tests were performed. (1) Link mechanism was studied to apply to the transportation of the heavy cassette in the restricted space. A compact mechanism with links for transportation of heavy cassette is designed through the optimization of the link angle taking account of space requirement and force efficiency. As a test result, the lifting capacity of 30 tons (larger than the cassette weight of 25 tons) using two link mechanisms has been demonstrated in the limited space. (2) Compact link mechanism was also studied to apply for locking of the cassette through the optimization of the link angle taking account of space requirement and force efficiency. As a test result, the final positioning accuracy of 0.03 mm for the 25 tons-cassette installation on the vacuum vessel from the initial positioning error of 5 mm has been demonstrated, so that the test result satisfies the requirement less than 2 mm using the link mechanisms in the limited space. (3) Sensor-based control using simple sensors such as optical fiber for divertor maintenance was tested using the full-scale mock-up divertor cassette and remote maintenance equipment. As a result, it is found that the positioning accuracy of 0.16 mm has been achieved by the optical fiber sensor and this value is sufficient for sensor-based control. In addition, the maintenance operation has been carried out through the human-machine interface

  17. 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)

  18. Tests on the integration of the ITER divertor dummy armour prototype on a simplified model of cassette body

    International Nuclear Information System (INIS)

    Dell'Orco, G.; Canneta, A.; Cattadori, G.; Gaspari, G.P.; Merola, M.; Polazzi, G.; Vieider, G.; Zito, D.

    2001-01-01

    In 1998, in the frame of the European R and D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body, designed with some mechanical and hydraulic simplifications with respect to the reference body, and the actively cooled Dummy Armour Prototype (DAP). This DAP consists of the Vertical Target, the Wing and the Dump Target, manufactured by the European industry, which are integrated with the Gas Box Liner supplied by the Russian Federation Home Team. In order to simplify the manufacturing, the DAP was layered with an equivalent CuCrZr thickness simulating the real armour (CFC or W tiles). In parallel with the manufacturing activity, the ITER European HT decided to assign to ENEA the Task EU-DV1 for the 'Component Integration and Thermal-Hydraulic Testing of the ITER Divertor Targets and Wing Dummy Prototypes and Cassette Body'

  19. Tests on the integration of the ITER divertor dummy armour prototype on a simplified model of cassette body

    Energy Technology Data Exchange (ETDEWEB)

    Dell' Orco, G. E-mail: dellorco@brasimone.enea.it; Canneta, A.; Cattadori, G.; Gaspari, G.P.; Merola, M.; Polazzi, G.; Vieider, G.; Zito, D

    2001-10-01

    In 1998, in the frame of the European R and D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body, designed with some mechanical and hydraulic simplifications with respect to the reference body, and the actively cooled Dummy Armour Prototype (DAP). This DAP consists of the Vertical Target, the Wing and the Dump Target, manufactured by the European industry, which are integrated with the Gas Box Liner supplied by the Russian Federation Home Team. In order to simplify the manufacturing, the DAP was layered with an equivalent CuCrZr thickness simulating the real armour (CFC or W tiles). In parallel with the manufacturing activity, the ITER European HT decided to assign to ENEA the Task EU-DV1 for the 'Component Integration and Thermal-Hydraulic Testing of the ITER Divertor Targets and Wing Dummy Prototypes and Cassette Body'.

  20. Plasma position and shape control for ITER

    International Nuclear Information System (INIS)

    Portone, A.; Gribov, Y.; Huguet, M.

    1995-01-01

    Key features and main results about the control of the plasma shape in ITER are presented. A control algorithm is designed to control up to 6 gaps between the plasma separatrix and the plasma facing components during the reference burn phase. Nonlinear simulations show the performances of the controller in the presence of plasma vertical position offsets, beta drops and power supply voltage saturation

  1. 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.)

  2. Manufacturing and testing of a prototypical divertor vertical target for ITER

    Science.gov (United States)

    Merola, M.; Plöchl, L.; Chappuis, Ph; Escourbiac, F.; Grattarola, M.; Smid, I.; Tivey, R.; Vieider, G.

    2000-12-01

    After an extensive R&D activity, a medium-scale divertor vertical target prototype has been manufactured by the EU Home Team. This component contains all the main features of the corresponding ITER divertor design and consists of two units with one cooling channel each, assembled together and having an overall length and width of about 600 and 50 mm, respectively. The upper part of the prototype has a tungsten macro-brush armour, whereas the lower part is covered by CFC monoblocks. A number of joining techniques were required to manufacture this component as well as an appreciable effort in the development of suitable non-destructive testing methods. The component was high heat flux tested in FE200 electron beam facility at Le Creusot, France. It endured 100 cycles at 5 MW/m 2, 1000 cycles at 10 MW/m 2 and more then 1000 cycles at 15-20 MW/m 2. The final critical heat flux test reached a value in excess of 30 MW/m 2.

  3. Manufacturing and testing of a prototypical divertor vertical target for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Merola, M. E-mail: merolam@ipp.mpg.de; Ploechl, L.; Chappuis, Ph.; Escourbiac, F.; Grattarola, M.; Smid, I.; Tivey, R.; Vieider, G

    2000-12-01

    After an extensive R and D activity, a medium-scale divertor vertical target prototype has been manufactured by the EU Home Team. This component contains all the main features of the corresponding ITER divertor design and consists of two units with one cooling channel each, assembled together and having an overall length and width of about 600 and 50 mm, respectively. The upper part of the prototype has a tungsten macro-brush armour, whereas the lower part is covered by CFC monoblocks. A number of joining techniques were required to manufacture this component as well as an appreciable effort in the development of suitable non-destructive testing methods. The component was high heat flux tested in FE200 electron beam facility at Le Creusot, France. It endured 100 cycles at 5 MW/m{sup 2}, 1000 cycles at 10 MW/m{sup 2} and more then 1000 cycles at 15-20 MW/m{sup 2}. The final critical heat flux test reached a value in excess of 30 MW/m{sup 2}.

  4. Beryllium mock-ups development and ultrasonic testing for ITER divertor conditions

    International Nuclear Information System (INIS)

    Barabash, V.R.; Bykov, V.A.; Giniyatulin, R.N.; Gervash, A.A.; Gurieva, T.M.; Egorov, K.E.; Komarov, V.L.; Korolkov, M.D.; Mazul, I.V.; Gitarsky, L.S.; Strulia, I.L.; Sizenev, V.S.; Pronyakin, V.T.

    1995-01-01

    At the present time beryllium is considered as the most suitable armour material for the ITER divertor application. Different types of Be-divertor mock-up construction are compared in the report. Two different technologies of beryllium tiles joining to a heat sink body are analysed: high temperature brazing and thermodiffusion bonding. The comparative analysis of different constructions has been performed on the basis of 2-D finite element calculation for temperatures and stresses. The main parameters and diagnostic capabilities of electron beam facility for HHF testing of beryllium mock-ups are described. The first results of HHF tests of ''beryllium-copper saddle-MAGT tube'' and ''beryllium-copper plate-SS body'' mock-ups are presented. The reasons of the damages during the HHF are analysed. The technique of ultrasonic testing of the thermodifussion bonding and brazing quality for beryllium-copper joints is presented. The recorded results are prepared in the form of ultrasound grams. The testing results are compared with the metallographic analysis. (orig.)

  5. 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.)

  6. Numerical simulation of CFC and tungsten target erosion in ITER-FEAT divertor

    International Nuclear Information System (INIS)

    Filatov, V.

    2003-01-01

    Physical, chemical and thermal surface erosion for water-cooled target armoured by CFC and tungsten is simulated by numerical code ERosion OF Immolated Layer (EROFIL-1). Some calculation results on the CFC and tungsten vertical target (VT) erosion in the ITER-FEAT divertor are presented for various operation modes (normal operations, slow transients, ELMs and disruptions). The main erosion mechanisms of CFC armour are the chemical and sublimation ones. Maximum erosion depth per 3000 cycles during normal operations and slow transients is of 2.7 mm at H phase and of 13.5 mm at DT phase. An evaluation of VT tungsten armour erosion per 3000 cycles of H and DT operations shows that no physical or chemical erosion as well as no melting are expected for tungsten armour at normal operations and slow transients. The tungsten armour melting at 2x10 6 ELMs is not allowable. The 300 disruptions are not dangerous in view of evaporation

  7. 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)

  8. 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)

  9. 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

  10. 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)

  11. 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

  12. Modeling of divertor particle and heat loads during application of resonant magnetic perturbation fields for ELM control in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Schmitz, O., E-mail: o.schmitz@fz-juelich.de [Forschungszentrum Jülich, IEK-4, Association EURATOM-FZJ, Jülich (Germany); Becoulet, M. [CEA/IRFM, Cadarache, 13108 St. Paul-lez-Durance Cedex (France); Cahyna, P. [IPP AS CR, Za Slovankou 3, 18200 Prague 8 (Czech Republic); Evans, T.E. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Feng, Y. [Max-Planck-Institut für Plasmaphysik, Greifswald (Germany); Frerichs, H.; Kirschner, A. [Forschungszentrum Jülich, IEK-4, Association EURATOM-FZJ, Jülich (Germany); Kukushkin, A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Laengner, R. [Forschungszentrum Jülich, IEK-4, Association EURATOM-FZJ, Jülich (Germany); Lunt, T. [Max-Planck-Institut für Plasmaphysik, Greifswald (Germany); Loarte, A.; Pitts, R. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Reiser, D.; Reiter, D. [Forschungszentrum Jülich, IEK-4, Association EURATOM-FZJ, Jülich (Germany); Saibene, G. [Fusion for Energy Joint Undertaking, Barcelona (Spain); Samm, U. [Forschungszentrum Jülich, IEK-4, Association EURATOM-FZJ, Jülich (Germany)

    2013-07-15

    First results from three-dimensional modeling of the divertor heat and particle flux pattern during application of resonant magnetic perturbation fields as ELM control scheme in ITER with the EMC3-Eirene fluid plasma and kinetic neutral transport code are discussed. The formation of a helical magnetic footprint breaks the toroidal symmetry of the heat and particle fluxes. Expansion of the flux pattern as far as 60 cm away from the unperturbed strike line is seen with vacuum RMP fields, resulting in a preferable heat flux spreading. Inclusion of plasma response reduces the radial extension of the heat and particle fluxes and results in a heat flux peaking closer to the unperturbed level. A strong reduction of the particle confinement is found. 3D flow channels are identified as a consistent reason due to direct parallel outflow from inside of the separatrix. Their radial inward expansion and hence the level of particle pump out is shown to be dependent on the perturbation level.

  13. 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

  14. Optimization for steady-state and hybrid operations of ITER by using scaling models of divertor heat load

    International Nuclear Information System (INIS)

    Murakami, Yoshiki; Itami, Kiyoshi; Sugihara, Masayoshi; Fujieda, Hirobumi.

    1992-09-01

    Steady-state and hybrid mode operations of ITER are investigated by 0-D power balance calculations assuming no radiation and charge-exchange cooling in divertor region. Operation points are optimized with respect to divertor heat load which must be reduced to the level of ignition mode (∼5 MW/m 2 ). Dependence of the divertor heat load on the variety of the models, i.e., constant-χ model, Bohm-type-χ model and JT-60U empirical scaling model, is also discussed. The divertor heat load increases linearly with the fusion power (P FUS ) in all models. The possible highest fusion power much differs for each model with an allowable divertor heat load. The heat load evaluated by constant-χ model is, for example, about 1.8 times larger than that by Bohm-type-χ model at P FUS = 750 MW. Effect of reduction of the helium accumulation, improvements of the confinement capability and the current-drive efficiency are also investigated aiming at lowering the divertor heat load. It is found that NBI power should be larger than about 60 MW to obtain a burn time longer than 2000 s. The optimized operation point, where the minimum divertor heat load is achieved, does not depend on the model and is the point with the minimum-P FUS and the maximum-P NBI . When P FUS = 690 MW and P NBI = 110 MW, the divertor heat load can be reduced to the level of ignition mode without impurity seeding if H = 2.2 is achieved. Controllability of the current-profile is also discussed. (J.P.N.)

  15. Results and analysis of high heat flux tests on a full-scale vertical target prototype of ITER divertor

    International Nuclear Information System (INIS)

    Missirlian, M.; Escourbiac, F.; Merola, M.; Bobin-Vastra, I.; Schlosser, J.; Durocher, A.

    2005-01-01

    After an extensive R and D development program, a full-scale divertor target prototype, manufactured with all the main features of the corresponding ITER divertor, was intensively tested in the high heat flux FE200 facility. The prototype consists of four units having a full monoblock geometry. The lower part (CFC armour) and the upper part (W armour) of each monoblock were joined to the solution annealed, quenched and cold worked CuCrZr tube by HIP technique. This paper summarises and analyses the main test results obtained on this prototype

  16. 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)

  17. 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)

  18. 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.

  19. 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

  20. 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

  1. Micro-/nano-characterization of the surface structures on the divertor tiles from JET ITER-like wall

    Energy Technology Data Exchange (ETDEWEB)

    Tokitani, M., E-mail: tokitani.masayuki@LHD.nifs.ac.jp [National Institute for Fusion Science, Oroshi, Toki, Gifu 509-5292 (Japan); Miyamoto, M. [Shimane University, Matsue, Shimane 690-8504 (Japan); Masuzaki, S. [National Institute for Fusion Science, Oroshi, Toki, Gifu 509-5292 (Japan); Fujii, Y. [Shimane University, Matsue, Shimane 690-8504 (Japan); Sakamoto, R. [National Institute for Fusion Science, Oroshi, Toki, Gifu 509-5292 (Japan); Oya, Y. [Shizuoka University, Shizuoka 422-8529 (Japan); Hatano, Y. [University of Toyama, Toyama 930-8555 (Japan); Otsuka, T. [Kindai University, Higashi-Osaka, Osaka, 577-8502 (Japan); Oyaidzu, M.; Kurotaki, H.; Suzuki, T.; Hamaguchi, D.; Isobe, K.; Asakura, N. [National Institute for Quantum and Radiological Science and Technology (QST), Rokkasho Aomori 039-3212 (Japan); Widdowson, A. [EUROfusion Consortium, JET, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Rubel, M. [Royal Institute of Technology (KTH), 100 44 Stockholm (Sweden)

    2017-03-15

    Highlights: • Micro-/nano-characterization of the surface structures on the divertor tiles from JET ITER-like wall were studied. • The stratified mixed-material deposition layer composed by W, C, O, Mo and Be with the thickness of ∼1.5 μm was formed on the apron of Tile 1. • The study revealed the micro- and nano-scale modification of the inner tile surface of the JET ILW. - Abstract: Micro-/nano-characterization of the surface structures on the divertor tiles used in the first campaign (2011–2012) of the JET tokamak with the ITER-like wall (JET ILW) were studied. The analyzed tiles were a single poloidal section of the tile numbers of 1, 3 and 4, i.e., upper, vertical and horizontal targets, respectively. A sample from the apron of Tile 1 was deposition-dominated. Stratified mixed-material layers composed of Be, W, Ni, O and C were deposited on the original W-coating. Their total thickness was ∼1.5 μm. By means of transmission electron microscopy, nano-size bubble-like structures with a size of more than 100 nm were identified in that layer. They could be related to deuterium retention in the layer dominated by Be. The surface microstructure of the sample from Tile 4 also showed deposition: a stratified mixed-material layer with the total thickness of 200–300 nm. The electron diffraction pattern obtained with transmission electron microscope indicated Be was included in the layer. No bubble-like structures have been identified. The surface of Tile 3, originally coated by Mo, was identified as the erosion zone. This is consistent with the fact that the strike point was often located on that tile during the plasma operation. The study revealed the micro- and nano-scale modification of the inner tile surface of the JET ILW. In particular, a complex mixed-material deposition layer could affect hydrogen isotope retention and dust formation.

  2. 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

  3. Thermal transients due to sweeping of the separatrix on the monoblock divertor concept for ITER

    International Nuclear Information System (INIS)

    Renda, V.; Papa, L.; Soria, A.

    1991-01-01

    The ITER divertor plate considered in the present study is the monoblock design option, consisting of an armour of CFC-SEP-Carb graphite tiles, crossed by the tubes of the water cooling system made in Mo-Re alloy. Preliminary steady-state calculations for a peak flux of 15 MW/m 2 showed that the allowable thickness to limit the maximum temperature to 1273 K (1000degC) is about 5 mm. This small value reduces the lifetime of the plate, due to the expected erosion rate, to an unacceptable value from the engineering standpoint. A sweeping of the separatrix has been proposed to reduce the erosion of the protective armour and to lessen the thermomechanical effects of the localized peak surface heat flux. A rotation of the null points of the separatrix of 30 mm radius with a frequency of 0.3 Hz for a surface heat flux of 15 MW/m 2 was assumed as nominal working condition. Several scenarios were considered as off-normal conditions: the loss of sweeping accident, the change in frequency from 0.3 to 0.1 Hz and the change of the peak of the surface heat flux from 15 to 30 MW/m 2 . The results related to the nominal condition show that a 16 mm thick armour could be allowed; this value should ensure an acceptable lifetime for the divertor plate. The loss of sweeping accident leads the surface temperature to reach about 2273 K in few seconds; the change in frequency raises the maximum temperature of 423 K, but its range doubles; the change in peak flux leads to a maximum temperature of about 2373 K. (author)

  4. 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.

  5. 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

  6. Experimental evaluation of brazed molybdenum-graphite bonds for the divertor of the NET/ITER nuclear fusion device

    International Nuclear Information System (INIS)

    Smid, I.; Linke, J.; Nickel, H.; Kny, E.; Reheis, N.; Kneringer, G.; Bolt, H.

    1995-01-01

    Composites consisting of plasma-facing carbon material brazed to molybdenum (TZM) substrates are a promising system for the divertor of the Next European Torus (NET) and the International Thermonuclear Experimental Reactor (ITER). Isotropic graphite and a refractory metal (molybdenum or TZM, a high temperature alloy of molybdenum), two dissimilar substrate materials, yet closely matched in their thermal expansivities, were joined with the use of four different high-temperature brazes: Zr, 90Ni-10Ti, 90Cu- 10Ti, and 70Ag-27Cu-3Ti (compositions in wt%). A summary is given of experiments on mechanical strength, heat transfer capability, structural changes, and failure modes under high heat loads of brazed bonds. Tensile-strength tests on the brazing interface prove the suitability of the brazes up to their melting point. The expected enhancement in thermal contact compared with graphite is confirmed. Passively cooled tiles of dimensions 25 mm x 25 mm were subjected to thermal cycling in electron-beam simulations. Heat fluxes of up to 10 MW m -2 were applied. (author)

  7. Experimental evaluation of brazed molybdenum-graphite bonds for the divertor of the NET/ITER nuclear fusion device

    International Nuclear Information System (INIS)

    Smid, Ivica; Linke, Jochen; Nickel, Hubertus; Kny, Erich; Reheis, Nikolaus; Kneringer, Guenther; Bolt, Harald

    1990-01-01

    Composites consisting of plasma-facing carbon material brazed to molybdenum (TZM) substrates are a promising system for the divertor of the Next European Torus (NET) and the International Thermonuclear Experimental Reactor (ITER). Isotropic graphite and a refractory metal (molybdenum or TZM, a high temperature alloy of molybdenum), two dissimilar substrate materials, yet closely matched in their thermal expansivities, were joined with the use of four different high-temperature brazes: Zr,90Ni-10Ti,90Cu-10Ti, and 70Ag-27Cu-3Ti(compositions in wt%). A summary is given of experiments on mechanical strength, heat transfer capability, structural changes, and failure modes under high heat loads of brazed bonds. Tensile-strength tests on the brazing interface prove the suitability of the brazes up to their melting point. The expected enhancement in thermal contact compared with graphite is confirmed. Passively cooled tiles of dimensions 25 mm x 25 mm were subjected to thermal cycling in electron-beam simulations. Heat fluxes of up to 10 MW m -2 were applied. (author)

  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. ITER plasma safety interface models and assessments

    International Nuclear Information System (INIS)

    Uckan, N.A.; Bartels, H-W.; Honda, T.; Amano, T.; Boucher, D.; Post, D.; Wesley, J.

    1996-01-01

    Physics models and requirements to be used as a basis for safety analysis studies are developed and physics results motivated by safety considerations are presented for the ITER design. Physics specifications are provided for enveloping plasma dynamic events for Category I (operational event), Category II (likely event), and Category III (unlikely event). A safety analysis code SAFALY has been developed to investigate plasma anomaly events. The plasma response to ex-vessel component failure and machine response to plasma transients are considered

  10. 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.)

  11. 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.

  12. Evaluation of the Erosion on the CFC tiles of the ITER Divertor by means o f FE calculations

    International Nuclear Information System (INIS)

    Schlosser, J.; Bouvet, J.; Riccardi, B.

    2007-01-01

    Full text of publication follows: The vertical target of the ITER divertor is armoured with Carbon Fibre Composite (CFC) mono-blocks in the lower part. This part is subjected to the maximum power and particles loads and, consequently, has a risk of high erosion and a significant risk of failure. In order to calculate the erosion during operation an original methodology has been developed using the CASTEM CEA finite element code. The calculation is based on a series of steady states the mesh being updated at each step of the iteration taking into account the rate of erosion between two steps. The model was developed thanks to the routines developed 10 years ago for the toroidal pump limiter of Tore Supra and takes into account shadowing effect and possible penetration of power into the gap between two mono-blocks. Both physical and chemical sputtering together with sublimation have been included in the code to describe the loss of material by the thermal and particle loads envisaged for ITER normal operation regime. This model has been validated by comparison with analytical or other code results. As erosion instability in normal operation in case of one faulty mono-block besides good ones due to the balanced rate between the various erosion mechanisms at different temperatures can be expected, coherent plasma parameters, which represent the worse cases of erosion in normal operation, have been taken into account to analyse the erosion behaviour of the mono-blocks. The aim of the study was also to evaluate the influence of a mono-block defect on erosion behaviour and the impact of these phenomena on the mono-block acceptance criteria. The calculations have pointed out the occurrence of some erosion instabilities for the studied cases (neighbour mono-block with reduced conductivity or with 90 deg. defects). Moreover it was shown that, when applying 20 MW/m 2 to the erosion model already subjected to the normal condition loads for 10,000 s, the plasma shaping of the

  13. 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)

  14. Burning plasmas in ITER for energy source

    International Nuclear Information System (INIS)

    Inoue, Nobuyuki

    2002-01-01

    Fusion research and development has two aspects. One is an academic research on science and technology, i.e., discovery and understanding of unexpected phenomena and, development of innovative technology, respectively. The other is energy source development to realize fusion as a viable energy future. Fusion research has been made remarkable progress in the past several decades, and ITER will soon realize burning plasma that is essential for both academic research and energy development. With ITER, scientific research on unknown phenomena such as self-organization of the plasma in burning state will become possible and it contributes to create a variety of academic outcome. Fusion researchers will have a responsibility to generate actual energy, and electricity generation immediately after the success of burning plasma control experiment in ITER is the next important step that has to be discussed seriously. (author)

  15. Burning plasmas in ITER for energy source

    Energy Technology Data Exchange (ETDEWEB)

    Inoue, Nobuyuki [Atomic Energy Commission, Tokyo (Japan)

    2002-10-01

    Fusion research and development has two aspects. One is an academic research on science and technology, i.e., discovery and understanding of unexpected phenomena and, development of innovative technology, respectively. The other is energy source development to realize fusion as a viable energy future. Fusion research has been made remarkable progress in the past several decades, and ITER will soon realize burning plasma that is essential for both academic research and energy development. With ITER, scientific research on unknown phenomena such as self-organization of the plasma in burning state will become possible and it contributes to create a variety of academic outcome. Fusion researchers will have a responsibility to generate actual energy, and electricity generation immediately after the success of burning plasma control experiment in ITER is the next important step that has to be discussed seriously. (author)

  16. Large Area Divertor Temperature Measurements Using A High-speed Camera With Near-infrared FiIters in NSTX

    International Nuclear Information System (INIS)

    Lyons, B.C.; Scotti, F.; Zweben, S.J.; Gray, T.K.; Hosea, J.; Kaita, R.; Kugel, H.W.; Maqueda, R.J.; McLean, A.G.; Roquemore, A.L.; Soukhanovskii, V.A.; Taylor, G.

    2011-01-01

    Fast cameras already installed on the National Spherical Torus Experiment (NSTX) have be equipped with near-infrared (NIR) filters in order to measure the surface temperature in the lower divertor region. Such a system provides a unique combination of high speed (> 50 kHz) and wide fi eld-of-view (> 50% of the divertor). Benchtop calibrations demonstrated the system's ability to measure thermal emission down to 330 oC. There is also, however, signi cant plasma light background in NSTX. Without improvements in background reduction, the current system is incapable of measuring signals below the background equivalent temperature (600 - 700 oC). Thermal signatures have been detected in cases of extreme divertor heating. It is observed that the divertor can reach temperatures around 800 oC when high harmonic fast wave (HHFW) heating is used. These temperature profiles were fi t using a simple heat diffusion code, providing a measurement of the heat flux to the divertor. Comparisons to other infrared thermography systems on NSTX are made.

  17. 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)

  18. Progress in the design, R and D and procurement preparation of the ITER Divertor Remote Handling System

    Energy Technology Data Exchange (ETDEWEB)

    Esqué, Salvador, E-mail: Salvador.Esque@f4e.europa.eu [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Hille, Carine van; Ranz, Roberto; Damiani, Carlo [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Palmer, Jim; Hamilton, David [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-lez-Durance (France)

    2014-10-15

    Highlights: •The ITER Divertor Remote Handling System (DRHS) reference design is presented. •Different R and D activities that have contributed to the development and validation of the current reference design are reported. •The DRHS turns to be a unique system in terms of complexity due to size of the to-be-handled components, the novelty of the remote operations and the operational conditions. -- Abstract: The ITER Divertor Remote Handling System (DRHS) consists of a number of dedicated remote handling equipment and tooling that will provide the means to perform the exchange of the divertor system in a full-remote way. In order to achieve this objective the DRHS will need to perform a number of novel and complex remote operations in a contaminated and space-constrained environment, in rather poor lightening conditions. Fusion for Energy has recently launched the tendering phase for the in-kind procurement of the DRHS. The procurement is based on a set of system requirements and functional specifications supported by a reference design which are presented and discussed in this paper along with the main outcomes of the different R and D activities that have contributed to the development and validation of the current reference design.

  19. Model-based radiation scalings for the ITER-like divertors of JET and ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Aho-Mantila, L., E-mail: leena.aho-mantila@vtt.fi [VTT Technical Research Centre of Finland, FI-02044 VTT (Finland); Bonnin, X. [LSPM – CNRS, Université Paris 13, Sorbonne Paris Cité, F-93430 Villetaneuse (France); Coster, D.P. [Max-Planck Institut für Plasmaphysik, D-85748 Garching (Germany); Lowry, C. [EFDA JET CSU, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Wischmeier, M. [Max-Planck Institut für Plasmaphysik, D-85748 Garching (Germany); Brezinsek, S. [Forschungszentrum Jülich, Institut für Energie- und Klimaforschung Plasmaphysik, 52425 Jülich (Germany); Federici, G. [EFDA PPP& T Department, D-85748 Garching (Germany)

    2015-08-15

    Effects of N-seeding in L-mode experiments in ASDEX Upgrade and JET are analysed numerically with the SOLPS5.0 code package. The modelling yields 3 qualitatively different radiative regimes with increasing N concentration, when initially attached outer divertor conditions are studied. The radiation pattern is observed to evolve asymmetrically, with radiation increasing first in the inner divertor, then in the outer divertor, and finally on closed field lines above the X-point. The properties of these radiative regimes are observed to be sensitive to cross-field drifts and they differ between the two devices. The modelled scaling of the divertor radiated power with the divertor neutral pressure is similar to an experimental scaling law for H-mode radiation. The same parametric dependencies are not observed in simulations without drifts.

  20. 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)

  1. A computational procedure for the investigation of whipping effect on ITER High Energy Piping and its application to the ITER divertor primary heat transfer system

    Energy Technology Data Exchange (ETDEWEB)

    Spagnuolo, G.A., E-mail: Alessandro.Spagnuolo@kraftanlagen.com [Kraftanlagen Heidelberg Gmbh, Im Breitspiel 7, D-69126 Heidelberg (Germany); Dell’Orco, G. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Di Maio, P.A. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy); Mazzei, M. [Kraftanlagen Heidelberg Gmbh, Im Breitspiel 7, D-69126 Heidelberg (Germany)

    2015-10-15

    Highlights: • High Energy Piping (HEP) are components containing water or steam with P ≥ 2.0 MPa and/or T ≥ 100 °C. • The whipping effect in HEP may cause dangerous domino effect with relative rupture propagation. • The rapture is envisaged or postulated according to the stress state of piping. • A FEM analysis is performed in order to study the dynamic of whipping effect. • Study of special support to avoid and/or mitigate the whipping effect. - Abstract: The Tokamak Cooling Water System of nuclear facility has the function to remove heat from plasma facing components maintaining coolant temperatures, pressures and flow rates as required and, depending on thermal-hydraulic requirements, its systems are defined as High Energy Piping (HEP) because they contain fluids, such as water or steam, at a pressure greater than or equal to 2.0 MPa and/or at a temperature greater than or equal to 100 °C, or even gas at pressure above the atmospheric one. The French standards contemplate the need to consider the whipping effect on HEP design. This effect happens when, after a double ended guillotine break, the reaction force could create a displacement of the piping which might affect adjacent components. A research campaign has been performed, in cooperation by ITER Organization and University of Palermo, to outline the procedure to check whether whipping effect might occur and assess its potential damage effects so to allow their mitigation. This procedure is based on the guidelines issued by U.S. Nuclear Regulatory Commission. The proposed procedure has been applied to the analysis of the whipping effect of divertor primary heat transfer system HEP, using a theoretical–computational approach based on the finite element method.

  2. 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)

  3. Manufacturing, testing and post-test examination of ITER divertor vertical target W small scale mock-ups

    International Nuclear Information System (INIS)

    Visca, Eliseo; Cacciotti, Emanuele; Komarov, Anton; Libera, Stefano; Litunovsky, Nikolay; Makhankov, Alexey; Mancini, Andrea; Merola, Mario; Pizzuto, Aldo; Riccardi, Bruno; Roccella, Selanna

    2011-01-01

    ENEA is involved in the International Thermonuclear Experimental Reactor (ITER) R and D activities. During the last years ENEA has set up and widely tested a manufacturing process, named Hot Radial Pressing (HRP), suitable for the construction of high heat flux plasma-facing components, such as the divertor targets. In the frame of the EFDA contract six mock-ups were manufactured by HRP in the ENEA labs using W monoblocks supplied by the Efremov Institute in St. Petersburg, Russian Federation and IG CuCrZr tubes. According to the technical specifications the mock-ups were examined by ultrasonic technique and after their acceptance they were delivered to the Efremov Institute TSEFEY-M e-beam facility for the thermal fatigue testing. The test consisted in 3000 cycles of 15 s heating and 15 s cooling at 10 MW/m 2 and finally 1000 cycles at 20 MW/m 2 . After the testing the ultrasonic non-destructive examination was repeated and the results compared with the investigation performed before the testing. A microstructure modification of the W monoblock material due to the overheating of the surfaces and the copper interlayer structure modification were observed in the high heat flux area. The leakage points of the mock-ups that did not conclude the testing were localized in the middle of the monoblock while they were expected between two monoblocks. This paper reports the manufacturing route, the thermal fatigue testing, the pre and post non destructive examination and finally the results of the destructive examination performed on the monoblock small scale mock-ups.

  4. 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)

  5. 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

  6. Ultrasonic test of carbon composite/copper joints in the ITER divertor

    International Nuclear Information System (INIS)

    Roccella, S.; Cacciotti, E.; Candura, D.; Mancini, A.; Pizzuto, A.; Reale, A.; Tatì, A.; Visca, E.

    2013-01-01

    Highlights: • ENEA developed and tested a specimen for the simulation of defects at the interface between CFC and copper. • The use of an ultrasonic technique properly set permitted to highlight and size with high accuracy the defects. • The technology developed could be employed successfully in the production of these components for high heat flux applications. -- Abstract: The vertical targets of the ITER divertor consist of high flux units (HFU) actively cooled: CuCrZr tubes armoured by tungsten and carbon/carbon fibre composite (CFC). The armour is obtained with holed parallelepiped blocks, called monoblocks, previously prepared and welded onto the tubes by means diffusion bonding. The monoblock preparation consists in the casting of a layer of copper oxygen free (Cu OFHC) inside the monoblock hole. Each HFU is covered with more than 100 monoblocks that have to be joined simultaneously to the tube. Therefore, it is very important to individuate any defects present in the casting of Cu OFHC or at the interface with the CFC before the monoblocks are installed on the units. This paper discusses the application of non-destructive testing by ultrasound (US) method for the control of the joining interfaces between CFC monoblocks and Cu OFHC, before the brazing on the CrCrZr tube. In ENEA laboratory an ultrasonic technique (UT) suitable for the control of these joints with size and geometry according to the ITER specifications has been developed and widely tested. Real defects in this type of joints are, however, still hardly detected by UT. The CFC surface has to be machined to improve the mechanical strength of the joint. This results in a surface not perpendicular to the ultrasonic wave. Moreover, CFC is characterized by high acoustic attenuation of the ultrasonic wave and then it is not easy to get information regarding the Cu/CFC bonding. Nevertheless, the UT sharpness and simplicity pushes to perform some further study. With this purpose, a sample with

  7. Ultrasonic test of carbon composite/copper joints in the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Roccella, S., E-mail: selanna.roccella@enea.it [Associazione ENEA-Euratom sulla Fusione, C.R. Frascati, 00044 Frascati, RM (Italy); Cacciotti, E. [Associazione ENEA-Euratom sulla Fusione, C.R. Frascati, 00044 Frascati, RM (Italy); Candura, D. [Ansaldo Nucleare S.p.A., C. so F.M. Perrone 25, 16152 Genoa (Italy); Mancini, A.; Pizzuto, A.; Reale, A. [Associazione ENEA-Euratom sulla Fusione, C.R. Frascati, 00044 Frascati, RM (Italy); Tatì, A. [Associazione Euratom-ENEA sulla Fusione, C.R. Casaccia, Via Anguillarese 301, 00123 Santa Maria di Galeria, RM (Italy); Visca, E. [Associazione ENEA-Euratom sulla Fusione, C.R. Frascati, 00044 Frascati, RM (Italy)

    2013-10-15

    Highlights: • ENEA developed and tested a specimen for the simulation of defects at the interface between CFC and copper. • The use of an ultrasonic technique properly set permitted to highlight and size with high accuracy the defects. • The technology developed could be employed successfully in the production of these components for high heat flux applications. -- Abstract: The vertical targets of the ITER divertor consist of high flux units (HFU) actively cooled: CuCrZr tubes armoured by tungsten and carbon/carbon fibre composite (CFC). The armour is obtained with holed parallelepiped blocks, called monoblocks, previously prepared and welded onto the tubes by means diffusion bonding. The monoblock preparation consists in the casting of a layer of copper oxygen free (Cu OFHC) inside the monoblock hole. Each HFU is covered with more than 100 monoblocks that have to be joined simultaneously to the tube. Therefore, it is very important to individuate any defects present in the casting of Cu OFHC or at the interface with the CFC before the monoblocks are installed on the units. This paper discusses the application of non-destructive testing by ultrasound (US) method for the control of the joining interfaces between CFC monoblocks and Cu OFHC, before the brazing on the CrCrZr tube. In ENEA laboratory an ultrasonic technique (UT) suitable for the control of these joints with size and geometry according to the ITER specifications has been developed and widely tested. Real defects in this type of joints are, however, still hardly detected by UT. The CFC surface has to be machined to improve the mechanical strength of the joint. This results in a surface not perpendicular to the ultrasonic wave. Moreover, CFC is characterized by high acoustic attenuation of the ultrasonic wave and then it is not easy to get information regarding the Cu/CFC bonding. Nevertheless, the UT sharpness and simplicity pushes to perform some further study. With this purpose, a sample with

  8. 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)

  9. Suppression of Tritium Retention in Remote Areas of ITER by Nonperturbative Reactive Gas Injection

    NARCIS (Netherlands)

    Tabares, F. L.; Ferreira, J. A.; Ramos, A.; van Rooij, G. J.; Westerhout, J.; Al, R.; Rapp, J.; Drenik, A.; Mozetic, M.

    2010-01-01

    A technique based on reactive gas injection in the afterglow region of the divertor plasma is proposed for the suppression of tritium-carbon codeposits in remote areas of ITER when operated with carbon-based divertor targets. Experiments in a divertor simulator plasma device indicate that a 4 nm/min

  10. 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.

  11. Prospective performances in JT-60SA towards the ITER and DEMO relevant plasmas

    International Nuclear Information System (INIS)

    Tamai, H.; Fujita, T.; Kikuchi, M.

    2006-01-01

    JT-60SA, the former JT-60SC and NCT, a superconducting tokamak positioned as the satellite machine of ITER, collaborating with Japan and EU fusion community, aims at contribution to ITER and DEMO through the demonstration of advanced plasma operation scenario and the plasma applicability test with advanced materials. After the discussions in JA-EU Satellite Tokamak Working Group in 2005, the increased heating power, higher heat removal capacity for the plasma facing components, improvement of the radiation shielding, the remote handling system for the maintenance of in-vessel components, and related equipment are planed to be additionally installed. With such full equipment towards the increased heating power of 41 MW (34 MW-NBI and 7 MW-ECH) with 100 s, the prospective plasma performances, analysed by the equilibrium and transport analysis codes, are rather improved in the view point of the contribution to ITER and DEMO relevant research. Accessibility for higher heating power in a higher density region enables the lower normalized Larmor radius and normalized collision frequency close to the reactor relevant plasma with the ITER-similar configuration of single null divertor plasma with the aspect ratio of A = 3.1, elongation of k95 = 1.7, triangularity of d95 (q95) in the plasma current of I p = 3.5 MA, toroidal magnetic field of B T = 2.59 T and the major radius of Rp=3.16 m. The increases in the electron temperature, beam driven and bootstrap current fraction by the increase of the power of Negative ion based NBI (10 MW) reduce the loop voltage and contribute to increase the maximum plasma current of ITER similar shape. Full non-inductive current drive controllability is also extended into the high density and high plasma current operation towards high beta plasma. Flexibility in aspect ratio and shape parameter is kept the same as NCT, i.e. a double null divertor plasma with A = 2.6, k95 = 1.83, d95 = 0.57, I p = 5.5 MA, B T = 2.72 T, and R p = 3.01 m which

  12. Results and analysis of high heat flux tests on a full scale vertical target prototype of ITER divertor

    International Nuclear Information System (INIS)

    Missirlian, M.; Escourbiac, F.; Schlosser, J.; Durocher, A.; Bobin-Vastra, I.

    2004-01-01

    After an extensive development program, a Full-Scale Divertor Target prototype (VTFS) manufactured with all the main features of the corresponding ITER divertor, was intensively tested in the high heat flux FE200 facility. The prototype consists of four units having a full mono-block geometry. The lower part (CFC armour) and the upper part (W armour) of each mono-block were joined to the solution annealed, quenched and cold worked CuCrZr tube by HIP technique. The CFC mono-block was successfully tested up to 1000 cycles at 23 MW/m 2 without any indication of failure. This value is well beyond the ITER design target of 300 cycles at 20 MW/m 2 . The W mono-block endured ∼600 cycles at 10 MW/m 2 . This value of flux is one order of magnitude higher than the ITER design target for the upper part of the vertical target. Fatigue damage is observed when pursuing the cycling up to 15 MW/m 2 . A first stress analysis seems to predict these factual results. However, macro-graphic examinations should bring a better damage valuation. Meanwhile, the fatigue testing will continue on the W healthy part of the VTFS prototype with castellation located on the heated surface (reducing the stresses close to the W-Cu interface). (authors)

  13. Results and analysis of high heat flux tests on a full scale vertical target prototype of ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Missirlian, M.; Escourbiac, F.; Schlosser, J.; Durocher, A. [Association Euratom-CEA, Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Merola, M. [EFDA Close Support Unit, Garching (Germany); Bobin-Vastra, I. [Framatome, 71 - Le Creusot (France)

    2004-07-01

    After an extensive development program, a Full-Scale Divertor Target prototype (VTFS) manufactured with all the main features of the corresponding ITER divertor, was intensively tested in the high heat flux FE200 facility. The prototype consists of four units having a full mono-block geometry. The lower part (CFC armour) and the upper part (W armour) of each mono-block were joined to the solution annealed, quenched and cold worked CuCrZr tube by HIP technique. The CFC mono-block was successfully tested up to 1000 cycles at 23 MW/m{sup 2} without any indication of failure. This value is well beyond the ITER design target of 300 cycles at 20 MW/m{sup 2}. The W mono-block endured {approx}600 cycles at 10 MW/m{sup 2}. This value of flux is one order of magnitude higher than the ITER design target for the upper part of the vertical target. Fatigue damage is observed when pursuing the cycling up to 15 MW/m{sup 2}. A first stress analysis seems to predict these factual results. However, macro-graphic examinations should bring a better damage valuation. Meanwhile, the fatigue testing will continue on the W healthy part of the VTFS prototype with castellation located on the heated surface (reducing the stresses close to the W-Cu interface). (authors)

  14. Modeling of complex gas distribution systems operating under any vacuum conditions: Simulations of the ITER divertor pumping system

    International Nuclear Information System (INIS)

    Vasileiadis, N.; Tatsios, G.; Misdanitis, S.; Valougeorgis, D.

    2016-01-01

    Highlights: • An integrated s/w for modeling complex rarefied gas distribution systems is presented. • Analysis is based on kinetic theory of gases. • Code effectiveness is demonstrated by simulating the ITER divertor pumping system. • The present s/w has the potential to support design work in large vacuum systems. - Abstract: An integrated software tool for modeling and simulation of complex gas distribution systems operating under any vacuum conditions is presented and validated. The algorithm structure includes (a) the input geometrical and operational data of the network, (b) the definition of the fundamental set of network loops and pseudoloops, (c) the formulation and solution of the mass and energy conservation equations, (d) the kinetic data base of the flow rates for channels of any length in the whole range of the Knudsen number, supporting, in an explicit manner, the solution of the conservation equations and (e) the network output data (mainly node pressures and channel flow rates/conductance). The code validity is benchmarked under rough vacuum conditions by comparison with hydrodynamic solutions in the slip regime. Then, its feasibility, effectiveness and potential are demonstrated by simulating the ITER torus vacuum system with the six direct pumps based on the 2012 design of the ITER divertor. Detailed results of the flow patterns and paths in the cassettes, in the gaps between the cassettes and along the divertor ring, as well as of the total throughput for various pumping scenarios and dome pressures are provided. A comparison with previous results available in the literature is included.

  15. 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.)

  16. 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.)

  17. 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.

  18. Operation and control of ITER plasmas

    International Nuclear Information System (INIS)

    2001-01-01

    Features incorporated in the design of the International Thermonuclear Experimental Reactor (ITER) tokamak and its ancillary and plasma diagnostic systems that will facilitate operation and control of ignited and/or high-Q DT plasmas are presented. Control methods based upon straight-forward extrapolation of techniques employed in the present generation of tokamaks are found to be adequate and effective for DT plasma control with burn durations of ≥1000 s. Examples of simulations of key plasma control functions including magnetic configuration control and fusion burn (power) control are given. The prospects for the creation and control of steady-state plasmas sustained by non-inductive current drive are also discussed. (author)

  19. Operation and control of ITER plasmas

    International Nuclear Information System (INIS)

    1999-01-01

    Features incorporated in the design of the International Thermonuclear Experimental Reactor (ITER) tokamak and its ancillary and plasma diagnostic systems that will facilitate operation and control of ignited and/or high-Q DT plasmas are presented. Control methods based upon straight-forward extrapolation of techniques employed in the present generation of tokamaks are found to be adequate and effective for DT plasma control with burn durations of ≥1000 s. Examples of simulations of key plasma control functions including magnetic configuration control and fusion burn (power) control are given. The prospects for the creation and control of steady-state plasmas sustained by non-inductive current drive are also discussed. (author)

  20. 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

  1. 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)

  2. Experimental assessment of the effects of ELMs and disruptions on ITER divertor armour materials

    International Nuclear Information System (INIS)

    Zhitlukhin, A.; Federici, G.; Giniyatulin, R.; Landman, I.; Linke, J.; Loarte, A.; Merola, M.; Podkovyrov, V.; Safronov, V.

    2005-01-01

    The response of plasma protection materials to thermal energy deposited during simulated Type I Edge Localised Modes (ELMs) and disruptions was studied. The paper describes the design and manufacture of special CFC and tungsten macrobrush targets, the experimental conditions achievable at simulating facilities and results of selected experiments. Experiments are conducted primarily under an EU/RF research collaboration in two plasma guns (QSPA and MK-200UG) located in TRINITI, Troitsk, Russia. The targets were exposed to a large number of repetitive pulses in QSPA plasma gun with heat loads varying in a range of 1-2 MJ/m 2 lasting 0.1-0.5 ms, with the purpose to determine the total expected erosion rate in ITER. MK-200UG experiments were focused on studying mainly vapour plasma production and impurity transport during ELMs. Moderate tungsten erosion less than 0.3 microns per shot was demonstrated for 1.5 MJ/m 2 energy densities. Energy density increasing up to 1.8 MJ/m 2 resulted in sharp growth of tungsten erosion, caused by intensive droplet ejection from irradiated tungsten surface. The program of further experiments is discussed. (author)

  3. 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)

  4. 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)

  5. Evaporation and Vapor Shielding of CFC Targets Exposed to Plasma Heat Fluxes Relevant to ITER ELMs

    International Nuclear Information System (INIS)

    Safronov, V.; Arkhipov, N.I.; Toporkov, D.A.; Zhitlukhin, A.M.; Landman, I.

    2007-01-01

    Full text of publication follows: Carbon-fibre composite (CFC) is foreseen presently as armour material for the divertor target in ITER. During the transient processes such as instabilities of Edge Localized Modes (ELMs) the target as anticipated will be exposed to the plasma heat loads of a few MJ/m 2 on the time scale of a fraction of ms, which causes an intense evaporation at the target surface and contaminates tokamak plasma by evaporated carbon. The ITER transient loads are not achievable at existing tokamaks therefore for testing divertor armour materials other facilities, in particular plasma guns are employed. In the present work the CFC targets have been tested for ITER at the plasma gun facility MK- 200 UG in Troitsk by ELM relevant heat fluxes. The targets in the applied magnetic field up to 2 T were irradiated by hydrogen plasma streams of diameter 6 - 8 cm, impact ion energy 2 - 3 keV, pulse duration 0.05 ms and energy density varying in the range 0.05 - 1 MJ/m 2 . Primary attention has been focused on the measurement of evaporation threshold and investigation of carbon vapor properties. Fast infrared pyrometer, optical and VUV spectrometers, framing cameras and plasma calorimeters were applied as diagnostics. The paper reports the results obtained on the evaporation threshold of CFC, the evaporation rate of the carbon fibers oriented parallel and perpendicular to the exposed target surface, the velocity of carbon vapor motion along and across the magnetic field lines, and the parameters of carbon plasma such as temperature, density and ionization state measured up to the distance 15 cm at varying plasma load. First experimental results on investigation of the vapor shield onset conditions are presented also. (authors)

  6. 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.

  7. Modelling of steady state erosion of CFC actively water-cooled mock-up for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Ogorodnikova, O.V. [Departement de Recherches sur la Fusion Controlee, Association Euratom-CEA, CEA-Cadarache, F-13108 Saint Paul Lez Durance cedex (France)], E-mail: igra32@rambler.ru

    2008-04-15

    Calculations of the physical and chemical erosion of CFC (carbon fibre composite) monoblocks as outer vertical target of the ITER divertor during normal operation regimes have been done. Off-normal events and ELM's are not considered here. For a set of components under thermal and particles loads at glancing incident angle, variations in the material properties and/or assembly of defects could result in different erosion of actively-cooled components and, thus, in temperature instabilities. Operation regimes where the temperature instability takes place are investigated. It is shown that the temperature and erosion instabilities, probably, are not a critical point for the present design of ITER vertical target if a realistic variation of material properties is assumed, namely, the difference in the thermal conductivities of the neighbouring monoblocks is 20% and the maximum allowable size of a defect between CFC armour and cooling tube is +/-90{sup o} in circumferential direction from the apex.

  8. Modelling of steady state erosion of CFC actively water-cooled mock-up for the ITER divertor

    Science.gov (United States)

    Ogorodnikova, O. V.

    2008-04-01

    Calculations of the physical and chemical erosion of CFC (carbon fibre composite) monoblocks as outer vertical target of the ITER divertor during normal operation regimes have been done. Off-normal events and ELM's are not considered here. For a set of components under thermal and particles loads at glancing incident angle, variations in the material properties and/or assembly of defects could result in different erosion of actively-cooled components and, thus, in temperature instabilities. Operation regimes where the temperature instability takes place are investigated. It is shown that the temperature and erosion instabilities, probably, are not a critical point for the present design of ITER vertical target if a realistic variation of material properties is assumed, namely, the difference in the thermal conductivities of the neighbouring monoblocks is 20% and the maximum allowable size of a defect between CFC armour and cooling tube is +/-90° in circumferential direction from the apex.

  9. Magnetic configuration control of ITER plasmas

    International Nuclear Information System (INIS)

    Albanese, R.; Mattei, M.; Portone, A.; Ambrosino, G.; Artaserse, G.; Crisanti, F.; De Tommasi, G.; Fresa, R.; Sartori, F.; Villone, F.

    2007-01-01

    The aim of this paper is to present some new tools used to review the capability of the ITER Poloidal Field (PF) system in controlling the broad range of plasma configurations presently forecasted during ITER operation. The attention is focused on the axi-symmetric aspects of plasma magnetic configuration control since they pose the greatest challenges in terms of control power and they have the largest impact on machine capital cost. Some preliminary results obtained during ongoing activities in collaboration between ENEA/CREATE and EFDA are presented. The paper is divided in two main parts devoted, respectively, to the presentation of a procedure for the PF current optimisation during the scenario, and of a software environment for the study of the PF system capabilities using the plasma linearized response. The proposed PF current optimisation procedure is then used to assess Scenario 2 design, also taking into account the presence of axisymmetric eddy currents and possible variations of poloidal beta and internal inductance. The numerical linear model based tool derived from the JET oriented eXtreme Shape Controller (XSC) tools is finally used to obtain results on the strike point sweeping in ITER

  10. 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

  11. Plasma cleaning of ITER first mirrors

    Science.gov (United States)

    Moser, L.; Marot, L.; Steiner, R.; Reichle, R.; Leipold, F.; Vorpahl, C.; Le Guern, F.; Walach, U.; Alberti, S.; Furno, I.; Yan, R.; Peng, J.; Ben Yaala, M.; Meyer, E.

    2017-12-01

    Nuclear fusion is an extremely attractive option for future generations to compete with the strong increase in energy consumption. Proper control of the fusion plasma is mandatory to reach the ambitious objectives set while preserving the machine’s integrity, which requests a large number of plasma diagnostic systems. Due to the large neutron flux expected in the International Thermonuclear Experimental Reactor (ITER), regular windows or fibre optics are unusable and were replaced by so-called metallic first mirrors (FMs) embedded in the neutron shielding, forming an optical labyrinth. Materials eroded from the first wall reactor through physical or chemical sputtering will migrate and will be deposited onto mirrors. Mirrors subject to net deposition will suffer from reflectivity losses due to the deposition of impurities. Cleaning systems of metallic FMs are required in more than 20 optical diagnostic systems in ITER. Plasma cleaning using radio frequency (RF) generated plasmas is currently being considered the most promising in situ cleaning technique. An update of recent results obtained with this technique will be presented. These include the demonstration of cleaning of several deposit types (beryllium, tungsten and beryllium proxy, i.e. aluminium) at 13.56 or 60 MHz as well as large scale cleaning (mirror size: 200 × 300 mm2). Tests under a strong magnetic field up to 3.5 T in laboratory and first experiments of RF plasma cleaning in EAST tokamak will also be discussed. A specific focus will be given on repetitive cleaning experiments performed on several FM material candidates.

  12. 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

  13. ICRF specific plasma wall interactions in JET with the ITER-like wall

    Energy Technology Data Exchange (ETDEWEB)

    Bobkov, Vl., E-mail: bobkov@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Garching (Germany); Arnoux, G. [Culham Science Centre, Association EURATOM-CCFE, Abingdon, Oxon (United Kingdom); Brezinsek, S.; Coenen, J.W. [Institute of Energy and Climate Research, Association EURATOM-FZJ (Germany); Colas, L. [CEA, IRFM, F-13108 St. Paul-lez-Durance (France); Clever, M. [Institute of Energy and Climate Research, Association EURATOM-FZJ (Germany); Czarnecka, A. [Association EURATOM-IPPLM, Hery 23, 01-497 Warsaw (Poland); Braun, F.; Dux, R. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Garching (Germany); Huber, A. [Institute of Energy and Climate Research, Association EURATOM-FZJ (Germany); Jacquet, P. [Culham Science Centre, Association EURATOM-CCFE, Abingdon, Oxon (United Kingdom); Klepper, C. [CEA, IRFM, F-13108 St. Paul-lez-Durance (France); Lerche, E. [LPP-ERM/KMS, Association Euratom-Belgian State, TEC Partners, Brussels (Belgium); Maggi, C. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Garching (Germany); Marcotte, F. [CEA, IRFM, F-13108 St. Paul-lez-Durance (France); Maslov, M.; Matthews, G.; Mayoral, M.L. [Culham Science Centre, Association EURATOM-CCFE, Abingdon, Oxon (United Kingdom); McCormick, K. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Garching (Germany); Meigs, A. [Culham Science Centre, Association EURATOM-CCFE, Abingdon, Oxon (United Kingdom); and others

    2013-07-15

    A variety of plasma wall interactions (PWIs) during operation of the so-called A2 ICRF antennas is observed in JET with the ITER-like wall. Amongst effects of the PWIs, the W content increase is the most significant, especially at low plasma densities. No increase of W source from the main divertor and entrance of the outer divertor during ICRF compared to NBI phases was found by means of spectroscopic and WI (400.9 nm) imaging diagnostics. In contrary, the W flux there is higher during NBI. Charge exchange neutrals of hydrogen isotopes could be excluded as considerable contributors to the W source. The high W content in ICRF heated limiter discharges suggests the possibility of other W sources than the divertor alone. Dependencies of PWIs to individual ICRF antennas during q{sub 95}-scans, and intensification of those for the −90° phasing, indicate a link between the PWIs and the antenna near-fields. The PWIs include heat loads and Be sputtering pattern on antenna limiters. Indications of some PWIs at the outer divertor entrance are observed which do not result in higher W flux compared to the NBI phases, but are characterized by small antenna-specific (up to 25% with respect to ohmic phases) bipolar variations of WI emission. The first TOPICA calculations show a particularity of the A2 antennas compared to the ITER antenna, due to the presence of long antenna limiters in the RF image current loop and thus high near-fields across the most part of the JET outer wall.

  14. An experimental investigation of the post-CHF enhancement factor for a prototypical ITER divertor plate with water coolant

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, T.D. [Rensselaer Polytechnic Institute, Troy, NY (United States); Watson, R.D.; McDonald, J.M. [Sandia National Lab., Albuquerque, NM (United States)] [and others

    1995-09-01

    In an off-normal event, water-cooled copper divertor plates in the International Thermonuclear Experimental Reactor (ITER) may either experience heat loads beyond their design basis, or the normal heat loads may be accompanied by low coolant pressure and velocity. The purpose of this experiment was to illustrate that during one-sided heating, as in ITER, a copper divertor plate with the proper side wall thickness, at low system pressure and velocity can absorb without failing an incident heat flux, q{sub i}, that significantly exceed the value, q{sub i}{sup CHF}, which is associated with local CHF at the wall of the coolant channel. The experiment was performed using a 30 kW electron beam test system for heating of a square cross-section divertor heat sink with a smooth circular channel of 7.63 mm diameter. The heated width, length, and wall thickness were 16, 40, and 3 mm, respectively. Stable surface temperatures were observed at incident heat fluxes greater than the local CHF point, presumably due to circumferential conduction around the thick tube walls when q{sub i}{sup CHF} was exceeded. The Post-CHF enhancement factor, {eta}, is defined as the ratio of the incident burnout heat flux, q{sub i}{sup BO}, to q{sub i}{sup CHF}. For this experiment with water at inlet conditions of 70{degrees}C, 1 m/s, and 1 MPa, q{sub i}{sup CHF} and q{sub i}{sup BO} were 600 and 1100 W/cm{sup 2}, respectively, which gave an {eta} of 1.8.

  15. 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

  16. 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.

  17. Overview of the engineering design of the ITER divertor improvements towards manufacture

    International Nuclear Information System (INIS)

    Tivey, R.; D'Agata, E.; Chuyanov, V.; Heidl, H.

    2005-01-01

    A divertor design, supported by R and D, capable of sustaining high heat loads and large electro-magnetic disturbances has been reported previously . This paper reports on design improvements that, in response to reaction from researchers and industry, focus on cost reductions, holding to a minimum the number of component variants and pursuing the establishment of workable acceptance criteria for divertor armour joints (the latter reported in ). In addition, comment from remote assembly experts has prompted improvements of the in-vessel handling and cassette to vessel attachments

  18. Bi-directional reflectance distribution function of a tungsten block for ITER divertor

    International Nuclear Information System (INIS)

    Iwamae, Atsushi; Ogawa, Hiroaki; Sugie, Tatsuo; Kusama, Yoshinori

    2012-02-01

    In order to investigate reflection properties on plasma-facing material in ITER, the bi-directional reflectance distribution function (BRDF) of a tungsten block sample has been measured. On the machining surface of the block, one-directional machining lines are engraved. Two laser diodes λ652 nm and λ473 nm were used to simulate H α and H β emissions, respectively. The reflected light is affected by the machining surface. The reflected light traces an arc when the incident light is injected in the parallel direction to the engraved line. On the other hand the reflected light traces a line shape when the incident light is injected in the perpendicular direction to the engraved lines. Ray tracing simulation qualitatively explains the experimental results. (author)

  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. Tritium analysis of divertor tiles used in JET ITER-like wall campaigns by means of β-ray induced x-ray spectrometry

    Science.gov (United States)

    Hatano, Y.; Yumizuru, K.; Koivuranta, S.; Likonen, J.; Hara, M.; Matsuyama, M.; Masuzaki, S.; Tokitani, M.; Asakura, N.; Isobe, K.; Hayashi, T.; Baron-Wiechec, A.; Widdowson, A.; contributors, JET

    2017-12-01

    Energy spectra of β-ray induced x-rays from divertor tiles used in ITER-like wall campaigns of the Joint European Torus were measured to examine tritium (T) penetration into tungsten (W) layers. The penetration depth of T evaluated from the intensity ratio of W(Lα) x-rays to W(Mα) x-rays showed clear correlation with poloidal position; the penetration depth at the upper divertor region reached several micrometers, while that at the lower divertor region was less than 500 nm. The deep penetration at the upper part was ascribed to the implantation of high energy T produced by DD fusion reactions. The poloidal distribution of total x-ray intensity indicated higher T retention in the inboard side than the outboard side of the divertor region.

  1. 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.

  2. 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)

  3. Expected energy fluxes onto ITER Plasma Facing Components during disruption thermal quenches from multi-machine data comparisons

    International Nuclear Information System (INIS)

    Loarte, A.; Andrew, P.; Matthews, G.F.; Paley, J.; Riccardo, V.; Counsell, G.; Eich, T.; Fuchs, C.; Gruber, O.; Herrmann, A.; Pautasso, G.; Federici, G.; Finken, K.H.; Maddaluno, G.; Whyte, D.

    2005-01-01

    A comparison of the power flux characteristics during the thermal quench of plasma disruptions among various tokamak experiments has been carried out and conclusions for ITER have been drawn. It is generally observed that the energy of the plasma at the thermal quench is much smaller than that of a full performance plasma. The timescales for power fluxes onto PFCs during the thermal quench, as determined by IR measurements, are found to scale with device size but not to correlate with pre-disruptive plasma characteristics. The profiles of the thermal quench power fluxes are very broad for diverted discharges, typically a factor of 5-10 broader than that measured during 'normal' plasma operation, while for limiter discharges this broadening is absent. The combination of all the above factors is used to derive the expected range of power fluxes on the ITER divertor target during the thermal quench. The new extrapolation derived in this paper indicates that the average disruption in ITER will deposit an energy flux approximately one order of magnitude lower than previously thought. The evaluation of the ITER divertor lifetime with these revised specifications is carried out. (author)

  4. 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)

  5. 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)

  6. The ITER Plasma Control System Simulation Platform

    International Nuclear Information System (INIS)

    Walker, M.L.; Ambrosino, G.; De Tommasi, G.; Humphreys, D.A.; Mattei, M.; Neu, G.; Rapson, C.J.; Raupp, G.; Treutterer, W.; Welander, A.S.; Winter, A.

    2015-01-01

    Highlights: • A development and test environment called PCSSP has been constructed for the ITER PCS. • A description of requirements and use cases, a final design and software architecture design, users guide, and a prototype implementation have been delivered. • The prototype implementation was demonstrated at IO in December of 2013. • PCSSP will be deployed for alpha testing to the IO, the development group, and selected other ITER partners upon completion of the next development phase. - Abstract: The Plasma Control System Simulation Platform (PCSSP) is a highly flexible, modular, time-dependent simulation environment developed primarily to support development of the ITER Plasma Control System (PCS). It has been under development since 2011 and is scheduled for first release to users in the ITER Organization (IO) and at selected additional sites in 2015. Modules presently implemented in PCSSP enable exploration of axisymmetric evolution and control, basic kinetic control, and tearing mode suppression. A basic capability for generation of control-relevant events is included, enabling study of exception handling in the PCS, continuous controllers, and PCS architecture. While the control design focus of PCSSP applications tends to require only a moderate level of accuracy and complexity in modules, more complex codes can be embedded or connected to access higher accuracy if needed. This paper describes the background and motivation for PCSSP, provides an overview of the capabilities, architecture, and features of PCSSP, and discusses details of the PCSSP vision and its intended goals and application. Completed work, including architectural design, prototype implementation, reference documents, and IO demonstration of PCSSP, is summarized and example use of PCSSP is illustrated. Near-term high-level objectives are summarized and include preparation for release of an “alpha” version of PCSSP and preparation for the next development phase. High

  7. The ITER Plasma Control System Simulation Platform

    Energy Technology Data Exchange (ETDEWEB)

    Walker, M.L., E-mail: walker@fusion.gat.com [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Ambrosino, G.; De Tommasi, G. [CREATE/Università di Napoli Federico II, Napoli (Italy); Humphreys, D.A. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Mattei, M. [CREATE/Seconda Università di Napoli, Napoli (Italy); Neu, G.; Rapson, C.J.; Raupp, G.; Treutterer, W. [Max Planck Institute for Plasma Physics, Boltzmannstr. 2, 85748 Garching (Germany); Welander, A.S. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Winter, A. [ITER Organization, Route de Vinon-sur-Verdon, 13115 St. Paul-lez-Durance (France)

    2015-10-15

    Highlights: • A development and test environment called PCSSP has been constructed for the ITER PCS. • A description of requirements and use cases, a final design and software architecture design, users guide, and a prototype implementation have been delivered. • The prototype implementation was demonstrated at IO in December of 2013. • PCSSP will be deployed for alpha testing to the IO, the development group, and selected other ITER partners upon completion of the next development phase. - Abstract: The Plasma Control System Simulation Platform (PCSSP) is a highly flexible, modular, time-dependent simulation environment developed primarily to support development of the ITER Plasma Control System (PCS). It has been under development since 2011 and is scheduled for first release to users in the ITER Organization (IO) and at selected additional sites in 2015. Modules presently implemented in PCSSP enable exploration of axisymmetric evolution and control, basic kinetic control, and tearing mode suppression. A basic capability for generation of control-relevant events is included, enabling study of exception handling in the PCS, continuous controllers, and PCS architecture. While the control design focus of PCSSP applications tends to require only a moderate level of accuracy and complexity in modules, more complex codes can be embedded or connected to access higher accuracy if needed. This paper describes the background and motivation for PCSSP, provides an overview of the capabilities, architecture, and features of PCSSP, and discusses details of the PCSSP vision and its intended goals and application. Completed work, including architectural design, prototype implementation, reference documents, and IO demonstration of PCSSP, is summarized and example use of PCSSP is illustrated. Near-term high-level objectives are summarized and include preparation for release of an “alpha” version of PCSSP and preparation for the next development phase. High

  8. Modelling of the edge of a fusion plasma towards ITER and experimental validation on JET

    International Nuclear Information System (INIS)

    Guillemaut, Christophe

    2013-01-01

    The conditions required for fusion can be obtained in tokamaks. In most of these machines, the plasma wall-interaction and the exhaust of heating power are handled in a cavity called divertor. However, the high heat flux involved and the limitations of the materials of the plasma facing components (PFC) are problematic. Many researches are done this field in the context of ITER which should demonstrate 500 MW of DT fusion power during ∼ 400 s. Such operations could bring the heat flux on the PFC too high to be handled. Its reduction to manageable levels relies on the divertor detachment involving the reduction of the particle and heat fluxes on the PFC. Unfortunately, this phenomenon is still difficult to model. The aim of this PhD is to use the modelling of JET experiments with EDGE2D-EIRENE to make some progress in the understanding of the detachment. The simulations reproduce the observed detachment in C and Be/W environments. The distribution of the radiation is well reproduced by the code for C but with some discrepancies in Be/W. The comparison between different sets of atomic physics processes shows that ion-molecule elastic collisions are responsible for the detachment seen in EDGE2D-EIRENE. This process provides good neutral confinement in the divertor and significant momentum losses at low temperature, when the plasma is recombining. Comparison between EDGE2D-EIRENE and SOLPS4.3 shows similar detachment trends but the importance of the ion-molecule elastic collisions is reduced in SOLPS4.3. Both codes suggest that any process capable of improving the neutral confinement in the divertor should help to improve the modelling of the detachment. (author) [fr

  9. Structural impact of armor monoblock dimensions on the failure behavior of ITER-type divertor target components: Size matters

    Energy Technology Data Exchange (ETDEWEB)

    Li, Muyuan; You, Jeong-Ha, E-mail: you@ipp.mpg.de

    2016-12-15

    Highlights: • Quantitative assessment of size effects was conducted numerically for W monoblock. • Decreasing the width of W monoblock leads to a lower risk of failure. • The Cu interlayer was not affected significantly by varying armor thickness. • The predicted trends were in line with the experimental observations. - Abstract: Plenty of high-heat-flux tests conducted on tungsten monoblock type divertor target mock-ups showed that the threshold heat flux density for cracking and fracture of tungsten armor seems to be related to the dimension of the monoblocks. Thus, quantitative assessment of such size effects is of practical importance for divertor target design. In this paper, a computational study about the thermal and structural impact of monoblock size on the plastic fatigue and fracture behavior of an ITER-type tungsten divertor target is reported. As dimensional parameters, the width and thickness of monoblock, the thickness of sacrificial armor, and the inner diameter of cooling tube were varied. Plastic fatigue lifetime was estimated for the loading surface of tungsten armor and the copper interlayer by use of a cyclic-plastic constitutive model. The driving force of brittle crack growth through the tungsten armor was assessed in terms of J-integral at the crack tip. Decrease of the monoblock width effectively reduced accumulation of plastic strain at the armor surface and the driving force of brittle cracking. Decrease of sacrificial armor thickness led to decrease of plastic deformation at the loading surface due to lower surface temperature, but the thermal and mechanical response of the copper interlayer was not affected by the variation of armor thickness. Monoblock with a smaller tube diameter but with the same armor thickness and shoulder thickness experienced lower fatigue load. The predicted trends were in line with the experimental observations.

  10. 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

  11. 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

  12. High heat flux thermal-hydraulic analysis of ITER divertor and blanket systems

    International Nuclear Information System (INIS)

    Raffray, A.R.; Chiocchio, S.; Ioki, K.; Tivey, R.; Krassovski, D.; Kubik, D.

    1998-01-01

    Three separate cooling systems are used for the divertor and blanket components, based mainly on flow routing access and on grouping together components with the highest heat load levels and uncertainties: divertor, limiter/outboard baffle, and primary first wall/inboard baffle. The coolant parameters for these systems are set to accommodate peak heat load conditions with a reasonable critical heat flux (CHF) margin. Material temperature constraints and heat transport system space and cost requirements are also taken into consideration. This paper summarises the three cooling system designs and highlights the high heat flux thermal-hydraulic analysis carried out in converging on the design values for the coolant operating parameters. Application of results from on-going high heat flux R and D and a brief description of future R and D effort to address remaining issues are also included. (orig.)

  13. 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

  14. 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

  15. 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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  17. Simulation of burning plasma dynamics in ITER

    International Nuclear Information System (INIS)

    Wang, J.F.; Amano, T.; Ogawa, Y.; Inoue, N.

    1996-02-01

    Dynamics of burning plasma for various transient situations in ITER plasma has been simulated with a 1.5-dimensional up-down asymmetry Tokamak Transport Simulation Code (TTSC). We have mainly paid attention to intrinsic plasma transport processes such as the confinement improvement and the change of plasma profiles. It is shown that a large excursion of the fusion power takes place with a small improvement of the plasma confinement; e.g., an increase of the global energy confinement by a factor of 1.22 yields the fusion power excursion of ∼ 30% within a few seconds. Any feedback control of fueling D-T gas is difficult to respond to this short time scale of fusion power transient. The effect of the plasma profile on the fusion power excursion has been studied, by changing the particle transport denoted by the inward pinch parameter C V . It is found that the fusion power excursion is mild and slow, and the feedback control is quite effective in suppressing the fusion power excursion and in shortening the duration time of power transient in this case. The change in the pumping efficiency has also been studied and a large excursion of the fusion power has not been observed, because of the decrease in the fuel density itself in the case of the increase in the pumping efficiency, and the helium ash accumulation in the case of the decrease in the pumping efficiency. Finally it is shown that the MHD sawteeth activity leads to the fusion power fluctuation of ± 20%, although it is helpful for the helium ash exhaust. (author)

  18. Overview of the preliminary design of the ITER plasma control system

    Science.gov (United States)

    Snipes, J. A.; Albanese, R.; Ambrosino, G.; Ambrosino, R.; Amoskov, V.; Blanken, T. C.; Bremond, S.; Cinque, M.; de Tommasi, G.; de Vries, P. C.; Eidietis, N.; Felici, F.; Felton, R.; Ferron, J.; Formisano, A.; Gribov, Y.; Hosokawa, M.; Hyatt, A.; Humphreys, D.; Jackson, G.; Kavin, A.; Khayrutdinov, R.; Kim, D.; Kim, S. H.; Konovalov, S.; Lamzin, E.; Lehnen, M.; Lukash, V.; Lomas, P.; Mattei, M.; Mineev, A.; Moreau, P.; Neu, G.; Nouailletas, R.; Pautasso, G.; Pironti, A.; Rapson, C.; Raupp, G.; Ravensbergen, T.; Rimini, F.; Schneider, M.; Travere, J.-M.; Treutterer, W.; Villone, F.; Walker, M.; Welander, A.; Winter, A.; Zabeo, L.

    2017-12-01

    An overview of the preliminary design of the ITER plasma control system (PCS) is described here, which focusses on the needs for 1st plasma and early plasma operation in hydrogen/helium (H/He) up to a plasma current of 15 MA with moderate auxiliary heating power in low confinement mode (L-mode). Candidate control schemes for basic magnetic control, including divertor operation and kinetic control of the electron density with gas puffing and pellet injection, were developed. Commissioning of the auxiliary heating systems is included as well as support functions for stray field topology and real-time plasma boundary reconstruction. Initial exception handling schemes for faults of essential plant systems and for disruption protection were developed. The PCS architecture was also developed to be capable of handling basic control for early commissioning and the advanced control functions that will be needed for future high performance operation. A plasma control simulator is also being developed to test and validate control schemes. To handle the complexity of the ITER PCS, a systems engineering approach has been adopted with the development of a plasma control database to keep track of all control requirements.

  19. Beryllium layer response to ITER-like ELM plasma pulses in QSPA-Be

    Directory of Open Access Journals (Sweden)

    N.S. Klimov

    2017-08-01

    Full Text Available Material migration in ITER is expected to move beryllium (Be eroded from the first wall primarily to the tungsten (W divertor region and to magnetically shadowed areas of the wall itself. This paper is concerned with experimental study of Be layer response to ELM-like plasma pulses using the new QSPA-Be plasma gun (SRC RF TRINITI. The Be layers (1→50µm thick are deposited on special castellated Be and W targets supplied by the ITER Organization using the Thermionic Vacuum Arc technique. Transient deuterium plasma pulses with duration ∼0.5ms were selected to provide absorbed energy densities on the plasma stream axis for a 30° target inclination of 0.2 and 0.5MJm−2, the first well below and the second near the Be melting point. This latter value is close to the prescribed maximum energy density for controlled ELMs on ITER. At 0.2MJm−2 on W, all Be layer thicknesses tested retain their integrity up to the maximum pulse number, except at local defects (flakes, holes and cracks and on tile edges. At 0.5MJm−2 on W, Be layer melting and melt layer agglomeration are the main damage processes, they happen immediately in the first plasma impact. Melt layer movement was observed only near plasma facing edges. No significant melt splashing is observed in spite of high plasma pressure (higher than expected in ITER. Be layer of 10µm thick on Be target has higher resistance to plasma irradiation than 1 and 55µm, and retain their integrity up to the maximum pulse number at 0.2MJm−2. For 1µm and 55µm thick on Be target significant Be layer losses were observed at 0.2MJm−2.

  20. ITER operational space for full plasma current H-mode operation

    Energy Technology Data Exchange (ETDEWEB)

    Mattei, M. [Assoc. Euratom-ENEA-CREATE, Seconda University di Napoli, Aversa (Italy)], E-mail: massimiliano.mattei@unirc.it; Cavinato, M.; Saibene, G.; Portone, A. [Fusion for Energy Joint Undertaking, 08019 Barcelona (Spain); Albanese, R.; Ambrosino, G. [Assoc. Euratom-ENEA-CREATE, University Napoli Federico II, Napoli (Italy); Horton, L.D. [Max Planck-Institut fur Plasmaphysik, EURATOM-Association, Garching (Germany); Kessel, C. [Princeton Plasma Physics Laboratory, Princeton University (United States); Koechl, F. [Assoc. EURATOM-OAW/ATI, Vienna (Austria); Lomas, P.J. [Euratom/UKAEA Fusion Assoc., Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Nunes, I. [Assoc. EURATOM/IST, Centro de Fusao Nuclear, Lisbon (Portugal); Parail, V. [Max Planck-Institut fur Plasmaphysik, EURATOM-Association, Garching (Germany); Sartori, R. [Fusion for Energy Joint Undertaking, 08019 Barcelona (Spain); Sips, A.C.C. [Max Planck-Institut fur Plasmaphysik, EURATOM-Association, Garching (Germany); Thomas, P.R. [Fusion for Energy Joint Undertaking, 08019 Barcelona (Spain)

    2009-06-15

    Sensitivity studies performed as part of the ITER IO design review highlighted a very stiff dependence of the maximum Q attainable on the machine parameters. In particular, in the considered range, the achievable Q scales with I{sub p}{sup 4}. As a consequence, the achievement of the ITER objective of Q = 10 requires the machine to be routinely operated at a nominal current I{sub p} of 15 MA, and at full toroidal field BT of 5.3 T. This paper analyses the capabilities of the poloidal field (PF) system (including the central solenoid) of ITER against realistic full current plasma scenarios. An exploration of the ITER operational space for the 15 and 17 MA inductive scenario is carried out. An extensive analysis includes the evaluation of margins for the closed loop shape control action. The overall results of this analysis indicate that the control of a 15 MA plasma in ITER is likely to be adequate in the range of li 0.7-0.9 whereas, for a 17 MA plasma, control capabilities are strongly reduced. The ITER operational space, provided by the reference pre-2008 PF system, was rather limited if compared to the range of parameters normally observed in present experiment. Proposals for increasing the current and field limits on PF2, PF5 and PF6, adjustment on the number of turns in some of the PF coils, changes to the divertor dome geometry, to the conductor of PF6 to Nb3Sn, moving PF6 radially and/or vertically are described and evaluated in the paper. Some of them have been included in 2008 ITER revised configuration.

  1. Plasma scram in ITER L-mode ignited plasmas

    International Nuclear Information System (INIS)

    Villar Colome, J.; Johner, J.; Ane, J.M.

    1995-01-01

    The security of ITER will depend on the capability of the system in rapidly extinguishing the 1.5 GW of nominal fusion power without disruption. The local RLW transport model is used to simulate such a Plasma Scram. The conditions for a passively secure operation point in steady-state are discussed in terms of particle exhaust. The time scales of the process should determine the power supplies of both equilibrium coils and central solenoid. (authors). 6 refs., 4 figs., 2 tabs

  2. 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.

  3. Development of a radiative divertor for DIII-D

    International Nuclear Information System (INIS)

    Allen, S.L.; Brooks, N.H.; Campbell, R.B.; Fenstermacher, M.E.; Hill, D.N.; Hyatt, A.W.; Knoll, D.; Lasnier, C.J.; Lazarus, E.A.; Leonard, A.W.; Lippmann, S.I.; Mahdavi, M.A.; Maingi, R.; Meyer, W.; Moyer, R.A.; Petrie, T.W.; Porter, G.D.; Rensink, M.E.; Rognlien, T.D.; Schaffer, M.J.; Smith, J.P.; Staebler, G.M.; Stambaugh, R.D.; West, W.P.; Wood, R.D.

    1995-01-01

    We have used experiments and modeling to develop a new radiative divertor configuration for DIII-D. Gas puffing experiments with the existing open divertor have shown the creation of a localized ( similar 10 cm diameter) radiation zone which results in substantial reduction (3-10) in the divertor heat flux while τ E remains similar 2 times ITER-89P scaling. However, n e increases with D 2 puffing, and Z eff increases with neon puffing. Divertor structures are required to minimize the effects on the core plasma. The UEDGE fluid code, benchmarked with DIII-D data, and the DEGAS neutrals transport code are used to estimate the effectiveness of divertor configurations; slots reduce the core ionization more than baffles. The overall divertor shape is set by confinement studies which indicate that high triangularity (δ∼0.8) is important for high τ E VH-modes. Results from engineering feasibility studies, including diagnostic access, will be presented. ((orig.))

  4. Analysis of first wall and divertor cooling loop failures for the ITER plant

    International Nuclear Information System (INIS)

    Eriksson, J.; Sjoberg, A.; Collen, J.

    1998-01-01

    In this study the capability of the in-vessel heat transfer systems to maintain sufficiently low structure temperatures during certain events have been investigated. The findings are that in the case of blackout PWF/IBB structure temperatures remain low enough not to jeopardize the integrity. In an event of divertor pump trip generally lower copper temperatures are achieved and opening of the pressurizer safety valve is avoided if a prolonged pump coasting down period is selected. However, an adequate minimum thickness of protective CFC armour is still crucial for maintaining structure integrity. (authors)

  5. Status of technology R&D for the ITER tungsten divertor monoblock

    Energy Technology Data Exchange (ETDEWEB)

    Hirai, T., E-mail: takeshi.hirai@iter.org [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Escourbiac, F.; Barabash, V.; Durocher, A.; Fedosov, A.; Ferrand, L.; Jokinen, T.; Komarov, V.; Merola, M.; Carpentier-Chouchana, S. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Arkhipov, N. [Project Center ITER, 1, Building 3, Kurchatov Sq., 123182 Moscow (Russian Federation); Kuznetcov, V.; Volodin, A. [NIIEFA, 3 doroga na Metallostroy, Metallostroy, St. Petersburg 196641 (Russian Federation); Suzuki, S.; Ezato, K.; Seki, Y. [JAEA, Fusion Research and Development Directorate JAEA, 801-1 Mukouyama, Naka, Ibaragi 311-0193 (Japan); Riccardi, B.; Bednarek, M.; Gavila, P. [F4E, c/Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain)

    2015-08-15

    In order to develop and validate the high performance tungsten monoblock technology, the full-tungsten divertor qualification program was defined. As the first step, small-scale mock-ups were manufactured and successfully tested under the required high heat flux loads. The test results demonstrated that the technology is available in Japan and Europe. Post-tests observation of the loaded W monoblocks showed generation of self-castellation – a crack along coolant tube axis. The cause of the self-castellation was discussed and a tungsten material characterization program is being developed with the objective to understand mechanical properties that influence the occurrence of the self-castellation.

  6. Modelling of the material transport and layer formation in the divertor of JET: Comparison of ITER-like wall with full carbon wall conditions

    International Nuclear Information System (INIS)

    Kirschner, A.; Matveev, D.; Borodin, D.; Airila, M.; Brezinsek, S.; Groth, M.; Wiesen, S.; Widdowson, A.; Beal, J.; Esser, H.G.; Likonen, J.; Bekris, N.; Ding, R.

    2015-01-01

    Impurity transport within the inner JET divertor has been modelled with ERO to estimate the transport to and the resulting deposition at remote areas. Various parametric studies involving divertor plasma conditions and strike point position have been performed. In JET-ILW (beryllium main chamber and tungsten divertor) beryllium, flowing from the main chamber into the divertor and then effectively reflected at the tungsten divertor tiles, is transported to remote areas. The tungsten flux to remote areas in L-Mode is in comparison to the beryllium flux negligible due to small sputtering. However, tungsten is sputtered during ELMs in H-Mode conditions. Nevertheless, depending on the plasma conditions, strike point position and the location of the remote area, the maximum resulting tungsten flux to remote areas is at least ∼3 times lower than the corresponding beryllium flux. Modelled beryllium and tungsten deposition on a rotating collector probe located below tile 5 is in good agreement with measurements if the beryllium influx into the inner divertor is assumed to be in the range of 0.1% relative to the deuterium ion flux and erosion due to fast charge exchange neutrals is considered. Comparison between JET-ILW and JET-C is presented

  7. Modelling of the material transport and layer formation in the divertor of JET: Comparison of ITER-like wall with full carbon wall conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kirschner, A., E-mail: a.kirschner@fz-juelich.de [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, Trilateral Euregio Cluster, 52425 Jülich (Germany); Matveev, D.; Borodin, D. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, Trilateral Euregio Cluster, 52425 Jülich (Germany); Airila, M. [VTT Technical Research Centre of Finland, 02044 VTT (Finland); Brezinsek, S. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, Trilateral Euregio Cluster, 52425 Jülich (Germany); Groth, M. [Aalto University, Otakaari 4, 02015 Espoo (Finland); Wiesen, S. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, Trilateral Euregio Cluster, 52425 Jülich (Germany); Widdowson, A. [Culham Centre for Fusion Energy, Abingdon OX14 3DB (United Kingdom); Beal, J. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Esser, H.G. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, Trilateral Euregio Cluster, 52425 Jülich (Germany); Likonen, J. [VTT Technical Research Centre of Finland, 02044 VTT (Finland); Bekris, N. [Karlsruhe Institute of Technology, Institute for Technical Physics, Hermann-von-Helmholtz-Platz 1, Bau 451, 76344 Eggenstein-Leopoldshafen (Germany); Ding, R. [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Hefei, Anhui 230031 (China)

    2015-08-15

    Impurity transport within the inner JET divertor has been modelled with ERO to estimate the transport to and the resulting deposition at remote areas. Various parametric studies involving divertor plasma conditions and strike point position have been performed. In JET-ILW (beryllium main chamber and tungsten divertor) beryllium, flowing from the main chamber into the divertor and then effectively reflected at the tungsten divertor tiles, is transported to remote areas. The tungsten flux to remote areas in L-Mode is in comparison to the beryllium flux negligible due to small sputtering. However, tungsten is sputtered during ELMs in H-Mode conditions. Nevertheless, depending on the plasma conditions, strike point position and the location of the remote area, the maximum resulting tungsten flux to remote areas is at least ∼3 times lower than the corresponding beryllium flux. Modelled beryllium and tungsten deposition on a rotating collector probe located below tile 5 is in good agreement with measurements if the beryllium influx into the inner divertor is assumed to be in the range of 0.1% relative to the deuterium ion flux and erosion due to fast charge exchange neutrals is considered. Comparison between JET-ILW and JET-C is presented.

  8. 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

  9. 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.

  10. An introduction to boundary plasma physics

    International Nuclear Information System (INIS)

    Shimizu, Katsuhiro; Takizuka, Tomonori

    2004-01-01

    History of tokamak experiments is briefly reviewed with a special focus on divertors. Two-point divertor model, which calculates plasma parameters up-stream and at the divertor plate for a given condition of particle flux and heat flux, is explained. The model is applied to ITER to discuss the heat flux onto the target plate. The important issues of divertor physics related to recycling, remote radiative cooling, detached plasma and MARFE are also introduced. (author)

  11. 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.

  12. The plasma position control of ITER EDA plasma

    International Nuclear Information System (INIS)

    Senda, Ikuo; Nishio, Satoshi; Tsunematsu, Toshihide; Nishino, Toru; Fujieda, Hirobumi.

    1994-09-01

    The study on the plasma position control of ITER EDA performed by Japan Home Team during the sensitivity study in 1994 is summarized. The controllabilities of plasmas in the Outline Design and elongated version are compared. The model used to describe the motion of the plasma is a rigid model. The PD feedback control is applied with respect to the displacements of the plasma from the equilibrium. Three types of fluctuations, which initiate the motion of the plasma, are examined, namely a finite horizontal fluctuation field, a small horizontal fluctuation field such that the motion of the plasma is governed by the passive structures and an abrupt change of the poloidal beta β p and internal inductance l i . In the simulations of finite horizontal fluctuation fields, controls depend on the strength of the fluctuations, for instance, 3-5V is needed for 5-10G of fluctuation fields in the Outline Design. When the fluctuation field is small and the plasma displacement grows in a characteristic time of the passive structures, a few volt of the control voltage is enough to obtain good controllability. It is shown that the control when (β p , l i ) changes simultaneously is demanding and a large control voltage is required to maintain satisfactory control. Comparing the elongated version with the Outline Design, the control voltage which is larger than the Outline Design by a factor of 2-3 is required to obtain the same controllability in the elongated version. (author)

  13. 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)

  14. 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

  15. 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)

  16. 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

  17. 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)

  18. 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.

  19. Thermal and radiation loads on the first wall and divertor plates in the KTM tokamak

    International Nuclear Information System (INIS)

    Azizov, Eh.A.; Buzhinskij, O.I.; Gladush, G.G.; Darmagraj, V.V.; Priyampol'skij, I.R.; Dvorkin, N.Ya.; Lejkin, I.N.; Tazhibaeva, I.L.; Shestakov, V.P.

    2001-01-01

    The constructing of the KTM tokamak is intended for wide scale studies of behavior both inner-chamber element materials and structures (first wall, limiters, divertor, hf-antennas, etc.) under conditions approaching to the ITER-FEAT and a future thermonuclear reactors. The KTM tokamak is designed for maintain of interaction conditions of plasma-wall, plasma flows and divertor field, stimulating conditions of ITER-FEAT; and for examination of a future tokamaks' materials. In the work the thermal loads on the first wall, divertor plates are presented

  20. Detailed Particle and Power Fluxes Into ITER Castellated Divertor Gaps During ELMs

    Czech Academy of Sciences Publication Activity Database

    Dejarnac, Renaud; Komm, M.; Tskhakaya, D.; Gunn, J. P.; Pekarek, Z.

    2010-01-01

    Roč. 38, č. 4 (2010), s. 1042-1046 ISSN 0093-3813 Institutional research plan: CEZ:AV0Z20430508 Keywords : Edge modeling * Ion-surface interactions * ITER * Sheaths Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.070, year: 2010

  1. 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

  2. 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.)

  3. 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

  4. 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)

  5. Thermal fatigue testing of a diffusion-bonded beryllium divertor mock-up under ITER relevant conditions

    International Nuclear Information System (INIS)

    Youchison, D.L.; Guiniiatouline, R.; Watson, R.D.

    1994-01-01

    Thermal response and thermal fatigue tests of four 5 mm thick beryllium tiles on a Russian divertor mock-up were completed on the Electron Beam Test System at Sandia National Laboratories. The beryllium tiles were diffusion bonded onto an OFHC copper saddleblock and a DSCu (MAGT) tube containing a porous coating. Thermal response tests were performed on the tiles to an absorbed heat flux of 5 MW/m 2 and surface temperatures near 300 degrees C using 1.4 MPa water at 5.0 m/s flow velocity and an inlet temperature of 8-15 degrees C. One tile was exposed to incrementally increasing heat fluxes up to 9.5 MW/m 2 and surface temperatures up to 690 degrees C before debonding at 10 MW/m 2 . A third tile debonded after 9200 thermal fatigue cycles at 5 MW/m 2 , while another debonded after 6800 cycles. In all cases, fatigue failure occurred in the intermetallic layers between the beryllium and copper. No fatigue cracking of the bulk beryllium was observed. During thermal cycling, a gradual loss of porous coating produced increasing sample temperatures. These experiments indicate that diffusion-bonded beryllium tiles can survive several thousand thermal cycles under ITER relevant conditions without failure. However, the reliability of the diffusion bonded Joint remains a serious issue

  6. 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

  7. Divertor radiation in the ASDEX upgrade tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Sehmer, Till; Bernert, Matthias; Koll, Juergen; Meister, Hans; Wischmeier, Marco; Fantz, Ursel [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2, 85748 Garching (Germany); Reimold, Felix [Forschungszentrum Juelich GmbH, Institut fuer Energie- und Klimaforschung - Plasmaphysik, 52425 Juelich (Germany); Collaboration: The ASDEX Upgrade Team

    2016-07-01

    To reduce in ITER the expected power flux density onto the divertor target, the plasma-wall interaction in the divertor needs to be strongly reduced. The fundamental path to achieve this is using radiation from seeded impurities, whereas the localization of this radiation (e.g. inside/outside confined region), which could have an impact onto the power balance, is a key challenge. The absolute radiated power distribution can be measured by foil bolometers. To study at the ASDEX Upgrade tungsten divertor the localization and quantification of radiation, the respective line of sight density of the bolometers has been improved by two additional cameras. The divertor radiation enhanced by nitrogen (N{sub 2}) seeding has been investigated, using variations of (1) the external heating power or (2) the N{sub 2} seeding rate. While in both cases the inner divertor stays fully detached, measurements indicate that the region of dominant radiation moves from the inner divertor through the X-Point into the confined region. In the outer divertor however, the measurements indicate either an immediate upwards shift or a continuous movement of the radiation away from the target, depending on experimental conditions.

  8. JET with a pumped divertor -- Technical issues and main results

    International Nuclear Information System (INIS)

    Bertolini, E.

    1995-01-01

    The most recent modification to JET has been the installation of a single-null pumped divertor, for active control of plasma impurities. This is to address central physics issues relevant to the design of a next step tokamak. Experiments conducted during the 1994--95 campaign, with plasma currents up to 6MA, have shown that the Mark I divertor, which makes use of strike point sweeping across the target plates, is a suitable tool to control the influx of impurities in the plasma core. The operation of a tokamak with a pumped divertor has been characterized in detail. However the divertor configuration must be optimized to better meet ITER requirements. Therefore an improved (more closed) divertor structure, which may not require sweeping, is under assembly at present (Mark II). It is designed, in addition, to allow divertor tile structures to be fully replaceable by remote handling techniques, following D-T fusion experiments. New types of events involving electromechanical interactions of plasma with the vessel and in-vessel structural components have been encountered, due to plasma vertical instabilities and disruptions (such as toroidal asymmetries of vacuum vessel forces and side-ways vessel displacements). The physics and engineering experimental work performed in JET is primarily dedicated to the finalization of the ITER design

  9. 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)

  10. 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.

  11. 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)

  12. 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

  13. 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

  14. 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.

  15. Variation of particle exhaust with changes in divertor magnetic balance

    International Nuclear Information System (INIS)

    Petrie, T.W.; Allen, S.L.; Brooks, N.H.

    2006-01-01

    Recent experiments on DIII-D point to the importance of two factors in determining how effectively the deuterium particle inventory in a tokamak plasma can be controlled through pumping at the divertor target(s): (1) the divertor magnetic balance, i.e. the degree to which the divertor topology is single-null or double-null (DN) and (2) the direction of the of B x ∇B ion drift with respect to the X-point(s). Changes in divertor magnetic balance near the DN shape have a much stronger effect on the particle exhaust rate at the inner divertor target(s) than on the particle exhaust rate at the outer divertor target(s). The particle exhaust rate for the DN shape is strongest at the outer strike point opposite the B x ∇B ion particle drift direction. Our data suggests that the presence of B x ∇B and E x B ion particle drifts in the scrape-off layer and divertor(s) play an important role in the particle exhaust rates of DN and near-DN plasmas. Particle exhaust rates are shown to depend strongly on the edge (pedestal) density. These results have implications for particle control in ITER and other future tokamaks

  16. MAGNUM-PSI, a plasma generator for plasma-surface interaction research in ITER-like conditions

    International Nuclear Information System (INIS)

    Goedheer, W.J.; Rooij, G.J. van; Veremiyenko, V.; Ahmad, Z.; Barth, C.J.; Eck, H.J.N. van; Groot, B. de; Hellermann, M.G. von; Kruijtzer, G.L.; Wolff, J.C.; Brezinsek, S.; Philipps, V.; Pospieszczyk, A.; Samm, U.; Schweer, B.; Dahiya, R.P.; Engeln, R.A.H.; Schram, D.C.; Fantz, U.; Kleyn, A.W.; Lopes Cardozo, N.J.

    2005-01-01

    The FOM-Institute for Plasma Physics - together with its TEC partners - is preparing the construction of Magnum-psi, a magnetized (3 T), steady-state, large area (100 cm 2 ), high-flux (up to 10 24 H + ions m -2 s -1 ) plasma generator. The research programme of Magnum-psi will address the questions for the ITER divertor: erosion, redeposition and hydrogen retention with carbon substrates, melting of metal surfaces, erosion and redeposition with mixed materials. In order to explore and develop the techniques to be applied in Magnum-psi, a pilot experiment (Pilot-psi), operating at a magnetic field up to 1.6 Tesla, has been constructed. Pilot-psi produces a hydrogen plasma beam with the required parameters (T e ≤ 1eV and flux ≥ 10 23 m -2 s -1 ) over an area of 1 cm 2 . In this paper the results of extensive diagnostic measurements on Pilot-psi (a.o., Thomson Scattering and high-resolution spectroscopy), combined with numerical studies of the source and the expansion of the plasma will be presented to demonstrate the feasibility of the large Magnum-psi plasma generator. (author)

  17. 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

  18. Developing a Roadmap for US Divertor and PMI Research in the ITER Era

    Science.gov (United States)

    Hill, D. N.; Lipschultz, B.; Whyte, D. G.; Garofalo, A. M.; Leonard, A. W.; Maingi, R.

    2013-10-01

    The role of existing and candidate future facilities for developing driven core, boundary plasma and plasma-facing components (PFCs) solutions for burning plasma experiments will be discussed in light of scientific and technical challenges, testing capabilities, scheduling implications, and cost. Present experiments point to likely integrated core-edge solutions which may enable steady-state high-gain, high power density operation; focused research on existing tokamak facilities could strengthen confidence significantly. In parallel, both existing and new candidate materials suitable for testing under high neutron fluence can be developed and qualified. We will also discuss the potential role of new facilities in closing the knowledge gaps to a Fusion Nuclear Science Facility (FNSF), and what form the final step of integrating core and edge solutions will be (separate, or as part of an FNSF) in terms of size, goals and cost. Supported by the US DOE under DE-AC52-07NA27344, DE-FC02-99ER54512, DE-SC00-02060, DE-FG02-04ER54762, DE-FC-02-04ER54698, and DE-AC02-09CH11466.

  19. 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

  20. 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.

  1. 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

  2. Impact of the plasma response in three-dimensional edge plasma transport modelling for RMP ELM control scenarios at ITER

    Science.gov (United States)

    Schmitz, Oliver

    2014-10-01

    The constrains used in magneto-hydrodynamic (MHD) modeling of the plasma response to external resonant magnetic perturbation (RMP) fields have a profound impact on the three-dimensional (3-D) shape of the plasma boundary induced by RMP fields. In this contribution, the consequences of the plasma response on the actual 3D boundary structure and transport during RMP application at ITER are investigated. The 3D fluid plasma and kinetic neutral transport code EMC3-Eirene is used for edge transport modeling. Plasma response modeling is conducted with the M3D-C1 code using a single fluid, non-linear and a two fluid, linear MHD constrain. These approaches are compared to results with an ideal MHD like plasma response. A 3D plasma boundary is formed for all cases consisting of magnetic finger structures at the X-point intersecting the divertor surface in a helical footprint pattern. The width of the helical footprint pattern is largely reduced compared to vacuum magnetic fields when using the ideal MHD like screening model. This yields increasing peak heat fluxes in contrast to a beneficial heat flux spreading seen with vacuum fields. The particle pump out as well as loss of thermal energy is reduced by a factor of two compared to vacuum fields. In contrast, the impact of the plasma response obtained from both MHD constrains in M3D-C1 is nearly negligible at the plasma boundary and only a small modification of the magnetic footprint topology is detected. Accordingly, heat and particle fluxes on the target plates as well as the edge transport characteristics are comparable to the vacuum solution. This span of modeling results with different plasma response models highlights the importance of thoroughly validating both, plasma response and 3D edge transport models for a robust extrapolation towards ITER. Supported by ITER Grant IO/CT/11/4300000497 and F4E Grant GRT-055 (PMS-PE) and by Start-Up Funds of the University of Wisconsin - Madison.

  3. ELM mitigation with pellet ELM triggering and implications for PFCs and plasma performance in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Baylor, Larry R. [ORNL; Lang, P. [EURATOM / UKAEA, Abingdon, UK; Allen, S. L. [Lawrence Livermore National Laboratory (LLNL); Lasnier, C. J. [Lawrence Livermore National Laboratory (LLNL); Meitner, Steven J. [ORNL; Combs, Stephen Kirk [ORNL; Commaux, Nicolas JC [ORNL; Loarte, A. [ITER Organization, Cadarache, France; Jernigan, Thomas C. [ORNL

    2015-08-01

    The triggering of rapid small edge localized modes (ELMs) by high frequency pellet injection has been proposed as a method to prevent large naturally occurring ELMs that can erode the ITER plasma facing components (PFCs). Deuterium pellet injection has been used to successfully demonstrate the on-demand triggering of edge localized modes (ELMs) at much higher rates and with much smaller intensity than natural ELMs. The proposed hypothesis for the triggering mechanism of ELMs by pellets is the local pressure perturbation resulting from reheating of the pellet cloud that can exceed the local high-n ballooning mode threshold where the pellet is injected. Nonlinear MHD simulations of the pellet ELM triggering show destabilization of high-n ballooning modes by such a local pressure perturbation.A review of the recent pellet ELM triggering results from ASDEX Upgrade (AUG), DIII-D, and JET reveals that a number of uncertainties about this ELM mitigation technique still remain. These include the heat flux impact pattern on the divertor and wall from pellet triggered and natural ELMs, the necessary pellet size and injection location to reliably trigger ELMs, and the level of fueling to be expected from ELM triggering pellets and synergy with larger fueling pellets. The implications of these issues for pellet ELM mitigation in ITER and its impact on the PFCs are presented along with the design features of the pellet injection system for ITER.

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  5. ITER jako živý

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan

    2010-01-01

    Roč. 10, č. 6 (2010), s. 18-19 Institutional research plan: CEZ:AV0Z20430508 Keywords : Fusion * ITER * magnetic field * ELMs * cryo pumps * central solenoid * correction coils * superconducting coils * toroidal field coils * poloidal field coils * divertor * Cadarache Subject RIV: BL - Plasma and Gas Discharge Physics http://www.tretipol.cz/900-iter-jako-zivy

  6. 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.)

  7. 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)

  8. 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.

  9. The ITER poloidal field system

    Energy Technology Data Exchange (ETDEWEB)

    Wesley, J [General Atomics, San Diego, CA (USA); Beljakov, V; Kavin, A; Korshakov, V; Kostenko, A; Roshal, A; Zakharov, L [Kurchatov Inst. of Atomic Energy, Moscow (USSR); Bulmer, R; Kaiser, T; Miller, J R; Pearlstein, L D [Lawrence Livermore National Lab., CA (USA); Hogan, J [Oak Ridge National Lab., TN (USA); Kurihara, K; Shimomura, Y; Sugihara, M; Yoshino, R [Japan Atomic Energy Resea

    1990-12-15

    The ITER poloidal field (PF) system uses superconducting coils to provide the plasma equilibrium fields, slow equilibrium control and plasma flux linkage (V-s) needed for the ITER Operations and Research Program. Double-null (DN) divertor plasmas and operation scenarios for 22 MA Physics (high-Q/ignition) and 15 MA Technology (high-fluence testing) phases are provided. For 22 MA plasmas, total PF flux swing is 333 V-s. This provides inductive current drive (CD) for start-up with 66 V-s of resistive loss and 440-s (330-s minimum) sustained burn. The PF system also allows plasma start-up and shutdown scenarios, and can maintain the plasma configuration during burn over a range of current and pressure profiles. Other capabilities include increased plasma current (25 MA with inductive CD; 28 MA with non-inductive CD assist), divertor separatrix sweeping, and semi-DN and single-null plasmas.

  10. Multi-Level iterative methods in computational plasma physics

    International Nuclear Information System (INIS)

    Knoll, D.A.; Barnes, D.C.; Brackbill, J.U.; Chacon, L.; Lapenta, G.

    1999-01-01

    Plasma physics phenomena occur on a wide range of spatial scales and on a wide range of time scales. When attempting to model plasma physics problems numerically the authors are inevitably faced with the need for both fine spatial resolution (fine grids) and implicit time integration methods. Fine grids can tax the efficiency of iterative methods and large time steps can challenge the robustness of iterative methods. To meet these challenges they are developing a hybrid approach where multigrid methods are used as preconditioners to Krylov subspace based iterative methods such as conjugate gradients or GMRES. For nonlinear problems they apply multigrid preconditioning to a matrix-few Newton-GMRES method. Results are presented for application of these multilevel iterative methods to the field solves in implicit moment method PIC, multidimensional nonlinear Fokker-Planck problems, and their initial efforts in particle MHD

  11. Passive shut-down of ITER plasma by Be evaporation

    International Nuclear Information System (INIS)

    Amano, Tsuneo.

    1996-02-01

    In an accident event where the cooling system of first wall of the ITER fails, the first wall temperature continues to rise as long as the ignited state of the core plasma persists. In this paper, a passive shut-down scheme of the ITER from this accident by evaporated Be from the first wall is examined. It is shown the estimated Be influx 5 10 24 /sec is sufficient to quench the ignition. (author)

  12. 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)

  13. High Flux FRC Facility for the Stability, Confinement and ITER Divertor Studies

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, Alan L. [Univ. of Washington, Seattle, WA (United States). Aerospace and Energetics Research Program. Redmond Plasma Physics Lab.; Milroy, Richard D. [Univ. of Washington, Seattle, WA (United States). Aerospace and Energetics Research Program. Redmond Plasma Physics Lab.

    2014-01-31

    The TCS (Translation, Confinement, & Sustainment) program was begun on 7 August, 1996 to renew basic studies of the Field Reversed Configuration (FRC). The program made use of the old LSX (Large s Experiment) device, which was constructed at STI during the period from 1986 to 1990, but only operated for one year due to a DOE decision at the time to focus exclusively on the tokamak configuration. LSX was transferred to the University of Washington in 1992 and modified (LSX/mod) to perform Tokamak Refueling by Accelerated Plasmoids (TRAP) experiments. The TRAP program was funded from 7 August, 1992 until 6 August, 1996, but was utilized for an additional year while TCS was being constructed. During the first TCS funding period TCS was completed and initial experiments were begun. A large multi-megawatt RF power supply was built by Los Alamos National Laboratory (LANL) for use with a Rotating Magnetic Field (RMF) system, and LANL has been a continuing participant in our experimental program. A smaller prototype facility, called the Star Thrust Experiment (STX) was also built and operated in this period, partly with NASA funding, before TCS came on-line. A final report for this construction period was submitted in September 2000. A first renewal period (2.5 years) provided operating funds for the period between July 7, 2000 and January 6, 2003. A great deal of progress was made in understanding the use of RMF to both form and sustain FRCs during this period. The principal result of the experimental program was the formation of quasi steady-state (as long as RMF power was available) FRCs with densities in the 1-3x1019 m-3 range. However, the plasma temperature (Te or Ti) was limited to sub-25 eV, except transiently during start-up, by the rapid accumulation of impurities. This is not surprising since TCS was only designed to demonstrate RMF flux build-up and was not provided with either fueling capabilities or modern vacuum

  14. 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)

  15. 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.

  16. 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.

  17. 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

  18. 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.)

  19. 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.

  20. Numerical simulation of plasma vertical position stabilization in ITER

    International Nuclear Information System (INIS)

    Astapkovich, A.M.; Sadakov, S.N.

    1992-01-01

    The paper deals with numerical simulation of plasma vertical position stabilization in ITER. The calculations are performed using EDDY C-2 code by the method of direct numerical simulation of transient electromagnetic processes taking into account the evolution of plasma position, cross-section shape and full plasma current. When simulating free vertical plasma drift in ITER with twin passive stabilization loops, it was shown that account of the effects of cross-section deformation and plasma current alternations results in almost two fold degradation of passive stabilization parameters as compared to the calculations for 'rigid displacement' model. In terms of methodology, the account of the effects of cross section deformation and plasma current alternations requires clarification of the definitions for reverse increment of vertical instability and for stability margin coefficient. The simulation of plasma pinch return to equilibrium position after the closure of control coils allows to assess the required parameters of active control system and demonstrate the effect of screen current reverse in twin loops. The obtained results were used to develop the ITER conceptual design and affected the choice of the concept of twin passive loops and new positron of control coils as the basis approaches. 11 refs.; 12 figs.; 1 tab

  1. Repetitive plasma loads typical for ITER Type-I ELMS; simulation in QSPA Kh-50

    International Nuclear Information System (INIS)

    Tereshin, V.I.; Bandura, A.N.; Byrka, O.V.; Chebotarev, V.V.; Garkusha, I.E.; Makhlaj, V.A.; Solyakov, D.G.; Tsarenko, A.V.; Landman, I.

    2005-01-01

    The power loads on current tokamaks associated with the Type I ELMs generally do not affect the lifetime of divertor elements. However, the ITER ELMs may lead to unacceptable lifetime; their loads are estimated as QELM(1-3) MJ/m 2 at t = 0.1-1 ms and the repetition frequency of an order of 1 Hz (∼ 400 ELMs during each ITER pulse). Such plasma energy loads expected for ITER ELMs are not achieved in existing tokamaks. Therefore powerful plasma accelerators are used at present for study of plasma-target interaction and for numerical models validation. Quasi-steady-state plasma accelerators (QSPA), which characterized by essentially longer duration of plasma stream generation in comparison with pulsed plasma guns, became especially attractive facilities for investigations of plasma-surface interaction in conditions of high heat loads simulating the ITER disruptions and ELMs. The paper presents experimental study of energy characteristics of the plasma streams generated with quasi-steady-state plasma accelerator QSPA Kh-50 and the main features of plasma interaction with material surfaces in dependence on plasma heat loads. The samples of pure sintered tungsten of EU trademark have been exposed to hydrogen plasma streams produced by the accelerator. To estimate the range of tolerable loads the effects of ELMs on the lifetime of plasma facing components have been experimentally simulated for large numbers of impacts with varying energy density. The experiments were performed with up to 450 pulses of the duration of 0.25 ms and the heat loads in the range of 0.5 - 1.2 MJ/m 2 . At this calorimetry (both at plasma stream and at the target surface), piezo-detectors as well as spectroscopy and interferometry measurements were applied to determine the impacting plasma parameters in different regimes of operation. A threshold character of morphological changes on the tungsten surface under the melting in respect to the pulses number is demonstrated. The number of initial

  2. 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)

  3. Investigation of erosion mechanisms and erosion products in divertor armour materials under conditions relevant to elms and mitigated disruptions in ITER

    International Nuclear Information System (INIS)

    Safronov, V.M.; Arkhipov, N.I.; Klimov, N.S.; Kovalenko, D.V.; Moskaleva, A.A.; Podkovyrov, V.L.; Toporkov, D.A.; Zhitlukhin, A.M.; Landman, I.S.; Poznyak, I.M.

    2008-01-01

    Carbon fibre composite (CFC) and tungsten were irradiated by intense plasma streams at plasma gun facilities MK-200UG and QSPA-T. The targets were tested by plasma loads relevant to Edge Localised Modes (ELM) and mitigated disruptions in ITER. Onset condition of material erosion and properties of erosion products have been studied

  4. 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.

  5. 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.

  6. 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.

  7. 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)

  8. 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

  9. 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

  10. 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

  11. Implementation strategy for the ITER plasma control system

    International Nuclear Information System (INIS)

    Winter, A.; Ambrosino, G.; Bauvir, B.; De Tommasi, G.; Humphreys, D.A.; Mattei, M.; Neto, A.; Raupp, G.; Snipes, J.A.; Stephen, A.V.; Treutterer, W.; Walker, M.L.; Zabeo, L.

    2015-01-01

    This paper gives an overview of the scope and context of the CODAC high-level real-time applications (Supervision and Plasma Control) and presents the strategy and current state of design of the tools to support the implementation. A real-time framework, which is currently under development with strong support of the worldwide fusion community will not only support the implementation of plasma control strategies with the extensive exception handling and forecasting functionality foreseen for ITER, but also integrated commissioning, orchestration and supervision as well as the real-time needs of ITER plant system developers. A second cornerstone in the implementation strategy is the development of a powerful simulation environment (Plasma Control System Simulation Platform – PCSSP) to design and verify control strategies, event handling and orchestration and automation. The development of PCSSP is currently under contract and this paper will also give an overview of its current state of development.

  12. Implementation strategy for the ITER plasma control system

    Energy Technology Data Exchange (ETDEWEB)

    Winter, A., E-mail: axel.winter@iter.org [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Ambrosino, G. [CREATE/Università di Napoli Federico II, Dip. Ingegneria Elettrica e delle Tecnologie dell’Informazione (Italy); Bauvir, B. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); De Tommasi, G. [CREATE/Università di Napoli Federico II, Dip. Ingegneria Elettrica e delle Tecnologie dell’Informazione (Italy); Humphreys, D.A. [General Atomics, San Diego, CA (United States); Mattei, M. [CREATE/Seconda Università di Napoli, Dip. Ingegneria Industriale e dell’Informazione (Italy); Neto, A. [Fusion for Energy, Barcelona (Spain); Raupp, G. [Max Planck Institute for Plasma Physics, EURATOM Association, Garching (Germany); Snipes, J.A. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Stephen, A.V. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon (United Kingdom); Treutterer, W. [Max Planck Institute for Plasma Physics, EURATOM Association, Garching (Germany); Walker, M.L. [General Atomics, San Diego, CA (United States); Zabeo, L. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

    2015-10-15

    This paper gives an overview of the scope and context of the CODAC high-level real-time applications (Supervision and Plasma Control) and presents the strategy and current state of design of the tools to support the implementation. A real-time framework, which is currently under development with strong support of the worldwide fusion community will not only support the implementation of plasma control strategies with the extensive exception handling and forecasting functionality foreseen for ITER, but also integrated commissioning, orchestration and supervision as well as the real-time needs of ITER plant system developers. A second cornerstone in the implementation strategy is the development of a powerful simulation environment (Plasma Control System Simulation Platform – PCSSP) to design and verify control strategies, event handling and orchestration and automation. The development of PCSSP is currently under contract and this paper will also give an overview of its current state of development.

  13. Time parallelization of advanced operation scenario simulations of ITER plasma

    International Nuclear Information System (INIS)

    Samaddar, D; Casper, T A; Kim, S H; Houlberg, W A; Berry, L A; Elwasif, W R; Batchelor, D

    2013-01-01

    This work demonstrates that simulations of advanced burning plasma operation scenarios can be successfully parallelized in time using the parareal algorithm. CORSICA -an advanced operation scenario code for tokamak plasmas is used as a test case. This is a unique application since the parareal algorithm has so far been applied to relatively much simpler systems except for the case of turbulence. In the present application, a computational gain of an order of magnitude has been achieved which is extremely promising. A successful implementation of the Parareal algorithm to codes like CORSICA ushers in the possibility of time efficient simulations of ITER plasmas.

  14. 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

  15. 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

  16. 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)

  17. 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.

  18. 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.

  19. 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.

  20. 'EU divertor celebration day'

    International Nuclear Information System (INIS)

    Merola, M.

    2002-01-01

    The meeting 'EU divertor celebration day' organized on 16 January 2002 at Plansee AG, Reutte, Austria was held on the occasion of the completion of manufacturing activities of a complete set of near full-scale prototypes of divertor components including the vertical target, the dome liner and the cassette body. About 30 participants attended the meeting including Dr. Robert Aymar, ITER Director, representatives from EFDA, CEA, ENEA, IPP and others

  1. Turbulent Simulations of Divertor Detachment Based On BOUT + + Framework

    Science.gov (United States)

    Chen, Bin; Xu, Xueqiao; Xia, Tianyang; Ye, Minyou

    2015-11-01

    China Fusion Engineering Testing Reactor is under conceptual design, acting as a bridge between ITER and DEMO. The detached divertor operation offers great promise for a reduction of heat flux onto divertor target plates for acceptable erosion. Therefore, a density scan is performed via an increase of D2 gas puffing rates in the range of 0 . 0 ~ 5 . 0 ×1023s-1 by using the B2-Eirene/SOLPS 5.0 code package to study the heat flux control and impurity screening property. As the density increases, it shows a gradually change of the divertor operation status, from low-recycling regime to high-recycling regime and finally to detachment. Significant radiation loss inside the confined plasma in the divertor region during detachment leads to strong parallel density and temperature gradients. Based on the SOLPS simulations, BOUT + + simulations will be presented to investigate the stability and turbulent transport under divertor plasma detachment, particularly the strong parallel gradient driven instabilities and enhanced plasma turbulence to spread heat flux over larger surface areas. The correlation between outer mid-plane and divertor turbulence and the related transport will be analyzed. Prepared by LLNL under Contract DE-AC52-07NA27344. LLNL-ABS-675075.

  2. 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

  3. 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.

  4. Architectural concept for the ITER Plasma Control System

    Energy Technology Data Exchange (ETDEWEB)

    Treutterer, W., E-mail: Wolfgang.Treutterer@ipp.mpg.de [Max-Planck Institute for Plasma Physics, EURATOM Association, Garching (Germany); Humphreys, D., E-mail: humphreys@fusion.gat.com [General Atomics, San Diego, CA (United States); Raupp, G., E-mail: Gerhard.Raupp@ipp.mpg.de [Max-Planck Institute for Plasma Physics, EURATOM Association, Garching (Germany); Schuster, E., E-mail: schuster@lehigh.edu [Lehigh University, Bethlehem, PA (United States); Snipes, J., E-mail: Joseph.Snipes@iter.org [ITER Organization, 13115 St. Paul-lez-Durance (France); De Tommasi, G., E-mail: detommas@unina.it [CREATE/Università di Napoli Federico II, Napoli (Italy); Walker, M., E-mail: walker@fusion.gat.com [General Atomics, San Diego, CA (United States); Winter, A., E-mail: Axel.Winter@iter.org [ITER Organization, 13115 St. Paul-lez-Durance (France)

    2014-05-15

    The plasma control system is a key instrument for successfully investigating the physics of burning plasma at ITER. It has the task to execute an experimental plan, known as pulse schedule, in the presence of complex relationships between plasma parameters like temperature, pressure, confinement and shape. The biggest challenge in the design of the control system is to find an adequate breakdown of this task in a hierarchy of feedback control functions. But it is also important to foresee structures that allow handling unplanned exceptional situations to protect the machine. Also the management of the limited number of actuator systems for multiple targets is an aspect with a strong impact on system architecture. Finally, the control system must be flexible and reconfigurable to cover the manifold facets of plasma behaviour and investigation goals. In order to prepare the development of a control system for ITER plasma operation, a conceptual design has been proposed by a group of worldwide experts and reviewed by an ITER panel in 2012. In this paper we describe the fundamental principles of the proposed control system architecture and how they were derived from a systematic collection and analysis of use cases and requirements. The experience and best practices from many fusion devices and research laboratories, augmented by the envisaged ITER specific tasks, build the foundation of this collection. In the next step control functions were distilled from this input. An analysis of the relationships between the functions allowed sequential and parallel structures, alternate branches and conflicting requirements to be identified. Finally, a concept of selectable control layers consisting of nested “compact controllers” was synthesised. Each control layer represents a cascaded scheme from high-level to elementary controllers and implements a control hierarchy. The compact controllers are used to resolve conflicts when several control functions would use the same

  5. Architectural concept for the ITER Plasma Control System

    International Nuclear Information System (INIS)

    Treutterer, W.; Humphreys, D.; Raupp, G.; Schuster, E.; Snipes, J.; De Tommasi, G.; Walker, M.; Winter, A.

    2014-01-01

    The plasma control system is a key instrument for successfully investigating the physics of burning plasma at ITER. It has the task to execute an experimental plan, known as pulse schedule, in the presence of complex relationships between plasma parameters like temperature, pressure, confinement and shape. The biggest challenge in the design of the control system is to find an adequate breakdown of this task in a hierarchy of feedback control functions. But it is also important to foresee structures that allow handling unplanned exceptional situations to protect the machine. Also the management of the limited number of actuator systems for multiple targets is an aspect with a strong impact on system architecture. Finally, the control system must be flexible and reconfigurable to cover the manifold facets of plasma behaviour and investigation goals. In order to prepare the development of a control system for ITER plasma operation, a conceptual design has been proposed by a group of worldwide experts and reviewed by an ITER panel in 2012. In this paper we describe the fundamental principles of the proposed control system architecture and how they were derived from a systematic collection and analysis of use cases and requirements. The experience and best practices from many fusion devices and research laboratories, augmented by the envisaged ITER specific tasks, build the foundation of this collection. In the next step control functions were distilled from this input. An analysis of the relationships between the functions allowed sequential and parallel structures, alternate branches and conflicting requirements to be identified. Finally, a concept of selectable control layers consisting of nested “compact controllers” was synthesised. Each control layer represents a cascaded scheme from high-level to elementary controllers and implements a control hierarchy. The compact controllers are used to resolve conflicts when several control functions would use the same

  6. 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

  7. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-15

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

  9. Robustness of radiative mantle plasma power exhaust solutions for ITER

    International Nuclear Information System (INIS)

    Mandrekas, J.; Stacey, W.M.; Kelly, F.A.

    1997-01-01

    The robustness of impurity-seeded radiative mantle solutions for ITER to uncertainties in several physics and operating parameters is examined. The results indicate that ∼ 50--90% of the input power can be radiated from inside the separatrix with Ne, Ar and Kr injection, without significant detriment to the core power balance or collapse of the edge temperature profile, for a wide range of conditions on the impurity pinch velocity, edge temperature pedestal, and plasma density

  10. 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.)

  11. 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.

  12. 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

  13. Fusion Plasma Physics and ITER - An Introduction (1/4)

    CERN Multimedia

    CERN. Geneva

    2011-01-01

    In November 2006, ministers representing the world’s major fusion research communities signed the agreement formally establishing the international project ITER. Sited at Cadarache in France, the project involves China, the European Union (including Switzerland), India, Japan, the Russian Federation, South Korea and the United States. ITER is a critical step in the development of fusion energy: its role is to confirm the feasibility of exploiting magnetic confinement fusion for the production of energy for peaceful purposes by providing an integrated demonstration of the physics and technology required for a fusion power plant. The ITER tokamak is designed to study the “burning plasma” regime in deuterium-tritium (D-T) plasmas by achieving a fusion amplification factor, Q (the ratio of fusion output power to plasma heating input power), of 10 for several hundreds of seconds with a nominal fusion power output of 500MW. It is also intended to allow the study of steady-state plasma operation at Q≥5 by me...

  14. Erosion of beryllium under ITER - Relevant transient plasma loads

    Science.gov (United States)

    Kupriyanov, I. B.; Nikolaev, G. N.; Kurbatova, L. A.; Porezanov, N. P.; Podkovyrov, V. L.; Muzichenko, A. D.; Zhitlukhin, A. M.; Gervash, A. A.; Safronov, V. M.

    2015-08-01

    Beryllium will be used as a armor material for the ITER first wall. It is expected that erosion of beryllium under transient plasma loads such as the edge-localized modes (ELMs) and disruptions will mainly determine a lifetime of the ITER first wall. This paper presents the results of recent experiments with the Russian beryllium of TGP-56FW ITER grade on QSPA-Be plasma gun facility. The Be/CuCrZr mock-ups were exposed to up to 100 shots by deuterium plasma streams (5 cm in diameter) with pulse duration of 0.5 ms and heat loads range of 0.2-0.5 MJ/m2 at different temperature of beryllium tiles. The temperature of Be tiles has been maintained about 250 and 500 °C during the experiments. After 10, 40 and 100 shots, the beryllium mass loss/gain under erosion process were investigated as well as evolution of surface microstructure and cracks morphology.

  15. 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)

  16. Experimental modelling of plasma-graphite surface interaction in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Martynenko, Yu.V.; Guseva, M.I.; Gureev, V.M.; Danelyan, L.S.; Neumoin, V.E.; Petrov, V.B.; Khripunov, B.I.; Sokolov, Yu.A.; Stativkina, O.V.; Stolyarova, V.G. [Rossijskij Nauchnyj Tsentr ``Kurchatovskij Inst.``, Moscow (Russian Federation); Vasiliev, V.I.; Strunnikov, V.M. [TRINITI, Troizk (Russian Federation)

    1998-10-01

    The investigation of graphite erosion under normal operation ITER regime and disruption was performed by means of exposure of RGT graphite samples in a stationary deuterium plasma to a dose of 10{sup 22} cm{sup -2} and subsequent irradiation by power (250 MW/cm{sup 2}) pulse deuterium plasma flow imitating disruption. The stationary plasma exposure was carried out in the installation LENTA with the energy of deuterium ions being 200 eV at target temperatures of 770 C and 1150 C. The preliminary exposure in stationary plasma at temperature of physical sputtering does not essentially change the erosion due to a disruption, whereas exposure at the temperature of radiation enhanced sublimation dramatically increases the erosion due to disruption. In the latter case, the depth of erosion due to a disruption is determined by the depth of a layer with decreased strength. (orig.) 9 refs.

  17. The dynamical mechanical properties of tungsten under compression at working temperature range of divertors

    Science.gov (United States)

    Zhu, C. C.; Song, Y. T.; Peng, X. B.; Wei, Y. P.; Mao, X.; Li, W. X.; Qian, X. Y.

    2016-02-01

    In the divertor structure of ITER and EAST with mono-block module, tungsten plays not only a role of armor material but also a role of structural material, because electromagnetic (EM) impact will be exerted on tungsten components in VDEs or CQ. The EM loads can reach to 100 MN, which would cause high strain rates. In addition, directly exposed to high-temperature plasma, the temperature regime of divertor components is complex. Aiming at studying dynamical response of tungsten divertors under EM loads, an experiment on tungsten employed in EAST divertors was performed using a Kolsky bar system. The testing strain rates and temperatures is derived from actual working conditions, which makes the constitutive equation concluded by using John-Cook model and testing data very accurate and practical. The work would give a guidance to estimate the dynamical response, fatigue life and damage evolution of tungsten divertor components under EM impact loads.

  18. 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.

  19. '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.

  20. "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.

  1. 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

  2. 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

  3. 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.

  4. 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 wa