WorldWideScience

Sample records for be-cu divertor modules

  1. Pre-irradiation testing of actively cooled Be-Cu divertor modules

    Energy Technology Data Exchange (ETDEWEB)

    Linke, J.; Duwe, R.; Kuehnlein, W. [Forschungszentrum Juelich GmbH (Germany)] [and others

    1995-09-01

    A set of neutron irradiation tests is prepared on different plasma facing materials (PFM) candidates and miniaturized components for ITER. Beside beryllium the irradiation program which will be performed in the High Flux Reactor (HFR) in Petten, includes different carbon fiber composites (CFQ) and tungsten alloys. The target values for the neutron irradiation will be 0.5 dpa at temperatures of 350{degrees}C and 700{degrees}C, resp.. The post irradiation examination (PIE) will cover a wide range of mechanical tests; in addition the degradation of thermal conductivity will be investigated. To determine the high heat flux (HHF) performance of actively cooled divertor modules, electron beam tests which simulate the expected heat loads during the operation of ITER, are scheduled in the hot cell electron beam facility JUDITH. These tests on a selection of different actively cooled beryllium-copper and CFC-copper divertor modules are performed before and after neutron irradiation; the pre-irradiation testing is an essential part of the program to quantify the zero-fluence high heat flux performance and to detect defects in the modules, in particular in the brazed joints.

  2. Impurity radiation modulations in an ergodic divertor

    Energy Technology Data Exchange (ETDEWEB)

    Laugier, F. E-mail: laugier@pegase.cad.cea.fr; Becoulet, M.; De Michelis, C.; Ghendrih, Ph.; Gunn, J.P.; Monier-Garbet, P.; Reichle, R.; Vallet, J.C

    2001-03-01

    The 3-D geometry of radiation losses is investigated in the Tore Supra ergodic divertor. Measurements from passive bolometers located on the divertor coils show evidence of toroidal and poloidal radiation modulations. They were interpreted using a 3-D code solving heat transport equation that gives the whole geometry of plasma radiation in a divertor configuration close to Tore Supra. The results of the code are in qualitative agreement with the measurements and they show that the total radiated power is underestimated when inferred from standard bolometers located between divertor modules. Maximum of radiation in front of the modules is explained by the multiplication of radiative zones at this place due to the intersection of field lines with the vessel wall. This effect leads to non-monotonic temperature profiles along field lines in the boundary plasma.

  3. Impurity radiation modulations in an ergodic divertor

    International Nuclear Information System (INIS)

    The 3-D geometry of radiation losses is investigated in the Tore Supra ergodic divertor. Measurements from passive bolometers located on the divertor coils show evidence of toroidal and poloidal radiation modulations. They were interpreted using a 3-D code solving heat transport equation that gives the whole geometry of plasma radiation in a divertor configuration close to Tore Supra. The results of the code are in qualitative agreement with the measurements and they show that the total radiated power is underestimated when inferred from standard bolometers located between divertor modules. Maximum of radiation in front of the modules is explained by the multiplication of radiative zones at this place due to the intersection of field lines with the vessel wall. This effect leads to non-monotonic temperature profiles along field lines in the boundary plasma

  4. Research proposal on : amplitude modulated reflectometry system for JET divertor

    International Nuclear Information System (INIS)

    Amplitude Modulated reflectometry is presented here as a tool for density profile measurements in the JET divertor plasmas. One of the main problems which has been presented in most reflectometers during the last years is the need for a coherent tracking of the phase delay: fast density fluctuations and strong modulation on the amplitude of the reflected signal usually bring to fringe jumps' in the phase signal, which are a big problem when the phase values are much larger than 2 pi. The conditions in the JET divertor plasmas: plasma geometry, access and long oversized broad-band waveguide paths makes very difficult the phase measurements at the millimeter wave range. AM reflectometry is to some extension an intermediate solution between the classical phase delay reflectometry, so far applied to small distances, and the time domain reflectometry, used for ionospheric studies and recently also proposed for fusion plasma. the main advantage is to allow the use of millimeter wave reflectometry with moderate phase shifts (approx 2 pi). (author)

  5. Research proposal on: amplitude modulated reflectometry system for the JET divertor

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, J.; Branas, B.; Estrada, T.; Luna, E. de la

    1992-07-01

    Amplitude Modulated reflectometry is presented here as a tool for density profile measurements in the JET divertor plasmas. One of the main problems which has been present in most reflectometers during the last years is the need for a coherent tracking of the phase delay: fast density fluctuations and strong modulation on the amplitude of the reflected signal usually bring to fringe jumps in the phase signal, which are a big problem when the phase values are much larger than 2{pi} The conditions in the JET divertor plasmas: plasma geometry, access and long oversized broad- band waveguide paths makes very difficult the phase measurements at the millimeter wave range. AM reflectometry is to some extension an intermediate solution between the classical phase delay reflectometry, so far applied to small distances, and the time domain reflectometry, used for onospheric studies and recently also proposed for fusion plasmas. The main advantage is to allow the use of millimeter wave reflectometry with moderate phase shifts ( {approx} 2{pi} ). (Author) 2 refs.

  6. Spectroscopic observation of temperature and density modulations in the boundary layer during ergodic divertor operation in the Tore Supra tokamak

    International Nuclear Information System (INIS)

    Spatially resolved spectroscopic measurements of carbon impurity ion line brightness profiles, in ergodic divertor (ED) Tore Supra tokamak plasmas, have shown poloidal electron temperature and density modulations in the peripheral ED layer. These effects have been qualitatively reproduced by using a field line tracing code. (author). 14 refs., 3 figs

  7. 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.

  8. Interoperability of remote handling control system software modules at Divertor Test Platform 2 using middleware

    Energy Technology Data Exchange (ETDEWEB)

    Tuominen, Janne, E-mail: janne.m.tuominen@tut.fi [Tampere University of Technology, Department of Intelligent Hydraulics and Automation, Tampere (Finland); Rasi, Teemu; Mattila, Jouni [Tampere University of Technology, Department of Intelligent Hydraulics and Automation, Tampere (Finland); Siuko, Mikko [VTT, Technical Research Centre of Finland, Tampere (Finland); Esque, Salvador [F4E, Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla2, 08019, Barcelona (Spain); Hamilton, David [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: ► The prototype DTP2 remote handling control system is a heterogeneous collection of subsystems, each realizing a functional area of responsibility. ► Middleware provides well-known, reusable solutions to problems, such as heterogeneity, interoperability, security and dependability. ► A middleware solution was selected and integrated with the DTP2 RH control system. The middleware was successfully used to integrate all relevant subsystems and functionality was demonstrated. -- Abstract: This paper focuses on the inter-subsystem communication channels in a prototype distributed remote handling control system at Divertor Test Platform 2 (DTP2). The subsystems are responsible for specific tasks and, over the years, their development has been carried out using various platforms and programming languages. The communication channels between subsystems have different priorities, e.g. very high messaging rate and deterministic timing or high reliability in terms of individual messages. Generally, a control system's communication infrastructure should provide interoperability, scalability, performance and maintainability. An attractive approach to accomplish this is to use a standardized and proven middleware implementation. The selection of a middleware can have a major cost impact in future integration efforts. In this paper we present development done at DTP2 using the Object Management Group's (OMG) standard specification for Data Distribution Service (DDS) for ensuring communications interoperability. DDS has gained a stable foothold especially in the military field. It lacks a centralized broker, thereby avoiding a single-point-of-failure. It also includes an extensive set of Quality of Service (QoS) policies. The standard defines a platform- and programming language independent model and an interoperability wire protocol that enables DDS vendor interoperability, allowing software developers to avoid vendor lock-in situations.

  9. Interoperability of remote handling control system software modules at Divertor Test Platform 2 using middleware

    International Nuclear Information System (INIS)

    Highlights: ► The prototype DTP2 remote handling control system is a heterogeneous collection of subsystems, each realizing a functional area of responsibility. ► Middleware provides well-known, reusable solutions to problems, such as heterogeneity, interoperability, security and dependability. ► A middleware solution was selected and integrated with the DTP2 RH control system. The middleware was successfully used to integrate all relevant subsystems and functionality was demonstrated. -- Abstract: This paper focuses on the inter-subsystem communication channels in a prototype distributed remote handling control system at Divertor Test Platform 2 (DTP2). The subsystems are responsible for specific tasks and, over the years, their development has been carried out using various platforms and programming languages. The communication channels between subsystems have different priorities, e.g. very high messaging rate and deterministic timing or high reliability in terms of individual messages. Generally, a control system's communication infrastructure should provide interoperability, scalability, performance and maintainability. An attractive approach to accomplish this is to use a standardized and proven middleware implementation. The selection of a middleware can have a major cost impact in future integration efforts. In this paper we present development done at DTP2 using the Object Management Group's (OMG) standard specification for Data Distribution Service (DDS) for ensuring communications interoperability. DDS has gained a stable foothold especially in the military field. It lacks a centralized broker, thereby avoiding a single-point-of-failure. It also includes an extensive set of Quality of Service (QoS) policies. The standard defines a platform- and programming language independent model and an interoperability wire protocol that enables DDS vendor interoperability, allowing software developers to avoid vendor lock-in situations

  10. Investigation of Be/Cu joints via HHF tests of small-scale mockups

    Energy Technology Data Exchange (ETDEWEB)

    Giniatulin, R.; Gervash, A.; Komarov, V.L.; Litunovsky, N.; Mazul, I.; Yablokov, N. [Efremov Inst., St. Petersburg (Russian Federation)

    1998-01-01

    Beryllium-copper (Be/Cu) joints in divertor components work under cyclic heat loads. To develop reliable joints small-scale mockups are fabricated by divertor technologies and tested under the divertor conditions. One of the critical damaging factors that exist in the divertor and have to be simulated is thermocyclic heat loads in the range of 1-15 MW/m{sup 2}. This work presents the divertor mockups that have beryllium tiles with different dimensions (5 x 5 - 44 x 44) mm{sup 2} brazed with copper alloy heat sink. The electron beam was used to braze these mockups so as to decrease the formation of brittle intermetallic layers. The description of mockups design, geometry of armour tiles and fabrication techniques are presented in the paper. The results of screening and thermocyclic tests of these mockups in the heat flux range of 2-12 MW/m{sup 2} with a number of cycles {approx}10{sup 3} are presented. The results of metallographic analysis are also presented. The results of fabrication and testing with small-scale mockups for first wall application are also described. (author)

  11. Actively convected liquid metal divertor

    Science.gov (United States)

    Shimada, Michiya; Hirooka, Yoshi

    2014-12-01

    The use of actively convected liquid metals with j × B force is proposed to facilitate heat handling by the divertor, a challenging issue associated with magnetic fusion experiments such as ITER. This issue will be aggravated even more for DEMO and power reactors because the divertor heat load will be significantly higher and yet the use of copper would not be allowed as the heat sink material. Instead, reduced activation ferritic/martensitic steel alloys with heat conductivities substantially lower than that of copper, will be used as the structural materials. The present proposal is to fill the lower part of the vacuum vessel with liquid metals with relatively low melting points and low chemical activities including Ga and Sn. The divertor modules, equipped with electrodes and cooling tubes, are immersed in the liquid metal. The electrode, placed in the middle of the liquid metal, can be biased positively or negatively with respect to the module. The j × B force due to the current between the electrode and the module provides a rotating motion for the liquid metal around the electrodes. The rise in liquid temperature at the separatrix hit point can be maintained at acceptable levels from the operation point of view. As the rotation speed increases, the current in the liquid metal is expected to decrease due to the v × B electromotive force. This rotating motion in the poloidal plane will reduce the divertor heat load significantly. Another important benefit of the convected liquid metal divertor is the fast recovery from unmitigated disruptions. Also, the liquid metal divertor concept eliminates the erosion problem.

  12. Rapidly Moving Divertor Plates In A Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S. Zweben

    2011-05-16

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

  13. Rapidly Moving Divertor Plates In A Tokamak

    International Nuclear Information System (INIS)

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

  14. Preliminary characterization of interlayer for Be/Cu sintered compacts

    Energy Technology Data Exchange (ETDEWEB)

    Sakamoto, N.; Kawamura, H. [Oarai Research Establishment, Ibaraki-ken (Japan)

    1995-09-01

    At present, beryllium is under consideration as a main candidate material for plasma facing components of ITER, because of its many advantages such as low Z, high thermal conductivity, low tritium retention, low activation and so on. Among the different divertor design options, the duplex structure where the beryllium armor is bonded with heat sink structural materials (DS-copper, Cu-Cr-Zr and so on) is under consideration. And plasma facing components will be exposed to high heat load and high neutron flux generated by the plasma. Therefore, it is necessary to develop the reliable bonding technologies between beryllium and heat sink structural materials in order to fabricate plasma facing components which can resist those. Then, we started the bonding technology development of beryllium and copper alloy with FGM (functional gradient material) in order to reduce thermal stress due to the difference of thermal expansion between beryllium and copper alloy. As the interlayers for FGM, eleven kinds of sintered compacts in which the mixing ratio of beryllium powder and oxygen free copper powder is different, were fabricated by the hot press/HIP method. The dimension of each compact is 8mm in diameter, 2mm in thickness. Then, thermal diffusivity and specific heat of these compacts were measured by laser flash method, and thermal conductivity was calculated from those values. From metalographical observation, it became clear that the sintered compacts of mixture of beryllium powder and copper powder contain residual beryllium, copper and two kinds of intermetallic compounds, Be{sub 2}Cu({delta}) and BeCu({gamma}). From the results of thermal characterization, thermal diffusivity of interlayers increased with increase of copper containing ratio. And, specific heat gradually decreased with increase of copper containing ratio.

  15. Divertor parameters and divertor operation in ASDEX

    Science.gov (United States)

    Fussmann, G.; Ditte, U.; Eckstein, W.; Grave, T.; Keilhacker, M.; McCormick, K.; Murmann, H.; Röhr, H.; Elshaer, M.; Steuer, K.-H.; Szymanski, Z.; Wagner, F.; Becker, G.; Bernhardi, K.; Eberhagen, A.; Gehre, O.; Gernhardt, J.; Gierke, G. V.; Glock, E.; Gruber, O.; Haas, G.; Hesse, M.; Janeschitz, G.; Karger, F.; Kissel, S.; Klüber, O.; Kornherr, M.; Lisitano, G.; Mayer, H. M.; Meisel, D.; Müller, E. R.; Poschenrieder, W.; Ryter, F.; Rapp, H.; Schneider, F.; Siller, G.; Smeulders, P.; Söldner, F.; Speth, E.; Stäbler, A.; Vollmer, O.

    1984-12-01

    Recent measurements of plasma boundary and divertor scrape-off parameters for ohmically and neutral injection heated plasmas are presented. For these data the power flow onto the divertor plates and the sputtering rates at the plates are calculated and compared with separate measurements. The impurity behaviour in front of the plates is also discussed.

  16. Divertor efficiency in ASDEX

    Science.gov (United States)

    Engelhardt, W.; Becker, G.; Behringer, K.; Campbell, D.; Eberhagen, A.; Fussmann, G.; Gehre, O.; Gierke, G. V.; Glock, E.; Haas, G.; Huang, M.; Karger, F.; Keilhacker, M.; KlÜber, O.; Kornherr, M.; Lisitano, G.; Mayer, H.-M.; Meisel, D.; Müller, E. R.; Murmann, H.; Niedermeyer, H.; Poschenrieder, W.; Rapp, H.; Schneider, F.; Siller, G.; Steuer, K.-H.; Venus, G.; Vernickel, H.; Wagner, F.

    1982-12-01

    The divertor efficiency in ASDEX is discussed for ohmically heated plasmas. The parameters of the boundary layer both in the torus midplane and the divertor chamber have been measured. The results are reasonably well understood in terms of parallel and perpendicular transport. A high pressure of neutral hydrogen builds up in the divertor chamber and Franck-Condon particles recycle back through the divertor throat. Due to dissociation processes the boundary plasma is effectively cooled before it reaches the neutralizer plates. The shielding property of the boundary layer against impurity influx is comparable to that of a limiter plasma. The transport of iron is numerically simulated for an iron influx produced by sputtering of charge exchange neutrals at the wall. The results are consistent with the measured iron concentration. First results from a comparison of the poloidal divertor with toroidally closed limiters (stainless steel, carbon) are given. Diverted discharges are considerably cleaner and easier to create.

  17. The Dynamic Ergodic Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Lehnen, M.; Adbullaev, S.; Biel, W.; Bock, M. F. M.; Brezinsek, S.; Busch, C.; Classen, I.; Finken, K. H.; Hartin, D.; Hellermann, M. von; Jachmich, S.; Jakubowski, M.; Jaspers, R.; Koslowski, H. R.; Kramer-Flecken, A.; Kikuchi, Y.; Liang, Y.; Loozen, X.; Pospieszczyk, A.; Rompuy, T. van; Reiter, D.; Samm, U.; Schmitz, O.; Sergienko, G.; Tokar, M.; Unterberg, B.; Wolf, R.; Zimmermann, O.

    2005-07-01

    The concept of the Dynamic Ergodic Divertor (DED) is based on plasma edge ergodisation by a resonant perturbation. Such a divertor concept is closely related to helical or island divertors in stellarators. The base mode of the DED perturbation field can be m/n = 12 /4, 6/2 or 3/1. The 3/1 base mode with its deep penetration of the perturbation field provides the excitation of tearing modes. This topic was presented elsewhere. In this contribution we concentrate on the divertor properties of the DED. We report on the characterisation of the topology, transport properties in ergodic fields, divertor regimes, impurity transport and density limit behaviour. The 12/4 base mode where the perturbation is restricted to the plasma edge is suitable for divertor operation. With increasing perturbation field island chains are built up at the resonance layers. Overlapping islands lead to ergodisation. The plasma is guided in the laminar region via open field lines of short connection length to the divertor target. The magnetic topology is not only controlled by the coil current but especially by the edge safety factor. For appropriate edge safety factor we observe a strong temperature drop in the plasma edge, indicating an expanding laminar region, which is necessary to decouple the divertor plasma from the core plasma. This temperature drop is accompanied by a redistribution of the heat and particle flux on the divertor target which is measured by thermography, visible spectroscopy and Langmuir probes. The modifications of the magnetic topology by the DED are reflected in the distribution of the plasma edge density and temperature measured by atomic beams and can be directly seen for example from carbon emission lines. The magnetic structure is calculated by the ATLAS code and shows good agreement with the experimental findings. The particle and energy transport is modelled with the EMC3-EIRENE code package and is in qualitative agreement with the measured densities and

  18. 'EU divertor celebration day'

    International Nuclear Information System (INIS)

    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

  19. Hybrid bundle divertor design

    International Nuclear Information System (INIS)

    A hybrid bundle divertor design is presented that produces <0.3% magnetic ripple at the center of the plasma while providing adequate space for the coil shielding and structure for a tokamak fusion test reactor similar to the International Tokamak Reactor and the Engineering Test Facility (with R = 5 m, B = 5 T, and a /SUB wall/ = 1.5 m, in particular). This hybrid divertor consists of a set of quadrupole ''wing'' coils running tangent to the tokamak plasma on either side of a bundle divertor. The wing coils by themselves pull the edge of the plasma out 1.5 m and spread the thickness of the scrape-off layer from 0.1 to 0.7 m at the midplane. The clear aperture of the bundle divertor throat is 1.0 m high and 1.8 m wide. For maintenance or replacement, the hybrid divertor can be disassembled into three parts, with the bundle divertor part pulling straight out between toroidal field coils and the wing coils then sliding out through the same opening

  20. Kinetic divertor modeling

    International Nuclear Information System (INIS)

    Highlights: ► We have studied the coupling among gas, plasma and surface in the divertor region. ► A one-dimensional PIC-DSMC model has been developed. ► Profiles of density and temperature of all the species involved have been provided. ► MAR processes are effective in a region smaller than 1.5 mm from the divertor plate. ► For regions more distant, the ionization of atoms, produced by MAR, starts to occur. - Abstract: The coupled dynamics and kinetics between gas and plasma in the divertor region is studied by means of a one-dimensional Particle in Cell-Direct Simulation Monte Carlo (PIC-DSMC) model. In particular, the collision-induced vibrational excitation/relaxation of H2 molecules and particle–surface interaction (vibrational relaxation and recombinative desorption) have been considered in detail to estimate the importance of plasma volumetric recombination by molecular assisted reaction (MAR). Spatially resolved results show that MAR processes are effective very close to the divertor plate in a region smaller than 1.5 mm from the divertor plate. For regions more distant the ionization of atoms, produced by MAR, starts to make molecular assisted recombination an ineffective reaction.

  1. Innovative divertor concepts for LHD

    International Nuclear Information System (INIS)

    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)

  2. High conductivity Be-Cu alloys for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lilley, E.A. [NGK Metals Corp., Reading, PA (United States); Adachi, Takao; Ishibashi, Yoshiki [NGK Insulators, Ltd., Aichi-ken (Japan)

    1995-09-01

    The optimum material has not yet been identified. This will result in heat from plasma to the first wall and divertor. That is, because of cracks and melting by thermal power and shock. Today, it is considered to be some kinds of copper, alloys, however, for using, it must have high conductivity. And it is also needed another property, for example, high strength and so on. We have developed some new beryllium copper alloys with high conductivity, high strength, and high endurance. Therefore, we are introducing these new alloys as suitable materials for the heat sink in fusion reactors.

  3. Divertor Coil Design and Implementation on Pegasus

    Science.gov (United States)

    Shriwise, P. C.; Bongard, M. W.; Cole, J. A.; Fonck, R. J.; Kujak-Ford, B. A.; Lewicki, B. T.; Winz, G. R.

    2012-10-01

    An upgraded divertor coil system is being commissioned on the Pegasus Toroidal Experiment in conjunction with power system upgrades in order to achieve higher β plasmas, reduce impurities, and possibly achieve H-mode operation. Design points for the divertor coil locations and estimates of their necessary current ratings were found using predictive equilibrium modeling based upon a 300 kA target plasma. This modeling represented existing Pegasus coil locations and current drive limits. The resultant design calls for 125 kA-turns from the divertor system to support the creation of a double null magnetic topology in plasmas with IpIGBT power supply modules to provide IDIV<=4 kA. The resulting 20 kA-turn capability of the existing divertor coil will be augmented by a new coil providing additional A-turns in series. Induced vessel wall current modeling indicates the time response of a 28 turn augmentation coil remains fast compared to the poloidal field penetration rate through the vessel. First results operating the augmented system are shown.

  4. Divertor for a torsatron

    International Nuclear Information System (INIS)

    The divertor for a torsatron comprising a toroidal vacuum chamber embracing the toroidal chamber of torsatron trap and communicating with it through the gaps between helical conductors of the system for creation of the trap magnetic field is described. The divertor comprises also a collector realized in a form of plates crossing magnetic field force lines. With the purpose of decreasing the plasma contamination level the collector plates realized curvilinear and embrace conductors at full their length and have the curvature less than that of the magnetic field force lines in the plate mounting point. The invention permits to decrease the plasma contamination by decreasing the particles flux formed as a result of collector plates errosion and accordingly increase plasma temperature in the trap

  5. Divertor plasma detachment

    Science.gov (United States)

    Krasheninnikov, S. I.; Kukushkin, A. S.; Pshenov, A. A.

    2016-05-01

    Regime with the plasma detached from the divertor targets (detached divertor regime) is a natural continuation of the high recycling conditions to higher density and stronger impurity radiation loss. Both the theoretical considerations and experimental data show clearly that the increase of the impurity radiation loss and volumetric plasma recombination causes the rollover of the plasma flux to the target when the density increases, which is the manifestation of detachment. Plasma-neutral friction (neutral viscosity effects), although important for the sustainment of high density/pressure plasma upstream and providing the conditions for efficient recombination and power loss, is not directly involved in the reduction of the plasma flux to the targets. The stability of detachment is also discussed.

  6. Be-Cu gradient materials through controlled segregation. Basic investigations

    Energy Technology Data Exchange (ETDEWEB)

    Muecklich, F.; Lorinser, M.; Hartmann, S.; Beinstingel, S. [Saarland Univ., Saarbruecken (Germany); Linke, J.; Roedig, M.

    1998-01-01

    The joining of materials has a fundamental problematic nature: Creating a sharp interface between two different materials causes a more or less extreme jump in the properties at this point. This may result in the failure of the component under mechanical or thermal loads. In some cases there are further difficulties caused by using a third component (e.g. the transformation of Ag-lead into Cd by neutron beams). The solution may be the creating of a functionally gradient material (FGM) Be-Cu. We discuss the advantage of such a FGM and the probabilities of an new procedure for manufacturing 1-dimensional FGMs. (author)

  7. The JET divertor coil

    International Nuclear Information System (INIS)

    The divertor coil is mounted inside the Jet vacuum vessel and is able to carry 1 MA turns. It is of conventional construction - water cooled copper, epoxy glass insulation -and is contained in a thin stainless steel case. The coil has to be assembled, insulated and encased inside the Jet vacuum vessel. A description of the coil is given, together with technical information (including mechanical effects on the vacuum vessel), an outline of the manufacture process and a time schedule. (author)

  8. Numerical studies on divertor experiments

    International Nuclear Information System (INIS)

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

  9. Design, R&D and commissioning of EAST tungsten divertor

    Science.gov (United States)

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

    2016-02-01

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

  10. Advanced divertor configurations with large flux expansion

    NARCIS (Netherlands)

    Soukhanovskii, V. A.; R.E. Bell,; Diallo, A.; S. Gerhardt,; S. Kaye,; E. Kolemen,; B.P. LeBlanc,; McLean, A.; Menard, J. E.; S.F. Paul,; Podesta, M.; Raman, R.; D.D. Ryutov,; F. Scotti,; Kaita, R.; Maingi, R.; D.M. Mueller,; Roquemore, A. L.; Reimerdes, H.; G.P. Canal,; Labit, B.; Vijvers, W.; Coda, S.; Duval, B. P.; Morgan, T.; Zielinski, J.; De Temmerman, G.; Tal, B.

    2013-01-01

    Experimental studies of the novel snowflake divertor concept (D. Ryutov, Phys. Plasmas 14 (2007) 064502) performed in the NSTX and TCV tokamaks are reviewed in this paper. The snowflake divertor enables power sharing between divertor strike points, as well as the divertor plasma-wetted area, effecti

  11. Understanding impurity retention by divertors

    International Nuclear Information System (INIS)

    Simple, 1-D fluid model prescriptions are developed to predict under what circumstances impurities released at divertor targets would be expected to leak to the main plasma. The prescriptions are tested by comparison with results using the DIVIMP (divertor impurity) Monte Carlo code and are found to be well satisfied under strongly collisional conditions. The transition to collisionlessness degrades the agreement with the simple model. Usually, the simple model predicts a more-or-less catastrophic buildup of impurities outside the divertor. This, however, is an artificial result arising from the assumption of strictly one-dimensional, along B, motion; even weak cross-field transport can stop such impurity accumulation. ((orig.))

  12. R.H. divertor maintenance-the divertor refurbishment platform

    International Nuclear Information System (INIS)

    The ITER divertor assembly consists in 60 cassettes located in the bottom region of the vacuum vessel. Because of erosion and damage during, reactor operations, their replacement is expected to be required eight times during the machine lifetime. The cassettes will be withdrawn from the vessel through dedicated ducts and they will be transported to a hot cell for refurbishment. The divertor refurbishment platform (DRP) simulates the arrangement in the divertor hot cell for cassette inspection, component replacement and repair, measuring, and testing. The DRP had to demonstrate the feasibility of divertor cassette refurbishment, procedures, and the use of conventional remote handling equipment in a hot cell, for the refurbishment of high heat flux components (also called plasma facing components PFC), cassette locking systems, water feeds and post-repair, integrity testing. The true environmental conditions (temperature, atmosphere, radiation, contamination) have not been replicated in the DRP, but they were taken into account in the development of the mock ups, the remote handling equipment, and the operating procedures. The results permit to validate the hot cell operations for the cassette refurbishment and to specify the hot cell requirements. This paper describes the objectives, lay-out, test programme, test results, and future activities of the divertor refurbishment platform

  13. 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.

  14. Advanced divertor configurations with large flux expansion

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-15

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

  15. R.H. divertor maintenance -- the divertor test platform

    International Nuclear Information System (INIS)

    The ITER divertor assembly consists in 60 cassettes located in the bottom region of the vacuum vessel. Because of erosion and damage, their replacement is expected to be required eight times during the machine lifetime. The cassettes will be remotely withdrawn from the vessel through dedicated ducts and they will be transported to a hot cell for refurbishment. To demonstrate the feasibility of the withdrawal operations, and to optimise the maintenance scenario and the handling equipment design, a test facility has been set-up at the ENEA Research Centre of Brasimone (Italy), i.e. the divertor test platform (DTP) that allows to simulate, in full scale, all handling operations inside the vacuum vessel. This paper describes the objectives, test programme, layout, test results and future activities of the DTP

  16. Current state-of-the-art manufacturing technology for He-cooled divertor finger

    International Nuclear Information System (INIS)

    A divertor concept for DEMO has been investigated at Karlsruhe Institute of Technology (KIT) which has to withstand a heat flux of 10 MW/m2. The design utilizes small finger module composed of a small tungsten tile brazed on a thimble made from tungsten alloy. The divertor finger is cooled by helium jet impingement at 10 MPa and 600 deg. C. The subject of this paper is technological studies on machining and braze joining the divertor components. Goal of this task, which is considered an important R and D issue, is to find out appropriate manufacturing methods to ensure high functionality and high reliability of the divertor as well as to meet the economic aspect. One of the major requirements for manufacturing is micro-crack-free surface of tungsten parts, since crack propagations in tungsten were observed in the previous high-heat-flux tests at Efremov. Different manufacturing methods and the corresponding results are discussed in the following report.

  17. PFC integration on Tore-Supra WEST divertor

    International Nuclear Information System (INIS)

    Full text of publication follows. In the context of the Tokamak Tore Supra evolution, the CEA Cadarache aims at transforming it into a test bench for ITER plasma facing components. 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, and an X-point close to the lower divertor. This environment will allow exposing the divertor components up to 20 MW/m2 heat flux during long pulse. These specifications are well suited to test the actively cooled tungsten 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. The divertors coils are designed regarding the magnetic specifications, the plasma facing components are placed according to the plasma shape, and then the interfaces have to be managed regarding the remaining space. Moreover in this layer, many important smaller components have to be integrated as cooling pipes, magnetic diagnostics, gas injection, Langmuir probe, etc. This paper deals with the integrated design of ITER tungsten 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 PFC will be replaced several times. The ports size allows entering a 30 degrees sector of pre-installed tungsten targets which will be plugged as quickly and easily has possible. The main feature of steady state operations is the active cooling, which lead to have many embedded cooling channels and bulky pipes on the PFC module. It means to take care of the many connections and sealing between vacuum and water. (authors)

  18. Progress in ergodic divertor operation on Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Ghendrih, Ph.; Becoulet, M.; Colas, L.; Grosman, A.; Guirlet, R.; Gunn, J.; Loarer, T.; Azeroual, A.; Basiuk, V.; Beaumont, B.; Becoulet, A.; Bremond, S.; Bucalossi, J.; Capes, H.; Corre, Y.; Costanzo, L.; Michelis, C. de; Devynck, P.; Feron, S.; Friant, C.; Garbet, X.; Giannella, R.; Grisolia, C.; Hess, W.; Hogan, J.; Ladurelle, L.; Laugier, F.; Martin, G.; Mattioli, M.; Meslin, B.; Monier-Garbet, P.; Moulin, D.; Nguyen, F.; Pascal, J.Y.; Pecquet, A.L.; Pegourie, B.; Reichle, R.; Saint-Laurent, F.; Vallet, J.C.; Zabiego, M

    1999-09-01

    Upgrade of the Tore ergodic divertor has led to significant progress in ergodic divertor physics. The disruptive limit governed by the stochastization of the outer magnetic surfaces is found to occur for a value of the Chirikov parameter reaching 2 on the magnetic surface q = 2 + 3 / 12. This experimentally observed robustness allows one to operate at very low safety factor on the separatrix (q {approx} 2). Numerical analysis of ballooning turbulence in a stochastic layer indicates that the decay of the density fluctuations is in associated with an increase of the fluctuating electric drift velocity. The bottom line is then an enhanced cross-field transport in the vicinity of the target plates. This lowering of confinement appears to be compensated by an intrinsic transport barrier on the electron temperature. The 3-D response of the temperature field is computed with a fluid code. The intrinsic transport barrier at the separatrix, reported experimentally, can be recovered together with small amplitude temperature modulations in the divertor volume. Experimental evidence of the 3 density regimes (linear, high recycling and detachment) is reported. The low critical density values for these transitions indicate that similar parallel physics govern the axisymmetric and ergodic divertor, despite the open configuration of the latter. Measurement and understanding of these density regimes provide a means for feedback control of plasma density and an improvement in ICRH coupling scenarios. Experimental data also indicated that particle control with the vented target plates is effective. Increase of impurity control and radiation efficiency are recalled. Global power balance has been analysed. These results confirm the enhanced radiation capacity of the ergodic divertor. (author)

  19. Control of divertor geometry and performance of the ergodic divertor of Tore Supra

    International Nuclear Information System (INIS)

    Experimental evidence of the location of the ergodic divertor separatrix is shown to agree with the predicted value given by codes. Variation of this position modifies the divertor tightness, defined as the ratio of the divertor to core density. This effect is governed by laminar transport, i.e., transport proportional to the magnitude of the perturbation. Operation with feedback control of the divertor temperature allows one to optimise the choice of injected impurity species. At 10 eV divertor temperature, nitrogen is shown to lead to the largest decrease in energy flux to the divertor at lowest contribution to Zeff. Parallel energy fluxes as low as 2 MW m-2 are thus achieved on the target plates. For this impurity, radiation is localised in the divertor volume thus leading to radiation compression close to 10. The ergodic divertor appears as a powerful tool to control plasma-wall interaction with no loss of core confinement or plasma current

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  1. Simulation Analysis of Divertor Performance in EAST

    Institute of Scientific and Technical Information of China (English)

    Zhu Sizheng; Zha Xuejun

    2005-01-01

    A detailed study of the divertor performance in the EAST has been conducted for both its double null and single null configurations. The results of the application of the SOLPS (B2/Eirene) code package to the analysis of the EAST divertor are summarized. Here we concentrate on the effects of the increased geometrical closure and variation in the magnetic topology on the behavior of divertor plasmas. The results of numerical predictions for the EAST divertor's operational window are also described in this paper.

  2. Effect of Divertor Shaping on Divertor Plasma Behavior on DIII-D

    Science.gov (United States)

    Petrie, T. W.; Leonard, A. W.; Luce, T. C.; Mahdavi, M. A.; Holcomb, C. T.; Fenstermacher, M. E.; Hill, D. N.; Lasnier, C. J.; Watkins, J. G.; Moyer, R. A.; Stangeby, P. C.

    2012-10-01

    Recent experiments examined the dependence of divertor density (nTAR), temperature (TTAR), and heat flux at the outer divertor separatrix target on changes in the divertor separatrix geometry. The responses of nTAR and TTAR to changes in the parallel connection length in the scrape-off layer (SOL) (L||) are consistent with the predictions of the Two Point Model (TPM). However, nTAR and TTAR display a more complex response to changes in the radial location of the outer divertor strike point (RTAR) than expected based on the TPM. SOLPS transport analysis indicates that small differences in divertor geometry can change neutral trapping sufficient to explain differences between experiment and TPM predictions. The response of the core and divertor plasmas to changes in L|| and RTAR, under both radiating and non-radiating divertor conditions, will be shown.

  3. A simple model for biased divertors

    Energy Technology Data Exchange (ETDEWEB)

    Lachambre, J.-L.; Quirion, B.; Gunn, J.; Boucher, C.; Stansfield, B.; Gauvreau, J.-L. [Centre canadien de fusion magnetique, 1804, boulevard Lionel-Boulet, Varennes, Quebec, J3X 1S1 (Canada)

    1997-12-01

    Ionization near the target plate is shown to play an important role in biasing experiments. Our previous SOL model, which calculates the induced radial electric field, is found to be inadequate to treat the new divertor geometry of TdeV. When recycling is included via the measured D{sub {alpha}} emission near the plate, the upgraded model correctly reproduces all the observed electric currents and fields during biasing in the new divertor configuration. A simple divertor model using this calculated field has been developed to simulate the evolution of the divertor ion and neutral parameters under the action of neutralization plate biasing. Using a 1D adiabatic fluid model for the divertor ions, a 1D convective representation for the SOL neutrals and a 0D calculation for the plenum pressure, this divertor model satisfactorily simulates most of the TdeV biasing experiments at all biasing voltages and all toroidal field directions at low line-averaged densities. The weaker agreement at high densities is largely a consequence of the crudeness of the general divertor physics rather than of the deficiency of the biasing physics implemented in the model. The model is finally used to explain the polarity asymmetries observed in divertor efficiencies during biasing, and to demonstrate that no mechanism other than plate current saturation is required to interpret the saturation of toroidal rotation observed in the SOL at large biasing voltages of either polarity. (author)

  4. Optimization of a bundle divertor for FED

    International Nuclear Information System (INIS)

    Optimal double-T bundle divertor configurations have been obtained for the Fusion Engineering Device (FED). On-axis ripple is minimized, while satisfying a series of engineering constraints. The ensuing non-linear optimization problem is solved via a sequence of quadratic programming subproblems, using the VMCON algorithm. The resulting divertor designs are substantially improved over previous configurations

  5. ITER-FEAT divertor maintenance and integration

    International Nuclear Information System (INIS)

    This paper presents the design status of the maintenance and integration of the ITER-FEAT divertor. It also includes the first results of a study showing how the in-vessel viewing system could be integrated at the divertor level. The studies are on-going, but already preliminary practical layouts have been produced

  6. Moving Divertor Plates in a Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S.J. Zweben, H. Zhang

    2009-02-12

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

  7. Reactor application of an improved bundle divertor

    International Nuclear Information System (INIS)

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

  8. Divertor and gas blanket impurity control study

    Energy Technology Data Exchange (ETDEWEB)

    El Derini, Z; Stacey, Jr, W M

    1979-04-01

    A simple calculational model for the transport of particles across the scrap off region between the plasma and the wall in the presence of a divertor or a gas blanket has been developed. The model departs from previous work in including: (a) the entire impurity transport as well as its effect on the energy balance equations; (b) the recycling neutrals from the divertor, and (c) the reflected neutrals from the wall. Results obtained with this model show how the steady state impurity level in the plasma depends on the divertor parameters such as the neutral backflow from the divertor, the particle residence time and the scrape off thickness; and on the gas blanket parameters such as the neutral source strength and the gas blanket thickness. The variation of the divertor or gas blanket performance as a function of the heat and particle fluxes escaping from the plasma, the wall material and the cross field diffusion is examined and numerical examples are given.

  9. Ergodic divertor effect on low-Z impurity transport for inner-wall limited plasmas in Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Hogan, J. [Fusion Energy Division, Oak Ridge National Laboratory, Oak Ridge, TN (United States); De Michelis, C.; Monier-Garbet, P.; Corre, Y.; Guirlet, R. [Association EURATOM-CEA sur la Fusion Controlee, CEA-Cadarache, St-Paul-lez-Durance (France)

    2002-06-01

    Observations of systematic spatial modulation of low-Z impurity radiative emissions are analysed for Tore Supra ergodic divertor discharges limited on the inner wall. Some similarities to modulations previously observed with Marfe-like conditions are observed, but significant differences are also seen in the cases studied here. A simulation of the modulations is made, using the three-dimensional edge transport code BBQ. The simulations suggest that an important role is played by charge exchange with neutral deuterium, in addition to the ergodic divertor-induced modulations of the electron temperature. The interpretation highlights the important role of intermediate-Z states in impurity transport processes. (author)

  10. Magnetic divertors for experimental Tokamaks and fusion reactors

    International Nuclear Information System (INIS)

    Brief reports of working group discussions. These covered the requirements for a divertor in a fusion reactor including reducing impurities, exhausting the plasma and controlling the plasma-wall interactions. Divertor configurations were also reviewed and their merits and disadvantages compared. Existing divertor experiments were summarised and recommendations for further work made. Then the problems anticipated in designing a divertor for a conceptual reactor were considered. The physics of divertors and the scrape-off layer was discussed with reference to present models of plasma in divertors. Finally, experiments needed to demonstrate the feasibility of divertors for reactors and the development of specialised diagnostics for such experiments were considered. (U.K.)

  11. MAST-Upgrade Divertor Facility and Assessing Performance of Long-Legged Divertors

    CERN Document Server

    Fishpool, G; Cunningham, G; Harrison, J; Katramados, I; Kirk, A; Kovari, M; Meyer, H; Scannell, R

    2013-01-01

    A potentially important feature in a divertor design for a high-power tokamak is an extended and expanded divertor leg. The upgrade to MAST will allow a wide range of such divertor leg geometries to be produced, and hence will allow the roles of greatly increased connection length and flux expansion to be experimentally tested. This will include testing the potential of the Super-X configuration [1]. The design process for the upgrade has required analysis of producing and controlling the magnetic configurations, and has included consideration of the roles that divertor closure and increasing magnetic connection length will play.

  12. Overview of Bore Tools Systems for divertor remote maintenance of ITER

    International Nuclear Information System (INIS)

    Because of the radiation levels preventing direct, hands-on access to the machine components, maintenance work on ITER will eventually require the use of Remote Handling techniques. In particular, the replacement of components such as divertor and blanket modules will require the use of remote cutting, welding and Non Destructive Testing of water cooling pipes

  13. High flux expansion divertor studies in NSTX

    CERN Document Server

    Soukhanovskii, V A; Bell, R E; Gates, D A; Kaita, R; Kugel, H W; LeBlanc, B P; Maqueda, R; Menard, J E; Mueller, D; Paul, S F; Raman, R; Roquemore, A L

    2009-01-01

    High flux expansion divertor studies have been carried out in the National Spherical Torus Experiment using steady-state X-point height variations from 22 to 5-6 cm. Small-ELM H-mode confinement was maintained at all X-point heights. Divertor flux expansions from 6 to 26-28 were obtained, with associated reduction in X-point connection length from 5-6 m to 2 m. Peak divertor heat flux was reduced from 7-8 MW/m$^2$ to 1-2 MW/m$^2$. In low X-point configuration, outer strike point became nearly detached. Among factors affecting deposition of parallel heat flux in the divertor, the flux expansion factor appeared to be dominant

  14. Divertor Thomson scattering on DIII-D

    International Nuclear Information System (INIS)

    In this paper we describe the newly installed divertor Thomson scattering system for the DIII-D tokamak and present initial results from plasma discharges. Measured plasma densities have ranged from 5 x 1018 to 5 x 1020 m-3 and divertor plasma temperatures from 1 to 500 eV. These data are compared with earlier Langmuir probe data and qualitatively compared with UEDGE computer simulations. The divertor Thomson system uses one of the eight existing core Thomson scattering lasers (1 J, 20 Hz) which has been re-directed to probe the divertor region of the DIII-D vessel. Scattered light from this multipulse Nd:Yag laser is viewed with an f/6.8 collection optics system which provides eight spatial channels from 1-21 cm above the vessel floor (divertor target), each with 1.5 cm vertical resolution. Translating the plasma across the vessel floor using position controls provides a full scan of the divertor plasma. (orig.)

  15. Radiative power loading in the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Guillemaut, C., E-mail: christophe.guillemaut@cea.fr [ITER Organization, Route de Vinon CS 90 046, 13067 Saint-Paul-Lez-Durance (France); Pitts, R.A.; Kukushkin, A.S. [ITER Organization, Route de Vinon CS 90 046, 13067 Saint-Paul-Lez-Durance (France); O' Mullane, M. [Department of Physics, University of Strathclyde, Glasgow G4 0NG (United Kingdom)

    2011-12-15

    In ITER, steady state burning plasma operation will require a partially detached divertor state in order to reduce the peak power flux density to technologically achievable values at the actively cooled target plates ({approx}10 MW m{sup -2}). Such partially detached solutions require high radiative power dissipation in the divertor volume, with 60-70 MW expected in the baseline H-mode operating scenario. Power levels of this magnitude pose potential difficulties for divertor substructures, which, although also actively cooled, are not designed to withstand very high heat fluxes. This paper estimates the radiative power flux densities falling on critical divertor substructures during ITER burning plasma operation using commercial optical ray-tracing software to project radiation distributions simulated with the SOLPS plasma boundary simulation code onto a full 3D description of the divertor. The results indicate that inclusion of the real geometry provides heat flux densities due to photon illumination not higher than quasi-analytic estimates used in the original divertor design stages, and in some cases lower. When applied to the specific simple geometries used to develop the analytic expressions, the raytracing fully validates the analytic approach.

  16. FINAL REPORT FOR THE DIII-D RADIATIVE DIVERTOR PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    O' NEIL, RC; STAMBAUGH, RD

    2002-06-01

    OAK A271 FINAL REPORT FOR THE DIII-D RADIATIVE DIVERTOR PROJECT. The Radiative Divertor Project originated in 1993 when the DIII-D Five Year Plan for the period 1994--1998 was prepared. The Project Information Sheet described the objective of the project as ''to demonstrate dispersal of divertor power by a factor of then with sufficient diagnostics and modeling to extend the results to ITER and TPX''. Key divertor components identified were: (1) Carbon-carbon and graphite armor tiles; (2) The divertor structure providing a gas baffle and cooling; and (3) The divertor cryopumps to pump fuel and impurities.

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

    International Nuclear Information System (INIS)

    We survey the results of recent DIII-D experiments that tested the effectiveness of three innovative tokamak divertor concepts in reducing divertor heat flux while still maintaining acceptable energy confinement under neon/deuterium-based radiating divertor (RD) conditions: (1) magnetically unbalanced high performance double-null divertor (DND) plasmas, (2) high performance double-null “Snowflake” (SF-DN) plasmas, and (3) single-null H-mode plasmas having different isolation from their divertor targets. In general, all three concepts adapt well to RD conditions, achieving significant reduction in divertor heat flux (q⊥p) and maintaining high performance metrics, e.g., 50–70% reduction in peak divertor heat flux for DND and SF-DN plasmas that are characterized by βN ≅ 3.0 and H98(y,2) ≈ 1.35. It is also demonstrated that q⊥p could be reduced ≈50% by extending the parallel connection length (L||-XPT) in the scrape-off layer between the X-point and divertor targets over a variety of the RD and non-RD environments tested

  18. Control of divertor geometry and performance of the ergodic divertor of Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Ghendrih, Ph. E-mail: ghendrih@drfc.cad.cea.fr; Becoulet, M.; Costanzo, L.; Corre, Y.; Grisolia, C.; Grosman, A.; Guirlet, R.; Gunn, J.; Loarer, T.; Monier-Garbet, P.; Mank, G.; Reichle, R.; Vallet, J.-C.; Zabiego, M.; Azeroual, A.; Bucalossi, J.; Devynck, P.; De Michelis, C; Finken, K.H.; Hogan, J.; Laugier, F.; Nguyen, F.; Pegourie, B.; Saint-Laurent, F.; Schunke, B

    2001-03-01

    Experimental evidence of the location of the ergodic divertor separatrix is shown to agree with the predicted value given by codes. Variation of this position modifies the divertor tightness, defined as the ratio of the divertor to core density. This effect is governed by laminar transport, i.e., transport proportional to the magnitude of the perturbation. Operation with feedback control of the divertor temperature allows one to optimise the choice of injected impurity species. At 10 eV divertor temperature, nitrogen is shown to lead to the largest decrease in energy flux to the divertor at lowest contribution to Z{sub eff}. Parallel energy fluxes as low as 2 MW m{sup -2} are thus achieved on the target plates. For this impurity, radiation is localised in the divertor volume thus leading to radiation compression close to 10. The ergodic divertor appears as a powerful tool to control plasma-wall interaction with no loss of core confinement or plasma current.

  19. 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)

  20. 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.

  1. Microturbulence measurements during divertor biasing

    International Nuclear Information System (INIS)

    The application of a bias voltage to a neutralization plate of the upper divertor with respect to the vacuum chamber in the Tokamak de Varennes (TdeV) influences the plasma well inside the separatrix. In particular, the unbiased Ohmic poloidal rotation edge velocity measured by visible spectroscopy is found to be in the electron diamagnetic drift direction (2-3 km/s) and increases by a factor of two for Vbias = 100 V. This coincides with a major reduction of the microturbulence signal at low frequencies (50 kHz -1 -1), as determined from coherent laser scattering measurements. One possible explanation is that the turbulence signal is simply Doppler shifted to frequencies outside the accessible range. This scenario is, however, difficult to reconcile with some observations. Another explanation invokes a reduction of the turbulence level. The variation of the turbulence signal as a function of the applied bias voltage can indeed be reproduced with a theoretical model based on radial and poloidal decorrelation mechanisms, the latter corresponding to poloidal velocity shear stabilization. This model also explains the observed steepening of the k-spectrum decay during biasing. Biasing also modifies the electron density profile inside the separatrix. These changes of nabla ne cannot explain the behaviour of microturbulence behaviour, when explained in terms of stabilization, would agree with the plasma maintaining a steeper electron density gradient. (author). 17 refs, 9 figs

  2. First results from the dynamic ergodic divertor at TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Lehnen, M. [Institut fuer Plasmaphysik, Forschungszentrum Juelich GmbH, EURATOM Association, Trilateral Euregio Cluster, 52425 Juelich (Germany)]. E-mail: m.lehnen@fz-juelich.de; Abdullaev, S.S.; Biel, W.; Brezinsek, S.; Finken, K.H.; Harting, D.; Hellermann, M. von; Jakubowski, M.; Jaspers, R.; Kobayashi, M.; Koslowski, H.R.; Kraemer-Flecken, A.; Matsunaga, G.; Pospieszczyk, A.; Reiter, D.; Van Rompuy, T.; Samm, U.; Schmitz, O.; Sergienko, G.; Unterberg, B.; Wolf, R.; Zimmermann, O. [Institut fuer Plasmaphysik, Forschungszentrum Juelich GmbH, EURATOM Association, Trilateral Euregio Cluster, 52425 Juelich (Germany)

    2005-03-01

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

  3. Impurity-induced divertor plasma oscillations

    Science.gov (United States)

    Smirnov, R. D.; Kukushkin, A. S.; Krasheninnikov, S. I.; Pigarov, A. Yu.; Rognlien, T. D.

    2016-01-01

    Two different oscillatory plasma regimes induced by seeding the plasma with high- and low-Z impurities are found for ITER-like divertor plasmas, using computer modeling with the DUSTT/UEDGE and SOLPS4.3 plasma-impurity transport codes. The oscillations are characterized by significant variations of the impurity-radiated power and of the peak heat load on the divertor targets. Qualitative analysis of the divertor plasma oscillations reveals different mechanisms driving the oscillations in the cases of high- and low-Z impurity seeding. The oscillations caused by the high-Z impurities are excited near the X-point by an impurity-related instability of the radiation-condensation type, accompanied by parallel impurity ion transport affected by the thermal and plasma friction forces. The driving mechanism of the oscillations induced by the low-Z impurities is related to the cross-field transport of the impurity atoms, causing alteration between the high and low plasma temperature regimes in the plasma recycling region near the divertor targets. The implications of the impurity-induced plasma oscillations for divertor operation in the next generation tokamaks are also discussed.

  4. Impurity-induced divertor plasma oscillations

    Energy Technology Data Exchange (ETDEWEB)

    Smirnov, R. D., E-mail: rsmirnov@ucsd.edu; Krasheninnikov, S. I.; Pigarov, A. Yu. [University of California, San Diego, La Jolla, California 92093 (United States); Kukushkin, A. S. [NRC “Kurchatov Institute”, Moscow 123182 (Russian Federation); National Research Nuclear University MEPhI, Moscow 115409 (Russian Federation); Rognlien, T. D. [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States)

    2016-01-15

    Two different oscillatory plasma regimes induced by seeding the plasma with high- and low-Z impurities are found for ITER-like divertor plasmas, using computer modeling with the DUSTT/UEDGE and SOLPS4.3 plasma-impurity transport codes. The oscillations are characterized by significant variations of the impurity-radiated power and of the peak heat load on the divertor targets. Qualitative analysis of the divertor plasma oscillations reveals different mechanisms driving the oscillations in the cases of high- and low-Z impurity seeding. The oscillations caused by the high-Z impurities are excited near the X-point by an impurity-related instability of the radiation-condensation type, accompanied by parallel impurity ion transport affected by the thermal and plasma friction forces. The driving mechanism of the oscillations induced by the low-Z impurities is related to the cross-field transport of the impurity atoms, causing alteration between the high and low plasma temperature regimes in the plasma recycling region near the divertor targets. The implications of the impurity-induced plasma oscillations for divertor operation in the next generation tokamaks are also discussed.

  5. Impurity-induced divertor plasma oscillations

    International Nuclear Information System (INIS)

    Two different oscillatory plasma regimes induced by seeding the plasma with high- and low-Z impurities are found for ITER-like divertor plasmas, using computer modeling with the DUSTT/UEDGE and SOLPS4.3 plasma-impurity transport codes. The oscillations are characterized by significant variations of the impurity-radiated power and of the peak heat load on the divertor targets. Qualitative analysis of the divertor plasma oscillations reveals different mechanisms driving the oscillations in the cases of high- and low-Z impurity seeding. The oscillations caused by the high-Z impurities are excited near the X-point by an impurity-related instability of the radiation-condensation type, accompanied by parallel impurity ion transport affected by the thermal and plasma friction forces. The driving mechanism of the oscillations induced by the low-Z impurities is related to the cross-field transport of the impurity atoms, causing alteration between the high and low plasma temperature regimes in the plasma recycling region near the divertor targets. The implications of the impurity-induced plasma oscillations for divertor operation in the next generation tokamaks are also discussed

  6. Melt damage to the JET ITER-like Wall and divertor

    Science.gov (United States)

    Matthews, G. F.; Bazylev, B.; Baron-Wiechec, A.; Coenen, J.; Heinola, K.; Kiptily, V.; Maier, H.; Reux, C.; Riccardo, V.; Rimini, F.; Sergienko, G.; Thompson, V.; Widdowson, A.; Contributors, JET

    2016-02-01

    In October 2014, JET completed a scoping study involving high power scenario development in preparation for DT along with other experiments critical for ITER. These experiments have involved intentional and unintentional melt damage both to bulk beryllium main chamber tiles and to divertor tiles. This paper provides an overview of the findings of concern for machine protection in JET and ITER, illustrating each case with high resolution images taken by remote handling or after removal from the machine. The bulk beryllium upper dump plate tiles and some other protection tiles have been repeatedly flash melted by what we believe to be mainly fast unmitigated disruptions. The flash melting produced in this way is seen at all toroidal locations and the melt layer is driven by j × B forces radially outward and upwards against gravity. In contrast, the melt pools caused while attempting to use MGI to mitigate deliberately generated runaway electron beams are localized to several limiters and the ejected material appears less influenced by j × B forces and shows signs of boiling. In the divertor, transient melting of bulk tungsten by ELMs was studied in support of the ITER divertor material decision using a specially prepared divertor module containing an exposed edge. Removal of the module from the machine in 2015 has provided improved imaging of the melt and this confirms that the melt layers are driven by ELMs. No other melt damage to the other 9215 bulk tungsten lamellas has yet been observed.

  7. First experiments on the TO-2 tokamak with a divertor

    International Nuclear Information System (INIS)

    Long stable discharges have been obtained in a recetrack tokamak with toroidal divertors in low plasma density regime. Divertors sharply limit plasma filament cross section, plasma density decreasing by an order at 1 cm length near the separatrix. 8 mm thick well formed flux of plasma appears at the divertor plate. Divertor power efficiency at different modes of operation is 50- 70 %. As compared to the TO-1 nondivertor tokamak some plasma filament hot zone expansion is recorded in the TO-2 tokamak

  8. Metallographic anlaysis and strength investigation of different Be-Cu joints in the temperature range RT-3500C

    Energy Technology Data Exchange (ETDEWEB)

    Gervash, A.A.; Giniatouline, R.N.; Mazul, I.V. [Efremov Research Institute, St. Petersburg (Russian Federation)] [and others

    1995-09-01

    The goal of this work is to estimate the strength and structure of different Be-Cu joining techniques. Brazing, diffusion bonding and joint rolling methods were chosen as ITER Be-Cu joint method candidates. Selected for ITER application Be-Cu joints were produced as technological plates (30-50 mm x 50-100 mm x thickness). AR samples for farther investigations were cutted out from initial technological plates. To compare mechanical strength of selected Be-Cu joints tensile and shearing tests of chosen candidates were carried out in the temperature range RT - 350{degrees}C. The metallographic analysis of Be-Cu crosssection was also done. Preliminary results of these tests as well as metallographic analysis data are presented. The industrial possibilities of producing required for ITER full scale Be-Cu joints are discussed.

  9. Analysis of particle transport in a gas target divertor

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

  10. Designing divertor targets for uniform power load

    Science.gov (United States)

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

    2015-08-01

    Divertor design for next step fusion reactors heavily relies on 2D edge plasma modeling with codes as e.g. B2-EIRENE. While these codes are typically used in a design-by-analysis approach, in previous work we have shown that divertor design can alternatively be posed as a mathematical optimization problem, and solved very efficiently using adjoint methods adapted from computational aerodynamics. This approach has been applied successfully to divertor target shape design for more uniform power load. In this paper, the concept is further extended to include all contributions to the target power load, with particular focus on radiation. In a simplified test problem, we show the potential benefits of fully including the radiation load in the design cycle as compared to only assessing this load in a post-processing step.

  11. MAST-Upgrade Divertor Facility and Assessing Performance of Long-Legged Divertors

    OpenAIRE

    Fishpool, G.; Canik, J.; Cunningham, G.; Harrison, J.; Katramados, I.; Kirk, A.; Kovari, M.; H. Meyer; Scannell, R.; Team, the MAST-Upgrade

    2013-01-01

    A potentially important feature in a divertor design for a high-power tokamak is an extended and expanded divertor leg. The upgrade to MAST will allow a wide range of such divertor leg geometries to be produced, and hence will allow the roles of greatly increased connection length and flux expansion to be experimentally tested. This will include testing the potential of the Super-X configuration [1]. The design process for the upgrade has required analysis of producing and controlling the mag...

  12. Divertor detachment and exhaust on the TdeV tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Decoste, R.; Stansfield, B.L.; Gauvreau, J.L. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada)] [and others

    1996-12-01

    Experimental data, analysis and simulations are used to describe the physics of divertor detachment and He exhaust under detached conditions in TdeV. With increasingly density, the plasma is found to detach progressively from the outboard divertor plates with a marked reduction of the ion flux to the plates, the generation of a pressure gradient between an ionization front and the target plate, and strong cross-field transport in the divertor. Local interactions between the divertor plasma and the plates are described, with evidence for carbon sputtering and molecular processes near the divertor plates. Divertor exhaust and retention continue to increase through detachment and He exhaust is not affected although the divertor He enrichment remains low but constant at about 0.2. A moderate density of n-bar{sub e} {approx} 5 x 10{sup 19} m{sup -3} seems to be sufficient both for efficient peak power load reduction at the divertor plate and good He exhaust through the divertor. Simulation of the edge and divertor plasmas using the B2/EIRENE and DIVIMP codes give reasonable agreement with the measurements and indicate possible divertor geometry improvements. (author).

  13. Divertor detachment and exhaust on the TdeV tokamak

    International Nuclear Information System (INIS)

    Experimental data, analysis and simulations are used to describe the physics of divertor detachment and He exhaust under detached conditions in TdeV. With increasingly density, the plasma is found to detach progressively from the outboard divertor plates with a marked reduction of the ion flux to the plates, the generation of a pressure gradient between an ionization front and the target plate, and strong cross-field transport in the divertor. Local interactions between the divertor plasma and the plates are described, with evidence for carbon sputtering and molecular processes near the divertor plates. Divertor exhaust and retention continue to increase through detachment and He exhaust is not affected although the divertor He enrichment remains low but constant at about 0.2. A moderate density of n-bare ∼ 5 x 1019 m-3 seems to be sufficient both for efficient peak power load reduction at the divertor plate and good He exhaust through the divertor. Simulation of the edge and divertor plasmas using the B2/EIRENE and DIVIMP codes give reasonable agreement with the measurements and indicate possible divertor geometry improvements. (author)

  14. Detailed Radiative Transport Modeling of a Radiative Divertor

    CERN Document Server

    Wan, A S; Scott, H A; Post, D; Rognlien, T D

    1995-01-01

    An effective radiative divertor maximizes the utilization of atomic processes to spread out the energy deposition to the divertor chamber walls and to reduce the peak heat flux. Because the mixture of neutral atoms and ions in the divertor can be optically thick to a portion of radiated power, it is necessary to accurately model the magnitude and distribution of line radiation in this complex region. To assess their importance we calculate the effects of radiation transport using CRETIN, a multi-dimensional, non-local thermodynamic equilibrium simulation code that includes the atomic kinetics and radiative transport processes necessary to model the complex environment of a radiative divertor. We also include neutral transport to model radiation from recycling neutral atoms. This paper presents a case study of a high-recycling radiative divertor with a typical large neutral pressure at the divertor plate to estimate the impact of H line radiation on the overall power balance in the divertor region with conside...

  15. Divertor asymmetry and scrape-off layer flow in various divertor configurations in Experimental Advanced Superconducting Tokamak

    DEFF Research Database (Denmark)

    Liu, S. C.; Guo, H. Y.; Xu, Guandong;

    2012-01-01

    plasmas exhibit the usual in-out asymmetry in particle and heat fluxes in LSN with the ion del B direction toward the lower X-point, favoring the outer divertor, especially at high density. The in-out asymmetry is reversed when changing the divertor configuration from LSN to USN, thus clearly...... demonstrating the effect of classical drifts. DN exhibits an even stronger in-out divertor asymmetry, favoring the outer divertor. A significant top-down asymmetry is also seen for DN, with greater particle and heat fluxes to the bottom divertor. In addition, the parallel plasma flow has been measured by a fast...

  16. The tungsten divertor experiment at ASDEX Upgrade

    Science.gov (United States)

    Neu, R.; Asmussen, K.; Krieger, K.; Thoma, A.; Bosch, H.-S.; Deschka, S.; Dux, R.; Engelhardt, W.; García-Rosales, C.; Gruber, O.; Herrmann, A.; Kallenbach, A.; Kaufmann, M.; Mertens, V.; Ryter, F.; Rohde, V.; Roth, J.; Sokoll, M.; Stäbler, A.; Suttrop, W.; Weinlich, M.; Zohm, H.; Alexander, M.; Becker, G.; Behler, K.; Behringer, K.; Behrisch, R.; Bergmann, A.; Bessenrodt-Weberpals, M.; Brambilla, M.; Brinkschulte, H.; Büchl, K.; Carlson, A.; Chodura, R.; Coster, D.; Cupido, L.; de Blank, H. J.; de Peña Hempel, S.; Drube, R.; Fahrbach, H.-U.; Feist, J.-H.; Feneberg, W.; Fiedler, S.; Franzen, P.; Fuchs, J. C.; Fußmann, G.; Gafert, J.; Gehre, O.; Gernhardt, J.; Haas, G.; Herppich, G.; Herrmann, W.; Hirsch, S.; Hoek, M.; Hoenen, F.; Hofmeister, F.; Hohenöcker, H.; Jacobi, D.; Junker, W.; Kardaun, O.; Kass, T.; Kollotzek, H.; Köppendörfer, W.; Kurzan, B.; Lackner, K.; Lang, P. T.; Lang, R. S.; Laux, M.; Lengyel, L. L.; Leuterer, F.; Manso, M. E.; Maraschek, M.; Mast, K.-F.; McCarthy, P.; Meisel, D.; Merkel, R.; Müller, H. W.; Münich, M.; Murmann, H.; Napiontek, B.; Neu, G.; Neuhauser, J.; Niethammer, M.; Noterdaeme, J.-M.; Pasch, E.; Pautasso, G.; Peeters, A. G.; Pereverzev, G.; Pitcher, C. S.; Poschenrieder, W.; Raupp, G.; Reinmüller, K.; Riedl, R.; Röhr, H.; Salzmann, H.; Sandmann, W.; Schilling, H.-B.; Schlögl, D.; Schneider, H.; Schneider, R.; Schneider, W.; Schramm, G.; Schweinzer, J.; Scott, B. D.; Seidel, U.; Serra, F.; Speth, E.; Silva, A.; Steuer, K.-H.; Stober, J.; Streibl, B.; Treutterer, W.; Troppmann, M.; Tsois, N.; Ulrich, M.; Varela, P.; Verbeek, H.; Verplancke, Ph; Vollmer, O.; Wedler, H.; Wenzel, U.; Wesner, F.; Wolf, R.; Wunderlich, R.; Zasche, D.; Zehetbauer, T.; Zehrfeld, H.-P.

    1996-12-01

    Tungsten-coated tiles, manufactured by plasma spray on graphite, were mounted in the divertor of the ASDEX Upgrade tokamak and cover almost 90% of the surface facing the plasma in the strike zone. Over 600 plasma discharges have been performed to date, around 300 of which were auxiliary heated with heating powers up to 10 MW. The production of tungsten in the divertor was monitored by a W I line at 400.8 nm. In the plasma centre an array of spectral lines at 5 nm emitted by ionization states around W XXX was measured. From the intensity of these lines the W content was derived. Under normal discharge conditions W-concentrations around 0741-3335/38/12A/013/img12 or even lower were found. The influence on the main plasma parameters was found to be negligible. The maximum concentrations observed decrease with increasing heating power. In several low power discharges accumulation of tungsten occurred and the temperature profile was flattened. The concentrations of the intrinsic impurities carbon and oxygen were comparable to the discharges with the graphite divertor. Furthermore, the density and the 0741-3335/38/12A/013/img13 limits remained unchanged and no negative influence on the energy confinement or on the H-mode threshold was found. Discharges with neon radiative cooling showed the same behaviour as in the graphite divertor case.

  17. Divertor erosion in DIII-D

    International Nuclear Information System (INIS)

    Net erosion rates of carbon target plates have been measured in situ for the DIII-D lower divertor. The principal method of obtaining this data is the DiMES sample probe. Recent experiments have focused on erosion at the outer strike-point of two divertor plasma conditions: (1) attached (Te > 40 eV) ELMing plasmas and (2) detached (Te 10 cm/year, even with incident heat flux 2. In this case, measurements and modeling agree for both gross and net carbon erosion, showing the near-surface transport and redeposition of the carbon is well understood and that effective sputtering yields are > 10%. In ELM-free discharges, this erosion rate can account for the rate of carbon accumulation in the core plasma. Divertor plasma detachment eliminates physical sputtering, while spectroscopically measured chemical erosion yields are also found to be low (Y(C/D+) ≤ 2.0 x 10-3). This leads to suppression of net erosion at the outer strike-point, which becomes a region of net redeposition (∼ 4 cm/year). The private flux wall is measured to be a region of net redeposition with dense, high neutral pressure, attached divertor plasmas. Leading edges intercepting parallel heat flux (∼ 50 MW/m2) have very high net erosion rates (∼ 10 microm/s) at the OSP of an attached plasma. Leading edge erosion, and subsequent carbon redeposition, caused by tile gaps can account for half of the deuterium codeposition in the DIII-D divertor

  18. Development of plasma control system for divertor configuration on QUEST

    International Nuclear Information System (INIS)

    A plasma control system to sustain divertor configurations is developed on QUEST (Q-shu university experiment with steady-state spherical tokamak). Magnetic fluxes are numerically integrated at 100 kHz using FPGA (Field-Programmable Gate Array) modules and transferred to a main calculation loop at 4 kHz. With these signals, plasma shapes are identified in real time at 2 kHz under the assumption that the plasma current can be represented as one filament current. This calculation is done in another calculation loop in parallel by taking advantage of a multi-core processor of the plasma control system. The inside and outside plasma edge positions are controlled to their target positions using PID (proportional-integral-derivative) control loops. Whereas the outside edge position can not be controlled by the outer PF coil current, the inside edge position can be controlled by the inner PF coil current

  19. Modeling detachment physics in the NSTX snowflake divertor

    Energy Technology Data Exchange (ETDEWEB)

    Meier, E.T., E-mail: emeier@wm.edu [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Bell, R.E.; Diallo, A.; Kaita, R.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, Princeton, NJ 08540 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Podestà, M. [Princeton Plasma Physics Laboratory, Princeton, NJ 08540 (United States); Rognlien, T.D.; Scotti, F. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States)

    2015-08-15

    The snowflake divertor is a proposed technique for coping with the tokamak power exhaust problem in next-step experiments and eventually reactors, where extreme power fluxes to material surfaces represent a leading technological and physics challenge. In lithium-conditioned National Spherical Torus Experiment (NSTX) discharges, application of the snowflake divertor typically induced partial outer divertor detachment and severalfold heat flux reduction. UEDGE is used to analyze and compare conventional and snowflake divertor configurations in NSTX. Matching experimental upstream profiles and divertor measurements in the snowflake requires target recycling of 0.97 vs. 0.91 in the conventional case, implying partial saturation of the lithium-based pumping mechanism. Density scans are performed to analyze the mechanisms that facilitate detachment in the snowflake, revealing that increased divertor volume provides most of the parallel heat flux reduction. Also, neutral gas power loss is magnified by the increased wetted area in the snowflake, and plays a key role in generating volumetric recombination.

  20. 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.

  1. Divertor and scoop limiter experiments on PDX

    Energy Technology Data Exchange (ETDEWEB)

    McGuire, K.; Beiersdorfer, P.; Bell, M.; Bol, K.; Boyd, D.; Buchenauer, D.; Budny, R.; Cavallo, A.; Couture, P.; Crowley, T.

    1985-01-01

    Routine operation in the enhanced energy confinement (or H-mode) regime during neutral beam injection was achieved by modifying the PDX divertor hardware to inhibit the influx of neutral gas from the divertor region to the main plasma chamber. A particle scoop limiter has been studied as a mechanical means of controlling particles at the plasma edge, and neutral beam heated discharges with this limiter show similar confinement times (normalized to tau/sub E//I/sub p/) to average H-mode plasmas. Two new instabilities are observed near the plasma edge in PDX during H-mode operation. The first, a quasicoherent fluctuation, occurred in bursts at well-defined frequencies (..delta omega../..omega.. less than or equal to 0.1) in the range 50 to 180 kHz, and had no obvious effects on confinement. The second instability, the edge relaxation phenomena (ERP), did cause deterioration in the global confinement time. The ERP's are characterized by sharp spikes in the divertor plasma density, H/sub ..cap alpha../ emission, and on the x-ray signals they appear as sawtoothlike relaxations at the plasma edge with an inversion radius near the separatrix. Attempts to obtain high ..beta../sub T/ in the H-mode discharges were hampered by a deterioration in the H-mode confinement and major disruptions which limited the achievable ..beta../sub T/. A study of the stability of both the limiter L-mode and divertor H-mode discharges close to the theoretical ..beta.. boundary, showed that the major disruptions observed there are sometimes caused by a fast growing m/n = 1/1 mode with no observable external precursor oscillations.

  2. Divertor and scoop limiter experiments on PDX

    International Nuclear Information System (INIS)

    Routine operation in the enhanced energy confinement (or H-mode) regime during neutral beam injection was achieved by modifying the PDX divertor hardware to inhibit the influx of neutral gas from the divertor region to the main plasma chamber. A particle scoop limiter has been studied as a mechanical means of controlling particles at the plasma edge, and neutral beam heated discharges with this limiter show similar confinement times (normalized to tau/sub E//I/sub p/) to average H-mode plasmas. Two new instabilities are observed near the plasma edge in PDX during H-mode operation. The first, a quasicoherent fluctuation, occurred in bursts at well-defined frequencies (Δω/ω less than or equal to 0.1) in the range 50 to 180 kHz, and had no obvious effects on confinement. The second instability, the edge relaxation phenomena (ERP), did cause deterioration in the global confinement time. The ERP's are characterized by sharp spikes in the divertor plasma density, H/sub α/ emission, and on the x-ray signals they appear as sawtoothlike relaxations at the plasma edge with an inversion radius near the separatrix. Attempts to obtain high β/sub T/ in the H-mode discharges were hampered by a deterioration in the H-mode confinement and major disruptions which limited the achievable β/sub T/. A study of the stability of both the limiter L-mode and divertor H-mode discharges close to the theoretical β boundary, showed that the major disruptions observed there are sometimes caused by a fast growing m/n = 1/1 mode with no observable external precursor oscillations

  3. Divertor Development for a Future Fusion Power Plant

    OpenAIRE

    Norajitra, Prachai

    2011-01-01

    The thesis begins by describing the fusion process and operation of a fusion reactor, the approach in the conceptual development of a helium-cooled divertor, and leads to the KIT helium-cooled modular divertor design. Then the methods of verification and validation of the design by tests are described, results presented and discussed. The developed divertor concept has demonstrated its principal functionality and hence the used design process and tools can be conceived as verified and validated.

  4. ADX - Advanced Divertor and RF Tokamak Experiment

    Science.gov (United States)

    Greenwald, Martin; Labombard, Brian; Bonoli, Paul; Irby, Jim; Terry, Jim; Wallace, Greg; Vieira, Rui; Whyte, Dennis; Wolfe, Steve; Wukitch, Steve; Marmar, Earl

    2015-11-01

    The Advanced Divertor and RF Tokamak Experiment (ADX) is a design concept for a compact high-field tokamak that would address boundary plasma and plasma-material interaction physics challenges whose solution is critical for the viability of magnetic fusion energy. This device would have two crucial missions. First, it would serve as a Divertor Test Tokamak, developing divertor geometries, materials and operational scenarios that could meet the stringent requirements imposed in a fusion power plant. By operating at high field, ADX would address this problem at a level of power loading and other plasma conditions that are essentially identical to those expected in a future reactor. Secondly, ADX would investigate the physics and engineering of high-field-side launch of RF waves for current drive and heating. Efficient current drive is an essential element for achieving steady-state in a practical, power producing fusion device and high-field launch offers the prospect of higher efficiency, better control of the current profile and survivability of the launching structures. ADX would carry out this research in integrated scenarios that simultaneously demonstrate the required boundary regimes consistent with efficient current drive and core performance.

  5. Westinghouse compact poloidal divertor reference design

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-08-01

    A feasible compact poloidal divertor system has been designed as an impurity control and vacuum vessel first-wall protection option for the TNS tokamak. The divertor coils are inside the TF coil array and vacuum vessel. The poloidal divertor is formed by a pair of coil sets with zero net current. Each set consists of a number of coils forming a dish-shaped washer-like ring. The magnetic flux in the space between the coil sets is compressed vertically to limit the height and to expand the horizontal width of the particle and energy burial chamber which is located in the gap between the coil sets. The intensity of the poloidal field is increased to make the pitch angle of the flux lines very large so that the diverted particles can be intercepted by a large number of panels oriented at a small angle with respect to the flux lines. They are carefully shaped and designed such that the entire surfaces are exposed to the incident particles and are not shadowed by each other. Large collecting surface areas can be obtained. Flowing liquid lithium film and solid metal panels have been considered as the particle collectors. The power density for the former is designed at 1 MW/m/sup 2/ and for the latter 0.5 MW/m/sup 2/. The major mechanical, thermal, and vacuum problems have been evaluated in sufficient detail so that the advantages and difficulties are identified. A complete functional picture is presented.

  6. FLP: a field line plotting code for bundle divertor design

    International Nuclear Information System (INIS)

    A computer code was developed to aid in the design of bundle divertors. The code can handle discrete toroidal field coils and various divertor coil configurations. All coils must be composed of straight line segments. The code runs on the PDP-10 and displays plots of the configuration, field lines, and field ripple. It automatically chooses the coil currents to connect the separatrix produced by the divertor to the outer edge of the plasma and calculates the required coil cross sections. Several divertor designs are illustrated to show how the code works

  7. Divertor bypass in the Alcator C-Mod tokamak

    Science.gov (United States)

    Pitcher, C. S.; LaBombard, B.; Danforth, R.; Pina, W.; Silveira, M.; Parkin, B.

    2001-01-01

    The Alcator C-Mod divertor bypass has for the first time allowed in situ variations to the mechanical baffle design in a tokamak. The design utilizes small coils which interact with the ambient magnetic field inside the vessel to provide the torque required to control small flaps of a Venetian blind geometry. Plasma physics experiments with the bypass have revealed the importance of the divertor baffling to maintain high divertor gas pressures. These experiments have also indicated that the divertor baffling has only a limited effect on the main chamber pressure in C-Mod.

  8. Influence of stray light for divertor spectroscopy in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Kajita, Shin, E-mail: kajita.shin@nagoya-u.jp [EcoTopia Science Institute, Nagoya University, Nagoya 464-8603 (Japan); Veshchev, Evgeny; Lisgo, Steve; Barnsley, Robin; Morgan, Philip; Walsh, Michael [ITER Organization, 13115 St. Paul-lez-Durance (France); Ogawa, Hiroaki; Sugie, Tatsuo; Itami, Kiyoshi [Japan Atomic Energy Agency, Naka, Ibaraki 801-1 (Japan)

    2015-08-15

    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.

  9. Magnetic geometry and particle source drive of supersonic divertor regimes

    International Nuclear Information System (INIS)

    We present a comprehensive picture of the mechanisms driving the transition from subsonic to supersonic flows in tokamak plasmas. We demonstrate that supersonic parallel flows into the divertor volume are ubiquitous at low density and governed by the divertor magnetic geometry. As the density is increased, subsonic divertor plasmas are recovered. On detachment, we show the change in particle source can also drive the transition to a supersonic regime. The comprehensive theoretical analysis is completed by simulations in ITER geometry. Such results are essential in assessing the divertor performance and when interpreting measurements and experimental evidence. (technical note)

  10. Divertor Heat Flux Mitigation in the National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V A; Maingi, R; Gates, D A; Menard, J E; Paul, S F; Raman, R; Roquemore, A L; Bell, M G; Bell, R E; Boedo, J A; Bush, C E; Kaita, R; Kugel, H W; LeBlanc, B P; Mueller, D

    2008-08-04

    Steady-state handling of divertor heat flux is a critical issue for both ITER and spherical torus-based devices with compact high power density divertors. Significant reduction of heat flux to the divertor plate has been achieved simultaneously with favorable core and pedestal confinement and stability properties in a highly-shaped lower single null configuration in the National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 2000] using high magnetic flux expansion at the divertor strike point and the radiative divertor technique. A partial detachment of the outer strike point was achieved with divertor deuterium injection leading to peak flux reduction from 4-6 MW m{sup -2} to 0.5-2 MW m{sup -2} in small-ELM 0.8-1.0 MA, 4-6 MW neutral beam injection-heated H-mode discharges. A self-consistent picture of outer strike point partial detachment was evident from divertor heat flux profiles and recombination, particle flux and neutral pressure measurements. Analytic scrape-off layer parallel transport models were used for interpretation of NSTX detachment experiments. The modeling showed that the observed peak heat flux reduction and detachment are possible with high radiated power and momentum loss fractions, achievable with divertor gas injection, and nearly impossible to achieve with main electron density, divertor neutral density or recombination increases alone.

  11. Plasma flow in the DIII-D divertor

    Energy Technology Data Exchange (ETDEWEB)

    Boedo, J.A. [Univ. of California, San Diego, CA (United States); Porter, G.D. [Lawrence Livermore National Lab., CA (United States); Schaffer, M.J. [General Atomics, San Diego, CA (United States)] [and others

    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.

  12. A time dependent 2D divertor code with TVD scheme for complex divertor configurations

    Science.gov (United States)

    Shimizu, K.; Takizuka, T.; Hirayama, T.

    1999-11-01

    In order to study the transport of heat and particles in the SOL and divertor plasmas, a two-dimensional divertor code, SOLDOR has been developed. The model used in this code is identical to the B2-code. Fluid equations are discretized in space under a non orthogonal mesh to treat accurately the W shape divertor configuration of JT-60U. The total variation diminishing scheme (TVD), which is a most familiar one in computational fluid dynamics, is applied for convective terms. The equations obtained by a finite volume method (FVM) are discretized in time with a full implicit scheme and are solved time-dependently using the Newton-Raphson method. The discretized equations are solved efficiently using approximate factorization method (AF). Test calculations in the slab geometry successfully reproduced the B2 results (B.J. Braams, NET report 1987) . We are going to apply this code to JT-60U divertor plasma and investigate the flow reversal and impurity transport.

  13. Overview of experiments with the dynamic ergodic divertor on TEXTOR

    NARCIS (Netherlands)

    Finken, K.H.; Abdullaev, S.; Biel, W.; de Bock, M. F. M.; Brezinsek, S.; Busch, C.; Classen, I.; Harting, D.; von Hellermann, M.; Jachmich, S.; Jakubowski, M.; R. Jaspers,; Koslowski, H. R.; Kramer-Flecken, A.; Kikuchi, Y.; Lehnen, M.; Liang, Y.; Kobayashi, M.; Nicolai, A.; Pospieszczyk, A.; Reiter, D.; Van Rompuy, T.; Samm, U.; Schmitz, O.; Sergienko, G.; Unterberg, B.; Westerhof, E.; R C Wolf,; Zimmermann, O.

    2006-01-01

    The Dynamic Ergodic Divertor (DED) has recently been taken into operation on TEXTOR. The device is rather flexible and allows the investigation of very different questions. In the present context we concentrate on the divertor aspect and on results of the m/n=12/4 base mode. The DED-field generates

  14. DIII-D radiative divertor project, status and plans

    International Nuclear Information System (INIS)

    New divertor hardware is being designed and fabricated for the Radiative Divertor modification of the DIII-D tokamak. The installation of the hardware has been separated into two phases, the first phase starting in October of 1996 and the and second final phase, in 1988. When completed, the Radiative Divertor Project hardware will provide pumping at all four strike points of a double-null, high triangularity discharge and provide baffling of the neutral particles from transport back to the core plasma. Radiative Divertor diagnostics are being designed to provide comprehensive measurements for diagnosing the divertor. Minimal modifications are required to diagnostics for the Phase I installation. More extensive diagnostic changes are planned for the Phase 2 installation. 3 refs., 6 figs

  15. Researches on the Neutral Gas Pressure in the Divertor Chamber of the HL-2A Tokamak

    Institute of Scientific and Technical Information of China (English)

    WANGMingxu; LIBo; YANGZhigang; YANLongwen; HONGWenyu; YUANBaoshan; LIULi; CAOZeng; CUIChenghe; LIUYong; WANGEnyao; ZHANGNianman

    2003-01-01

    The neutral gas pressure in divertor chamber is a very basic and important physics parameter because it determines the temperature of charged particles, the thermal flux density onto divertor plates, the erosion of divertor plates, impurity retaining and exhausting, particle transportation and confinement performance of plasma in tokamaks. Therefore, the pressure measurement in divertor chamber is taken into account in many large tokamaks.

  16. SOLPS Modeling of Slot Divertor Configuration on DIII-D

    Science.gov (United States)

    Sang, C. F.; Stangeby, P. C.; Guo, H. Y.; Lao, L. L.

    2015-11-01

    A major thrust of the DIII-D boundary/PMI initiative is to develop an advanced divertor configuration for next-step devices, such as FNSF and DEMO. We are adopting an integrated approach by optimizing both divertor structure and magnetic shape. Initial SOLPS modeling was carried out to optimize divertor structure shape to enhance divertor power dissipation, focusing on slot configurations. In particular, four different slot divertor structures, i.e., orthogonal-target slot, slanted-target slot, very narrow slot and v-shaped slot have been analyzed and comparisons made with an open divertor structure. It is found that the slot helps to trap recycling neutrals and impurities thus increasing radiative power dissipation in the divertor, reducing the electron temperature Te and the perpendicular heat flux q⊥ at the target plate. As expected, a narrower slot leads to lower Te and q⊥ than a less narrow one. The v-shaped slot appears to be especially effective at redirecting and concentrating recycling neutrals and impurities near the separatrix, thus promoting detachment at a lower upstream density than the other configurations. Work supported by US DOE under DE-FC02-04ER54698.

  17. Modeling of extinguishing ELMs in detached divertor plasmas

    Science.gov (United States)

    Pigarov, A.; Krasheninnikov, S.; Hollmann, E.; Rognlien, T.

    2015-11-01

    Detached plasmas, the primary operational regime for divertors in next-step fusion devices, should be compatible with both good H-mode confinement and relatively small ELMs providing tolerable heat power loads on divertor targets. Here, dynamics of boundary plasma, impurities and material walls over a sequence of many type-I ELM events under detached divertor plasma conditions is studied with UEGDE-MB-W, the newest version of 2D edge plasma transport code, which incorporates Macro-Blob (MB) approach to simulate non-diffusive filamentary transport and various ``Wall'' (W) models for time-dependent hydrogen wall inventory and recycling. We present the results of multi-parametric analysis on the impact of the size and frequency of ELMs on the divertor plasma parameters where we vary the MB characteristics under different pedestals and divertor configurations. We discuss the conditions, under which small but frequent type-I ELMs (typical for high-power H-mode discharges on current tokamaks with hard deuterium gas puff) are not ``burning through'' the formed detached divertor plasma. In this case, the inner and outer divertors are filled by sub-eV, recombining, highly-impure plasma. Variations of impurity plasma content, radiation pattern, and deuterium wall inventory over the ELM cycle are analyzed. UEDGE-MB-W modeling results are compared to available experimental data.

  18. EMC3-EIRENE modeling of toroidally-localized divertor gas injection experiments on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Lore, J.D., E-mail: lorejd@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Reinke, M.L. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); LaBombard, B. [Plasma Science and Fusion Center, MIT, Cambridge, MA 02139 (United States); Lipschultz, B. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Churchill, R.M. [Plasma Science and Fusion Center, MIT, Cambridge, MA 02139 (United States); Pitts, R.A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Feng, Y. [Max Planck Institute for Plasma Physics, Greifswald (Germany)

    2015-08-15

    Experiments on Alcator C-Mod with toroidally and poloidally localized divertor nitrogen injection have been modeled using the three-dimensional edge transport code EMC3-EIRENE to elucidate the mechanisms driving measured toroidal asymmetries. In these experiments five toroidally distributed gas injectors in the private flux region were sequentially activated in separate discharges resulting in clear evidence of toroidal asymmetries in radiated power and nitrogen line emission as well as a ∼50% toroidal modulation in electron pressure at the divertor target. The pressure modulation is qualitatively reproduced by the modeling, with the simulation yielding a toroidal asymmetry in the heat flow to the outer strike point. Toroidal variation in impurity line emission is qualitatively matched in the scrape-off layer above the strike point, however kinetic corrections and cross-field drifts are likely required to quantitatively reproduce impurity behavior in the private flux region and electron temperatures and densities directly in front of the target.

  19. DIVERT: a divertor magnetic field line following code

    International Nuclear Information System (INIS)

    The computer code DIVERT has been written to trace magnetic field lines in the presence of a divertor. Its purpose is to allow a user to estimate the thickness of the plasma scrapeoff region and to provide a visual mapping of the magnetic field lines near the divertor. Included in the code is the capability to provide auxiliary graphics and compute the field ripple. The code can handle a divertor made up of any arrangement of straight line coil segments and will provide a graph of the field line configuration on output

  20. ICRF system in the JET pumped divertor configuration

    International Nuclear Information System (INIS)

    After one year of operation the new A2 ICRF system installed for the pumped divertor phase of JET has achieved a coupled power of 16 MW, a coupled energy of 70 MJ, and combined heating with NB of 32 MW, each in ITER relevant high density, highly radiating, ELMy H-mode plasmas. The generators and antenna system now operates over 30 kV, with rapid, on-line matching and phase control of four coupled current straps. The coupling to the plasma and heating efficiency are phase dependent-coupling improves but heating efficiency falls with decreasing k parallel; good heating in monopole phasing has been observed after installation of a new separator between the set of antennae in one of the modules. Cross-coupling between straps enhances the imbalance in plasma coupling of the inner and outer straps of the array due to a mismatch in the feed lines. Modifications to reduce this imbalance and improve low k parallel operation are described. The A2 array is similar in size to one row of the current ITER in-port antenna design. The implications for such a design are discussed

  1. Ergodic divertor impact on Tore Supra edge

    International Nuclear Information System (INIS)

    The present ergodic divertor experiments in Tore Supra have been devoted to benchmarking the operational regimes of the apparatus. Two major effects are reported: on the one hand, strong changes occur in the ergodized boundary layer (up to 20% of the minor radius), and on the other hand, the central plasma and especially the confinement is not directly affected, i.e. the observed modifications are induced by edge effects. The basic trends, which are recorded are a decrease of both the edge electronic temperature and the edge density gradient while the radiated power is increased at the very edge of the ergodic region. The latter feature is in agreement with the impurity line emission characterized by an increase of the peripheral lines with a strong decrease of the central lines. (orig.)

  2. Neutral gas blanket effects in a gaseous divertor

    International Nuclear Information System (INIS)

    The gaseous divertor employs a neutral gas blanket to absorb the plasma heat flux in the divertor chamber. This novel method for resolving the heat loading problem in a conventional divertor system is simulated experimentally. In our operational range (nsub(e) 13 cm-3, Tsub(e) <= 5 eV) it is demonstrated that the localized plasma heat flux is scattered relatively uniformly with neutral pressures of a few microns. At large neutral pressures the plasma stream is neutralized without touching a material wall. Plasma pumping inhibits neutral backflow and can sustain a neutral pressure difference comparable to the plasma pressure. Effective divertor channel conductance is measured to be reduced by a factor of six. (orig.)

  3. Thermal Fatigue Study on the Divertor Plate Materials

    Institute of Scientific and Technical Information of China (English)

    吴继红; 张斧; 许增裕; 严建成

    2002-01-01

    Thermal fatigue property of the divertor plate is one of the key issues that governs the lifetime of the divertor plate. Taking tungsten as surface material, a small-mock-up divertor plate was made by hot isostatic press welding (HIP). A thermal cycling experiment for divertor mock-up was carried out in the vacuum, where a high-heat-flux electronic gun was used as the thermal source. A cyclic heat flux of 9 MW/m2 was loaded onto the mock-up, a heating duration of 20 s was selected, the cooling water flow rate was 80 ml/s. After 1000 cycles, the surface and the W/Cu joint of the mock-up did not show any damage. The SEM was used to analyze the microstructure of the welding joint, where no cracks were found also.

  4. Divertor IR thermography on Alcator C-Moda)

    Science.gov (United States)

    Terry, J. L.; LaBombard, B.; Brunner, D.; Payne, J.; Wurden, G. A.

    2010-10-01

    Alcator C-Mod is a particularly challenging environment for thermography. It presents issues that will similarly face ITER, including low-emissivity metal targets, low-Z surface films, and closed divertor geometry. In order to make measurements of the incident divertor heat flux using IR thermography, the C-Mod divertor has been modified and instrumented. A 6° toroidal sector has been given a 2° toroidal ramp in order to eliminate magnetic field-line shadowing by imperfectly aligned divertor tiles. This sector is viewed from above by a toroidally displaced IR camera and is instrumented with thermocouples and calorimeters. The camera provides time histories of surface temperatures that are used to compute incident heat-flux profiles. The camera sensitivity is calibrated in situ using the embedded thermocouples, thus correcting for changes and nonuniformities in surface emissivity due to surface coatings.

  5. Impact of a poloidal divertor in ignition tokamak design

    International Nuclear Information System (INIS)

    System design studies were performed to assess the effect of assuming a poloidal divertor instead of a limiter as a means of impurity control for ignition tokamak configurations. Results show that for the nominal Tokamak Fusion Core Experiment (TFCX) device with superconducting TF coils, a feasible poloidal divertor configuration can be obtained without increasing the major radius. In the TFCX nominal copper TF coil device, however, field limits at the PF coils are exceeded when the effects of asymmetry associated with a poloidal divertor are included. It was found that a 12% increase in the major radius of this device is necessary to simultaneously satisfy the plasma-shaping requirements of a poloidal divertor and the magnetics constraints at the superconducting PF coils

  6. Status of National Spherical Torus Experiment Liquid Lithium Divertor

    Science.gov (United States)

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

    2009-11-01

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

  7. Study of the feasibility of installing a toroidal or bundle divertor in EBT-S. Phase I: EBT-S divertor project. Final report

    International Nuclear Information System (INIS)

    The following chapters are included: (1) magnetic field analysis of the basic EBT-S geometry with and without aspect ratio enhancement coils; (2) analyses of a toroidal divertor for EBT-S; (3) analysis of a bundle divertor for EBT-S; (4) engineering; and (5) divertor vacuum pumping

  8. Design of divertor impurity monitoring system for ITER. 2

    Energy Technology Data Exchange (ETDEWEB)

    Sugie, Tatsuo; Ogawa, Hiroaki; Ebisawa, Katsuyuki; Ando, Toshiro; Kasai, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Katsunuma, Atsushi; Maruo, Mitsumasa; Kita, Yoshio

    1998-11-01

    The divertor impurity monitoring system of ITER has been designed. The main functions of this system are to identify impurity species and to measure the two-dimensional distributions of the particle influxes in the divertor plasmas. The wavelength range is 200 nm to 1000 nm. The viewing fans are realized by molybdenum mirrors located in the divertor cassette. With additional viewing fans seeing through the gap between the divertor cassettes, the region approximately from the divertor leg to the x-point will be observed. The light from the divertor region passes through the quartz windows on the divertor port plug and the cryostat, and goes through the dog-leg optics in the biological shield. Three different type of spectrometers: (i) survey spectrometers for impurity species monitoring, (ii) filter spectrometers for the particle influx measurement with the spatial resolution of 10 mm and the time resolution of 1 ms and (iii) high dispersion spectrometers for high resolution wavelength measurements are designed. These spectrometers are installed just behind the biological shield (for {lambda} < 450 nm) to prevent the transmission loss in fiber and in the diagnostic room (for {lambda} {>=} 450 nm) from the point of view of accessibility and flexibility. The optics have been optimized by a ray trace analysis. As a result, 10-15 mm spatial resolution will be achieved in all regions of the divertor. In addition, the measurable limit, the neutron and {gamma}-ray irradiation effect on windows, a calibration method, an alignment method, a remote handling method and a data acquisition method are considered. (author)

  9. Design of divertor impurity monitoring system for ITER

    International Nuclear Information System (INIS)

    The divertor impurity monitoring system of ITER has been designed. The main objectives of this system are to identify impurity species and to measure two-dimensional distributions of particle influxes in the divertor plasma. This system, which is one of the most important diagnostic systems for plasma control of ITER, is nominated for the start-up set of ITER diagnostics. The conceptual design, the optical design and the mechanical design are mainly carried out. In order to satisfy the required measurements, three deferent type of spectral systems are selected corresponding to each objectives. First is the spectral system for impurity species monitoring. Second is the spectral system for particle influx measurement with spatial and time resolution. Third is the spectral system with high dispersion for particle energy distribution measurement in the divertor. The divertor impurity monitoring system is composed of these three systems. The two-dimensional measurement in the divertor is carried out with two viewing fans intersected each other. These viewing fans are realized by metallic mirrors (made of molybdenum or copper) sitting in the divertor cassette. In the optical design, the optimization of the optical system from the divertor to the spectrometer are carried out by using ray trace analysis. As the result, it is difficult to satisfy the spatial resolution of 3 mm in the divertor region. About 10 mm resolution will be reasonable. In addition, the measurable limit, the neutron and γ-ray irradiation effect on the optical fiber, the remote handling concept and the space requirement are considered preliminarily. The necessary design works during EDA, and necessary R and D are also listed. (author)

  10. Non-ambipolar divertor flows in heliotron E

    International Nuclear Information System (INIS)

    The object of the work is to find out (1) the poloidal distributions of PEC in different poloidal cross-sections of the torus within one field period; (2) the link between PEC in the divertor flows (DF) and the characteristics of the divertor field lines; (3) the effect of different methods and regimes of heating on PEC. The data having been obtained enable us to understand at least partially the nature of PEC in the diverted plasma of H-E

  11. Diagnostics for the DIII-D radiative divertor

    Energy Technology Data Exchange (ETDEWEB)

    Nilson, D.G. [Lawrence Livermore National Lab., CA (United States); Brooks, N.H.; Smith, J.P.; Snider, R.T.

    1995-10-01

    This paper reviews the design of new diagnostics and the modifications to existing diagnostics needed to carry out radiative divertor experiments in DIII-D following installation in late 1996 of a set of baffle structures that will restrict the backflow to the core plasma of neutral deuterium atoms and impurity gases. The divertor slots formed by the new baffle structures will inhibit the easy view of the divertor legs and target plates that the open divertor geometry in DIII-D currently affords. We review a basic set of diagnostics that are needed to demonstrate the reduction of divertor heat loading and radiative dissipation of energy within the divertor. This will include IR cameras, bolometry, foil bolometers, and Langmuir probes. Within the limits of available funding, we will implement a supplemental set of instruments which provide a more detailed understanding of the underlying physical processes. Many existing diagnostics require only re-aiming to provide proper coverage of the initial 23 cm long divertor plasma configuration (X- point to floor distance). Other diagnostics need extensive reconfiguration using in-vessel fiber-optic bundles or high power laser mirrors. The new divertor baffle panels provide a protective shelf for diagnostic hardware mounted underneath them, but the water cooling channels in the panels limit the permissible size of through holes and, thereby, restrict the available views of under-the- baffle diagnostics. The successful resolution of the design and implementation of these diagnostic modifications is dependent on a strong coordination between GA and its many diagnostic collaborators.

  12. 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.

  13. Analytic 1D Approximation of the Divertor Broadening S in the Divertor Region for Conductive Heat Transport

    CERN Document Server

    Nille, Dirk; Eich, Thomas

    2016-01-01

    Topic is the divertor broadening $S$, being a result of perpendicular transport in the scrape-off layer and resulting in a better distribution of the power load onto the divertor target. Recent studies show a scaling of the divertor broadening with an inverse power law to the target temperature $T_t$, promising its reduction to be a way of distributing the power entering the divertor volume onto a large surface area. It is shown that for pure conductive transport in the divertor region the suggested inverse power law scaling to $T_t$ is only valid for high target electron temperatures. For decreasing target temperatures ($T_t < 20\\,$eV) the increase of $S$ stagnates and the conductive model results in a finite value of $S$ even for zero target temperature. It is concluded that the target temperature is no valid parameter for a power law scaling, as it is not representative for the entire divertor volume. This is shown in simulations solving the 2D heat diffusion equation, which is used as reference for an ...

  14. Variation of divertor plasma parameters with divertor depth for H-mode discharges in DIII-D

    International Nuclear Information System (INIS)

    We report here the results of experiments aimed at quantifying the advantages of increasing the X-point to target-plate distance in a divertor tokamak operating with H-mode confinement. Larger distances should lower the peak electron temperature at the target plates, thereby reducing sputtering and lowering the impurity concentration in the core plasma. When gas puffing is used to reduce the divertor heat flux, extra field-line length may increase the volume available for radiation and increase gas isolation between the core and divertor regions. These experiments were carried out using a lower single-null open divertor configuration (IP = 1.4 MA, BT = 2.1 T) with neutral beam heating (PNBI = 4.8 and 6.8 MW) to produce ELMing H-mode discharges lasting about 3 s. The X-point height (zx) was varied from 1.5-32 cm above the target plates by changing the plasma elongation on a shot by shot basis; the X-point radius was also varied in order to keep the outer strike point aligned with divertor Langmuir probe tips. Though there was no gas fueling during the H-mode phase of the discharge, the plasma density remained constant for all Zx obtained. Additional D2 gas puffing for radiative divertor experiments was applied for the last 1.5 s of the H-mode period. (author) 5 refs., 4 figs

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

    Science.gov (United States)

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

    2014-05-01

    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.

  16. Spectroscopic investigations of divertor detachment in TCV

    CERN Document Server

    Verhaegh, K; Duval, B P; Harrison, J R; Reimerdes, H; Theiler, C; Labit, B; Maurizio, R; Marini, C; Nespoli, F; Sheikh, U; Tsui, C K; Vianello, N; Vijvers, W A J

    2016-01-01

    The aim of this work is to provide an understanding of detachment at TCV with emphasis on analysis of the Balmer line emission. A new Divertor Spectroscopy System has been developed for this purpose. Further development of Balmer line analysis techniques has allowed detailed information to be extracted on free-free and three-body recombination. During density ramps, the plasma at the target detaches as inferred from a drop in density at, and ion current to, the target. At the same time the Balmer $6\\rightarrow2$ and $7\\rightarrow2$ line emission near the target is dominated by recombination, indicating that the ionization region has also detached from the target to be replaced by a recombining region with densities more than a factor 2 higher than at the target. As the core density increases further, the density and recombination rate are rising all along the outer leg to the x-point while remaining highest at the target. Even at the highest core densities accessed (Greenwald fraction 0.7) the peaks in recomb...

  17. Interpretation of the impurity distribution in the divertor during divertor plate biasing using the DIVIMP code

    Energy Technology Data Exchange (ETDEWEB)

    Haddad, E. E-mail: haddad@ccfm.ireq.ca; Meo, F.; Marchand, R.; Ratel, G.; Stansfield, B.L.; Gunn, J.; Stangeby, P.C.; Elder, J.D.; Lisgo, S.; Krieger, K

    2000-02-01

    Simulations of carbon transport using the DIVIMP code [P.C. Stangeby, J.D. Elder, J. Nucl. Mater. 196-198 (1992) 258] are compared with 2D toroidal images of CII and CIII radiation near the external divertor plates in TdeV ohmic plasmas (I{sub p}=170 kA, n{sub e}=3 x 10{sup 19} m{sup -3}, B{sub T}=1.4T). The main plasma parameters in the SOL and divertor are calculated by the onion skin model (OSM) [K. Shimizu et al., J. Nucl. Mater. 196-198 (1992) 476] included in DIVIMP, the neutrals being calculated by EIRENE [D. Reiter, Internal Report, KFA, Julich, 1947 (1984), 2599 (1992)] in an iterative loop. The results show that the carbon is mainly created by chemical sputtering, with a considerable fraction coming from the external oblique plate. By interpreting experimental CII and CIII distributions, it is found that carbon is affected by the biasing (-125 to +125 V) through a combination of at least three processes: the ion flux to the plates, the ExB drift velocity, and the cross field diffusion.

  18. Liquid Lithium Divertor Characteristics and Plasma-Material Interactions in NSTX High-Performance Plasmas

    International Nuclear Information System (INIS)

    Full text: ITER and future fusion experiments are hampered by erosion and degradation of plasma- facing components, forcing regular replacement. The conventional approach has been the use of high-Z walls (e.g., W) which can undergo permanent modification due to erosion and melting. One novel approach to solving these issues in the tokamak edge is the usage of liquid metal plasma facing components. The National Spherical Torus Experiments (NSTX) is the only US confinement device operating a liquid metal divertor target to examine the technological and scientific aspects of this innovative approach. The Liquid Lithium Divertor (LLD) module formed a nearly toroidally continuous surface in the outer, lower divertor. NSTX H-mode discharges were repeatedly run with the outer strike-point directly on the LLD plates. Peak heat fluxes of ∼ 5 MW/m2 were regularly applied to the LLD surfaces alongside significant ion fluxes. No molybdenum line radiation was observed in these plasma [3] indicating protection of the substrate material. During these experiments, no macroscopic ejection was observed from the LLD contrary to experiments conducted in the DIII-D tokamak, where lithium ejection exposed the substrate [4]. Quiescent scrape-off layer current (SOLC) densities were ∼ 10 kA/m2, with peak SOLCs > 100 kA/m2 . Stability analyses for the liquid metal layers show that despite the large current densities, capillary and viscous forces are effective at reducing motion demonstrating stable operation of the liquid metal PFC. The strong chemical reactivity of lithium results in the steady accumulation of impurities in the PFC material, mitigating the low-Z benefits of the lithium. Eroded material from the carbon PFCs in NSTX can redeposit onto the LLD, and background vacuum gases are also gettered onto the surface. Flowing systems are under study and are designed to allow one to obtain a low-Z, replenishable PFC by removing gettered materials and eliminating the accumulation

  19. Upgraded divertor Thomson scattering system on DIII-D

    Science.gov (United States)

    Glass, F.; Carlstrom, T. N.; Du, D.; McLean, A. G.; Taussig, D. A.; Boivin, R. L.

    2016-11-01

    A design to extend the unique divertor Thomson scattering system on DIII-D to allow measurements of electron temperature and density in high triangularity plasmas is presented. Access to this region is selectable on a shot-by-shot basis by redirecting the laser beam of the existing divertor Thomson system inboard — beneath the lower floor using a moveable, high-damage threshold, in-vacuum mirror — and then redirecting again vertically. The currently measured divertor region remains available with this mirror retracted. Scattered light is collected from viewchords near the divertor floor using in-vacuum, high temperature optical elements and relayed through the port window, before being coupled into optical fiber bundles. At higher elevations from the floor, measurements are made by dynamically re-focusing the existing divertor system collection optics. Nd:YAG laser timing, analysis of the scattered light spectrum via polychromators, data acquisition, and calibration are all handled by existing systems or methods of the current multi-pulse Thomson scattering system. Existing filtered polychromators with 7 spectral channels are employed to provide maximum measurement breadth (Te in the range of 0.5 eV-2 keV, ne in the range of 5 × 1018-1 × 1021 m3) for both low Te in detachment and high Te measurement up beyond the separatrix.

  20. Coherence imaging of flows in the DIII-D divertor

    Energy Technology Data Exchange (ETDEWEB)

    Howard, J.; Diallo, A.; Creese, M. [Plasma Research Laboratory, The Australian National University, Canberra (Australia); Allen, S.L.; Ellis, R.M.; Meyer, W.; Fenstermacher, M.E.; Porter, G.D. [Lawrence Livermore National Laboratory at General Atomics, San Diego (United States); Brooks, N.H.; Van Zeeland, M.E.; Boivin, R.L. [General Atomics, San Diego (United States)

    2011-03-15

    Various spatial heterodyne polarization interferometers for spectrally-resolved optical imaging of edge and core parameters in high temperature magnetized plasmas are described. Applications for such ''coherence imaging'' (CI) systems include imaging motional Stark effect and Zeeman effect polarimetry for determination of the magnetic field pitch angle, and passive and active (charge exchange recombination spectroscopy - CXRS) Doppler imaging of plasma temperature and flow. In this paper we describe spatial heterodyne coherence imaging systems and present first results of Doppler flow imaging in the DIII-D divertor. Instruments have been installed for imaging flows in the divertor and scrape-off-layer in the DIII-D tokamak and also for Doppler imaging on the H-1 heliac [1]. In the former case, single snapshot interferometric images of the plasma in CII 514nm, and CIII 465nm emission have been demodulated to obtain flow and ion temperature projections in both the scrape-off-layer and divertor. Flow field amplitudes in the divertor are found to be broad agreement with UEDGE modeling [2], and point the way towards experiments that address important divertor transport issues in future (copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  1. Plasma transport in a simulated magnetic-divertor configuration

    Energy Technology Data Exchange (ETDEWEB)

    Strawitch, C. M.

    1981-03-01

    The transport properties of plasma on magnetic field lines that intersect a conducting plate are studied experimentally in the Wisconsin internal ring D.C. machine. The magnetic geometry is intended to simulate certain aspects of plasma phenomena that may take place in a tokamak divertor. It is found by a variety of measurements that the cross field transport is non-ambipolar; this may have important implications in heat loading considerations in tokamak divertors. The undesirable effects of nonambipolar flow make it preferable to be able to eliminate it. However, we find that though the non-ambipolarity may be reduced, it is difficult to eliminate entirely. The plasma flow velocity parallel to the magnetic field is found to be near the ion acoustic velocity in all cases. The experimental density and electron temperature profiles are compared to the solutions to a one dimensional transport model that is commonly used in divertor theory.

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

    International Nuclear Information System (INIS)

    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

  3. Understanding atomic hydrogen behaviour in pumped divertor plasmas

    International Nuclear Information System (INIS)

    In order to set up a data base and diagnostic capability for understanding atomic hydrogen behaviour in pumped divertor plasmas, an experiment and a feasibility study using a novel laser-induced fluorescence (LIF) technique were performed. For the former, combined measurements of LIF tuned to Hα and emission intensities at Hα/Hβ were carried out on the compact helical system (CHS). The comparison of the measured data and a particle simulation code revealed atomic hydrogen behaviour quantitatively, providing a full estimate of toroidally and poloidally asymmetric distributions of hydrogen atoms. In order to supplement data base around the pumped divertor region, the applicability of an LIF technique which uses two-photon excitation from the ground state examined, based on the real optical constraints of the envisaged JET pumped divertor. It was concluded that ii is feasible and the only remaining problem is not a serious one. (orig.)

  4. Modeling of Alcator C-Mod Divertor Baffling Experiments

    Energy Technology Data Exchange (ETDEWEB)

    D. P. Stotler; C. S. Pitcher; C. J. Boswell; T. K. Chung; B. LaBombard; B. Lipschultz; J. L. Terry; R. J. Kanzleiter

    2000-11-29

    A specific Alcator C-Mod discharge from the series of divertor baffling experiments is simulated with the DEGAS 2 Monte Carlo neutral transport code. A simple two-point plasma model is used to describe the plasma variation between Langmuir probe locations. A range of conductances for the bypass between the divertor plenum and the main chamber are considered. The experimentally observed insensitivity of the neutral current flowing through the bypass and of the D alpha emissions to the magnitude of the conductance is reproduced. The current of atoms in this regime is being limited by atomic physics processes and not the bypass conductance. The simulated trends in the divertor pressure, bypass current, and D alpha emission agree only qualitatively with the experimental measurements, however. Possible explanations for the quantitative differences are discussed.

  5. Plasma transport in a simulated magnetic-divertor configuration

    International Nuclear Information System (INIS)

    The transport properties of plasma on magnetic field lines that intersect a conducting plate are studied experimentally in the Wisconsin internal ring D.C. machine. The magnetic geometry is intended to simulate certain aspects of plasma phenomena that may take place in a tokamak divertor. It is found by a variety of measurements that the cross field transport is non-ambipolar; this may have important implications in heat loading considerations in tokamak divertors. The undesirable effects of nonambipolar flow make it preferable to be able to eliminate it. However, we find that though the non-ambipolarity may be reduced, it is difficult to eliminate entirely. The plasma flow velocity parallel to the magnetic field is found to be near the ion acoustic velocity in all cases. The experimental density and electron temperature profiles are compared to the solutions to a one dimensional transport model that is commonly used in divertor theory

  6. Radiative divertor plasmas with convection in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Leornard, A.W. [General Atomics, San Diego, CA (United States); Porter, G.D.; Wood, R.D. [Lawrence Livermore National Lab., CA (United States)] [and others

    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.

  7. Hydrogen recycling and transport in the helical divertor of TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Clever, Meike

    2010-07-01

    The aim of this thesis was to investigate the hydrogen recycling at the target plates of the helical divertor in TEXTOR and by this the capability of this divertor configuration to access such favourable operational regimes. In order to study the different divertor density regimes in TEXTOR, discharges were performed in which the total plasma density was increased continuously up to the density limit. The recycling was investigated in a fixed helical divertor structure where four helical strike points with a poloidal width of about 8-10 cm are created at the divertor target plates. The experimental investigation of the hydrogen recycling was carried out using mainly spectroscopic methods supplemented by Langmuir probe, interferometric and atomic beam measurements. In the framework of this thesis a spectroscopic multi camera system has been built that facilitates the simultaneous observation of four different spectral lines, recording images of the divertor target plates and the plasma volume close to the target. The system facilitates the simultaneous measurement of the poloidal and toroidal pattern of the recycling flux at the divertor target without the need for sweeping the plasma structure. The simultaneous observation of different spectral lines reduces the uncertainty in the analysis based on several lines, as the contribution from uncertainties in the reproducibility of plasma parameters in different discharges are eliminated and only the uncertainty of the measurement method limits the accuracy. The spatial resolution of the system in poloidal and toroidal direction (0.8 mm{+-}0.01 mm) is small compared to the separation of the helical strike points, the capability of the measurement method to resolve these structures is therefore limited by the line-of-sight integration and the penetration depth of the light emitting species. The measurements showed that the recycling flux increases linearly with increasing plasma density, a high recycling regime is not

  8. Operational boundaries on the stellarator W7-AS at the beginning of the divertor experiments

    International Nuclear Information System (INIS)

    During the last shutdown the stellarator W7-AS underwent two major modifications: First, the limiters were replaced by ten divertor modules, and the diagnostic set associated with the plasma boundary and target plate regions was greatly expanded. Secondly, the previously counter tangential neutral beam injector box was shifted to a co-position. Thus, the heating efficiency should be considerably increased at low magnetic fields and high densities. After resuming experiments these improvements will be used to test the boundary island divertor concept and further expand operational boundaries during the remaining experimental time until permanent shutdown in 2002. The present operational boundaries are reviewed with respect to the stability of high β and density limit discharges. Discharges with good confinement properties will be discussed where further progress was achieved after installing control coils to modify the size and properties of vacuum field islands. In contrast to the usual net-current free mode, W7-AS also allows operation at large toroidal currents. In this way disruption-like events in the presence of rather large external poloidal fields can be produced. (author)

  9. Thermal Analysis of the Divertor Primary Heat Transfer System Piping During the Gas Baking Process

    Energy Technology Data Exchange (ETDEWEB)

    Yoder Jr, Graydon L [ORNL; Harvey, Karen [ORNL; Ferrada, Juan J [ORNL

    2011-02-01

    A preliminary analysis has been performed examining the temperature distribution in the Divertor Primary Heat Transfer System (PHTS) piping and the divertor itself during the gas baking process. During gas baking, it is required that the divertor reach a temperature of 350 C. Thermal losses in the piping and from the divertor itself require that the gas supply temperature be maintained above that temperature in order to ensure that all of the divertor components reach the required temperature. The analysis described in this report was conducted in order to estimate the required supply temperature from the gas heater.

  10. 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

    to those encountered with the axisymmetric divertors. The inhomogeneities have been discussed thanks to the probe measurements at 14 locations of the divertor modules. (author) 71 refs.

  11. Particle collection by the ergodic divertor of Tore Supra: high recycling and partially detached plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Azeroual, A.; Pegourie, B.; Chareyre, E.; Guirlet, R.; Gunn, J.P.; Loarer, T.; Ghendrih, P.; Granata, G

    1999-10-15

    The detached plasma regime is envisaged as an operating condition for next step tokamaks. In this regime, the edge incident ion flux decreases dramatically and goes down to nearly zero on the wall. A proposed method of pumping is to capture backscattered neutrals du atomic processes, namely FC (Franck Condon) dissociation and CX (charge exchange), with a dedicated vented structure. This solution has been tested with the Ergodic Divertor (ED) of Tore Supra. In this configuration, the field lines connect to neutralizers located between the divertor current bars (7 neutralizers per module - 6 identical modules distributed toroidally around the vacuum chamber). These neutralizers are vented structures which are semi-transparent to neutrals and can be used for particle pumping. A comprehensive set of diagnostics has been installed: 14 Langmuir probes poloidally and toroidally distributed, pressure gauges in the modules plenum and D{sub {alpha}} measurements along the neutralizer located in the equatorial plane. In the present study, we analyse transient experiments performed for density scan studies and concentrate on high recycling and partially detached plasmas. For known density and temperature profiles at the edge, two quantities are sufficient to describe the neutral recirculation and particle balance: the pressure in the ED plenum (which characterizes the particle collection) and the distribution of the D{sub {alpha}} emission line in front of the neutralizers (which characterizes the ion source due to recycling). Their behaviour during the high recycling and semi-detached phases is described in the next sections. Scaling laws of the edge parameters and pressure in the pumping chamber with volume-averaged density and total power are given in the last section. (authors)

  12. Effects of divertor plate biasing on radial and poloidal edge fluxes in the TdeV

    International Nuclear Information System (INIS)

    The divertor plates of TdeV, a tokamak with a double-null divertor and closed divertor chambers, have been electrically biased with respect to the walls. The paper discusses the resulting effects on the edge electron density profile, on the neutral pressures and impurity fluxes in the main vacuum chamber and the divertor chambers, and on the plasma flow to the divertors. As a function of the bias voltage, which was varied between - 180 V and + 160 V, the electron density scrape-off width and the wall impurity influxes increase monotonically; the flows to the top and bottom divertors vary strongly, in qualitative agreement with an E-vector x B-vector/B2 rotation, but not symmetrically. With negative biasing, the electrostatic barrier and the rotation combine to give a strong improvement of the divertor efficiency. (author). 30 refs, 10 figs

  13. Controlled detachment and particle transport in the divertor plasma in TdeV

    International Nuclear Information System (INIS)

    At high densities, the plasma detaches from the outboard divertor plates in TdeV. The signatures are a reduction of the ion flux to the divertor plate, movement of the radiating zone from the plate toward the X-point, a pressure gradient between an ionization front and the target plate, and strong cross-field transport in the divertor. A toroidally-viewing TV imaging system allows us to observe local interactions between the divertor plasma and the different divertor plates. As the plasma detaches, the gas pressure in the divertor continues to rise, and there is evidence for molecular processes in the cold plasma near the divertor plates. Auxiliary heating increases the power and particle flow across the separatrix; our results suggest that detachment depends on the energy transported per particle. Simulations using the B2/EIRENE and DIVIMP codes give reasonable agreement with the measurements for the attached phase. (orig.)

  14. Controlled detachment and particle transport in the divertor plasma in TdeV

    Energy Technology Data Exchange (ETDEWEB)

    Stansfield, B.L. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Meo, F. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Abel, G. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Boucher, C. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Gauvreau, J.-L. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Gunn, J.P. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Haddad, E. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Lachambre, J.-L. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Mailloux, J. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Marchand, R. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Ratel, G. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Richard, N. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Shoucri, M.M. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Terreault, B. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Beaudry, S. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Decoste, R. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Pacher, G.W. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Zuzak, W. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Elder, J.D. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Stangeby, P.C. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada)

    1997-02-01

    At high densities, the plasma detaches from the outboard divertor plates in TdeV. The signatures are a reduction of the ion flux to the divertor plate, movement of the radiating zone from the plate toward the X-point, a pressure gradient between an ionization front and the target plate, and strong cross-field transport in the divertor. A toroidally-viewing TV imaging system allows us to observe local interactions between the divertor plasma and the different divertor plates. As the plasma detaches, the gas pressure in the divertor continues to rise, and there is evidence for molecular processes in the cold plasma near the divertor plates. Auxiliary heating increases the power and particle flow across the separatrix; our results suggest that detachment depends on the energy transported per particle. Simulations using the B2/EIRENE and DIVIMP codes give reasonable agreement with the measurements for the attached phase. (orig.).

  15. 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.

  16. Material and design considerations for the carbon armored ITER divertor

    International Nuclear Information System (INIS)

    The properties of materials for the carbon armored ITER divertor were evaluated from literature and manufacturers' documentation. Most of these data, however, have been not known or not published yet. We have evaluated an optimum data set of the candidate materials of the ITER divertor, which were needed for finite element analyses (FEM). The materials evaluated are as follows; MFC-1, CX2002U, SEP-N112, P-130, IG-430U for the carbon based materials, and Oxygen Free Copper (OFCu), Dispersion Strengthened Copper (DSCu), TZM, W5Re and W-Cu as a heat sink material. It should be noted that W-Cu is first proposed for a heat sink application of the ITER divertor plate. The finite element analyses were performed for the residual stress induced by brazing, thermal response and thermal stresses under a uniform heat flux of 15 MW/m2 to the plasma facing surface. The stress free temperature of 750degC is assumed for the residual stress by brazing. Ten different geometries of the divertor were considered in the analyses including possible material combinations. The FEM results show that the material combinations of MFC-1 and W-30Cu or DSUc in the flat-plate geometry satisfy the presently accepted ITER requirements. The combinations of CX2002U and TZM or W5Re is considered a good choice in terms of residual and thermal stresses, whereas the surface temperature exceeds the ITER requirements. (author) 106 refs

  17. High confinement dissipative divertor operation on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Goetz, J.A.; LaBombard, B.; Lipschultz, B.; Pitcher, C.S.; Terry, J.L.; Boswell, C.; Gangadhara, S.; Pappas, D.; Weaver, J.; Welch, B.; Boivin, R.L.; Bonoli, P.; Fiore, C.; Granetz, R.; Greenwald, M.; Hubbard, A.; Hutchinson, I.; Irby, J.; Marmar, E.; Mossessian, D.; Porkolab, M.; Rice, J.; Rowan, W.L.; Schilling, G.; Snipes, J.; Takase, Y.; Wolfe, S.; Wukitch, S. [Plasma Science Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States)

    1999-05-01

    Alcator C-Mod [I. H. Hutchinson {ital et al.}, Phys. Plasmas {bold 1}, 1511 (1994)] has operated a High-confinement-mode (H-mode) plasma together with a dissipative divertor and low core Z{sub eff}. The initially attached plasma is characterized by steady-state enhancement factor, H{sub ITER89P} [P. N. Yushmanov {ital et al.}, Nucl. Fusion {bold 30}, 1999 (1990)], of 1.9, central Z{sub eff} of 1.1, and a radiative fraction of {approximately}50{percent}. Feedback control of a nitrogen gas puff is used to increase radiative losses in both the core/edge and divertor plasmas in almost equal amounts. Simultaneously, the core plasma maintains H{sub ITER89P} of 1.6 and Z{sub eff} of 1.4 in this nearly 100{percent} radiative state. The power and particle flux to the divertor plates have been reduced to very low levels while the core plasma is relatively unchanged by the dissipative nature of the divertor. {copyright} {ital 1999 American Institute of Physics.}

  18. Modeling results for a linear simulator of a divertor

    Energy Technology Data Exchange (ETDEWEB)

    Hooper, E.B.; Brown, M.D.; Byers, J.A.; Casper, T.A.; Cohen, B.I.; Cohen, R.H.; Jackson, M.C.; Kaiser, T.B.; Molvik, A.W.; Nevins, W.M.; Nilson, D.G.; Pearlstein, L.D.; Rognlien, T.D.

    1993-06-23

    A divertor simulator, IDEAL, has been proposed by S. Cohen to study the difficult power-handling requirements of the tokamak program in general and the ITER program in particular. Projections of the power density in the ITER divertor reach {approximately} 1 Gw/m{sup 2} along the magnetic fieldlines and > 10 MW/m{sup 2} on a surface inclined at a shallow angle to the fieldlines. These power densities are substantially greater than can be handled reliably on the surface, so new techniques are required to reduce the power density to a reasonable level. Although the divertor physics must be demonstrated in tokamaks, a linear device could contribute to the development because of its flexibility, the easy access to the plasma and to tested components, and long pulse operation (essentially cw). However, a decision to build a simulator requires not just the recognition of its programmatic value, but also confidence that it can meet the required parameters at an affordable cost. Accordingly, as reported here, it was decided to examine the physics of the proposed device, including kinetic effects resulting from the intense heating required to reach the plasma parameters, and to conduct an independent cost estimate. The detailed role of the simulator in a divertor program is not explored in this report.

  19. Modeling results for a linear simulator of a divertor

    International Nuclear Information System (INIS)

    A divertor simulator, IDEAL, has been proposed by S. Cohen to study the difficult power-handling requirements of the tokamak program in general and the ITER program in particular. Projections of the power density in the ITER divertor reach ∼ 1 Gw/m2 along the magnetic fieldlines and > 10 MW/m2 on a surface inclined at a shallow angle to the fieldlines. These power densities are substantially greater than can be handled reliably on the surface, so new techniques are required to reduce the power density to a reasonable level. Although the divertor physics must be demonstrated in tokamaks, a linear device could contribute to the development because of its flexibility, the easy access to the plasma and to tested components, and long pulse operation (essentially cw). However, a decision to build a simulator requires not just the recognition of its programmatic value, but also confidence that it can meet the required parameters at an affordable cost. Accordingly, as reported here, it was decided to examine the physics of the proposed device, including kinetic effects resulting from the intense heating required to reach the plasma parameters, and to conduct an independent cost estimate. The detailed role of the simulator in a divertor program is not explored in this report

  20. Electron and molecular ion collisions relevant to divertor plasma

    International Nuclear Information System (INIS)

    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)

  1. Plasma/neutral gas transport in divertors and limiters

    International Nuclear Information System (INIS)

    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/m2 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

  2. Overview of experiments with the dynamic ergodic divertor on TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Finken, K.H.; Abdullaev, S.; Biel, W.; Brezinsek, S.; Busch, C.; Harting, D.; Jakubowski, M.; Koslowski, H.R.; Kraemer-Flecken, A.; Kikuchi, Y.; Lehnen, M.; Liang, Y.; Nicolai, A.; Pospieszczyk, A. [Institut fuer Plasmaphysik, Forschungszentrum Juelich GmbH, EURATOM Association, Trilateral Euregio Cluster, D-52425 Juelich (Germany); Bock, M.F.M. de; Classen, I.; Hellermann, M. von; Jaspers, R. [FOM-Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, Trilateral Euregio Cluster, P.O. Box: 1207, NL-3430 BE Nieuwegein (Netherlands); Jachmich, S. [Laboratory for Plasma Physics, Association EURATOM - Belgian State, KMS - ERM, Trilateral Euregio Cluster, B-1000 Brussels (Belgium); Kobayashi, M. [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki-shi 509-52 Toki (Japan); Reiter, D.; Rompuy, T. van; Samm, U.; Schmitz, O.; Sergienko, G.; Unterberg, B.; Westerhof, E.; Wolf, R.C.; Zimmermann, O.

    2006-09-15

    The Dynamic Ergodic Divertor (DED) has recently been taken into operation on TEXTOR. The device is rather flexible and allows the investigation of very different questions. In the present context we concentrate on the divertor aspect and on results of the m/n=12/4 base mode. The DED-field generates the proper ergodic zone and an area of open magnetic field lines, the laminar zone and the tangle structure. The properties of the laminar zone resemble the divertor region of a poloidal divertor. However, the distribution of the density and temperature is highly 3D and strongly related to the structure of the laminar and ergodic zones. The structures of the heat and particle fluxes to the wall agree well with the predicted patterns. A prominent feature of the ergodization is the creation of an edge electric field which results in a rotation of the plasma. (copyright 2006 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  3. Feedback system for divertor impurity seeding based on real-time measurements of surface heat flux in the Alcator C-Mod tokamak

    Science.gov (United States)

    Brunner, D.; Burke, W.; Kuang, A. Q.; LaBombard, B.; Lipschultz, B.; Wolfe, S.

    2016-02-01

    Mitigation of the intense heat flux to the divertor is one of the outstanding problems in fusion energy. One technique that has shown promise is impurity seeding, i.e., the injection of low-Z gaseous impurities (typically N2 or Ne) to radiate and dissipate the power before it arrives to the divertor target plate. To this end, the Alcator C-Mod team has created a first-of-its-kind feedback system to control the injection of seed gas based on real-time surface heat flux measurements. Surface thermocouples provide real-time measurements of the surface temperature response to the plasma heat flux. The surface temperature measurements are inputted into an analog computer that "solves" the 1-D heat transport equation to deliver accurate, real-time signals of the surface heat flux. The surface heat flux signals are sent to the C-Mod digital plasma control system, which uses a proportional-integral-derivative (PID) algorithm to control the duty cycle demand to a pulse width modulated piezo valve, which in turn controls the injection of gas into the private flux region of the C-Mod divertor. This paper presents the design and implementation of this new feedback system as well as initial results using it to control divertor heat flux.

  4. Feedback system for divertor impurity seeding based on real-time measurements of surface heat flux in the Alcator C-Mod tokamak.

    Science.gov (United States)

    Brunner, D; Burke, W; Kuang, A Q; LaBombard, B; Lipschultz, B; Wolfe, S

    2016-02-01

    Mitigation of the intense heat flux to the divertor is one of the outstanding problems in fusion energy. One technique that has shown promise is impurity seeding, i.e., the injection of low-Z gaseous impurities (typically N2 or Ne) to radiate and dissipate the power before it arrives to the divertor target plate. To this end, the Alcator C-Mod team has created a first-of-its-kind feedback system to control the injection of seed gas based on real-time surface heat flux measurements. Surface thermocouples provide real-time measurements of the surface temperature response to the plasma heat flux. The surface temperature measurements are inputted into an analog computer that "solves" the 1-D heat transport equation to deliver accurate, real-time signals of the surface heat flux. The surface heat flux signals are sent to the C-Mod digital plasma control system, which uses a proportional-integral-derivative (PID) algorithm to control the duty cycle demand to a pulse width modulated piezo valve, which in turn controls the injection of gas into the private flux region of the C-Mod divertor. This paper presents the design and implementation of this new feedback system as well as initial results using it to control divertor heat flux.

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

    International Nuclear Information System (INIS)

    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

  6. Preliminary concept design of the divertor remote handling system for DEMO power plant

    International Nuclear Information System (INIS)

    Highlights: • Concept design of the RH system for the DEMO fusion power plant. • Divertor Mover: Hydraulic telescopic boom concept design. An alternative solution to ITER rack and pinion divertor mover (CMM). • Divertor cassettes end effector studies. • Transportation cask conceptual studies and logistic. - Abstract: This paper is based on the remote maintenance system project (WPRM) for the demonstration fusion power reactor (DEMO). Following ITER, DEMO aims to confirm the capability of generating several hundred of MW of net electricity by 2050. The main objective of these activities is to develop an efficient and reliable remote handling (RH) system for replacing the divertor cassettes. This paper presents the preliminary results of the concept design of the divertor RH system. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections of 4 m each, and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel. Two alternative design of the end effector to grip and manipulate the divertor cassette are also presented in this work. Both the concepts are hydraulically actuated, basing on the ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate. The main objective of this paper is to illustrate the feasibility of DEMO divertor remote maintenance operations

  7. Diagnostic options for radiative divertor feedback control on NSTX-Ua)

    Science.gov (United States)

    Soukhanovskii, V. A.; Gerhardt, S. P.; Kaita, R.; McLean, A. G.; Raman, R.

    2012-10-01

    A radiative divertor technique is used in present tokamak experiments and planned for ITER to mitigate high heat loads on divertor plasma-facing components (PFCs) to prevent excessive material erosion and thermal damage. In NSTX, a large spherical tokamak with lithium-coated graphite PFCs and high divertor heat flux (qpeak ⩽ 15 MW/m2), radiative divertor experiments have demonstrated a significant reduction of divertor peak heat flux simultaneously with good core H-mode confinement using pre-programmed D2 or CD4 gas injections. In this work diagnostic options for a new real-time feedback control system for active radiative divertor detachment control in NSTX-U, where steady-state peak divertor heat fluxes are projected to reach 20-30 MW/m2, are discussed. Based on the NSTX divertor detachment measurements and analysis, the control diagnostic signals available for NSTX-U include divertor radiated power, neutral pressure, spectroscopic deuterium recombination signatures, infrared thermography of PFC surfaces, and thermoelectric scrape-off layer current. In addition, spectroscopic "security" monitoring of possible confinement or pedestal degradation is recommended. These signals would be implemented in a digital plasma control system to manage the divertor detachment process via an actuator (impurity gas seeding rate).

  8. Diagnostic options for radiative divertor feedback control on NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V. A.; McLean, A. G. [Lawrence Livermore National Laboratory, Livermore, California, 94550 (United States); Gerhardt, S. P.; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Raman, R. [University of Washington, Seattle, Washington 98195 (United States)

    2012-10-15

    A radiative divertor technique is used in present tokamak experiments and planned for ITER to mitigate high heat loads on divertor plasma-facing components (PFCs) to prevent excessive material erosion and thermal damage. In NSTX, a large spherical tokamak with lithium-coated graphite PFCs and high divertor heat flux (q{sub peak} Less-Than-Or-Slanted-Equal-To 15 MW/m{sup 2}), radiative divertor experiments have demonstrated a significant reduction of divertor peak heat flux simultaneously with good core H-mode confinement using pre-programmed D{sub 2} or CD{sub 4} gas injections. In this work diagnostic options for a new real-time feedback control system for active radiative divertor detachment control in NSTX-U, where steady-state peak divertor heat fluxes are projected to reach 20-30 MW/m{sup 2}, are discussed. Based on the NSTX divertor detachment measurements and analysis, the control diagnostic signals available for NSTX-U include divertor radiated power, neutral pressure, spectroscopic deuterium recombination signatures, infrared thermography of PFC surfaces, and thermoelectric scrape-off layer current. In addition, spectroscopic 'security' monitoring of possible confinement or pedestal degradation is recommended. These signals would be implemented in a digital plasma control system to manage the divertor detachment process via an actuator (impurity gas seeding rate).

  9. Diagnostic options for radiative divertor feedback control on NSTX-U

    International Nuclear Information System (INIS)

    A radiative divertor technique is used in present tokamak experiments and planned for ITER to mitigate high heat loads on divertor plasma-facing components (PFCs) to prevent excessive material erosion and thermal damage. In NSTX, a large spherical tokamak with lithium-coated graphite PFCs and high divertor heat flux (qpeak⩽ 15 MW/m2), radiative divertor experiments have demonstrated a significant reduction of divertor peak heat flux simultaneously with good core H-mode confinement using pre-programmed D2 or CD4 gas injections. In this work diagnostic options for a new real-time feedback control system for active radiative divertor detachment control in NSTX-U, where steady-state peak divertor heat fluxes are projected to reach 20–30 MW/m2, are discussed. Based on the NSTX divertor detachment measurements and analysis, the control diagnostic signals available for NSTX-U include divertor radiated power, neutral pressure, spectroscopic deuterium recombination signatures, infrared thermography of PFC surfaces, and thermoelectric scrape-off layer current. In addition, spectroscopic “security” monitoring of possible confinement or pedestal degradation is recommended. These signals would be implemented in a digital plasma control system to manage the divertor detachment process via an actuator (impurity gas seeding rate).

  10. Divertor detachment, He exhaust and compact toroid injection on TdeV

    International Nuclear Information System (INIS)

    Progressive detachment with increasing density is shown to proceed with a marked reduction of the ion flux to the divertor plates, a pressure gradient between a ionization front and the plate, and strong cross-field transport in the divertor. The divertor He exhaust is not affected by detachment although the He enrichment remains low but constant. A moderate density of n-bare ∼ 5 x 1019 m-3 seems to be sufficient both for efficient peak power load reduction at the plate and good He exhaust through the divertor. Simulations indicate possible divertor geometry improvements which will soon be verified experimentally in the new TdeV-96 divertor upgrade. Finally, central fuelling with compact toroid injection is reported with no detrimental effects on the plasma. (author). 16 refs, 8 figs

  11. Tungsten spectroscopy relevant to the diagnostics development of ITER divertor plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Clementson, J; Beiersdorfer, P; Magee, E W; McLean, H S; Wood, R D

    2009-12-01

    The ITER tokamak will have tungsten divertor tiles and, consequently, the divertor plasmas are expected to contain tungsten ions. The spectral emission from these ions can serve to diagnose the divertor for plasma parameters such as tungsten concentrations, densities, ion and electron temperatures, and flow velocities. The ITER divertor plasmas will likely have densities around 10{sup 14-15} cm{sup -3} and temperatures below 150 eV. These conditions are similar to the plasmas at the Sustained Spheromak Physics Experiment (SSPX) in Livermore. To simulate ITER divertor plasmas, a tungsten impurity was introduced into the SSPX spheromak by prefilling it with tungsten hexacarbonyl prior to the usual hydrogen gas injection and initiation of the plasma discharge. The possibility of using the emission from low charge state tungsten ions to diagnose tokamak divertor plasmas has been investigated using a high-resolution extreme ultraviolet spectrometer.

  12. An automated approach to magnetic divertor configuration design

    Science.gov (United States)

    Blommaert, M.; Dekeyser, W.; Baelmans, M.; Gauger, N. R.; Reiter, D.

    2015-01-01

    Automated methods based on optimization can greatly assist computational engineering design in many areas. In this paper an optimization approach to the magnetic design of a nuclear fusion reactor divertor is proposed and applied to a tokamak edge magnetic configuration in a first feasibility study. The approach is based on reduced models for magnetic field and plasma edge, which are integrated with a grid generator into one sensitivity code. The design objective chosen here for demonstrative purposes is to spread the divertor target heat load as much as possible over the entire target area. Constraints on the separatrix position are introduced to eliminate physically irrelevant magnetic field configurations during the optimization cycle. A gradient projection method is used to ensure stable cost function evaluations during optimization. The concept is applied to a configuration with typical Joint European Torus (JET) parameters and it automatically provides plausible configurations with reduced heat load.

  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. An automated approach to magnetic divertor configuration design

    OpenAIRE

    Blommaert, Maarten; Dekeyser, Wouter; Baelmans, Martine; Gauger, Nicolas Ralph; Reiter, Detlev

    2015-01-01

    Automated methods based on optimization can greatly assist computational engineering design in many areas. In this paper an optimization approach to the magnetic design of a nuclear fusion reactor divertor is proposed and applied to a tokamak edge magnetic configuration in a first feasibility study. The approach is based on reduced models for magnetic field and plasma edge, which are integrated with a grid generator into one sensitivity code. The design objective chosen here for demonstrat...

  15. Modeling of a poloidally symmetric toroidal field divertor in a reversed--field-pinch plasma machine

    International Nuclear Information System (INIS)

    Magnetic divertors have been shown to be successful in minimizing plasma-wall interactions and in leading to high confinement regimes in Tokamaks. This leads to the hope that similar benefits may occur in an Reversed-Field-Pinch (RPF) fitted with a divertor. Previous experiments using divertors in a RFP have used a poloidal field divertor configuration such as is used in Tokamaks. This study investigates another approach; namely a toroidal field divertor. In this study a simple model of a poloidally symmetric toroidal field divertor is developed and used in a study of stochastic effects due to the divertor and in a 3-D magnetohydrodynamic (MHD) code to study the response of the plasma to the large poloidal m = 0 perturbations caused by the divertor coils. It is found that the topology of the RFP-divertor system is much more complex than had been expected. Stochasticity is enhanced in the outer edge region of the plasma because of this geometrical complexity. The way of the RFP reaches an equilibrium in this complex system is investigated with the 3-D relaxation code, DEBS (authored by Dalton Schnack). This code showed that the divertor will not hinder the formation of a reversed toroidal field in the plasma, and that the dynamics of its formation is altered when toroidal effects are considered. The plasma develops flows and currents in the throat of the divertor in response to the vacuum-like divertor fields. These flows and currents help to restore the force free character of the plasma

  16. Effect of nozzle sizes on jet impingement heat transfer in He-cooled divertor

    OpenAIRE

    Končar, Boštjan; Norajitra, Prachai; Oblak, Klemen

    2009-01-01

    Abstract The use of impinging jets for divertor cooling in the conceptual fusion power plant is attracting much attention due to its very high heat removal capability and moderate pumping power requirement. The latest and the most advanced divertor concept is based on modular design cooled by helium impinging jets. To reduce the thermal stresses, the plasma-facing side of the divertor is build up of numerous small cooling fingers cooled by an array of helium jets. In this study the...

  17. Radiative divertor plasmas with convection in DIII-D

    International Nuclear Information System (INIS)

    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

  18. 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.)

  19. Particle recirculation in the ergodic divertor of Tore Supra

    International Nuclear Information System (INIS)

    The present paper addresses the issue of particle recirculation in discharges where low-energy flux to ergodic divertor target plates is achieved in highly-radiating detached ohmic plasmas. Plasma temperature and particle flux are measured by flush-mounted probes in the divertor plates and by an upstream fast scanning Mach probe. The scalings with core density of the ion flux and electron temperature are well described by the simple two-point model used in axisymmetric poloidal divertors. The detachment signature is a pressure drop that occurs when the edge temperature falls below 10 eV. The parallel ion flux gradient is always positive, indicating that recombination is unlikely to play an important role in detachment. Visible spectroscopy of a neutralizer plate shows that attainment of cold detached plasmas near the density limit coincides with an abrupt increase of fuelling efficiency for both deuterium and impurities. A feedback algorithm based on real-time Langmuir probe measurements has been developed to monitor detachment and avoid disruptions. (author)

  20. Copper matrix composites as heat sink materials for water-cooled divertor target

    OpenAIRE

    Jeong-Ha You

    2015-01-01

    According to the recent high heat flux (HHF) qualification tests of ITER divertor target mock-ups and the preliminary design studies of DEMO divertor target, the performance of CuCrZr alloy, the baseline heat sink material for DEMO divertor, seems to only marginally cover the envisaged operation regime. The structural integrity of the CuCrZr heat sink was shown to be affected by plastic fatigue at 20 MW/m². The relatively high neutron irradiation dose expected for the DEMO divertor target is ...

  1. The effect of the magnetic topology on particle recycling in the ergodic divertor of TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Lehnen, M. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany)]. E-mail: m.lehnen@fz-juelich.de; Abdullaev, S.S. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Brezinsek, S. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Finken, K.H. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Harting, D. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Hellermann, M. von [FOM-Rijnhuizen, Association EURATOM-FOM (Netherlands); Jakubowski, M.W. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Jaspers, R. [FOM-Rijnhuizen, Association EURATOM-FOM (Netherlands); Kirschner, A. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Pospieszczyk, A. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Reiter, D. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Samm, U. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Schmitz, O. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Sergienko, G. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Unterberg, B. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Wolf, R. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany)

    2007-06-15

    The influence of the divertor geometry of the dynamic ergodic divertor (DED) in TEXTOR on particle recycling is discussed. The geometry can be varied by the choice of the base mode, the edge safety factor and the divertor coil current. The divertor volume is split into the upstream and the downstream area. Strong plasma flows in the downstream area, essential for high screening efficiency, are predicted. The source strength of deuterium and carbon in the downstream area is estimated by using the two-dimensional distribution of D{sub {alpha}} and CIII emission in front of the target. The results are compared to EMC3 and ERO-code calculations.

  2. HHF test with 80x80x1 Be/Cu/SS Mock-ups for verifying the joining technology of the ITER blanket First Wall

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Bae, Young Dug; Kim, Suk Kwon; Hong, Bong Guen; Jeong, Yong Hwan; Park, Jeong Yong; Choi, Byung Kwon; Jung, Hyun Kyu

    2008-11-15

    Through the fabrication of the Cu/SS and Be/Cu joint specimens, fabrication procedure such as material preparation, canning, degassing, HIP (Hot Isostatic Pressing), PHHT (Post HIP heat treatment) was established. The HIP conditions (1050 .deg. C, 100 MPa 2 hr for Cu/SS, 580 .deg. C 100 MPa 2 hr for Be/Cu) were developed through the investigation on joint specimen fabricated with the various HIP conditions; the destructive tests of joint include the microstructure observation of the interface with the examination of the elemental distribution, tension test, bend test, Charpy impact test and fracture toughness test. However, since the joint should be tested under the High Heat Flux (HHF) conditions like the ITER operation for verifying its joint integrity, several HHF tests were performed like the previous HHF test with the Cu/SS, Be/Cu, Be/Cu/SS Mock-ups. In the present study, the HHF test with Be/Cu/SS Mock-ups, which have 80 mm x 80 mm single Be tile and each material depths were kept to be the same as the ITER blanket FW. The Mock-ups fabricated with three kinds of interlayers such as Cr/Ti/Cu, Ti/Cr/Cu, Ti/Cu, which were different from the developed interlayer (Cr/Cu), total 6 Mock-ups were fabricated. Preliminary analysis were performed to decide the test conditions; they were tested with up to 2.5 MW/m2 of heat fluxes and 20 cycles for each Mock-up in a given heat flux. They were tested with JUDITH-1 at FZJ in Germany. During tests, all Mock-ups showed delamination or full detachment of Be tile and it can be concluded that the joints with these interlayers have a bad joining but it can be used as a good data for developing the Be/Cu joint with HIP.

  3. HHF test with 80x80x1 Be/Cu/SS Mock-ups for verifying the joining technology of the ITER blanket First Wall

    International Nuclear Information System (INIS)

    Through the fabrication of the Cu/SS and Be/Cu joint specimens, fabrication procedure such as material preparation, canning, degassing, HIP (Hot Isostatic Pressing), PHHT (Post HIP heat treatment) was established. The HIP conditions (1050 .deg. C, 100 MPa 2 hr for Cu/SS, 580 .deg. C 100 MPa 2 hr for Be/Cu) were developed through the investigation on joint specimen fabricated with the various HIP conditions; the destructive tests of joint include the microstructure observation of the interface with the examination of the elemental distribution, tension test, bend test, Charpy impact test and fracture toughness test. However, since the joint should be tested under the High Heat Flux (HHF) conditions like the ITER operation for verifying its joint integrity, several HHF tests were performed like the previous HHF test with the Cu/SS, Be/Cu, Be/Cu/SS Mock-ups. In the present study, the HHF test with Be/Cu/SS Mock-ups, which have 80 mm x 80 mm single Be tile and each material depths were kept to be the same as the ITER blanket FW. The Mock-ups fabricated with three kinds of interlayers such as Cr/Ti/Cu, Ti/Cr/Cu, Ti/Cu, which were different from the developed interlayer (Cr/Cu), total 6 Mock-ups were fabricated. Preliminary analysis were performed to decide the test conditions; they were tested with up to 2.5 MW/m2 of heat fluxes and 20 cycles for each Mock-up in a given heat flux. They were tested with JUDITH-1 at FZJ in Germany. During tests, all Mock-ups showed delamination or full detachment of Be tile and it can be concluded that the joints with these interlayers have a bad joining but it can be used as a good data for developing the Be/Cu joint with HIP

  4. Divertor plasma conditions and neutral dynamics in horizontal and vertical divertor configurations in JET-ILW low confinement mode plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Groth, M., E-mail: mathias.groth@aalto.fi [Aalto University, Association EURATOM-Tekes, Otakaari 4, Espoo (Finland); Brezinsek, S. [Forschungszentrum Jülich, IEK4 – Plasma Physik, Jülich (Germany); Belo, P. [Institute of Plasmas and Nuclear Fusion, Association EURATOM/IST, Lisbon (Portugal); Culham Centre for Fusion Energy, Association EURATOM-CCFE, Abingdon (United Kingdom); Brix, M. [Culham Centre for Fusion Energy, Association EURATOM-CCFE, Abingdon (United Kingdom); Calabro, G. [Association EURATOM-ENEA, Frascati (Italy); Chankin, A. [Max-Planck Institute for Plasma Physics, EURATOM Association, Garching (Germany); Clever, M.; Coenen, J.W. [Forschungszentrum Jülich, IEK4 – Plasma Physik, Jülich (Germany); Corrigan, G. [Culham Centre for Fusion Energy, Association EURATOM-CCFE, Abingdon (United Kingdom); Drewelow, P. [Max-Planck-Institute for Plasma Physics, EURATOM Association, Greifswald (Germany); Guillemaut, C. [Association EURATOM CEA, CEA/DSM/IRFM, Cadarache (France); Harting, D. [Culham Centre for Fusion Energy, Association EURATOM-CCFE, Abingdon (United Kingdom); Huber, A. [Forschungszentrum Jülich, IEK4 – Plasma Physik, Jülich (Germany); Jachmich, S. [Association ‘Euratom-Belgian state’, Ecole Royale Militaire, Brussels (Belgium); Järvinen, A. [Aalto University, Association EURATOM-Tekes, Otakaari 4, Espoo (Finland); Kruezi, U.; Lawson, K.D. [Culham Centre for Fusion Energy, Association EURATOM-CCFE, Abingdon (United Kingdom); Lehnen, M. [Forschungszentrum Jülich, IEK4 – Plasma Physik, Jülich (Germany); ITER Organisation, 13115 Saint-Paul-Lez-Durance (France); and others

    2015-08-15

    Measurements of the plasma conditions at the low field side target plate in JET ITER-like wall ohmic and low confinement mode plasmas show minor differences in divertor plasma configurations with horizontally and vertically inclined targets. Both the reduction of the electron temperature in the vicinity of the strike points and the rollover of the ion current to the plates follow the same functional dependence on the density at the low field side midplane. Configurations with vertically inclined target plates, however, produce twice as high sub-divertor pressures for the same upstream density. Simulations with the EDGE2D-EIRENE code package predict significantly lower plasma temperatures at the low field side target in vertical than in horizontal target configurations. Including cross-field drifts and imposing a pumping by-pass leak at the low-field side plate can still not recover the experimental observations.

  5. Investigation of scrape-off layer and divertor heat transport in ASDEX Upgrade L-mode

    Science.gov (United States)

    Sieglin, B.; Eich, T.; Faitsch, M.; Herrmann, A.; Scarabosio, A.; the ASDEX Upgrade Team

    2016-05-01

    Power exhaust is one of the major challenges for the development of a fusion power plant. Predictions based upon a multimachine database give a scrape-off layer power fall-off length {λq}≤slant 1 mm for large fusion devices such as ITER. The power deposition profile on the target is broadened in the divertor by heat transport perpendicular to the magnetic field lines. This profile broadening is described by the power spreading S. Hence both {λq} and S need to be understood in order to estimate the expected divertor heat load for future fusion devices. For the investigation of S and {λq} L-Mode discharges with stable divertor conditions in hydrogen and deuterium were conducted in ASDEX Upgrade. A strong dependence of S on the divertor electron temperature and density is found which is the result of the competition between parallel electron heat conductivity and perpendicular diffusion in the divertor region. For high divertor temperatures it is found that the ion gyro radius at the divertor target needs to be considered. The dependence of the in/out asymmetry of the divertor power load on the electron density is investigated. The influence of the main ion species on the asymmetric behaviour is shown for hydrogen, deuterium and helium. A possible explanation for the observed asymmetry behaviour based on vertical drifts is proposed.

  6. L-H power threshold studies with tungsten/carbon divertor on the EAST tokamak

    DEFF Research Database (Denmark)

    Chen, L.; Xu, G. S.; Gao, W.;

    2016-01-01

    The power threshold for low (L) to high (H) confinement mode transition achieved by radio-frequency heating and molybdenum first wall with lithium coating has been experimentally investigated on the EAST tokamak for two sets of divertor geometries and materials: tungsten/carbon divertor and full...

  7. Divertor coil power supply in Aditya Tokamak for improved plasma operation

    International Nuclear Information System (INIS)

    The existing Aditya tokamak, a medium sized tokamak with limiter configuration is being upgraded to a Tokamak with divertor configuration. This moderate field Tokamak is capable of producing 250 kA of plasma current with 300 ms duration. Two new sets of diverter coils will be added to the system with an objective of producing double null plasmas in Aditya Upgrade Tokamak. Diverter coils, made up of continuously transposed conductor, are low voltage high current carrying poloidal field coils. One set of inner divertor coil has radius of 460 mm containing 6 turns and the other set of 1075 mm radius coil with 1 turn makes the outer divertor coils. The simulated plasma double null equilibrium demands 150 kAT of NI for the inner divertor coils and 10 - 20 kAT of NI for outer divertor coils. To energize the divertor coils with required power, a pulsed DC power supply of 3 MW (100V, 30 kA) has been designed. The designed pulsed DC power supply will be a 3-phase, 12-pulse rectifier based convertor power supply having a duty cycle of 300 ms on-time and 15 minutes off-time. The current rise time in the divertor coils will be ∼ 0.6 MA/sec. Detailed design of the divertor power supply with active controls for real time control of the plasma shape will be discussed in this paper. (author)

  8. Compatibility of the radiating divertor with high performance plasmas in DIII-D

    International Nuclear Information System (INIS)

    Full text: We report on recent DIII-D experiments that successfully applied a radiating divertor scenario to high performance 'hybrid' plasmas [T.C. Luce, et al., Nucl. Fusion 43 (2003) 321]. In the puff-and-pump approach [M.J. Schaffer, et al., Nucl. Mater. 241-243 (1997) 585] used here, argon was injected near the outer divertor target, plasma flows into both the inner and outer divertors were enhanced by a combination of particle pumping near both divertor targets and deuterium gas puffing upstream of the divertor targets, and a 'dome' structure in the private flux region isolated the inner divertor from the outer divertor. Good hybrid conditions were maintained (e.g. energy confinement time normalized to ITER89p ≥ 2 and normalized plasma β ≅ 2.4), and the argon accumulation in the main plasma was modest. The peak heat flux at the outer divertor target was reduced by a factor of ≅ 2.5, while the peak heat flux at the inner target fell by only ∼20%. This was largely due to a much higher argon concentration near the outer divertor target than near the inner target (∼7 times). Exhaust enrichment (ER) as high as 64 were obtained, and ER was insensitive to the argon injection rate. (ER is defined as the ratio of the neutral argon pressure in the baffle plenum to the atomic-equivalent pressure of deuterium in the baffle plenum, divided by the ratio of argon density to electron density in the main plasma.) The asymmetry in the argon distribution and the favorable enrichment values arose largely from the closed and partitioned divertor geometry and from the frictional forces due to the enhanced divertor flow, which impeded the escape of argon from the outer divertor. Although the argon density profiles were more peaked than the electron profiles at high argon injection rates, the emissivity profiles in the main plasma remained 'hollow'. Our results suggest that independent control of both the radiating properties at the inner and outer divertor targets can be

  9. Alcator C-Mod: A high-field divertor tokamak

    Science.gov (United States)

    Lipschultz, B.; Becker, H.; Bonoli, P.; Coleman, J.; Fiore, C.; Golovato, S.; Granetz, R.; Greenwald, M.; Gwinn, D.; Humphries, D.; Hutchinson, I.; Irby, J.; Marmar, E.; Montgomery, D. B.; Najmabadi, F.; Parker, R.; Porkolab, M.; Rice, J.; Sevillano, E.; Takase, Y.; Terry, J.; Watterson, R.; Wolfe, S.

    1989-04-01

    The Alcator C-Mod tokamak is a new device presently under construction at Massachusetts Institute of Technology (M.I.T.) which is scheduled to begin operation in mid-1990. The projected operating parameters are as follows: Toroidal field of 9 T; Ip ≤ 3 MA, R = 66.5 cm, a = 21 cm, κ ≤ 2.0, δ ≤ 0.5, ne ≤ 10 21m-3, PICRF ≤ 6 MW. The divertor configuration includes mechanical baffling as opposed to an 'open' geometry. Under strictly ohmic heating conditions, central Ti and Te are predicted to be in the range 2.5-3.5 keV over the density range (4-8) × 10 20m-3. With the addition of 6 MW of ICRF heating, Ti should vary from 4-8 keV over the same density range (assuming either Kaye-Goldston or Neo-Alcator scalings for electron confinement). Based on edge plasma characterizations from Alcator-C and divertor tokamaks, the scrape-off layer (SOL) properties are predicted to be: λn ≈ 10mm, density at the divertor plate < 2 × 10 21m-3, H 0 ionization mean free path between 1 and 10 mm. Maximum heat loads on various internal components are predicted to be in the range 5-10 MW/m 2. The flexibility of the poloidal field system in forming a number of flux surface geometries will provide further comparisons of the relative impurity control capabilities of double-null, single-null and limiter plasmas.

  10. Divertor retention of metallic impurities during neutralization plate biasing on TdeV

    International Nuclear Information System (INIS)

    Laser ablation injection of aluminium is used to measure the retention of metallic impurities in the lower poloidal divertor of TdeV. A detailed calibration of the ablation process allows the determination of the quantity and velocity distribution of the injected particles. The experiment measures the flow of the injected particles from the divertor to the main plasma. Negative biasing of the divertor neutralization plates is shown to improve the retention in the active divertor by a factor of at least four at -200 V. A simple model is developed to show that the improved confinement is due to the increased poloidal flux to the divertor during biasing. (author). 32 refs, 9 figs

  11. Analysis on EAST LHCD operation space by using simple Core-SOL-Divertor model

    International Nuclear Information System (INIS)

    A simple Core-SOL-Divertor model (CSD model) has been developed to investigate qualitatively the overall features of the operational space for the integrated core and edge plasma. In the CSD model, the core plasma model of ITER physics guidelines and the two-point SOL-divertor model are applied. This CSD model is validated by the two dimensional divertor transport code (B2-EIRINE) and by the JT-60U divertor recycling database, and this model is applicable to the low- and high-recycling state of the divertor plasma. The CSD model is applied to the study of the EAST operational space with lower hybrid current drive experiments under various kinds of trade-off for the basic plasma parameters, and the relationship between the operational space and the plasma discharge duration is also discussed. (author)

  12. Analysis on EAST LHCD Operation Space by Using Simple Core-SOL-Divertor Model

    International Nuclear Information System (INIS)

    A simple core-SOL-divertor model (CSD model) was developed to investigate qualitatively the overall features of the operational space for the integrated core and edge plasma. In the CSD model, the core plasma model of ITER physics guidelines and the two-point SOL-divertor model are applied. This CSD model is validated by the two dimensional divertor transport code (B2-EIRINE) and by the JT-60U divertor recycling database, and this model is applicable to the low- and high-recycling state of the divertor plasma. The CSD model is applied to the study of the EAST operational space with lower hybrid current drive under various kinds of trade-off for the basic plasma parameters, and the relationship between the operational space and the plasma discharge duration is also discussed. (magnetically confined plasma)

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

    International Nuclear Information System (INIS)

    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

  14. 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.

  15. Equilibrium configuration for a high current pumped divertor

    International Nuclear Information System (INIS)

    A realistic design of a pumped divertor plasma configuration to be fitted to the JET vessel can be obtained as a compromise among various geometrical, physical and technical constraints. The possibility of reaching a satisfactory solution has been analysed for plasmas up to 6 MA. Optimisation of the plasma coupling to the RF antennae requires a largely asymmetric distribution of ampere turns in the PF coils and some mechanical flexibility. The calculations presented were carried out using the specially developed JET equilibrium and configuration analysis codes. (U.K.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Allen, S.L.; Hill, D.N.; Porter, G.D. [and others

    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 ({delta} {approximately} 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{sub 2} injection in DN high-{delta} ELMing H-mode have shown that this configuration is more sensitive to gas puffing ({tau} decreases). Moving the X-point away from the target plate (to {approximately}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-{delta} DN divertor. To date, however, both the DIII-D divertor diagnostics and pump were optimized for lower single-null (LSN) low-{delta} ({delta}{approximately} 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-{delta} USN divertor baffle and pump; these results are discussed in Section 3. Both divertor and core parameters are discussed in each case.

  17. Power balance in the divertor-tokamak DIVA

    International Nuclear Information System (INIS)

    Power balances of Ohmically and radio-frequency (RF) heated plasmas including a boundary (scrape-off layer) plasma are investigated in the divertor-tokamak DIVA. First, methods of measurement of the boundary plasma are described. These are applied to the divertor plasma in the case of Ohmic heating. The results clarify characteristics of the boundary plasma of a conventional tokamak. The scaling law for the boundary plasma is derived in consideration of the power balance including the boundary plasma. Heat flux to material surfaces is investigated in detail; the relationship between heat flux, particle flux and electron velocity distribution is clarified. Gross power balance is investigated by measurements of total heat flux to the wall and total radiation loss including charge-exchange loss. These results provide experimental evidence for the above scaling law. Finally, power balance during the Ion-Cyclotron Range of Frequency (ICRF) heating is described. Optimum heating conditions of the ICRF heating in the two-ion hybrid regime are surveyed. For the optimum heating conditions, gross power balance including the boundary plasma is considered, in which the heating efficiency is derived. Radial profile of the RF-heating power, the ratio of the heating power to each species and the transport of RF-heated ions are clarified in the power balance. (author)

  18. Tore Supra divertor screening efficiency during density regime experiments

    International Nuclear Information System (INIS)

    The Tore Supra ergodic divertor (ED) screening efficiency has been investigated in density regime experiments. The ED screening efficiency is analysed by using the 'tightness' concept, which is the ratio of the density on the ED neutraliser plates to the volume averaged plasma density. Tightness is studied as a function of different plasma edge parameters, such as Tdiv, ED magnetic perturbation (Δ), plasma composition, location of recycling source, and additional power. Tightness is shown to increase with Δ, Pdiv0.55/(1-Fr)1.22, and 1/Tdiv0.5. These trends are well explained by a simple 0-D model, where the particle confinement time in the ergodized peripheral region is very small. Finally, tightness increases with the power conducted onto the ED plates. Since ED plasmas have low Pdiv, their tightness value remains low compared to that obtained with axisymmetric divertors for which Pdiv is considerably larger. Increasing Pdiv will result in an improved tightness and a better particle control

  19. Tore Supra divertor screening efficiency during density regime experiments

    Energy Technology Data Exchange (ETDEWEB)

    Grisolia, C. E-mail: grisolia@drfc.cad.cea.fr; Ghendrih, Ph.; Gunn, J.; Loarer, T.; Monier-Garbet, P.; De Michelis, C.; Costanzo, L.; Pascal, J.Y

    2001-03-01

    The Tore Supra ergodic divertor (ED) screening efficiency has been investigated in density regime experiments. The ED screening efficiency is analysed by using the 'tightness' concept, which is the ratio of the density on the ED neutraliser plates to the volume averaged plasma density. Tightness is studied as a function of different plasma edge parameters, such as T{sub div}, ED magnetic perturbation ({delta}), plasma composition, location of recycling source, and additional power. Tightness is shown to increase with {delta}, P{sub div}{sup 0.55}/(1-Fr){sup 1.22}, and 1/T{sub div}{sup 0.5}. These trends are well explained by a simple 0-D model, where the particle confinement time in the ergodized peripheral region is very small. Finally, tightness increases with the power conducted onto the ED plates. Since ED plasmas have low P{sub div}, their tightness value remains low compared to that obtained with axisymmetric divertors for which P{sub div} is considerably larger. Increasing P{sub div} will result in an improved tightness and a better particle control.

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

    International Nuclear Information System (INIS)

    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

  1. Divertor Experiments with MBI and Strong Gas Puffing on HL-2A

    Science.gov (United States)

    Duan, Xuru; Ding, Xuantong; Yang, Qingwei; Yan, Longwen; Yao, Lianghua; Hong, Wenyu; Xuan, Weimin; Liu, Dequan; Chen, Liaoyuan; Song, Xianming; Zhang, Jinhua; Cao, Zeng; Cui, Zhengying; Li, Wei; Liu, Yi; Pan, Yudong; Pan, Li; Zheng, Yinjia; Zhou, Yan; Mao, Weicheng; Liu, Yong; HL-2A Team

    2006-01-01

    In the HL-2A 2004 experiment campaign, pulsed molecular beam injection (MBI) and strong hydrogen gas puffing under the divertor configuration were used for gas fueling. The experimental results show that the MBI of hydrogen can reduce the heat flux to the divertor target plate. The electron temperature measured by the Langmuir probe array decreases significantly during the injection of the molecular beam whereas the electron density increases. This indicates that the plasma pressure near the target plates tends to be constant at a new equilibrium level. In the divertor plasmas with strong hydrogen gas puffing a high plasma density up to 4.4 × 1019 m-3 was achieved. In addition, a phenomenon similar to the partially detached divertor regime was observed, which is being studied in open divertor tokamaks such as DIII-D to reduce the peak heat flux on the target plates near the separatrix. After a strong gas puffing the electron temperature measured on the outer divertor target plate near the separatrix decreases till below 5 eV or even lower, but that of the farther outer divertor target plate does not change obviously; and the CIII and the Hα emissions at the plasma edge decrease as expected, but the Hα emission near the X-point increases. These results reflects some interesting characteristics, which needs to be studied by further modeling and experiments.

  2. Investigation of conventional and Super-X divertor configurations of MAST Upgrade using SOLPS

    CERN Document Server

    Havlickova, E; Wischmeier, M; Fishpool, G; Morris, A W

    2014-01-01

    One of the first studies of MAST Upgrade divertor configurations with SOLPS5.0 are presented. We focus on understanding main prospects associated with the novel geometry of the Super-X divertor (SXD). This includes a discussion of the effect of magnetic flux expansion and volumetric power losses on the reduction of target power loads, the effect of divertor geometry on the divertor closure and distribution of neutral species and radiation in the divertor, the role of the connection length in broadening the target wetted area. A comparison in conditions typical for MAST inter-ELM H-mode plasmas confirms improved performance of the Super-X topology resulting in significantly better divertor closure with respect to neutrals (the atomic flux from the target increased by a factor of 6, but the atomic flux from the divertor to the upper SOL reduced by a factor of 2), increased radiation volume and increased total power loss (a factor of 2) and a reduction of target power loads through both magnetic flux expansion a...

  3. Simulation of tokamak SOL and divertor region including heat flux mitigation by gas puffing

    Science.gov (United States)

    Park, Jin-Woo; Na, Yong-Su; Hong, Sang Hee; Ahn, Joon-Wook; Kim, Deok-Kyu; Han, Hyunsun; Shim, Seong Bo; Lee, Hae June

    2012-08-01

    Two-dimensional (2D), scrape-off layer (SOL)-divertor transport simulations are performed using the integrated plasma-neutral-impurity code KTRAN developed at Seoul National University. Firstly, the code is applied to reproduce a National Spherical Torus eXperiment (NSTX) discharge by using the prescribed transport coefficients and the boundary conditions obtained from the experiment. The plasma density, the heat flux on the divertor plate, and the D α emission rate profiles from the numerical simulation are found to follow experimental trends qualitatively. Secondly, predictive simulations are carried out for the baseline operation mode in Korea Superconducting Tokamak Advanced Research (KSTAR) to predict the heat flux on the divertor target plates. The stationary peak heat flux in the KSTAR baseline operation mode is expected to be 6.5 MW/m2 in the case of an orthogonal divertor. To study the mitigation of the heat flux, we investigated the puffing effects of deuterium and argon gases. The puffing position is assumed to be in front of the strike point at the outer lower divertor plate. In the simulations, mitigation of the peak heat flux at the divertor target plates is found to occur when the gas puffing rate exceeds certain values, ˜1.0 × 1020 /s and ˜5.0 × 1018 /s for deuterium and argon, respectively. Multi-charged impurity transport is also investigated for both NSTX and KSTAR SOL and divertor regions.

  4. Improvement of the divertor bolometer diagnostic in the ASDEX Upgrade tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Sehmer, Till; Meister, Hans; Bernert, Matthias; Koll, Juergen; Reimold, Felix; Wischmeier, Marco; Fantz, Ursel [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Collaboration: ASDEX Upgrade Team

    2015-05-01

    For future fusion devices such as ITER, the radiation balance in the divertor region will have a significant impact on the power exhaust balance. Therefore, scenarios with strongly localized radiation, like radiation in the high field side high density (HFSHD) region, X-Point radiation or radiation in the divertor legs during detachment, will be investigated in the next ASDEX Upgrade (AUG) operation campaign 2015. To obtain accurately the absolute divertor radiation out of these measurements, the AUG foil bolometer diagnostic system in the divertor region has been enhanced; two new cameras have been designed and manufactured. One will be mounted below the roof baffle and contains 28 lines of sight (LOS), which will observe the mentioned regions of particular physical interest. The second camera consists of 4 LOS and will be mounted at the high field side above the inner divertor nose. It will observe radiation arising from the X-Point region and from the outer divertor. The data will be analysed with a tomographic reconstruction algorithm to localize and quantify the divertor radiation.

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

    Energy Technology Data Exchange (ETDEWEB)

    Costanzo, L

    2001-10-01

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

  6. A bulk tungsten divertor row for the outer strike point in JET

    International Nuclear Information System (INIS)

    In the frame of the ITER-like wall project, a new row of divertor tiles has been developed which consists of 96 bulk tungsten load-bearing septum replacement plates (LB-SRP). Exposed to the outer strike point for most ITER-relevant, high triangularity configurations, they shall be subject to high power loads (locally 10 MW/m2 and above). These conditions are demanding, particularly for an inertially cooled design as prescribed. The expected erosion rates are high as well as the risk of melting, especially with transients and repetitive ELM loads. The development is also a real challenge with respect to the inevitable excursions of the tungsten material through the so-called DBTT, ductile-to-brittle transition temperature. A lamella design has been selected to fulfil the requirements with respect to the thermo-mechanical and electromagnetic loads during disruptions (∂T/∂z ≤ 5 x 104 K/m vertically, induction rate of change ∂B/∂t ≤ 100 T/s, and Ihalo ≤ 18 kA/module). Care is taken to act on refractory metals solely with compressive forces to a large extent. The dedicated clamping concept is described. Results of a test exposure to an electron beam around 70 MJ/m2 substantiate the resort to 'high temperature' materials like - among others - high-grade Nimonic alloys, molybdenum or ceramic coatings.

  7. Plasma performance control during ergodic divertor experiments in Tore Supra

    International Nuclear Information System (INIS)

    Ohmic plasma particle confinement times are controlled during magnetic perturbation and stochastic boundary layer experiments in TORE SUPRA with small currents in the ergodic divertor coils. Particle confinement may be improved or degraded depending on the plasma configuration and base parameters used. The magnitude of these steady state confinement changes are controlled by changing led and the base plasma parameters. Plasma confinement changes manifest either density increase with a reduction in the wall fueling flux or density decreases with an increase in the fueling flux depending on the geometric configuration. In addition, the effective thermal insulation of the boundary layer is controlled. Impurity and radiated power profiles are readily modified in the boundary layer

  8. Structural evaluation of a DTHR bundle divertor particle collector

    International Nuclear Information System (INIS)

    The purpose of this report is to present a structural evaluation of the current bundle divertor particle collector BDPC design under a peak heat flux in relation to criteria that protect against coolant leakage into the plasma over replacement schedules planned during DTHR operation. In addition, an assessment of the BDPC structural integrity at higher heat fluxes is presented. Further, recommendations for modifications in the current BDPC design that would improve design reliability to be considered in future design studies are described. Finally, experimental test programs directed to establishing materials data necessary in providing greater confidence in subsequent structural evaluations of BDPC designs in relation to coolant leakage over planned replacement schedules are identified

  9. Ion Beam Analysis methods applied to the examination of Be//Cu joints in hipped Be tiles for ITER first wall mock- ups

    International Nuclear Information System (INIS)

    A proposed fabrication route for ITER first wall components implies a diffusion welding step of Be tiles onto a Cu-based substrate. However, Be has a tendency to form particularly brittle intermetallics with Cu and a lot of other elements. Insertion of interlayers may be a solution to increase bond quality. Applying traditional analyses to this study can be problematic because of Be toxicity and low atomic number Z. Ion Beam Analysis methods have thus been considered together with scanning electron microscopy (SEM) and electron back-scattering diffraction (EBSD) as complementary techniques. The following work aims at demonstrating how such techniques (used in micro-beam mode), and in particular NRA (Nuclear Reaction Analysis) and PIXE (Particle Induced X-ray Emission) techniques, coupled with SEM/EBSD data, can bring valuable information in this area. Quantification of data allow to obtain concentration values (provided the hypotheses on the initial junction composition are valuable), then phase diagrams give clues about the composition and structure of the junction. SEM retro-diffused electrons chemical contrast images and EBSD allow to characterize the presence of the awaited intermetallics, and finally confirm or refine the conclusions of Ion Beam Analysis data quantification. A series of reference first wall mock-ups have been analysed. Interlayer-free mock-ups reveal intermetallics which are mainly BeCu (apparently mixed with lower quantities of BeCu2 compound). While Cr or Ti interlayers seem to behave as good Be diffusion barriers in the sense that they prevent the formation of BeCu, they strongly interact with Cu to form CuTi2 or Cr2Ti intermetallics. In the case of Cr, Be seems to be incorporated into the Cr layer. PIXE analysis has however been unable to characterize Al-based interlayers (Z=13, close to the lower PIXE sensibility limit) and emphasizes one limitation of Ion Beam Analysis methods for lighter metals, justifying the use of other complementary

  10. Neutron yields and emission rates in the forward direction for 50MeV/u 18O—ion on thick Be,Cu,Au targets

    Institute of Scientific and Technical Information of China (English)

    LiGui-Sheng; ZhangTian-Mei; 等

    1997-01-01

    Total neutron yields and neutron emission rates in the forward direction for 50MeV/u 18O-ion on thick Be,Cu,Au targets have been measured using an activation technique.The results indicate that neutron yields and emission rates in the forward direction depend on the atomic number of target nuclei,i.e.the lighter target the greater neurtron yield and neutron emission rate.Meanwhile,the neutron yield of 18O-ion is greater than that of 12C-ion when target nucleus and incident energy per nucleon are identical.

  11. Velocity dependence of ionization probability of Be, Cu, Ag, W, Pb and Sn atoms sputtered by 5.5 keV Ar+ ions

    International Nuclear Information System (INIS)

    The energy distributions of both ions and neutral atoms sputtered during ion bombardment of a few polycrystalline metals have been measured. The absolute values of the ionization probability P+ as a function of the emitted particle velocity ν were obtained. The ionization probability for Be, Cu, Ag and W targets is found to depend exponentially on the particle velocity in accordance with the electron tunneling model. In the case of Pb and Sn the ionization probability shows strong deviation from an exponential dependence at high emission energies. (author)

  12. Effects of magnetic configuration on divertor power and particle deposition for long pulse operation in EAST

    International Nuclear Information System (INIS)

    The magnetic configuration exhibits a strong influence on the dynamics of Edge Localized Modes (ELMs), as demonstrated in the EAST superconducting tokamak. We find that poloidal drifts play an important role in particle deposition during the ELMs, leading to a strong up/down asymmetry in the double null divertor configuration, favoring the upper divertor for normal toroidal field, Bt, i.e., with the ion ∇B drift towards the bottom, while the heat flux distribution appears to be rather uniform during ELMs. These observations are well reproduced by the boundary plasma turbulence code, BOUT++. As divertor pumping was only available at the bottom, the preferential particle flow towards the bottom divertor associated with reverse Bt led to a preferred scenario for long pulse operation in EAST

  13. Optimization design study of an innovative divertor concept for future experimental tokamak-type fusion reactors

    International Nuclear Information System (INIS)

    The design optimization study of an innovative divertor concept for future experimental tokamak-type fusion devices is both an answer to the actual problems encountered in the multilayer divertor proposals and an illustration of a rational modelling philosophy and optimization strategy for the development of a new divertor structure. Instead of using mechanical attachment or metallurgical bonding of the protective material to the heat sink as in most actual divertor concepts, the so-called brush divertor in this study uses an array of unidirectional fibers penetrating in both the protective armor and the underling composite heat sink. Although the approach is fully concentrated on the divertor performance, including both a description of its function from the theoretical point of view and an overview of the problems related to the materials choice and evaluation, both the approach followed in the numerical modelling and the judgment of the results are thought to be valid also for other applications. Therefore the spin-off of the study must be situated in both the technological progress towards a feasible divertor solution, which introduces no additional physical uncertainties, and in the general area of the thermo-mechanical finite-element modelling on both macro-and microscale. The brush divertor itself embodies the use, and thus the modelling, of advanced materials such as tailor-made metal matrix composites and dispersion strengthened metals, and is shown to offer large potential advantages, demanding however and experimental validation under working conditions. It is clearly indicated where the need originates for an integrated experimental program which must allow to verify the basic modelling assumptions in order to arrive at the use of numerical computation as a powerful and realistic tool of structural testing and life-time prediction

  14. Two-dimensional analysis for a scrapeoff and divertor regions with an MHD model

    International Nuclear Information System (INIS)

    With a two-dimensional time dependent fluid code for transport processes in the edge plasma in a tokamak, coupled with Monte-Carlo method for neutral gas behavior, preliminary numerical study has been carried out for the FER divertor. Design base data such as energy flux, particle flux and so on which are essentially important to make an divertor design reliable have been obtained. (author)

  15. Feasibility study for a multi-channel pulsed radar reflectometer for the jet divertor region

    International Nuclear Information System (INIS)

    In this report, the feasibility of a pulsed radar system for measuring the electron density profile in the divertor region of JET is studied. Some dedicated experiments are performed with a four-channel system, which was designed for the Rijnhuizen Tokamak Project. To simulate divertor plasmas the measurements are performed in ECRH induced plasmas without current. The parameters of these kinds of plasmas are: ne19 m-3, Te<100 eV, and a diameter of ∼30 cm. (HSI)

  16. Numerical simulations of resistive magnetohydrodynamic instabilities in a poloidal divertor tokamak

    International Nuclear Information System (INIS)

    A new 3-D resistive MHD initial value code RPD has been successfully developed from scratch to study the linear and nonlinear evolution of long wavelength resistive MHD instabilities in a square cross-section tokamak with or without a poloidal divertor. The code numerically advances the full set of compressible resistive MHD equations in a toroidal geometry, with an important option of permitting the divertor separatrix and the region outside it to be in the computational domain. A severe temporal step size restriction for numerical stability imposed by the fast compressional waves was removed by developing and implementing a new, efficient semi-implicit scheme extending one first proposed by Harned and Kerner. As a result, the code typically runs faster than that with a mostly explicit scheme by a factor of about the aspect ratio. The equilibrium input for RPD is generated by a new 2-D code EQPD that is based on the Chodura-Schluter method. The RPD code, as well as the new semi-implicit scheme, has passed very extensive numerical tests in both divertor and divertorless geometries. Linear and nonlinear simulations in a divertorless geometry have reproduced the standard, previously known results. In a geometry with a four-node divertor the m = 2,n = 1 (2/1) tearing mode tends to be linearly stabilized as the q = 2 surface approaches the divertor separatrix. However, the m = 1,n = 1 (1/1) resistive kink mode remains relatively unaffected by the nearness of the q = 1 surface to the divertor separatrix. When plasma current is added to the region outside the divertor separatrix, the 2/1 tearing mode is linearly stabilized not by this current, but by the profile modifications induced near the q = 2 surface and the divertor separatrix. A similar stabilization effect is seen for the 1/1 resistive kink mode, but to a lesser extent. 77 refs., 91 figs

  17. Favorable effects of turbulent plasma mixing on the performance of innovative tokamak divertors

    Science.gov (United States)

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

    2013-10-01

    The problem of reducing the heat load on plasma-facing components is one of the most demanding issues for MFE devices. The general approach to the solution of this problem is the use of a specially configured poloidal magnetic field, so called magnetic divertors. In recent years, novel divertors possessing the 2-nd and 3-rd order nulls of the poloidal field (PF) have been proposed. They are called a ``snowflake'' (SF) and a ``cloverleaf'' (CL) divertor, respectively, due to characteristic shape of the magnetic separatrix. Among several beneficial features of such divertors is an effect of strong turbulent plasma mixing that is intrinsic to the zone of weak PF near the null-point. The turbulence spreads the heat flux between multiple divertor exhaust channels and increases the heat flux width within each channel. Among physical processes affecting the onset of convection the curvature-driven mode of axisymmetric rolls is most prominent. The effect is quite significant for the SF and is even stronger for the CL divertor. Projections to future ITER-scale facilities are discussed. Work performed for U.S. DoE by LLNL under Contract DE-AC52-07NA27344.

  18. Critical need for MFE: the Alcator DX advanced divertor test facility

    Science.gov (United States)

    Vieira, R.; Labombard, B.; Marmar, E.; Irby, J.; Wolf, S.; Bonoli, P.; Fiore, C.; Granetz, R.; Greenwald, M.; Hutchinson, I.; Hubbard, A.; Hughes, J.; Lin, Y.; Lipschultz, B.; Parker, R.; Porkolab, M.; Reinke, M.; Rice, J.; Shiraiwa, S.; Terry, J.; Theiler, C.; Wallace, G.; White, A.; Whyte, D.; Wukitch, S.

    2013-10-01

    Three critical challenges must be met before a steady-state, power-producing fusion reactor can be realized: how to (1) safely handle extreme plasma exhaust power, (2) completely suppress material erosion at divertor targets and (3) do this while maintaining a burning plasma core. Advanced divertors such as ``Super X'' and ``X-point target'' may allow a fully detached, low temperature plasma to be produced in the divertor while maintaining a hot boundary layer around a clean plasma core - a potential game-changer for magnetic fusion. No facility currently exists to test these ideas at the required parallel heat flux densities. Alcator DX will be a national facility, employing the high magnetic field technology of Alcator combined with high-power ICRH and LHCD to test advanced divertor concepts at FNSF/DEMO power exhaust densities and plasma pressures. Its extended vacuum vessel contains divertor cassettes with poloidal field coils for conventional, snowflake, super-X and X-point target geometries. Divertor and core plasma performance will be explored in regimes inaccessible in conventional devices. Reactor relevant ICRF and LH drivers will be developed, utilizing high-field side launch platforms for low PMI. Alcator DX will inform the conceptual development and accelerate the readiness-for-deployment of next-step fusion facilities.

  19. Exploration of magnetic perturbation effects on advanced divertor configurations in NSTX-U

    Science.gov (United States)

    Frerichs, H.; Schmitz, O.; Waters, I.; Canal, G. P.; Evans, T. E.; Feng, Y.; Soukhanovskii, V. A.

    2016-06-01

    The control of divertor heat loads - both steady state and transient - remains a key challenge for the successful operation of ITER and FNSF. Magnetic perturbations provide a promising technique to control ELMs (Edge Localized Modes) (transients), but understanding their detailed impact is difficult due to their symmetry breaking nature. One approach for reducing steady state heat loads is so called "advanced divertors" which aim at optimizing the magnetic field configuration: the snowflake and the (super-)X-divertor. It is likely that both concepts - magnetic perturbations and advanced divertors - will have to work together, and we explore their interaction based on the NSTX-U setup. An overview of different divertor configurations under the impact of magnetic perturbations is presented, and the resulting impact on plasma edge transport is investigated with the EMC3-EIRENE code. Variations in size of the magnetic footprint of the perturbed separatrix are found, which are related to the level of flux expansion on the divertor target. Non-axisymmetric peaking of the heat flux related to the perturbed separatrix is found at the outer strike point, but only in locations where flux expansion is not too large.

  20. Spectroscopic imaging system for quantitative analysis of the divertor plasma of the Tokamak de Varennes

    International Nuclear Information System (INIS)

    A toroidally viewing spectroscopic imaging system has been developed for the Tokamak de Varennes providing measurements of the poloidal distribution of the absolute radiated power of deuterium and impurity species in the upper divertor region. Real time digitization is achieved using a low cost PC based digital imaging system. This system is used to obtain measurements of the divertor strike point as well as the shape of the flux surfaces in the divertor. The diagnostic close-quote s excellent spatial resolution and toroidal view provides an opportunity to quantitatively compare the measured two dimensional (2D) radiated power distribution to that calculated from 2D Monte Carlo transport codes. These 2D images provide unique and valuable information on the physics of local plasma interactions with divertor components and particle transport in a closed divertor. Additionally, by using two cameras simultaneously, the line ratio technique can be applied to the images to estimate plasma parameters in the divertor. copyright 1997 American Institute of Physics

  1. Spectroscopic imaging system for quantitative analysis of the divertor plasma of the Tokamak de Varennes

    Energy Technology Data Exchange (ETDEWEB)

    Meo, F.; Stansfield, B.L.; Chartre, M.; de Villers, P.; Marchand, R.; Ratel, G. [Centre Canadien de Fusion Magnetique, 1804 Boulevard Lionel-Boulet, Varennes, Quebec, J3X 1S1 (CANADA)

    1997-09-01

    A toroidally viewing spectroscopic imaging system has been developed for the Tokamak de Varennes providing measurements of the poloidal distribution of the absolute radiated power of deuterium and impurity species in the upper divertor region. Real time digitization is achieved using a low cost PC based digital imaging system. This system is used to obtain measurements of the divertor strike point as well as the shape of the flux surfaces in the divertor. The diagnostic{close_quote}s excellent spatial resolution and toroidal view provides an opportunity to quantitatively compare the measured two dimensional (2D) radiated power distribution to that calculated from 2D Monte Carlo transport codes. These 2D images provide unique and valuable information on the physics of local plasma interactions with divertor components and particle transport in a closed divertor. Additionally, by using two cameras simultaneously, the line ratio technique can be applied to the images to estimate plasma parameters in the divertor. {copyright} {ital 1997 American Institute of Physics. }

  2. Flow reversal, convection, and modeling in the DIII-D divertor

    International Nuclear Information System (INIS)

    Measurements of the parallel Mach number of background plasma in the DIII-D tokamak divertor [M. A. Mahdavi et al. in Proceedings, 16th International Conference, Montreal, 1996 (International Atomic Energy Agency, Vienna, 1997) Vol. I, p. 397] were performed using a fast scanning Mach probe. The parallel particle flow shows evidence of complex behavior such as reverse flow, i.e., flow away from the target plate, stagnant flow, and large scale convection. For detached discharges, measurements confirm predictions of convective flow towards the divertor target plate at near sound speed over large regions in the divertor. The resulting convected heat flux is a dominant heat transport mechanism in the divertor. For attached discharges with high recycling, particle flow reversal in a thin region at or near the outer separatrix, thereby confirming the existence of a mechanism by which impurities can be transported away from the divertor target plates. Modeling results from the two-dimensional fluid code UEDGE [G. D. Porter and the DIII-D Team, open-quotes Divertor characterization experiments and modelling in DIII-D,close quotes in Proceedings of the 23rd European Conference on Controlled Fusion and Plasma Physics, 24 endash 28 June 1996, Kiev, Ukraine (European Physical Society, Petit-Lancy, Switzerland, 1996), Vol. 20C, Part II, p. 699] can reproduce the main features of the experimental observations. copyright 1998 American Institute of Physics

  3. Thermofluid analysis of free surface liquid divertor in tokamak fusion reactor

    International Nuclear Information System (INIS)

    To attain high fusion power density, the divertor must suffer a high heat flux from the fusion plasma. It is very difficult to remove the high heat flux from the fusion plasma more than 20 MW/m2 using the only solid divertor plate due to the severe mechanical condition such as thermal stress and crack growth. Therefore, the concept of a liquid divertor is proposed to remove the high heat flux and neutron flux from the plasma by liquid films flowing on a solid wall. Feasibility study on the liquid divertor is being examined what kind of necessary condition should be satisfied if it was applied to the tokamak fusion reactor. There are many uncertain physics and techniques to apply the liquid divertor to the tokamak fusion reactor. This paper mainly descries a preliminary thermofluid analysis of a free surface liquid, made of FLiBe molten salt, flow suffering the high heat flux using the finite element analysis code ADINA-F. To realize the liquid divertor, two techniques of thermal hydraulics promotion using a secondary flow and liquid-solid multi-phase flow are proposed in this paper

  4. Development of a compact W-shaped pumped divertor in JT-60U

    International Nuclear Information System (INIS)

    In JT-60U, the modification to a W-shaped pumped divertor will be completed in May 1997, aiming to realize sufficient reduction in heat flux to the targets and good H-mode confinement simultaneously. W-shaped geometry is optimized not only for forming radiative divertor plasmas and reducing the back flow of neutral particles but also for allowing various experimental configurations. Toroidally and poloidally segmented divertor plates, dome and baffles are arranged in a W-shaped poloidal configuration. The pumping speed can be changed during a shot by variable shutter valves in the three pumping ports under the outer baffle. The net throughput is enough for particle control in the steady radiative operations with high power NBI heating. Carbon fiber composite (CFC) tiles are used for the divertor targets and the divertor throat where large heat flux is expected. Gaps between two adjacent segments are carefully sealed to suppress the leak of neutral gas from the exhaust duct below the divertor and baffles. The strength of the whole structure is confirmed by an electromagnetic force analysis and structural analysis carried out for disruptions of 3 MA discharges with a halo current. (orig.)

  5. Effects of radial losses of particle and energy on the stability of detachment front in a divertor plasma

    International Nuclear Information System (INIS)

    Operation under partially detached divertor (PDD) plasmas is a hopeful way in order to reduce the divertor heat load in the next generation tokamaks. The physical mechanism of PDD plasmas, however, has not fully been understood yet. We have studied them with a multi-layer one-dimensional divertor model. The PDD plasmas are successfully reproduced by introducing a neutral gas puffing model. Effect of the cross-field heat transport on the PDD plasmas is investigated. It is found that cross-field heat transport both in the SOL region and in the divertor region prevents detachment fronts from moving upstream in a detached flux tube. (author)

  6. Progress of divertor simulation research toward the realization of detached plasma using a large tandem mirror device

    Energy Technology Data Exchange (ETDEWEB)

    Nakashima, Y., E-mail: nakashma@prc.tsukuba.ac.jp [Plasma Research Center, University of Tsukuba, Tsukuba, Ibaraki 305-8577 (Japan); Takeda, H.; Ichimura, K.; Hosoi, K.; Oki, K.; Sakamoto, M.; Hirata, M.; Ichimura, M.; Ikezoe, R.; Imai, T.; Iwamoto, M.; Hosoda, Y.; Katanuma, I.; Kariya, T.; Kigure, S.; Kohagura, J.; Minami, R.; Numakura, T.; Takahashi, S.; Yoshikawa, M. [Plasma Research Center, University of Tsukuba, Tsukuba, Ibaraki 305-8577 (Japan); and others

    2015-08-15

    This paper describes the results of the experiments performed on Tandem Mirror device GAMMA 10/PDX mainly using a new “divertor simulation experimental module (D-module)” installed on one of the end mirror exits which is specially designed to investigate the physics of plasma detachment. The additional ICRF heating in the anchor-cells, connected to both ends of the central-cell, significantly increases the density in the both cells, which attained the generation of the highest particle flux up to 10{sup 23} particles/s m{sup 2} at the end-mirror exit. H{sub 2} and noble gas injection to enhance the radiation cooling in D-module was performed and a remarkable reduction of the electron temperature (from few tens eV to <3 eV) on the target plate were successfully achieved associated with the strong reduction of particle and heat flux. A significant effect of simultaneous injection with hydrogen and noble gases for detached plasma formation was recognized for the first time.

  7. Experimental study of the topological aspect of the ergodic divertor in Tore-supra tokamak

    International Nuclear Information System (INIS)

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

  8. The two-dimensional structure of radiative divertor plasmas in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Recent measurements of the two-dimensional (2-D) spatial profiles of divertor plasma density, temperature, and emissivity in the DIII-D tokamak [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] under highly radiating conditions are presented. Data are obtained using a divertor Thomson scattering system and other diagnostics optimized for measuring the high electron densities and low temperatures in these detached divertor plasmas (ne≤1021m-3, 0.5eV≤Te). D2 gas injection in the divertor increases the plasma radiation and lowers Te to less than 2 eV in most of the divertor volume. Modeling shows that this temperature is low enough to allow ion endash neutral collisions, charge exchange, and volume recombination to play significant roles in reducing the plasma pressure along the magnetic separatrix by a factor of 3 endash 5, consistent with the measurements. Absolutely calibrated vacuum ultraviolet spectroscopy and 2-D images of impurity emission show that carbon radiation near the X-point, and deuterium radiation near the target plates contribute to the reduction in Te. Uniformity of radiated power (Prad) (within a factor of 2) along the outer divertor leg, with peak heat flux on the divertor target reduced fourfold, was obtained. A comparison with 2-D fluid simulations shows good agreement when physical sputtering and an ad hoc chemical sputtering source (0.5%) from the private flux region surface are used. copyright 1997 American Institute of Physics

  9. Response of NSTX liquid lithium divertor to high heat loads

    Energy Technology Data Exchange (ETDEWEB)

    Abrams, T., E-mail: tabrams@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Jaworski, M.A. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Kallman, J. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Foley, E.L. [Nova Photonics, Inc., Princeton, NJ 08543 (United States); Gray, T.K. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Kugel, H. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Levinton, F. [Nova Photonics, Inc., Princeton, NJ 08543 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

    2013-07-15

    Samples of the NSTX Liquid Lithium Divertor (LLD) with and without an evaporative Li coating were directly exposed to a neutral beam ex-situ at a power of ∼1.5 MW/m{sup 2} for 1–3 s. Measurements of front face and bulk sample temperature were obtained. Predictions of temperature evolution were derived from a 1D heat flux model. No macroscopic damage occurred when the “bare” sample was exposed to the beam but microscopic changes to the surface were observed. The Li-coated sample developed a lithium hydroxide (LiOH) coating, which did not change even when the front face temperature exceeded the pure Li melting point. These results are consistent with the lack of damage to the LLD surface and imply that heating alone may not expose pure liquid Li if the melting point of surface impurities is not exceeded. This suggests that flow and heat are needed for future PFCs requiring a liquid Li surface.

  10. Physical Engineering Test and First Divertor Plasma Configuration in EAST

    Institute of Scientific and Technical Information of China (English)

    WAN Baonian

    2007-01-01

    Physical engineering capability on the superconducting magnetic system of EAST was tested and first divertor plasma configuration in EAST was obtained.The extrapolation of the safety limit has verified the reliability of the system for long pulse operation.A stably controlled diverted plasmas configuration with an elongation κ in excess of 1.8 and plasma current of up to 500 kA,by using the (copper) internal coils to control the vertical displacement instability was obtained by an optimized plasma control algorithm.Highly shaped plasma at various configura-tions,which almost covers all designed configurations for EAST,was generated stably.A number of operational issues,such as plasma initiation,ramp up and configuration control with constraints of superconducting coils,were successfully investigated.All of the results obtained proved both the capability of the superconducting poloidal magnets for operation under steady-state condition and effectiveness of the plasma control algorithm for EAST.

  11. Heliumlike Mg XI in the divertor-injected tokamak experiment

    International Nuclear Information System (INIS)

    Electron-impact excitation rates for transitions in heliumlike Mg XI, calculated with the R-matrix code, are used to derive the electron-density-sensitive emission line ratio R (=f/i) and temperature-sensitive ratio G [=(f+i)/r], where f is the forbidden 1s21S--1s2s 3S transition, i the intercombination 1s21S--1s2p 3P1,2 lines, and r the resonance 1s21S--1s2p 1P transition. A comparison of these with R and G ratios determined from x-ray spectra of the divertor-injected tokamak experiment reveals excellent agreement between theory and observation, with discrepancies of typically 3% and 9% in R and G, respectively. These discrepancies correspond to variations in Ne and Te of approximately 0.1 and 0.15 dex, respectively, and hence it should be possible to use the theoretical results to derive plasma parameters to this level of accuracy for remote sources for which no independent electron temperature and density estimates exist, such as solar flares

  12. Response of NSTX Liquid Lithium divertor to High Heat Loads

    Energy Technology Data Exchange (ETDEWEB)

    Abrams, Tyler; Kallman, J; Kaitaa, R; Foley, E L; Grayd, T K; Kugel, H; Levinton, F; McLean, A G

    2012-07-18

    Samples of the NSTX Liquid Lithium Divertor (LLD) with and without an evaporative Li coating were directly exposed to a neutral beam ex-situ at a power of ~1.5 MW/m2 for 1-3 seconds. Measurements of front face and bulk sample temperature were obtained. Predictions of temperature evolution were derived from a 1D heat flux model. No macroscopic damage occurred when the "bare" sample was exposed to the beam but microscopic changes to the surface were observed. The Li-coated sample developed a lithium hydroxide (LiOH) coating, which did not change even when the front face temperature exceeded the pure Li melting point. These results are consistent with the lack of damage to the LLD surface and imply that heating alone may not expose pure liquid Li if the melting point of surface impurities is not exceeded. This suggests that flow and heat are needed for future PFCs requiring a liquid Li surface. __________________________________________________

  13. Investigation of limiter recycling in the divertor tokamak ASDEX

    International Nuclear Information System (INIS)

    A divertor experiment like the ASDEX tokamak is especially suited for studying ion recycling at a material limiter, because the plasma can alternatively be limited by a magnetic limiter (separatrix) or by a material limiter. The role of the material limiter in ion recycling is documented by observing the increase in charge exchange flux emitted at the limiter position, and the decrease in external gas input necessary to keep the plasma line density invariant, when the material limiter is moved to the plasma. Ion recycling occurs predominantly at the outside section of a ring limiter. The limiter material saturates shortly after the start of the discharge. About 60% of the total recycling occurs at the limiter, which is nearly 100% of the ion recycling. The remaining 40% of the total recycling is carried by charge exchange neutrals. Due to saturation, the recycling coefficient at the limiter is 1; the recycling coefficient of the charge exchange neutrals at the wall is approximately 0.5 giving rise to a total recycling coefficient of limiter discharges of 0.8-0.9. It is observed that the plasma resistivity increases when the material limiter is moved toward the separatrix. The increase in Zsub(eff) can tentatively be explained by proton sputtering. (orig.)

  14. Control of 3D edge radiation structure with resonant magnetic perturbation fields applied to the stochastic layer and stabilization of radiative divertor plasma in LHD

    International Nuclear Information System (INIS)

    It is found that resonant magnetic perturbation (RMP) fields have a stabilizing effect on the radiating edge plasma, realizing stable sustainment of radiative divertor (RD) operation in the Large Helical Device (LHD). Without RMP, thermal instability leads to radiative collapse. Divertor power load is reduced by a factor of 3–10 during the RMP-assisted RD phase, while maintaining relatively good core plasma confinement with confinement enhancement factor τEexp/frenτEISS04∼0.96. It has also been demonstrated that the RMP field itself can initiate transition to RD operation by increasing perturbation strength, while keeping constant density and injection power. The results show a possibility of a new control knob for divertor power load in a 3D magnetic field configuration. It is also found that after the transition to RD, the energy confinement enhancement factor based on ISS04 scaling increases by a factor of 1.4 compared with the attached phase. The operation range of the RMP-assisted RD is identified in terms of RMP strength and radial location of the resonance layer of the RMP. A 3D edge radiation structure is analysed using the edge transport code EMC3-EIRENE and the results are compared with experiments. The comparison indicates that the application of RMP modulates the 3D edge radiation structure such that an intense radiation appears around the X-point of the m/n = 1/1 island in the case with RMP, while it is located at the inboard side without RMP. (paper)

  15. Copper matrix composites as heat sink materials for water-cooled divertor target

    Directory of Open Access Journals (Sweden)

    Jeong-Ha You

    2015-12-01

    Full Text Available According to the recent high heat flux (HHF qualification tests of ITER divertor target mock-ups and the preliminary design studies of DEMO divertor target, the performance of CuCrZr alloy, the baseline heat sink material for DEMO divertor, seems to only marginally cover the envisaged operation regime. The structural integrity of the CuCrZr heat sink was shown to be affected by plastic fatigue at 20 MW/m². The relatively high neutron irradiation dose expected for the DEMO divertor target is another serious concern, as it would cause significant embrittlement below 250 °C or irradiation creep above 350 °C. Hence, an advanced design concept of the divertor target needs to be devised for DEMO in order to enhance the HHF performance so that the structural design criteria are fulfilled for full operation scenarios including slow transients. The biggest potential lies in copper-matrix composite materials for the heat sink. In this article, three promising Cu-matrix composite materials are reviewed in terms of thermal, mechanical and HHF performance as structural heat sink materials. The considered candidates are W particle-reinforced, W wire-reinforced and SiC fiber-reinforced Cu matrix composites. The comprehensive results of recent studies on fabrication technology, design concepts, materials properties and the HHF performance of mock-ups are presented. Limitations and challenges are discussed.

  16. Scaling and transport analysis of divertor conditions on the Alcator C-Mod tokamak

    Energy Technology Data Exchange (ETDEWEB)

    LaBombard, B.; Goetz, J.; Kurz, C.; Jablonski, D.; Lipschultz, B.; McCracken, G.; Niemczewski, A.; Boivin, R.L.; Bombarda, F.; Christensen, C.; Fairfax, S.; Fiore, C.; Garnier, D.; Graf, M.; Golovato, S.; Granetz, R.; Greenwald, M.; Horne, S.; Hubbard, A.; Hutchinson, I.; Irby, J.; Kesner, J.; Luke, T.; Marmar, E.; May, M.; O`Shea, P.; Porkolab, M.; Reardon, J.; Rice, J.; Schachter, J.; Snipes, J.; Stek, P.; Takase, Y.; Terry, J.; Tinios, G.; Watterson, R.; Welch, B.; Wolfe, S. [Plasma Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States)

    1995-06-01

    Detailed measurements and transport analysis of divertor conditions in Alcator C-Mod [Phys. Plasmas {bold 1}, 1511 (1994)] are presented for a range of line-averaged densities, 0.7{lt}{ital {bar n}}{sub {ital e}}{lt}2.2{times}10{sup 20} m{sup {minus}3}. Three parallel heat transport regimes are evident in the scrape-off layer: sheath-limited conduction, high-recycling divertor, and detached divertor, which can coexist in the same discharge. {ital Local} cross-field pressure gradients are found to scale simply with a {ital local} electron temperature. This scaling is consistent with classical electron parallel conduction being balanced by anomalous cross-field transport ({chi}{sub {perpendicular}}{similar_to}0.2 m{sup 2} s{sup {minus}1}) proportional to the local pressure gradient. A 60%--80% of divertor power is radiated in attached discharges, approaching 100% in detached discharges. Detachment occurs when the heat flux to the plate is low and the plasma pressure is high ({ital T}{sub {ital e}}{similar_to}5 eV). High neutral pressures in the divertor are nearly always present (1--20 mTorr), sufficient to remove parallel momentum via ion--neutral collisions.

  17. ADX: A high Power Density, Advanced RF-Driven Divertor Test Tokamak for PMI studies

    Science.gov (United States)

    Whyte, Dennis; ADX Team

    2015-11-01

    The MIT PSFC and collaborators are proposing an advanced divertor experiment, ADX; a divertor test tokamak dedicated to address critical gaps in plasma-material interactions (PMI) science, and the world fusion research program, on the pathway to FNSF/DEMO. Basic ADX design features are motivated and discussed. In order to assess the widest range of advanced divertor concepts, a large fraction (>50%) of the toroidal field volume is purpose-built with innovative magnetic topology control and flexibility for assessing different surfaces, including liquids. ADX features high B-field (>6 Tesla) and high global power density (P/S ~ 1.5 MW/m2) in order to access the full range of parallel heat flux and divertor plasma pressures foreseen for reactors, while simultaneously assessing the effect of highly dissipative divertors on core plasma/pedestal. Various options for efficiently achieving high field are being assessed including the use of Alcator technology (cryogenic cooled copper) and high-temperature superconductors. The experimental platform would also explore advanced lower hybrid current drive and ion-cyclotron range of frequency actuators located at the high-field side; a location which is predicted to greatly reduce the PMI effects on the launcher while minimally perturbing the core plasma. The synergistic effects of high-field launchers with high total B on current and flow drive can thus be studied in reactor-relevant boundary plasmas.

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

    International Nuclear Information System (INIS)

    Highlights: • SATIR tests on DEMO divertor fingers (integrating or not He cooling system). • Millimeter size artificial defects were manufactured. • Detectability of millimeter size artificial defects was evaluated. • SATIR can detect defect in DEMO divertor fingers. • Simulations are well correlated to SATIR tests. -- Abstract: Plasma facing components (PFCs) with tungsten (W) armor materials for DEMO divertor require a high heat flux removal capability (at least 10 MW/m2 in steady-state conditions). The reference divertor PFC concept is a finger with a tungsten tile as a protection and sacrificial layer brazed to a thimble made of tungsten alloy W – 1% La2O3 (WL10). Defects may be located at the W thimble to W tile interface. As the number of fingers is considerable (>250,000), it is then a major issue to develop a reliable control procedure in order to control with a non-destructive examination the fabrication processes. The feasibility for detecting defect with infrared thermography SATIR test bed is presented. SATIR is based on the heat transient method and is used as an inspection tool in order to assess component heat transfer capability. SATIR tests were performed on fingers integrating or not the complex He cooling system (steel cartridge with jet holes). Millimeter size artificial defects were manufactured and their detectability was evaluated. Results of this study demonstrate that the SATIR method can be considered as a relevant non-destructive technique examination for the defect detection of DEMO divertor fingers

  19. Impact of carbon and tungsten as divertor materials on the scrape-off layer conditions in JET

    NARCIS (Netherlands)

    Groth, M.; Brezinsek, S.; Belo, P.; Beurskens, M. N. A.; Brix, M.; Clever, M.; Coenen, J. W.; Corrigan, C.; Eich, T.; Flanagan, J.; Guillemaut, C.; Giroud, C.; Harting, D.; Huber, A.; Jachmich, S.; Kruezi, U.; Lawson, K. D.; Lehnen, M.; Lowry, C.; Maggi, C. F.; Marsen, S.; Meigs, A. G.; Pitts, R.A.; Sergienko, G.; Sieglin, B.; Silva, C.; Sirinelli, A.; Stamp, M. F.; van Rooij, G. J.; Wiesen, S.; JET-EFDA Contributors,

    2013-01-01

    The impact of carbon and beryllium/tungsten as plasma-facing components on plasma radiation, divertor power and particle fluxes, and plasma and neutral conditions in the divertors has been assessed in JET both experimentally and by edge fluid code simulations for plasmas in low-confinement mode. In

  20. Research of the capillary structure heat removal efficiency under divertor conditions

    Energy Technology Data Exchange (ETDEWEB)

    Pistunovich, V.I. [Rossijskij Nauchnyj Tsentr ``Kurchatovskij Inst.``, Moscow (Russian Federation); Vertkov, A.V. [Stock Corp. `Prana`, Moscow (Russian Federation); Evtikhin, V.A. [Stock Corp. `Prana`, Moscow (Russian Federation); Korjavin, V.M. [Fusion Dept. Ministry of Atomic Energy, Moscow (Russian Federation); Lyublinski, I.E. [Stock Corp. `Prana`, Moscow (Russian Federation); Petrov, V.B. [Rossijskij Nauchnyj Tsentr ``Kurchatovskij Inst.``, Moscow (Russian Federation); Khripunov, B.I. [Rossijskij Nauchnyj Tsentr ``Kurchatovskij Inst.``, Moscow (Russian Federation); Shapkin, V.V. [Rossijskij Nauchnyj Tsentr ``Kurchatovskij Inst.``, Moscow (Russian Federation)

    1996-10-01

    Experimental models of capillary structure for liquid metal fusion reactor divertor simulation have been designed, manufactured and tested in order to estimate the behaviour and possibilities of plasma-facing components based on lithium capillary system at long-pulse high heat load. The power load on the capillary target structures up to 50 MW/m{sup 2} was provided by electron beam with electron energy {<=}10 keV. The exposition-time was up to several minutes and was limited by the lithium quantity in the supply vessel. The operation parameters of the models determined in the experiments are in accordance with there design estimations. The tests of various model constructions at the divertor relevant power loads have shown promise for the new concept of a divertor taking into account long life and reliability. (orig.).

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

    International Nuclear Information System (INIS)

    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)

  2. Analysis of FAST snowflake divertor by EDGE2D/EIRENE

    Energy Technology Data Exchange (ETDEWEB)

    Viola, B., E-mail: bruno.viola@enea.it [ENEA Unità Tecnica Fusione, C.R. Frascati, Via E. Fermi 45, 00044 Frascati, Rome (Italy); Pericoli Ridolfini, V. [Consorzio CREATE, Università di Napoli Federico II, Via Claudio 21, 80125 Napoli (Italy); Visona, N. [Consorzio RFX, C.so Stati Uniti 4, Padova 35127 (Italy); Corrigan, G.; Harting, D. [Culham Centre of Fusion Energy, OX14 3DB Abingdon (United Kingdom); Maddaluno, G. [ENEA Unità Tecnica Fusione, C.R. Frascati, Via E. Fermi 45, 00044 Frascati, Rome (Italy); Zagórski, R. [Institute of Plasma Physics and Laser Microfusion, 01-497 Warsaw (Poland)

    2015-08-15

    The snowflake [1,2] divertor is a proposal for solving the heat and particle exhaust problem in fusion grade plasmas. Turning the X-point into a second order null gives the possibility of radially expanding the poloidal flux in the divertor region much more than in a SD, increasing the connection length, redistributing the power load on a larger area and enhancing radiative losses. Since the efforts associated to the design of reactor-relevant configurations, like the snowflake, are large, ENEA is studying this configuration using efficient and flexible numerical tools to design and optimise tokamak equilibrium configurations. Such studies are applied to the Divertor Test Tokamak FAST, a satellite tokamak proposed for the European roadmap towards fusion.

  3. Design and optimization of W/Cu divertor mock-ups

    Institute of Scientific and Technical Information of China (English)

    Qiong Li; Weiping Shen

    2007-01-01

    Tungsten is a promising candidate for plasma-facing materials to cover the surface of the divertor plate in the design of an international thermonuclear experimental reactor (ITER). Copper as a heat sink material serves to transfer heat excellently. Divertor mock-ups with W/Cu graded interlayers were designed to reduce thermal stresses. Thermally induced stresses and temperature in a W/Cu divertor mock-up were analyzed using the finite element method. The graded structures with different exponents p and thicknesses were designed and discussed. The conclusions drawn from these analyses are that thermal stresses reach the minimum and the temperature is suitable when exponent p is 1.5 and the thickness of five graded interlayers is 5 mm.

  4. Investigation of SOL parameters and divertor particle flux from electric probe measurements in KSTAR

    Energy Technology Data Exchange (ETDEWEB)

    Bak, J.G., E-mail: jgbak@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, H.S. [National Fusion Research Institute, Daejeon (Korea, Republic of); Bae, M.K. [Hanyang University, Seoul (Korea, Republic of); Juhn, J.W.; Seo, D.C.; Bang, E.N. [National Fusion Research Institute, Daejeon (Korea, Republic of); Shim, S.B. [Pusan National University, Pusan (Korea, Republic of); Chung, K.S. [Hanyang University, Seoul (Korea, Republic of); Lee, H.J. [Pusan National University, Pusan (Korea, Republic of); Hong, S.H. [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-08-15

    The upstream scrape-off layer (SOL) profiles and downstream particle fluxes are measured with a fast reciprocating Langmuir probe assembly (FRLPA) at the outboard mid-plane and a fixed edge Langmuir probe array (ELPA) at divertor region, respectively in the KSTAR. It is found that the SOL has a two-layer structure in the outboard wall-limited (OWL) ohmic and L-mode: a near SOL (∼5 mm zone) with a narrow feature and a far SOL with a broader profile. The near SOL width evaluated from the SOL profiles in the OWL plasmas is comparable to the scaling for the L-mode divertor plasmas in the JET and AUG. In the SOL profiles and the divertor particle flux profile during the ELMy H-modes, the characteristic e-folding lengths of electron temperature, plasma density and particle flux during an ELM phase are about two times larger than ones at the inter ELM.

  5. Co-deposited layers in the divertor region of JET-ILW

    Energy Technology Data Exchange (ETDEWEB)

    Petersson, P., E-mail: Per.Petersson@ee.kth.se [KTH Royal Institute of Technology, Association EURATOM – VR, SE-100 44 Stockholm (Sweden); Rubel, M. [KTH Royal Institute of Technology, Association EURATOM – VR, SE-100 44 Stockholm (Sweden); Esser, H.G. [Forschungszentrum Jülich, Association EURATOM, 52425 Jülich (Germany); Likonen, J.; Koivuranta, S. [VTT, Association EURATOM – TEKES, 02044 VTT (Finland); Widdowson, A. [CCFE/EURATOM Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)

    2015-08-15

    Tungsten-coated carbon tiles from a poloidal cross-section of the divertor and several types of erosion–deposition probes from the shadowed areas in the divertor were studied using heavy ion elastic recoil detection to obtain quantitative and depth-resolved deposition patterns. Deuterium, beryllium, carbon, nitrogen and oxygen along with tungsten and Inconel components are the main species detected in the studied surface region. The top of Tile 1 in the inner divertor is the main deposition area where the greatest amounts of deposited species are measured. Beryllium and tungsten-containing deposits on the probes (test mirrors and quartz microbalance) indicate that both low-Z and high-Z metals are transported to remote areas. Deposition of nitrogen-15 tracer used for edge cooling only at the end of experimental campaigns in 2012 was also detected giving evidence that nitrogen is effectively retained in wall components.

  6. Measurement of dust conversion factor for the JET carbon divertor phases

    Energy Technology Data Exchange (ETDEWEB)

    Likonen, J., E-mail: jari.likonen@vtt.fi [EUROfusion Consortium, JET, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Association EURATOM-TEKES, VTT, PO Box 1000, 02044 VTT, Espoo (Finland); Coad, J.P. [EUROfusion Consortium, JET, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Association EURATOM-TEKES, VTT, PO Box 1000, 02044 VTT, Espoo (Finland); EURATOM/CCFE Fusion Association, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); Hakola, A.; Karhunen, J.; Koivuranta, S. [EUROfusion Consortium, JET, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Association EURATOM-TEKES, VTT, PO Box 1000, 02044 VTT, Espoo (Finland); Pitts, R. [EUROfusion Consortium, JET, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-lez-Durance (France); Widdowson, A.M. [EUROfusion Consortium, JET, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); EURATOM/CCFE Fusion Association, Culham Science Centre, OX14 3DB Abingdon (United Kingdom)

    2015-08-15

    We compare the deposition of material on poloidal sets of divertor tiles that had been exposed in JET in 1998–2009 and 1998–2007. Post mortem analyses suggest toroidally integrated deposition being increased by 197.5 cm{sup 3} during 2007–2009. The analysis of dust collected from the divertor indicates the amount accumulated during the same period to be 248.4 g. Converting the weight of dust to volume, the fraction of material entering the divertor that was converted to dust and flakes is 43 ± 10%. The size of most dust particles ranged from 10 to 100 μm. The integrated amount of deposition on the “marker” tiles exposed in 2007–9 was found to be more than twice the amount expected from film growth on other tiles plus the dust because the plasma responds differently to the new tiles.

  7. A Snowflake Divertor: a Possible Way of Improving the Power Handling in Future Fusion Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Ryutov, D D; Bulmer, R H; Cohen, R H; Hill, D N; Lao, L; Menard, J E; Petrie, T W; Pearlstein, L D; Rognlien, T D; Snyder, P B; Soukhanovskii, V; Umansky, M V

    2008-09-17

    Handling high power loads on plasma facing components is one of the critical issues in developing an economically competitive fusion reactor based on tokamak. In this study, we provide a detailed analysis of a relatively unexplored approach to this problem based on the use of divertors with the poloidal magnetic field structure closely approaching a second-order null. We demonstrate that this geometry opens up new possibilities for radiative divertors, has favorable effect on the convective transport, and provides an additional control over ELM activity. In the ideal case where the null is exactly second order, the separatrix near the null acquires a characteristic hexagonal shape reminiscent of a snowflake, whence the name of this configuration. It can be created by a simple set of divertor coils situated outside the toroidal field coils.

  8. Local deposition of {sup 13}C tracer in the JET MKII-HD divertor

    Energy Technology Data Exchange (ETDEWEB)

    Likonen, Jari, E-mail: jari.likonen@vtt.fi [Association EURATOM-TEKES, VTT, PO Box 1000, 02044 VTT, Espoo (Finland); Airila, M.I.; Coad, J.P.; Hakola, A.; Koivuranta, S.; Ahonen, E. [Association EURATOM-TEKES, VTT, PO Box 1000, 02044 VTT, Espoo (Finland); Alves, E.; Barradas, N. [Instituto Tecnológico e Nuclear, Sacavém 2686-953 (Portugal); Widdowson, A. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Rubel, M. [Alfvén Laboratory, Royal Institute of Technology, Association EURATOM-VR, 100 44 Stockholm (Sweden); Brezinsek, S. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, Association EURATOM-FZJ, Partner in the Trilateral Euregio Cluster, D-52425 Jülich (Germany); Groth, M. [Association EURATOM-TEKES, Aalto University, 02015 Espoo (Finland)

    2013-07-15

    Migration and deposition of {sup 13}C have been investigated at JET by injecting {sup 13}C-labelled methane at the outer divertor base at the end of the 2009 campaign. The {sup 13}C deposition profile was measured with enhanced proton scattering (EPS) and secondary ion mass spectrometry (SIMS) techniques. A strong toroidal deposition band for {sup 13}C was observed experimentally on each of the analysed four outer divertor floor tiles. In addition, {sup 13}C was also found on the vertical edge of load bearing tile (LBT) and at the bottom of the LBT tile facing the puffing hole. Local {sup 13}C migration in the vicinity of the injection location was modelled by the ERO code. The ERO simulations also produced the strong toroidal {sup 13}C deposition band but there is strong deposition also on the vertical edge of the LBT tile and elsewhere on the horizontal part of the outer divertor floor tile.

  9. Experimental studies and modeling of complete H-mode divertor detachment in ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Reimold, F., E-mail: Felix.Reimold@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, Boltzmannstraß e 2, D-85748 Garching (Germany); Wischmeier, M.; Bernert, M.; Potzel, S.; Coster, D. [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, D-85748 Garching (Germany); Bonnin, X. [CNRS-LSPM, Université Paris 13, Sorbonne Paris Cité, F-93430 Villetaneuse (France); Reiter, D. [Institut für Energie- und Klimaforschung – Plasmaphysik, Forschungszentrum Jülich GmbH (Germany); Meisl, G.; Kallenbach, A. [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, D-85748 Garching (Germany); Aho-Mantila, L. [VTT, FI-02044 VTT (Finland); Stroth, U. [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, D-85748 Garching (Germany)

    2015-08-15

    Power exhaust in future fusion devices is critical and operation with a detached divertor is foreseen for ITER and DEMO. The evolution of detachment in nitrogen seeded H-mode discharges at ASDEX Upgrade is categorized in four phases. Complete detachment of the outer target is found to be correlated with a strongly localized radiation at the X-point and a pressure loss at the pedestal top at almost constant core plasma pressure. SOLPS modeling shows that enhanced radial transport in the divertor region is necessary to reconcile the experimental profiles with the simulations. The modeling supports the experimental observation of the correlation of complete detachment with an X-point radiation and a reduction of the pedestal top pressure. A remaining discrepancy are significantly lower neutral densities in the divertor compared to experiment. The effects of wall pumping, the particle reflection model and the boundary conditions on the plasma solution are discussed.

  10. Observation And Modeling Of Inner Divertor Re-attachment In Discharges With Lithium Coatings in NSTX

    International Nuclear Information System (INIS)

    In the National Spherical Torus Experiment (NSTX), modifications to the inner divertor plasma regimes are observed in high triangularity, H-mode, NBI heated discharges due to lithium coatings evaporated on the plasma facing components. In particular, the drop in the recombination rate, the reduced neutral pressure and the reduced electron density (inferred from Stark broadening measurements of high-n deuterium Balmer lines) suggested that the inner divertor, which is usually detached in discharges without lithium, re-attached. Experimental results are compared to simulations obtained with a 1D partially ionized plasma transport model integrated in the non-local thermodynamic equilibrium radiation transport code CRETIN to understand how the reduced recycling affects the divertor parameters in NSTX discharges with lithium coatings.

  11. JET ICRH antenna for pumped-divertor geometry

    International Nuclear Information System (INIS)

    The plasma configuration in the proposed JET programme extending up to 1996 will be a Single-Null (bottom) X-point with a pump divertor. This geometry has important limitations for coupling the RF power by the present ICRH antennas as the plasma size would be smaller and it will be significantly vertically asymmetric. It is clear that the present ICRH antenna (A1) system should be made compatible with the new proposed plasma configuration to utilise the full potential of the 32 MW (generator), 20 s pulse-length, 25-55 MHz JET ICRH installed facility for plasma heating and possible current drive applications in the proposed new phase of the JET programme. The present state-of-the-art knowledge of the antenna design at JET will be used for A2-antenna design which would also incorporate the ICRH current drive features as a prelude to the design of an ICRH launcher of the Next-Step devices. In this design, antennas would be made wider and deeper which would improve the coupling and it is estimated that more than 20 MW can be coupled to X-point plasmas from the ICRH plant. The current drive capability would be improved (≅ 1 MA) by the use of septums which allow arbitrary phasing between each central conductors. The design philosophy that is being followed in the design of JET A2-antennas is outlined and the present status and the main features of the physics and engineering design of A2-antenna are discussed. The antenna-plasma coupling and the antenna-directivity for the new antenna are then presented. Finally, a time-schedule for the design, construction and installation of the antennas is also given. (author)

  12. Optimization of tungsten castellated structures for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Litnovsky, A., E-mail: a.litnovsky@fz-juelich.de [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich, Trilateral Euregio Cluster, Association EURATOM-FZ Jülich, D 52425 Jülich (Germany); Hellwig, M. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich, Trilateral Euregio Cluster, Association EURATOM-FZ Jülich, D 52425 Jülich (Germany); Matveev, D. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich, Trilateral Euregio Cluster, Association EURATOM-FZ Jülich, D 52425 Jülich (Germany); Department of Applied Physics, Ghent University, Plateaustraat 22, B-9000 Ghent (Belgium); Komm, M. [Institute of Plasma Physics AS CR, v.v.i., Za Slovankou 3, 182 00 Prague 8 (Czech Republic); Berg, M. van den; De Temmerman, G. [FOM Institute DIFFER – Dutch Institute for Fundamental Energy Research, Postbus 1207, 3430BE Nieuwegein (Netherlands); Rudakov, D. [University of California, San Diego, La Jolla, CA 92093-0417 (United States); Ding, F. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, D-85748 Garching (Germany); Luo, G.-N. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Krieger, K.; Sugiyama, K. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, D-85748 Garching (Germany); Pitts, R.A. [ITER Organization, Route de Vinon sur Verdon, CS 90 046 – 13067, St. Paul Lez Durance Cedex (France); Petersson, P. [Royal Institute of Technology, SE-100, 44 Stockholm (Sweden)

    2015-08-15

    In ITER, the plasma-facing components (PFCs) of the first wall and the divertor armor will be castellated to improve their thermo-mechanical stability and to limit forces due to induced currents. The fuel accumulation in the gaps may significantly contribute to the in-vessel fuel inventory. Castellation shaping may be the most straightforward way to minimize the fuel inventory and to alleviate the thermal loads onto castellations. A new castellation shape was proposed and comparative modeling of conventional (rectangular) and shaped castellation was performed for ITER conditions. Shaped castellation was predicted to be capable to operate under stationary heat load of 20 MW/m{sup 2}. An 11-fold decrease of beryllium (Be) content in the gaps of the shaped cells alone with a 7-fold decrease of carbon content was predicted. In order to validate the predictive capabilities of modeling tools used for ITER conditions, the dedicated modeling with the same codes was made for existing tokamaks and benchmarked with the results of multi-machine experiments. For the castellations exposed in TEXTOR and DIII-D, the carbon amount in the gaps of shaped cells was 1.9–2.3 times smaller than that of rectangular ones. Modeling for TEXTOR conditions yielded to 1.5-fold decrease of carbon content in the gaps of shaped castellation outlining fair agreement with the experiment. At the same time, a number of processes, like enhanced erosion of molten layer yet need to be implemented in the codes in order to increase the accuracy of predictions for ITER.

  13. Consistency between current ramp-up/recharging scenario by non-inductive current drive and dense and cold divertor plasma

    International Nuclear Information System (INIS)

    Consistency between non-inductive current drive and the formation of cold and dense divertor plasma in phases of plasma current ramp-up and recharging. When we consider the current drive efficiency obtained in the experiments of JT-60 as the actual upper limit, it is difficult to realize the low plasma temperature below 50 eV near the divertor plate for the reasonable absorbed power (20MW) in FER. Divertor plasma temperature is reduced to about 20 eV for the absorbed power 30 MW. It is essentially important to increase the drive efficiency in order to attain the cold divertor plasma. When we use the slightly higher efficiency model than the experimental result of JT-60, the divertor plasma temperature will be reduced to 20 eV and about 10 eV for the absorbed power 20 MW and 30 MW respectively. (author)

  14. Impact of 3D magnetic field structure on boundary and divertor plasmas in stellarator/heliotron devices

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, M. [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292 (Japan); Feng, Y. [Max-Planck-Institute fuer Plasmaphysik, D-17491 Greifswald (Germany); Xu, Y. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Tabares, F.L. [Laboratorio Nacional de Fusion, Ciemat, Madrid (Spain); Ida, K. [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292 (Japan); Schmitz, O. [University of Wisconsin – Madison, WI (United States); Evans, T.E. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Frerichs, H. [University of Wisconsin – Madison, WI (United States); Liang, Y. [Forschungszentrum Jülich GmbH Institut für Energie- und Klimaforschung – Plasmaphysik, Jülich (Germany); Bader, A. [University of Wisconsin – Madison, WI (United States); Itoh, K.; Yamada, H. [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292 (Japan); Ghendrih, Ph.; Ciraolo, G. [IRFM, CEA Cadarache, St Paul Lez Durance (France); Tafalla, D.; Lopez-Fraguas, A. [Laboratorio Nacional de Fusion, Ciemat, Madrid (Spain); Guo, H.Y. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Institute of Plasma Physics, CAS, Hefei (China); Cui, Z.Y. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Reiter, D. [Forschungszentrum Jülich GmbH Institut für Energie- und Klimaforschung – Plasmaphysik, Jülich (Germany); Asakura, N. [Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan); and others

    2015-08-15

    This paper overviews recent progress on the experimental identification and physics interpretation of 3D effects of magnetic field geometry on divertor transport. The 3D effects are elucidated as a consequence of competition between transports parallel (||) and perpendicular (⊥) to magnetic field, in open field lines cut by divertor plates, or in magnetic islands. The competition has strong impacts on divertor functions, such as determination of density regime, impurity screening, and detachment control. The effects of magnetic perturbation on the edge electric field and turbulent transport are also discussed. Based on the experiments and numerical simulations, key parameters governing the 3D transport physics for the individual divertor functions, e.g. pumping efficiency through divertor density regime, impurity screening and detachment control, are discussed.

  15. Transport studies in boundary and divertor plasmas of JT-60U

    Energy Technology Data Exchange (ETDEWEB)

    Kumagai, Akira [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1999-03-01

    This thesis describes an investigation on transport of plasma, neutral particle and impurity in the boundary and divertor of the JT-60U tokamak to provide a better understanding of plasma-surface interactions and divertor physics. The asymmetry between the inboard and outboard divertor on plasma parameters (in-out asymmetry) are usually observed in tokamaks with the divertor. In this study, the in-out asymmetry was investigated under various plasma conditions and discharge parameters. The observed results were discussed with several mechanisms that can produce the in-out asymmetry. It was confirmed experimentally that the importance of each mechanism depends on the plasma parameters and discharge conditions. The current flowing in the scrape-off layer (SOL) due to the in-out asymmetry was observed. The SOL currents in the high density plasma with the occurrence of the plasma detachment were investigated for the first time in this study. The ion temperature in the divertor region is one of the most important factors for both generation and transport of impurity. However, the background ion temperature in the divertor region has not been measured in any tokamak so far. The ion temperature in the divertor region has been measured for the first time with the Doppler broading of the C{sup 3+} ion emission line. The measured temperature was analyzed by an impurity particle transport code. The code calculation showed that the measured temperature reflects the low temperature at the outside of the separatrix in the inboard region. The spectral profile of Balmer-{alpha} (D{sub {alpha}}) line emitted from the deuterium atoms reflects the velocity distribution of neutral particles by the Doppler effect and is effective for investigating the detailed neutral behavior and recycling process. The spatial variation of the D{sub {alpha}} line spectral profile in the divertor region has been measured for the first time in this study. The observed results were compared with the

  16. Combination of helical ferritic-steel inserts and flux-tube-expansion divertor for the heat control in tokamak DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takizuka, T., E-mail: takizuka.tomonori@gmail.com [Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita 565-0871 (Japan); Tokunaga, S.; Hoshino, K. [Japan Atomic Energy Agency, 2-166, Omotedate, Obuchi, Rokkasho, Aomori 039-3212 (Japan); Shimizu, K. [Japan Atomic Energy Agency, 801-1, Mukoyama, Naka 311-0193 (Japan); Asakura, N. [Japan Atomic Energy Agency, 2-166, Omotedate, Obuchi, Rokkasho, Aomori 039-3212 (Japan)

    2015-08-15

    Edge localized modes (ELMs) in the H-mode operation of tokamak reactors may be suppressed/mitigated by the resonant magnetic perturbation (RMP), but RMP coils are considered incompatible with DEMO reactors under the strong neutron flux. We propose an innovative concept of the RMP without installing coils but inserting ferritic steels of the helical configuration. Helically perturbed field is naturally formed in the axisymmetric toroidal field through the helical ferritic steel inserts (FSIs). When ELMs are avoided, large stationary heat load on divertor plates can be reduced by adopting a flux-tube-expansion (FTE) divertor like an X divertor. Separatrix shape and divertor-plate inclination are similar to those of a simple long-leg divertor configuration. Combination of the helical FSIs and the FTE divertor is a suitable method for the heat control to avoid transient ELM heat pulse and to reduce stationary divertor heat load in a tokamak DEMO reactor.

  17. SPIRAL field mapping on NSTX for comparison to divertor RF heat deposition

    Science.gov (United States)

    Hosea, J. C.; Perkins, R.; Jaworski, M. A.; Kramer, G. J.; Ahn, J.-W.; Bertelli, N.; Gerhardt, S.; Gray, T. K.; LeBlanc, B. P.; Maingi, R.; Phillips, C. K.; Roquemore, L.; Ryan, P. M.; Sabbagh, S.; Taylor, G.; Tritz, K.; Wilson, J. R.; NSTX Team

    2014-02-01

    Field-aligned losses of HHFW power in the SOL of NSTX have been studied with IR cameras and probes, but the interpretation of the data depends somewhat on the magnetic equilibrium reconstruction. Both EFIT02 and LRDFIT04 magnetic equilibria have been used with the SPIRAL code to provide field mappings in the scrape off layer (SOL) on NSTX from the midplane SOL in front of the HHFW antenna to the divertor regions, where the heat deposition spirals are measured. The field-line mapping spiral produced at the divertor plate with LRDFIT04 matches the HHFW-produced heat deposition best, in general. An independent method for comparing the field-line strike patterns on the outer divertor for the two equilibria is provided by measuring Langmuir probe characteristics in the vicinity of the outer vessel strike radius (OVSR) and observing the effect on floating potential, saturation current, and zero-probe-voltage current (IV=0) with the crossing of the OVSR over the probe. Interestingly, these comparisons also reveal that LRDFIT04 gives the more accurate location of the predicted OVSR, and confirm that the RF power flow in the SOL is essentially along the magnetic field lines. Also, the probe characteristics and IV=0 data indicate that current flows under the OVSR in the divertor tiles in most cases studied.

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

    DEFF Research Database (Denmark)

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

    2012-01-01

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

  19. Numerical analysis of high Mach flow and flow reversal in the experimental advanced superconducting tokamak divertor

    Institute of Scientific and Technical Information of China (English)

    Ou Jing; Yang Jin-Hong

    2011-01-01

    The B2-Eirene (SOLPS 4.0) code package is used to investigate the plasma parallel flow,i.e.,the scrape-off layer (SOL) flow,in the experimental advanced superconducting tokamak (EAST) divertor. Simulation results show that the SOL flow in the divertor region can exhibit complex behaviour,such as a high Mach flow and flow reversal in different plasma regimes. When the divertor plasma is in the detachment state,the high Mach flow with approaching or exceeding sonic speed is observed away from the target plate in our simulation. When the divertor plasma is in the high recycling The driving mechanisms for the high Mach flow and the reversed flow are analysed theoretically through momentum and continuity equations,respectively. The profile of the ionization sources is shown to be a possible formation condition causing the complex behaviour of the SOL flow. In addition,the effects of the high Mach flow and the flow reversal on the impurity transport are also discussed in this paper.

  20. Appearance of hot spots due to deposits in the JET MKII-HD outer divertor

    NARCIS (Netherlands)

    van Rooij, G. J.; Brezinsek, S.; Coad, J. P.; Fundamenski, W.; Philipps, V.; Arnoux, G.; Stamp, M. F.

    2009-01-01

    Deposited layers in the JET MKII-HD outer divertor have been investigated on the basis of their transient heating. The Planck radiation in the 400-600 nm wavelength range and IR thermography data were analyzed to correlate the appearance of the layers with plasma conditions. Both methods yielded sig

  1. JET contributions to the workshop on the new phase for JET: the pumped divertor proposal

    International Nuclear Information System (INIS)

    Contributions to the Workshop consist of 13 papers on the new phase of operation of JET, including an outline of the objectives of the study of impurity control and the operating domain relative to the next generation of tokamaks. Studies are presented on the pumped divertor proposed for JET, diagnostic measurements required, and the performance expectations in the new configuration. (U.K.)

  2. Impact of nitrogen seeding on carbon erosion in the JET divertor

    NARCIS (Netherlands)

    Brezinsek, S.; Jachmich, S.; Rapp, J.; Meigs, A. G.; Nicholas, C.; O' Mullane, M.; Pospieszczyk, A.; van Rooij, G. J.

    2011-01-01

    Nitrogen has been introduced in H-mode plasmas in JET in order to study its radiation cooling capability and impact on the erosion of divertor plasma-facing components made of carbon-fiber composites (CFC). Experiments in the ionizing plasma regime with low nitrogen injection show a reduction of the

  3. A procedure for generating quantitative 3-D camera views of tokamak divertors

    International Nuclear Information System (INIS)

    A procedure is described for precision modeling of the views for imaging diagnostics monitoring tokamak internal components, particularly high heat flux divertor components. These models are required to enable predictions of resolution and viewing angle for the available viewing locations. Because of the oblique views expected for slot divertors, fully 3-D perspective imaging is required. A suite of matched 3-D CAD, graphics and animation applications are used to provide a fast and flexible technique for reproducing these views. An analytic calculation of the resolution and viewing incidence angle is developed to validate the results of the modeling procedures. The calculation is applicable to any viewed surface describable with a coordinate array. The Tokamak Physics Experiment (TPX) diagnostics for infrared viewing are used as an example to demonstrate the implementation of the tools. For the TPX experiment the available locations are severely constrained by access limitations at the end resulting images are marginal in both resolution and viewing incidence angle. Full coverage of the divertor is possible if an array of cameras is installed at 45 degree toroidal intervals. Two poloidal locations are required in order to view both the upper and lower divertors. The procedures described here provide a complete design tool for in-vessel viewing, both for camera location and for identification of viewed surfaces. Additionally these same tools can be used for the interpretation of the actual images obtained by the actual diagnostic

  4. Enhanced -->E*-->B drift effects in the TCV snowflake divertor

    NARCIS (Netherlands)

    G.P. Canal,; Lunt, T.; Reimerdes, H.; Duval, B. P.; Labit, B.; Vijvers, W. A. J.; TCV team,

    2015-01-01

    Measurements of various plasma parameters at the divertor targets of snowflake (SF) and conventional single-null configurations indicate an enhanced effect of the -->E*-->B drift in the scrape-off layer of plasmas in the SF configuration. Plasma boundary transport simulations using the EMC3-Ei

  5. The Influence of Filaments in the Private Flux Region on Divertor Particle and Power Deposition

    CERN Document Server

    Harrison, J R; Thornton, A J; Walkden, N R

    2015-01-01

    The transport of particles via intermittent filamentary structures in the private flux region of plasmas in the MAST tokamak has been investigated using a fast framing camera recording visible light emission from the volume of the lower divertor, as well as Langmuir probes and IR thermography monitoring particle and power fluxes to plasma-facing surfaces in the divertor. The visible camera data suggests that, in the divertor volume, fluctuations in light emission above the X-point are strongest in the scrape-off layer (SOL). Conversely, in the region below the X-point, it is found that these fluctuations are strongest in the private flux region (PFR) of the inner divertor leg. Detailed analysis of the appearance of these filaments in the camera data suggests that they are approximately circular, around 1-2cm in diameter. The most probable toroidal mode number is between 2 and 3. These filaments eject plasma deeper into the private flux region, sometimes by the production of secondary filaments, moving at a sp...

  6. Flow Field and Thermal Analysis of the Divertor Target Plate for HL-2A Tokamak

    Institute of Scientific and Technical Information of China (English)

    2005-01-01

    In the initial phase of the physics experiment, the double-null divertor plates used consist of graphite armor tiles, Mo-alloy intermediate layers and Cu-alloy coolant tubes. In the later operating phase, tungsten will be used as armor tiles.A multi-physical field numerical analysis method is used in this paper. Its analysis model reflects more realistically the real divertor structure than other models. Two-dimensional (2D)and three-dimensional (3D) fluid flow field, temperature distribution and thermal stress analyses of the divertor plates are carried out by the ANSYS code. During the physics experimental phase with a heat flux of 1 MW/m2, a coolant velocity of 5.48 m/s, and a thermal stress of 750 kg/cm2,the graphite armor tiles successfully meet the requirements of temperature, thermal stress and sputtering erosion. The tungsten armor will be considered as a second candidate. The result of simulation can be used for upgrading the design parameters of the HL-2A poloidal divertor.

  7. Particle exhaust modeling for the collaborative DIII-D Advanced Divertor Program

    International Nuclear Information System (INIS)

    A principal objective of the collaborative DIII-D Divertor Program (ADP) is to achieve density control in H-mode discharges with edge biasing and with continuous particle exhaust at a rate determined by the external fueling sources (typically 20 Torr·L/s). The divertor baffle-bias ring system has been optimized for pumping speeds ∼50,000 L/s with the neutral transport code DEGAS. With an entrance slot conductance of 50,000 L/s, a pumping speed of the same order is required to remove half of the ∼40 Torr·L/s that enters the baffle chamber for typical H-mode discharges. Increasing the exhaust fraction with higher pumping speed is self-limiting, owing to the attendant reduction of the recycling flux. The effects of pumping on the plasma core, scrape-off layer (SOL), and divertor have been estimated with a model that self-consistently couples the transport in these regions. The required ∼50,000 L/s pumping speed can be achieved with either titanium getter pumps or cryopumps. Evaluation of both systems has led to the conclusion that cryopumps will be more compatible with the environment of the DIII-D divertor. 8 refs., 7 figs

  8. The WEST project: preparing power exhaust control for ITER tungsten divertor operation

    International Nuclear Information System (INIS)

    Full text of publication follows. Power exhaust in next step steady state fusion devices will require complex integrated control schemes. The seeding of impurity is foreseen to increase the radiation fraction but with a price to pay on energy confinement. To optimize the plasma performance one will want to minimize the radiation fraction and thus operate close to the technological limit of the plasma facing components (PFC) in terms of power handling. In order to do so, accurate knowledge of the PFC power load is required in real time. Underestimating it will lead to degradation of the PFC and eventually to water leaks while overestimating it will unnecessarily constrain access to high fusion performance. ITER baseline plans the use of a full tungsten (W) divertor for the nuclear phase and discussions to start divertor operation with the full W divertor are ongoing. Simulations have shown that, in the burning phase, the maximum allowable steady state heat flux for the actively cooled divertor can be largely exceeded, typically by a factor 4 if the radiated fraction in the divertor falls to 20%. Therefore, the control of the power exhaust will be mandatory for safe operation. In contrast with present day devices, the metallic environment and the accessibility in ITER will severely constrain power load measurement and further tools will have to be developed in order to properly master the steady state power exhaust. This control issue will be addressed in detail in the frame of the WEST project implementing an actively cooled W divertor representative of ITER PFC inside the long pulse tokamak Tore Supra. Large heat fluxes will be made available in steady state (above 20 MW/m2) and a set of relevant diagnostics will be installed (magnetics, infrared/visible thermography, water calorimetry, thermocouples, etc.). Steady state PFC heat patterns have been simulated (PFCflux code) as well as the associated reflections (SPEOS code) in the complex geometry for different WEST

  9. Divertor modelling for conceptual studies of tokamak fusion reactor FDS-III

    International Nuclear Information System (INIS)

    The tokamak fusion power plant FDS-III with major radius R-5.1 m, minor radius a=l.7 m, plasma current Ip-16.0 MA, toroidal field Bt=8.0 T, elongation k-1.7, triangularity 6=0.59, edge safe factor q95=3.33, toroidal β, βT=5.64%, poloidal β, βp=1.88 and normalized β, βN=4.8 has been proposed. The divertor simulation which aims at optimizing the conceptual design of divertor in the reactor FDS-III has been done by using the edge plasma code package SOLPS5.0 (B2.5-EIRENE). The simulation is performed self-consistently with the parameters in core plasma and the MHD equilibrium in the reactor, a MHD equilibrium code EFIT is employed for the equilibrium computation and the equilibrium configuration is used for the SOLPS5.0 (B2.5-EIRENE) simulation. The real reactor geometry, drive power, fusion power and a particle power are taken into account, the plasma species include D+, T+, He+2, impurity ions and the neutrals in the simulation. The distribution of plasma parameters and heat fluxes in the divertor region has been obtained with pumping and gas puffing, the possibility assessment of the He ash removal and heat exhaust of the divertor has been carried out based on the simulation. The simulation results can be used for the engineering design of divertor in the reactor. (authors)

  10. Design and tests of a simplified divertor dummy coil structure for the WEST project

    International Nuclear Information System (INIS)

    Full text of publication follows. In order to fully validate actively cooled tungsten plasma facing components (industrial fabrication, operation with long plasma duration), the implementation of a tungsten axisymmetric divertor structure in the tokamak Tore-Supra is studied. With this major upgrade, so called WEST (Tungsten Environment in Steady state Tokamak), Tore-Supra will be able to address the problematic of long plasma discharges with a metallic divertor target. To do so, it is planned to install two symmetric divertor coils inside the vacuum vessel. This assembly, called divertor structure, is made up of two stainless steel casings containing a copper winding pack cooled by a pressurized hot water circuit (up to 200 Celsius degrees, 4 MPa) and is designed to perform steady state plasma operation (up to 1000 s). The divertor structure will be a complex assembly of 4 meter diameter and 4 meter height representing a total weight of around 20 tonnes. The technical challenge of this component will be the implementation of angular sectors inside the vacuum vessel environment (TIG welding of the coil casing, induction brazing and electrical insulation of the copper winding). Moreover, this complex assembly must sustain harsh environmental conditions in terms of ultra high vacuum conditions, mechanical loads (induced by disruptions) and electrical isolation (13 kV test) under high temperature. In order to fully validate the feasibility, the mounting and the performance of this complex component, the production of a scale one dummy coil is in progress. The paper will illustrate, the technical developments performed during 2012 in order to finalise the design for the call for tender phase. The progress and the first results of the simplified dummy coils will be also addressed. (authors)

  11. In-pile thermocycling testing and post-test analysis of beryllium divertor mockups

    Energy Technology Data Exchange (ETDEWEB)

    Giniatulin, R.; Mazul, I. [Efremov Inst., St. Petersburg (Russian Federation); Melder, R.; Pokrovsky, A.; Sandakov, V.; Shiuchkin, A.

    1998-01-01

    The main damaging factors which impact the ITER divertor components are neutron irradiation, cyclic surface heat loads and hydrogen environment. One of the important questions in divertor mockups development is the reliability of beryllium/copper joints and the beryllium resistance under neutron irradiation and thermal cycling. This work presents the experiment, where neutron irradiation and thermocyclic heat loads were applied simultaneously for two beryllium/copper divertor mockups in a nuclear reactor channel to simulate divertor operational conditions. Two mockups with different beryllium grades were mounted facing each other with the tantalum heater placed between them. This device was installed in the active zone of the nuclear reactor SM-2 (Dimitrovgrad, Russia) and the tantalum block was heated by neutron irradiation up to a high temperature. The main part of the heat flux from the tantalum surface was transported to the beryllium surface through hydrogen, as a result the heat flux loaded two mockups simultaneously. The mockups were cooled by reactor water. The device was lowered to the active zone so as to obtain the heating regime and to provide cooling lifted. This experiment was performed under the following conditions: tantalum heater temperature - 1950degC; hydrogen environment -1000 Pa; surface heat flux density -3.2 MW/m{sup 2}; number of thermal cycles (lowering and lifting) -101; load time in each cycle - 200-5000 s; dwell time (no heat flux, no neutrons) - 300-2000 s; cooling water parameters: v - 1 m/s, Tin - 50degC, Pin - 5 MPa; neutron fluence -2.5 x 10{sup 20} cm{sup -2} ({approx}8 years of ITER divertor operation from the start up). The metallographic analysis was performed after experiment to investigate the beryllium and beryllium/copper joint structures, the results are presented in the paper. (author)

  12. Investigation of Main-Chamber and Divertor Recycling in DIII-D Using Tangentially Viewing CID Cameras

    International Nuclear Information System (INIS)

    Measurements of the Dα emission profiles from the divertor and main chamber region in DIII-D, performed in low-density L-mode, and low and high-density ELMy H-mode plasmas imply that core plasma fueling occurs through the divertor channel. Emission profiles of carbon, combined with UEDGE modeling of the L-mode plasmas, also suggests that chemical sputtering of carbon from the flux surface adjacent to the inner divertor walls, and temperature gradient forces in the scrape-off layer, determine the carbon content of the inner main chamber scrape-off layer

  13. INVESTIGATION OF MAIN-CHAMBER AND DIVERTOR RECYCING IN DIII-D USING TANGENTIALLY VIEWING CID CAMERAS

    International Nuclear Information System (INIS)

    OAK-B135 Measurements of the Dα emission profiles from the divertor and main chamber region in DIII-D, performed in low-density L-mode, and low and high-density ELMy H-mode plasmas imply that core plasma fueling occurs through the divertor channel. Emission profiles of carbon, combined with UEDGE modeling of the L-mode plasmas, also suggests that chemical sputtering of carbon from the flux surface adjacent to the inner divertor walls, and temperature gradient forces in the scrape-off layer, determine the carbon content of the inner scrape-off layer

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. 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.

  16. Approaches towards Steady-State Advanced Divertor Operations on EAST by Active Control of Plasma-Wall Interactions

    International Nuclear Information System (INIS)

    Full text: EAST will be one of the world's first magnetic confinement devices that must address Plasma- Wall Interaction (PWI) issues facing high power steady-state operations. EAST has recently significantly augmented its RF heating capabilities up to 10 MW, including LHCD and ICRH. It has also undertaken an extensive upgrade during the recent shutdown to replace the carbon tiles on the main chamber wall and divertor surface by the Mo tiles, except those near the strike points, allowing baking up to 250 deg C, with active water cooling. The divertor titles will further be upgraded to monoblock Tungsten, as to be used in ITER, to address PWI issues for ITER and DEMO. EAST demonstrated long pulse operation over 100 s, entirely driven by LHCD during the last experimental campaign. In order to achieve this, the following major means were applied to EAST to actively control PWI interactions: 1. Active divertor pumping using an in-vessel large capacity cryopump for facilitating density control. 2. Advanced wall conditioning with Lithium (Li) evaporation and real-time, in-situ Li powder injection for controlling neutral recycling. 3. Localized divertor gas puffing for reducing peak heat fluxes near the strike points. 4. Strike point sweeping to spread the heat loads on the divertor target plates. In addition, highly radiative impurity Ar was injected into the divertor to further reduce the peak divertor heat fluxes and mitigate the in-out divertor plasma asymmetries in EAST. Despite the injection of Ar, Zeff in the core plasma was little affected, suggesting strong divertor screening. Ar seeding has also been explored in the newly achieved H-modes in EAST, significantly increasing the frequency and decreasing the amplitude of ELMs, thus reducing the particle and heat loads on the divertor target plates. These first results are very promising, and will further be investigated in EAST for high power, long pulse operations. EAST has now just started a new experimental

  17. Challenges to radiative divertor/mantle operations in advanced, steady-state scenarios

    International Nuclear Information System (INIS)

    Full text of publication follows. Managing the heat exhaust problem is well recognized to be a major challenge in transforming present successes in magnetic confinement fusion experiments to demonstration of cost-effective, steady-state power generation from fusion [1][2]. One approach is to convert plasma thermal energy, normally directed to isolated surfaces, to isotropic photon emission, distributing exhaust power over a large surface area. Successful demonstrations of this technique on existing short pulse devices are shown, along with the inherent limitations; the collapse of core confinement with excessive radiation from the bulk plasma and restrictions to dissipation in the divertor volume. Feedback control of impurity seeding is discussed, showing recent examples from tokamaks [3]. For steady-state devices, additional constraints on divertor scenarios are driven by long-term plasma material interaction effects, with fuel recycling, net erosion limits and surface morphology changes forcing detached plasma operation where both heat and particle fluxes are substantially reduced. The instability of these detachment layers in standard X-point divertors with impurity seeding is outlined. Achieving these steady-state, high performance scenarios also restricts the divertor solution by requiring it be compatible with current-drive actuators and enhanced core confinement regimes. While ITER will operate with impurity seeding in a conventional tokamak geometry [4], it is not clear that this concept will reliably scale to a reactor and has been identified as a major risk factor in the development of fusion power [2]. Alternatives concepts are discussed, including the snowflake [5] and super-X divertor [6], along with their respective proof of principle experiments. The complications in convincingly scaling these concepts to a reactor are outlined, including challenges in validating numerical simulations of advanced, dissipative divertors. References: [1] Greenwald, M

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

    Energy Technology Data Exchange (ETDEWEB)

    Wuerz, H. [Kernforschungszentrum Karlsruhe, INR, Postfach 36 40, D-76021 Karlsruhe (Germany); Arkhipov, N.I. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Bakhtin, V.P. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Konkashbaev, I. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Landman, I. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Safronov, V.M. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Toporkov, D.A. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Zhitlukhin, A.M. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation)

    1995-04-01

    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.)).

  19. Upgrade of Langmuir probe diagnostic in ITER-like tungsten mono-block divertor on experimental advanced superconducting tokamak

    Science.gov (United States)

    Xu, J. C.; Wang, L.; Xu, G. S.; Luo, G. N.; Yao, D. M.; Li, Q.; Cao, L.; Chen, L.; Zhang, W.; Liu, S. C.; Wang, H. Q.; Jia, M. N.; Feng, W.; Deng, G. Z.; Hu, L. Q.; Wan, B. N.; Li, J.; Sun, Y. W.; Guo, H. Y.

    2016-08-01

    In order to withstand rapid increase in particle and power impact onto the divertor and demonstrate the feasibility of the ITER design under long pulse operation, the upper divertor of the EAST tokamak has been upgraded to actively water-cooled, ITER-like tungsten mono-block structure since the 2014 campaign, which is the first attempt for ITER on the tokamak devices. Therefore, a new divertor Langmuir probe diagnostic system (DivLP) was designed and successfully upgraded on the tungsten divertor to obtain the plasma parameters in the divertor region such as electron temperature, electron density, particle and heat fluxes. More specifically, two identical triple probe arrays have been installed at two ports of different toroidal positions (112.5-deg separated toroidally), which can provide fundamental data to study the toroidal asymmetry of divertor power deposition and related 3-dimension (3D) physics, as induced by resonant magnetic perturbations, lower hybrid wave, and so on. The shape of graphite tip and fixed structure of the probe are designed according to the structure of the upper tungsten divertor. The ceramic support, small graphite tip, and proper connector installed make it possible to be successfully installed in the very narrow interval between the cassette body and tungsten mono-block, i.e., 13.5 mm. It was demonstrated during the 2014 and 2015 commissioning campaigns that the newly upgraded divertor Langmuir probe diagnostic system is successful. Representative experimental data are given and discussed for the DivLP measurements, then proving its availability and reliability.

  20. Calculations of energy losses due to atomic processes in tokamaks with applications to the International Thermonuclear Experimental Reactor divertor

    International Nuclear Information System (INIS)

    Reduction of the peak heat loads on the plasma facing components is essential for the success of the next generation of high fusion power tokamaks such as the International Thermonuclear Experimental Reactor (ITER) [Rebut et al., Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, in press)]. Many present concepts for accomplishing this involve the use of atomic processes to transfer the heat from the plasma to the main chamber and divertor chamber walls and much of the experimental and theoretical physics research in the fusion program is directed toward this issue. The results of these experiments and calculations depend upon a complex interplay of many processes. In order to identify the key features of these experiments and calculations and the relative role of the primary atomic processes, simple quasianalytic models and the latest atomic physics rate coefficients and cross sections have been used to assess the relative roles of central radiation losses through bremsstrahlung, impurity radiation losses from the plasma edge, charge exchange and hydrogen radiation losses from the scrape-off layer, and divertor plasma and impurity radiation losses from the divertor plasma. This analysis indicates that bremsstrahlung from the plasma center and impurity radiation from the plasma edge and divertor plasma can each play a significant role in reducing the power to the divertor plates, and identifies many of the factors which determine the relative role of each process. For instance, for radiation losses in the divertor to be large enough to radiate the power in the divertor for high power experiments, a neutral fraction of 10-3 to 10-2 and an impurity recycling rate of neτrecycle of ∼1016 s m-3 will be required in the divertor

  1. Self-consistent treatment of the sheath boundary conditions by introducing anisotropic ion temperatures and virtual divertor model

    Science.gov (United States)

    Togo, Satoshi; Takizuka, Tomonori; Nakamura, Makoto; Hoshino, Kazuo; Ibano, Kenzo; Lang, Tee Long; Ogawa, Yuichi

    2016-04-01

    One-dimensional SOL-divertor plasma fluid simulation code which considers anisotropy of ion temperature has been developed so as to deal with sheath theory self-consistently. In our fluid modeling, explicit use of boundary condition for Mach number M at divertor plate, e.g., M = 1, becomes unnecessary. In order to deal with the Bohm condition and the sheath heat transmission factors at divertor plate self-consistently, we introduced a virtual divertor (VD) model which sets an artificial region beyond divertor plates and artificial sinks for particle, momentum and energy there to model the effects of the sheath region in front of the divertor plate. Validity of our fluid model with VD model is confirmed by showing that simulation results agree well with those from a kinetic code regarding the Bohm condition, ion temperature anisotropy and supersonic flow. We also show that the strength of artificial sinks in VD region does not affect profiles in plasma region at least in the steady state and that sheath heat transmission factors can be adjusted to theoretical values by VD model. Validity of viscous flux is also investigated.

  2. Plasma decontamination during ergodic divertor experiments in Tore Supra

    International Nuclear Information System (INIS)

    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 innerwall. 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 centimetres 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

  3. Upgrade of the infrared camera diagnostics for the JET ITER-like wall divertor.

    Science.gov (United States)

    Balboa, I; Arnoux, G; Eich, T; Sieglin, B; Devaux, S; Zeidner, W; Morlock, C; Kruezi, U; Sergienko, G; Kinna, D; Thomas, P D; Rack, M

    2012-10-01

    For the new ITER-like wall at JET, two new infrared diagnostics (KL9B, KL3B) have been installed. These diagnostics can operate between 3.5 and 5 μm and up to sampling frequencies of ∼20 kHz. KL9B and KL3B image the horizontal and vertical tiles of the divertor. The divertor tiles are tungsten coated carbon fiber composite except the central tile which is bulk tungsten and consists of lamella segments. The thermal emission between lamellae affects the surface temperature measurement and therefore KL9A has been upgraded to achieve a higher spatial resolution (by a factor of 2). A technical description of KL9A, KL9B, and KL3B and cross correlation with a near infrared camera and a two-color pyrometer is presented.

  4. 3D plasma fluid simulations in divertor tokamaks. Final technical report, 1993--1995

    International Nuclear Information System (INIS)

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

  5. Simple Core-SOL-Divertor model and its application to operational space of HT-7U

    International Nuclear Information System (INIS)

    We develop a simple Core-SOL-Divertor (C-S-D) model to investigate qualitatively the overall features of the operational space for the integrated core and edge plasma. To construct the simple C-S-D model, a simple core plasma model of ITER physics guidelines and a two-point SOL-divertor model are used. The simple C-S-D model is applied to the study of the HT-7U operational space with lower hybrid current drive experiments under various kinds of trade-off for the basic plasma parameters. Effective methods for extending the operation space are also presented. From this study for the HT-7U operation space, it is shown that the C-S-D model is a useful tool to understand qualitatively the overall features of the plasma operation space

  6. Overall feature of EAST operation space by using simple Core-SOL-Divertor model

    International Nuclear Information System (INIS)

    We have developed a simple Core-SOL-Divertor (C-S-D) model to investigate qualitatively the overall features of the operational space for the integrated core and edge plasma. To construct the simple C-S-D model, a simple core plasma model of ITER physics guidelines and a two-point SOL-divertor model are used. The simple C-S-D model is applied to the study of the EAST operational space with lower hybrid current drive experiments under various kinds of trade-off for the basic plasma parameters. Effective methods for extending the operation space are also presented. As shown by this study for the EAST operation space, it is evident that the C-S-D model is a useful tool to understand qualitatively the overall features of the plasma operation space. (author)

  7. Analytic Criteria for Power Exhaust in Divertors due to Impurity Radiation

    CERN Document Server

    Post, D; Perkins, F W; Nevins, W

    1995-01-01

    Present divertor concepts for next step experiments such ITER and TPX rely upon impurity and hydrogen radiation to transfer the energy from the edge plasma to the main chamber and divertor chamber walls. The efficiency of these processes depends strongly on the heat flux, the impurity species, and the connection length. Using a database for impurity radiation rates constructed from the ADPAK code package, we have developed criteria for the required impurity fraction, impurity species, connection length and electron temperature and density at the mid-plane. Consistent with previous work, we find that the impurity radiation from coronal equilibrium rates is, in general, not adequate to exhaust the highest expected heating powers in present and future experiments. As suggested by others, we examine the effects of enhancing the radiation rates with charge exchange recombination and impurity recycling, and develop criteria for the minimum neutral fraction and impurity recycling rate that is required to exhaust a s...

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

    International Nuclear Information System (INIS)

    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

  9. Modelling of passive spectroscopy in the ITER divertor: the first hydrogen Balmer lines

    International Nuclear Information System (INIS)

    The first lines of the hydrogen Balmer series are investigated in ITER divertor conditions using a line shape code and a plasma edge transport code. It is shown that most of the emissivity originates from a localized, cold and dense region close to the divertor target plates, where the plasma is in the recombining regime. We simulate the signal obtained by pointing a spectrometer at this zone. The physical processes which contribute to the spectral line formation are examined, with a special emphasis on the Stark effect, photon absorption and stimulated emission. It is shown that, even though the Stark effect is significant, local information on the Doppler atomic temperature can be obtained from a fitting analysis of the Dα spectral line shape.

  10. Modelling of passive spectroscopy in the ITER divertor: the first hydrogen Balmer lines

    Science.gov (United States)

    Rosato, J.; Kotov, V.; Reiter, D.

    2010-07-01

    The first lines of the hydrogen Balmer series are investigated in ITER divertor conditions using a line shape code and a plasma edge transport code. It is shown that most of the emissivity originates from a localized, cold and dense region close to the divertor target plates, where the plasma is in the recombining regime. We simulate the signal obtained by pointing a spectrometer at this zone. The physical processes which contribute to the spectral line formation are examined, with a special emphasis on the Stark effect, photon absorption and stimulated emission. It is shown that, even though the Stark effect is significant, local information on the Doppler atomic temperature can be obtained from a fitting analysis of the Dα spectral line shape.

  11. Free-boundary ideal MHD stability of W7-X divertor equilibria

    Science.gov (United States)

    Nührenberg, C.

    2016-07-01

    Plasma configurations describing the stellarator experiment Wendelstein 7-X (W7-X) are computationally established taking into account the geometry of the test-divertor unit and the high-heat-flux divertor which will be installed in the vacuum chamber of the device (Gasparotto et al 2014 Fusion Eng. Des. 89 2121). These plasma equilibria are computationally studied for their global ideal magnetohydrodynamic (MHD) stability properties. Results from the ideal MHD stability code cas3d (Nührenberg 1996 Phys. Plasmas 3 2401), stability limits, spatial structures and growth rates are presented for free-boundary perturbations. The work focusses on the exploration of MHD unstable regions of the W7-X configuration space, thereby providing information for future experiments in W7-X aiming at an assessment of the role of ideal MHD in stellarator confinement.

  12. High radiation from intrinsic and injected impurities in Tore Supra ergodic divertor plasmas

    International Nuclear Information System (INIS)

    We report experiments aimed at comparing several impurity mixtures (C, O, Cl, N, Ne, Ar) regarding their capability to reduce the power load on the divertor target plates. The divertor conditions required for each mixture to minimise the parallel power flux are determined, along with the resulting core effective charge Zeff and volume averaged density. The radiation efficiency (ratio of edge radiation to plasma core contamination) of intrinsic carbon is found to increase with the total injected power. In the impurity injection experiments, nitrogen is found to be the best choice to reduce the power flux to the target plates: it has the same characteristics as C/O radiation (low core contamination), and it can be controlled. The low Zeff observed in this case is attributed to the large value of the screening of the radiating ionisation stages of the impurity

  13. Spectroscopic study of neon emission and retention in the Tore Supra ergodic divertor

    Energy Technology Data Exchange (ETDEWEB)

    Guirlet, R. E-mail: guirlet@drfc.cad.cea.fr; Hogan, J.; Corre, Y.; Michelis, C. de; Escarguel, A.; Hess, W.; Monier-Garbet, P.; Schunke, B

    2001-03-01

    In order to assess the capability of the Tore Supra ergodic divertor (ED) to retain impurities in the low confinement edge region, spectroscopic observations of a divertor neutraliser plate are reported. The neutral neon density is deduced from these measurements; it increases strongly (up to 1.5x10{sup 17} m{sup -3} per injected Pa l) when the plasma approaches detachment. The central neon density is approximately independent of the plasma edge conditions. A 2D model confirms the relatively weak measured dependence of the neutral neon penetration on edge electron density and temperature. Comparison of BBQ (3D scrape-off layer Monte-Carlo code) results with 1D impurity radial transport modelling suggests a possible mechanism for the observed weak dependence of core content on edge impurity influx: enhanced exchange between the ergodized layer of the core and the neutraliser region.

  14. High radiation from intrinsic and injected impurities in Tore Supra ergodic divertor plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Monier-Garbet, P. E-mail: monier@drfc.cad.cea.fr; DeMichelis, C.; Ghendrih, Ph.; Grisolia, C.; Grosman, A.; Guirlet, R.; Gunn, J.; Loarer, T.; Bush, C.E.; Clement, C.; Corre, Y.; Costanzo, L.; Schunke, B.; Vallet, J.C

    2001-03-01

    We report experiments aimed at comparing several impurity mixtures (C, O, Cl, N, Ne, Ar) regarding their capability to reduce the power load on the divertor target plates. The divertor conditions required for each mixture to minimise the parallel power flux are determined, along with the resulting core effective charge Z{sub eff} and volume averaged density. The radiation efficiency (ratio of edge radiation to plasma core contamination) of intrinsic carbon is found to increase with the total injected power. In the impurity injection experiments, nitrogen is found to be the best choice to reduce the power flux to the target plates: it has the same characteristics as C/O radiation (low core contamination), and it can be controlled. The low Z{sub eff} observed in this case is attributed to the large value of the screening of the radiating ionisation stages of the impurity.

  15. Microscopically nonuniform deposition and deuterium retention in the divertor in JET with ITER-like wall

    Energy Technology Data Exchange (ETDEWEB)

    Bergsåker, H., E-mail: henricb@kth.se [Department of Fusion Plasma Physics, Association EURATOM-VR, School of Electrical Engineering, KTH Royal Institute of Technology, S-10044 Stockholm (Sweden); Bykov, I.; Petersson, P. [Department of Fusion Plasma Physics, Association EURATOM-VR, School of Electrical Engineering, KTH Royal Institute of Technology, S-10044 Stockholm (Sweden); Possnert, G. [Uppsala Universitet, Tandem Laboratory, Association EURATOM-VR, S-75105 Uppsala (Sweden); Likonen, J.; Koivuranta, S.; Coad, J.P. [VTT, Association Euratom-Tekes, PO Box 1000, FI-02044 VTT (Finland); Van Renterghem, W.; Uytdenhouwen, I. [SCK-CEN, Institute for Nuclear Material Sciences, Boeretang 200, 2400 Mol (Belgium); Widdowson, A.M. [JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom)

    2015-08-15

    The divertor surfaces in JET with ITER-like wall (ILW) have been studied using micro ion beam analysis (μ-IBA) methods and scanning electron microscopy (SEM). Deposited layers with beryllium as main constituent had been formed during plasma operations through 2011–2012. The deuterium trapping and impurity deposition were non-uniform, frequently enhanced within pits, cracks and valleys, regions reaching in size from 10 μm to 200 μm. The impurity deposition and fuel retention were correlated with the surface slope with respect to the direction of ion incidence. Typically more than 70% of the total measured areal density of trapped D was found in less than 30% of the surface area. This is of consequence for the interpretation of other surface analyses and in extrapolation from fuel retention in JET with ITER-like wall and rough divertor surfaces to ITER with smoother surfaces.

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

    Science.gov (United States)

    Griswold, M. E.; Korepanov, S.; Thompson, M. C.

    2016-11-01

    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.

  17. Observation of Non-Maxwellian Electron Distributions in th e NSTX Divertor

    Energy Technology Data Exchange (ETDEWEB)

    M.A. Jaworski, et. al.

    2013-03-07

    The scrape-off layer plasma at the tokamak region is characterized by open field lines and often contains large variations in plasma properties along these field-lines. Proper characterization of local plasma conditions is critical to assessing plasma-material interaction processes occuring at the target. Langmuir probes are frequently employed in tokamak divertors but are challenging to interpretation. A kinetic interpretation for Langmuir probes in NSTX has yielded non-Maxwellian electron distributions in the divertor characterized by cool bulk populations and energetic tail populations with temperatures of 2-4 times the bulk. Spectroscopic analysis and modeling confirms the bulk plasma temperature and density which can only be obtained with the kinetic interpretation

  18. Divertor heat flux footprints in EDA H-mode discharges on Alcator C-Mod

    International Nuclear Information System (INIS)

    The physics that sets the width of the power exhaust channel in a tokamak scrape-off layer and its scaling with engineering parameters is of fundamental importance for reactor design, yet it remains to be understood. An extensive array of divertor heat flux diagnostics was recently commissioned in Alcator C-Mod with the aim of improving our understanding. Initial results are reported from EDA H-mode discharges in which plasma current, input power, toroidal field and magnetic topology were varied. The integral width of the outer divertor heat flux footprint is found to lie in the range of 3-5 mm mapped to the mid-plane. Widths are insensitive to single versus double-null topology and the magnitude of toroidal field. Pedestal physics appears to largely determine these widths; a dependence of width on plasma thermal energy is noted, yielding a reduction in width as plasma current is increased for the best EDA H-modes.

  19. Divertor heat flux footprints in EDA H-mode discharges on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    LaBombard, B., E-mail: labombard@psfc.mit.edu [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Terry, J.L.; Hughes, J.W.; Brunner, D.; Payne, J.; Reinke, M.L.; Lin, Y.; Wukitch, S. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)

    2011-08-01

    The physics that sets the width of the power exhaust channel in a tokamak scrape-off layer and its scaling with engineering parameters is of fundamental importance for reactor design, yet it remains to be understood. An extensive array of divertor heat flux diagnostics was recently commissioned in Alcator C-Mod with the aim of improving our understanding. Initial results are reported from EDA H-mode discharges in which plasma current, input power, toroidal field and magnetic topology were varied. The integral width of the outer divertor heat flux footprint is found to lie in the range of 3-5 mm mapped to the mid-plane. Widths are insensitive to single versus double-null topology and the magnitude of toroidal field. Pedestal physics appears to largely determine these widths; a dependence of width on plasma thermal energy is noted, yielding a reduction in width as plasma current is increased for the best EDA H-modes.

  20. Particle exhaust with vented structures: application to the ergodic divertor of Tore Supra

    International Nuclear Information System (INIS)

    In a thermonuclear reactor, one must continuously fuel the discharge and extract the ashes resulting from fusion reactions. To avoid the risk of discharge poisoning, α-particle concentration is limited to ∼ 10 %. To allow for steady-state conditions requires then to extract ≥2 % of the helium out flux. In Tore Supra, the ergodic divertor is the main component managing the heat and particle fluxes at the edge. Its principle consists in generating a resonant perturbation able to destroy magnetic surfaces at the plasma periphery. In this region, the field lines are open and connected at both ends to neutralizers which are wetted by the major part of the heat and particle fluxes and are the structures through which a part of the plasma out flux is pumped for maintaining the discharge in steady-state conditions. This work describes the neutral recirculation around the ergodic divertor and is based on a data base of 56 discharges. One discuss the two processes allowing for particle exhaust: the ballistic collection of ions and that of neutrals backscattered by atomic reactions. These two processes are modelled accounting for a realistic description of the divertor geometry. A comparison between simulations and experiments is presented for measurements characterising the three main actors of plasma-wall interaction: the edge plasma, the Dα light emission and the neutral pressure in the divertor plenum. Last, one question how such a system can be extrapolated to next step machines, for which one must account for technical constraints linked to the presence of the shield protecting the coils from the high neutron flux. (author)

  1. Recent progress in the NSTX/NSTX-U lithium programme and prospects for reactor-relevant liquid-lithium based divertor development

    Science.gov (United States)

    Ono, M.; Jaworski, M. A.; Kaita, R.; Kugel, H. W.; Ahn, J.-W.; Allain, J. P.; Bell, M. G.; Bell, R. E.; Clayton, D. J.; Canik, J. M.; Ding, S.; Gerhardt, S.; Gray, T. K.; Guttenfelder, W.; Hirooka, Y.; Kallman, J.; Kaye, S.; Kumar, D.; LeBlanc, B. P.; Maingi, R.; Mansfield, D. K.; McLean, A.; Menard, J.; Mueller, D.; Nygren, R.; Paul, S.; Podesta, M.; Raman, R.; Ren, Y.; Sabbagh, S.; Scotti, F.; Skinner, C. H.; Soukhanovskii, V.; Surla, V.; Taylor, C. N.; Timberlake, J.; Zakharov, L. E.; the NSTX Research Team

    2013-11-01

    Developing a reactor-compatible divertor has been identified as a particularly challenging technology problem for magnetic confinement fusion. Application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and other plasma performance benefits. During the 2010 NSTX campaign, application of a relatively modest amount of Li (300 mg prior to the discharge) resulted in a ˜50% reduction in heat load on the liquid lithium divertor (LLD) attributable to enhanced divertor bolometric radiation. These promising Li results in NSTX and related modelling calculations motivated the radiative LLD concept proposed here. Li is evaporated from the liquid lithium (LL) coated divertor strike-point surface due to the intense heat flux. The evaporated Li is readily ionized by the plasma due to its low ionization energy, and the poor Li particle confinement near the divertor plate enables ionized Li ions to radiate strongly, resulting in a significant reduction in the divertor heat flux. This radiative process has the desired effect of spreading the localized divertor heat load to the rest of the divertor chamber wall surfaces, facilitating the divertor heat removal. The LL coating of divertor surfaces can also provide a ‘sacrificial’ protective layer to protect the substrate solid material from transient high heat flux such as the ones caused by the edge localized modes. By operating at lower temperature than the first wall, the LL covered large divertor chamber wall surfaces can serve as an effective particle pump for the entire reactor chamber, as impurities generally migrate towards lower temperature LL divertor surfaces. To maintain the LL purity, a closed LL loop system with a modest circulating capacity (e.g., ˜1 l s-1 for ˜1% level ‘impurities’) is envisioned for a steady-state 1 GW-electric class fusion power plant.

  2. Infrared thermography inspection for monoblock divertor target in JT-60SA

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Shigetoshi, E-mail: nakamura.shigetoshi@jaea.go.jp; Sakurai, Shinji; Ozaki, Hidetsugu; Seki, Yohji; Yokoyama, Kenji; Sakasai, Akira; Tsuru, Daigo

    2014-10-15

    Highlights: • Infrared thermography inspection is modified to inspect JT-60SA divertor targets. • Infrared thermography inspection is effective to detect joining defects of targets. • Numerical analysis is in good agreement with inspection results of mock-up targets. • Database for setting screening criteria has been constructed by numerical analysis. - Abstract: Carbon fiber composite (CFC) monoblock divertor target is required for power handling in JT-60SA. Quality of the targets depends on a joining technology in manufacturing process. To inspect the quality of more than 900 target pieces, efficient non-destructive inspection is needed. An infrared thermography inspection (IR inspection), has been proposed by ITER and IRFM, where the quality between CFC and a cooling tube is examined by a use of transient thermal response at a rapid switch from hot to cold water flow. In JT-60SA divertor target, a screw tube will be employed to obtain high heat transfer efficiency with simple structure. Since the time response of the screw tube is much faster than that of smooth tube, it is required to confirm the feasibility of this IR inspection. Thus, the effect of joining defects on transient thermal response of the targets has been investigated experimentally by using the mock-up targets containing defects which are artificially made. It was found that the IR inspection can detect the defects. Moreover, screening criteria of IR inspection for acceptable monoblock target is discussed.

  3. Measurements of flows in the DIII-D divertor by Mach probes

    Energy Technology Data Exchange (ETDEWEB)

    Boedo, J.A.; Lehmer, R.; Moyer, R.A. [Univ. of California, San Diego, CA (United States); Watkins, J.G. [Sandia National Labs., Albuquerque, NM (United States); Porter, G.D. [Lawrence Livermore National Lab., NM (United States); Evans, T.E.; Leonard, A.W.; Schaffer, M.J. [General Atomics, San Diego, CA (United States)

    1998-06-01

    First measurements of Mach number of background plasma in the DIII-D divertor are presented in conjunction with temperature T{sub e} and density n{sub e} using a fast scanning probe array. To validate the probe measurements, the authors compared the T{sub e}, n{sub e} and J{sub sat} data to Thomson scattering data and find good overall agreement in attached discharges and some discrepancy for T{sub e} and n{sub e} in detached discharges. The discrepancy is mostly due to the effect of large fluctuations present during detached plasmas on the probe characteristic; the particle flux is accurately measured in every case. A composite 2-D map of measured flows is presented for an ELMing H-mode discharge and they focus on some of the details. They have also documented the temperature, density and Mach number in the private flux region of the divertor and the vicinity of the X-point, which are important transition regions that have been little studied or modeled. Background parallel plasma flows and electric fields in the divertor region show a complex structure.

  4. Attainment of high confinement in neutral beam heated divertor discharges in the PDX tokamak

    International Nuclear Information System (INIS)

    The PDX divertor configuration has recently been converted from an open to a closed geometry to inhibit the return of neutral gas from the divertor region to the main chamber. Since then, operation in a regime with high energy confinement in neutral beam heated discharges (ASDEX H-mode) has been routine over a wide range of operating conditions. These H-mode discharges are characterized by a sudden drop in divertor density and H/sub α/ emission and a spontaneous rise in main chamber plasma density during neutral beam injection. The confinement time is found to scale nearly linearly with plasma current, but it can be degraded due to either the presence of edge instabilities or heavy gas puffing. Detailed Thomson scattering temperature profiles show high values of Te near the plasma edge (approx. 450 eV) with sharp radial gradients (approx. 400 eV/cm) near the separatrix. Density profiles are broad and also exhibit steep gradients close to the separatrix

  5. The snowflake divertor, physics of a new concept for power exhaust of fusion plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Lunt, Tilmann; Feng, Yuehe [Max-Planck-Institut fuer Plasmaphysik, Garching/Greifswald (Germany); Canal, Gustavo; Reimerdes, Holger [Ecole Polytechnique Federale de Lausanne, Lausanne (Switzerland)

    2014-07-01

    Fusion reactors based on the tokamak design will have to deal with very high heat loads on the divertor plates. One of the approaches to solve this heat load problem is the so called 'snowflake divertor', a magnetic configuration with two nearby x-points and two additional divertor legs. In this contribution we report on 'EMC3-Eirene' simulations of the plasma- and neutral particle transport in the scrape-off layer of the swiss tokamak TCV of a series of snowflake equilibria with different values of σ, the distance between the x-points normalized to the minor radius of the plasma. The constant anomalous transport coefficients were chosen such that the power- and particle deposition profiles at the primary inner strike point match the Langmuir probe measurements for the σ=0.1 case. At one of the secondary strike points, however, a significantly larger power flux than that predicted by the simulation was measured by the probes, indicating the presence of an enhanced transport across the primary separatrix. We discuss the possible reason for this enhanced transport as well as its scaling with machine size. Another prediction from the simulation is that the density as well as the radiation maximum are moving from the recycling region in front of the plates upwards to the x-point.

  6. ITER divertor maintenance L7 R and D project - results and perspectives

    International Nuclear Information System (INIS)

    The availability of International Thermonuclear Experimental Reactor (ITER) (or an 'ITER-like') reactor will be strongly influenced by the effectiveness of the in-vessel components remote handling strategy. In the present reactor concept, the relevant key components are the divertor cassettes, located in the lower region of the vacuum vessel. Due to erosion and damage to the divertor plasma facing components, estimated scheduled replacement of the cassettes will be required eight times during the machine lifetime. Moreover, for such an experimental tokamak where completely new plasma regimes will be established, unscheduled interventions cannot be excluded a priori. To test and optimise the divertor maintenance scenario, the so called ITER L7 R and D project has been realised at the ENEA Brasimone laboratories, with a full scale simulation of the in-vessel cassette maintenance environment and of the hot cell refurbishment operations. The basic demonstration of the validity of the current scenario has been established, and now new activities are in progress to optimise many aspects of the operations (procedures, hardware improvements, reliability, etc.). Based on the background and the results of these activities, this paper discusses the lessons learned during the project implementation, and identifies key points of the current strategy that should be maintained for any new design of ITER or an 'ITER-like' reactor

  7. ASDEX upgrade - definition of a tokamak experiment with a reactor compatible polaoidal divertor

    International Nuclear Information System (INIS)

    ASDEX Upgrade is intended as the next experimental step after ASDEX. It is designed to investigate the physics of a divertor tokamak as closely as possible to fusion reactor requirements, without thermonuclear heating. It is characterized by a poloidal divertor configuration with divertor coils located outside the toroidal field coils, by machine parameters which allow a line density within the plasma boundary sufficient to screen fast CX particles from the plasma core, by a scrape-off layer essentially opaque to neutrals produced at the target plates, and, finally, by an auxiliary heating power high enough for producing a reactor-like power flux density through the plasma boundary. Design considerations on the basis of physical and technical constraints yielded the tokamak system optimized with respect to effort and costs as described in the following. It uses normal-conducting coil systems, is the size of ASDEX, and has a field of 3.9 T, a plasma current of up to 1.5 MA, and a pulse duration of 10 s. To provide the required power flux density, an ICRH power of 10 MW is needed. For comparison, a superconducting version is under investigation. (orig.)

  8. Modelling of surface evolution of rough surface on divertor target in fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Dai, Shuyu, E-mail: daishuyu@dlut.edu.cn [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China); Liu, Shengguang; Sun, Jizhong [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China); Kirschner, A. [Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, D-52425 Jülich (Germany); Kawamura, G. [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki, Gifu 509-5292 (Japan); Tskhakaya, D. [Association EURATOM – öAW, Institute of Applied Physics, TU Wien, A-1040 Vienna (Austria); Ding, Rui; Luo, Guangnan [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Wang, Dezhen, E-mail: wangdez@dlut.edu.cn [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China)

    2015-08-15

    Highlights: • We study the surface evolution of rough surface on divertor target in fusion devices. • The effects of gyration motion and E × B drift affect 3D angular distribution. • A larger magnetic field angle leads to a reduced net eroded areal density. • The rough surface evolution affects the physical sputtering yield. - Abstract: The 3D Monte-Carlo code SURO has been used to study the surface evolution of rough surface on the divertor target in fusion devices. The edge plasma at divertor region is modelled by the SDPIC code and used as input data for SURO. Coupled with SDPIC, SURO can perform more sophisticated simulations to calculate the local angle and surface evolution of rough surface. The simulation results show that the incident direction of magnetic field, gyration and E × B force has a significant impact on 3D angular distribution of background plasma and accordingly on the erosion of rough surface. The net eroded areal density of rough surface is studied by varying the magnetic field angle with surface normal. The evolution of the microscopic morphology of rough surface can lead to a significant change in the physical sputtering yield.

  9. Divertor cassette locking system remote handling trials with WHMAN at DTP2

    Energy Technology Data Exchange (ETDEWEB)

    Lyytikäinen, Ville; Kinnunen, Pasi; Koivumäki, Janne; Mattila, Jouni [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Siuko, Mikko [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Esque, Salvador [F4E, Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla2, 08019, Barcelona (Spain); Palmer, Jim, E-mail: ville.lyytikainen@tut.fi [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: ► RH requirements were developed from operator feedback, potential problem analysis and task description. ► Tools were designed according to these RH specific requirements. ► Two RH capable were developed and their functionality was verified at DPT2. -- Abstract: A key ITER maintenance activity is the exchange of the divertor cassettes. The current major step in this programme involves the full scale physical test facility, namely divertor test platform 2 (DTP2), in Tampere, Finland. The objective of the DTP2 is the design and proof of concept studies of various remote handling (RH) device prototypes and their RH control systems, but is also important to define principles for standardizing control systems and methods around the ITER maintenance equipment. The development process of divertor cassette locking system (CLS) RH Tool prototypes is presented in this paper. The validation of the developed CLS Tool prototypes is accomplished in RH trials at DTP2. For this RH Trial, a CLS task description (TD) and tool prototypes were developed, manufactured and, finally, tested under remote operations. These tools, designed to be operated by water hydraulic manipulator (WHMAN), are water hydraulic jack (WHJ), pin tool (PT) and wrench tool (WT)

  10. Design and operation of a novel divertor cryopumping system in Alcator C-Mod

    Science.gov (United States)

    Labombard, B.; Beck, B.; Bosco, J.; Childs, R.; Gwinn, D.; Irby, J.; Leccacorvi, R.; Marazita, S.; Mucic, N.; Pierson, S.; Rokhman, Y.; Titus, P.; Vieira, R.; Zaks, J.; Zhukovsky, A.

    2007-11-01

    C-Mod's recently installed upper-divertor cryopump is unique among the world's tokamaks, employing an array of gas-pumping slots that penetrate the upper divertor target. This geometry enables the use of a single toroidal loop of liquid helium, operating in an efficient heat transfer regime with low or no helium flow. A system pumping speed of 9,600 l/sec for D2 gas has been achieved, matching that of a full-scale prototype system. Neutral pressures in the pumping slots during upper-null plasmas (USN) are found to meet or exceed pressures in the lower divertor's private flux region during lower-null (LSN) -- evidence that the pumping-slot geometry is performing as intended. Very high steady-state pumping throughputs (exceeding ˜140 torr-l/s) have been demonstrated in USN. Reliable and efficient operation of the pump has been established, synchronized with the C-Mod shot cycle and consuming 60 to 90 liters of liquid helium during a full day of operation.

  11. Infrared thermography inspection for monoblock divertor target in JT-60SA

    International Nuclear Information System (INIS)

    Highlights: • Infrared thermography inspection is modified to inspect JT-60SA divertor targets. • Infrared thermography inspection is effective to detect joining defects of targets. • Numerical analysis is in good agreement with inspection results of mock-up targets. • Database for setting screening criteria has been constructed by numerical analysis. - Abstract: Carbon fiber composite (CFC) monoblock divertor target is required for power handling in JT-60SA. Quality of the targets depends on a joining technology in manufacturing process. To inspect the quality of more than 900 target pieces, efficient non-destructive inspection is needed. An infrared thermography inspection (IR inspection), has been proposed by ITER and IRFM, where the quality between CFC and a cooling tube is examined by a use of transient thermal response at a rapid switch from hot to cold water flow. In JT-60SA divertor target, a screw tube will be employed to obtain high heat transfer efficiency with simple structure. Since the time response of the screw tube is much faster than that of smooth tube, it is required to confirm the feasibility of this IR inspection. Thus, the effect of joining defects on transient thermal response of the targets has been investigated experimentally by using the mock-up targets containing defects which are artificially made. It was found that the IR inspection can detect the defects. Moreover, screening criteria of IR inspection for acceptable monoblock target is discussed

  12. Understanding of Neutral Gas Transport in the Alcator C-Mod Tokamak Divertor

    Energy Technology Data Exchange (ETDEWEB)

    D.P. Stotler; C.S. Pitcher; C.J. Boswell; B. LaBombard; J.L. Terry; J.D. Elder; S. Lisgo

    2002-05-07

    A series of experiments on the effect of divertor baffling on the Alcator C-Mod tokamak provides stringent tests on models of neutral gas transport in and around the divertor region. One attractive feature of these experiments is that a trial description of the background plasma can be constructed from experimental measurements using a simple model, allowing the neutral gas transport to be studied with a stand-alone code. The neutral-ion and neutral-neutral elastic scattering processes recently added to the DEGAS 2 Monte Carlo neutral transport code permit the neutral gas flow rates between the divertor and main chamber to be simulated more realistically than before. Nonetheless, the simulated neutral pressures are too low and the deuterium Balmer-alpha emission profiles differ qualitatively from those measured, indicating an incomplete understanding of the physical processes involved in the experiment. Some potential explanations are examined and opportunities for future exploration a re highlighted. Improvements to atomic and surface physics data and models will play a role in the latter.

  13. Modelling of surface evolution of rough surface on divertor target in fusion devices

    International Nuclear Information System (INIS)

    Highlights: • We study the surface evolution of rough surface on divertor target in fusion devices. • The effects of gyration motion and E × B drift affect 3D angular distribution. • A larger magnetic field angle leads to a reduced net eroded areal density. • The rough surface evolution affects the physical sputtering yield. - Abstract: The 3D Monte-Carlo code SURO has been used to study the surface evolution of rough surface on the divertor target in fusion devices. The edge plasma at divertor region is modelled by the SDPIC code and used as input data for SURO. Coupled with SDPIC, SURO can perform more sophisticated simulations to calculate the local angle and surface evolution of rough surface. The simulation results show that the incident direction of magnetic field, gyration and E × B force has a significant impact on 3D angular distribution of background plasma and accordingly on the erosion of rough surface. The net eroded areal density of rough surface is studied by varying the magnetic field angle with surface normal. The evolution of the microscopic morphology of rough surface can lead to a significant change in the physical sputtering yield

  14. Understanding of Neutral Gas Transport in the Alcator C-Mod Tokamak Divertor

    International Nuclear Information System (INIS)

    A series of experiments on the effect of divertor baffling on the Alcator C-Mod tokamak provides stringent tests on models of neutral gas transport in and around the divertor region. One attractive feature of these experiments is that a trial description of the background plasma can be constructed from experimental measurements using a simple model, allowing the neutral gas transport to be studied with a stand-alone code. The neutral-ion and neutral-neutral elastic scattering processes recently added to the DEGAS 2 Monte Carlo neutral transport code permit the neutral gas flow rates between the divertor and main chamber to be simulated more realistically than before. Nonetheless, the simulated neutral pressures are too low and the deuterium Balmer-alpha emission profiles differ qualitatively from those measured, indicating an incomplete understanding of the physical processes involved in the experiment. Some potential explanations are examined and opportunities for future exploration a re highlighted. Improvements to atomic and surface physics data and models will play a role in the latter

  15. 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.

  16. ADX: a high field, high power density, advanced divertor and RF tokamak

    Science.gov (United States)

    LaBombard, B.; Marmar, E.; Irby, J.; Terry, J. L.; Vieira, R.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; Baek, S.; Beck, W.; Bonoli, P.; Brunner, D.; Doody, J.; Ellis, R.; Ernst, D.; Fiore, C.; Freidberg, J. P.; Golfinopoulos, T.; Granetz, R.; Greenwald, M.; Hartwig, Z. S.; Hubbard, A.; Hughes, J. W.; Hutchinson, I. H.; Kessel, C.; Kotschenreuther, M.; Leccacorvi, R.; Lin, Y.; Lipschultz, B.; Mahajan, S.; Minervini, J.; Mumgaard, R.; Nygren, R.; Parker, R.; Poli, F.; Porkolab, M.; Reinke, M. L.; Rice, J.; Rognlien, T.; Rowan, W.; Shiraiwa, S.; Terry, D.; Theiler, C.; Titus, P.; Umansky, M.; Valanju, P.; Walk, J.; White, A.; Wilson, J. R.; Wright, G.; Zweben, S. J.

    2015-05-01

    The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)—a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (⩾6.5 T, 1.5 MA), high power density facility (P/S ˜ 1.5 MW m-2) will test innovative divertor ideas, including an ‘X-point target divertor’ concept, at the required performance parameters—reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region—while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side—the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination—advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions—will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept

  17. Time and space-resolved energy flux measurements in the divertor of the ASDEX tokamak by computerized infrared thermography

    International Nuclear Information System (INIS)

    A new, fully computerized and automatic thermographic system has been developed. Its two central components are an AGA THV 780 infrared camera and a PDP-11/34 computer. A combined analytical-numerical method of solving the 1-dimensional heat diffusion equation for a solid of finite thickness bounded by two parallel planes was developed. In high-density (anti nsub(e) = 8 x 1013 cm-3) neutral-beam-heated (L-mode) divertor discharges in ASDEX, the power deposition on the neutralizer plates is reduced to about 10-15% of the total heating power, owing to the inelastic scattering of the divertor plasma from a neutral gas target. Between 30% and 40% of the power is missing in the global balance. The power flow inside the divertor chambers is restricted to an approximately 1-cm-thick plasma scrape-off layer. This width depends only weakly on the density and heating power. During H-phases free of Edge Localized Mode (ELM) activity the energy flow into the divertor is blocked. During H-phases with ELM activity the energy is expelled into the divertor in very short intense pulses (several MW for about one hundred μs). Sawtooth events are able to transport significant amounts of energy from the plasma core to the peripheral zones and the scrape-off layer, and they are frequently correlated with transitions from the L to the H mode. (orig./AH)

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

    International Nuclear Information System (INIS)

    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.

  19. High thermal performance divertor plate optimization of the monobloc divertor plate by the use of ultra-high thermal conductivity carbon fibres

    International Nuclear Information System (INIS)

    A conceptual study of an advanced divertor plate is presented. The essential feature of the new concept, apart from the use of ultrahigh conductivity carbon fibres, is the use of a single material, a CFC composite, for the whole structure. The coolant is helium gas. The main advantages of this solutions are: elimination of the severe joint-interface problems inherent in other multimaterial solutions, avoidance of the risk of burn-out, no damage caused by run-away electrons, low-activation properties, great tolerance towards off-normal operating conditions, great reduction of mechanical stresses induced by electromagnetic transient and the ease of baking at high temperature. The maximum computed temperature is about 1000 C and the required pumping power is approximately only 30 % higher than a corresponding cooling performed by water in swirl-tubes

  20. 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

  1. Divertor heat fluxes and profiles during mitigated and unmitigated Edge Localised Modes (ELMs) on the Mega Amp Spherical Tokamak (MAST)

    CERN Document Server

    Thornton, A J; Chapman, I T; Harrison, J R

    2013-01-01

    Edge localised modes (ELMs) are a concern for future devices as they can limit the operational lifetime of the divertor. The mitigation of ELMs can be performed by the application of resonant magnetic perturbations (RMPs) which act to degrade the pressure gradient in the edge of the plasma. Investigations of the effect of RMPs on MAST have been performed in a range of plasmas using perturbations with toroidal mode numbers of n=3, 4 and 6. It has been seen that the RMPs increase the ELM frequency, which gives rise to a corresponding decrease in the ELM energy. The reduced ELM energy decreases the peak heat flux to the divertor, with a three fold reduction in the ELM energy, generating a 1.5 fold reduction in the peak heat flux. Measurements of the divertor heat flux profile show evidence of strike point splitting consistent with modelling using the vacuum code ERGOS.

  2. Coherence imaging of scrape-off-layer and divertor impurity flows in the Mega Amp Spherical Tokamak (invited)

    Energy Technology Data Exchange (ETDEWEB)

    Silburn, S. A., E-mail: s.a.silburn@durham.ac.uk; Sharples, R. M. [Centre for Advanced Instrumentation, Department of Physics, Durham University, Durham DH1 3LE (United Kingdom); Harrison, J. R.; Meyer, H.; Michael, C. A. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Howard, J. [Plasma Research Laboratory, Australian National University, Canberra, ACT 0200 (Australia); Gibson, K. J. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom)

    2014-11-15

    A new coherence imaging Doppler spectroscopy diagnostic has been deployed on the UK’s Mega Amp Spherical Tokamak for scrape-off-layer and divertor impurity flow measurements. The system has successfully obtained 2D images of C III, C II, and He II line-of-sight flows, in both the lower divertor and main scrape-off-layer. Flow imaging has been obtained at frame rates up to 1 kHz, with flow resolution of around 1 km/s and spatial resolution better than 1 cm, over a 40° field of view. C III data have been tomographically inverted to obtain poloidal profiles of the parallel impurity flow in the divertor under various conditions. In this paper we present the details of the instrument design, operation, calibration, and data analysis as well as a selection of flow imaging results which demonstrate the diagnostic's capabilities.

  3. Inferring divertor plasma properties from hydrogen Balmer and Paschen series spectroscopy in JET-ILW

    Science.gov (United States)

    Lomanowski, B. A.; Meigs, A. G.; Sharples, R. M.; Stamp, M.; Guillemaut, C.; Contributors, JET

    2015-11-01

    A parametrised spectral line profile model is formulated to investigate the diagnostic scope for recovering plasma parameters from hydrogenic Balmer and Paschen series spectroscopy in the context of JET-ILW divertor plasmas. The separate treatment of Zeeman and Stark contributions in the line model is tested against the PPP-B code which accounts for their combined influence on the spectral line shape. The proposed simplified model does not fully reproduce the Stark-Zeeman features for the α and β transitions, but good agreement is observed in the line width and wing profiles, especially for n  >  5. The line model has been applied to infer radial density profiles in the JET-ILW divertor with generally good agreement between the D 5\\to 2 , 5\\to 3 , 6\\to 2 , 7\\to 2 and 9\\to 2 lines for high recycling and detached conditions. In an L-mode detached plasma pulse the Langmuir probe measurements typically underestimated the density by a factor 2-3 and overestimated the electron temperature by a factor of 5-10 compared to spectroscopically derived values. The line model is further used to generate synthetic high-resolution spectra for low-n transitions to assess the potential for parameter recovery using a multi-parametric fitting technique. In cases with 4 parameter fits with a single Maxwellian neutral temperature component the D 4\\to 3 line yields the best results with parameter estimates within 10% of the input values. For cases with 9 parameter fits inclusive of a multi-component neutral velocity distribution function the quality of the fits is degraded. Simultaneous fitting of the D 3\\to 2 and 4\\to 3 profiles improves the fit quality significantly, highlighting the importance of complementary spectroscopic measurements for divertor plasma emission studies.

  4. Production and fabrication of vanadium alloys for the radiative divertor program of DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, W.R.; Smith, J.P.; Stambaugh, R.D.

    1996-04-01

    V-4Cr-4-Ti alloy has been recently selected for use in the manufacture of a portion of the DIII-D Radiative Divertor modification, as part of an overall DIII-D vanadium alloy deployment effort developed by General Atomics (GA) in conjunction with the Argonne and Oak Ridge National Laboratories (ANL or ORNL). The goal of this work is to produce a production-scale heat of the alloy and fabricate it into product forms for the manufacture of a portion of the Radiative Divertor (RD) for the DIII-D tokamak, to develop the fabrications technology for manufacture of the vanadium alloy radiative Divertor components, and to determine the effects of typical tokamak environments in the behavior of the vanadium alloy. The production of a {approx}1300-kg heat of V-4Cr-4Ti alloy is currently in progress at Teledyne Wah Chang of Albany, oregon (TWCA) to provide sufficient material for applicable product forms. Two unalloyed vanadium ingots for the alloy have already been produced by electron beam melting of raw processes vanadium. Chemical compositions of one ingot and a portion of the second were acceptable, and Charpy V-Notch (CVN) impact test performed on processed ingot samples indicated ductile behavior. Material from these ingots are currently being blended with chromium and titanium additions, and will be vacuum-arc remelted into a V-4Cr-4Ti alloy ingot and converted into product forms suitable for components of the DIII-D RD structure. Several joining methods selected for specific applications in fabrication of the RD components are being investigated, and preliminary trials have been successful in the joining of V-alloy to itself by both resistance and inertial welding processes and to Inconel 625 by inertial welding.

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

    International Nuclear Information System (INIS)

    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/m2. • 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/m2 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/m2. 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

  6. 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.

  7. Divertor plate biasing effects on particle recycling and power loss distribution in TdeV during lower hybrid current drive and heating experiments

    International Nuclear Information System (INIS)

    Preliminary results concerning the influence of negative biasing of the divertor plates on particle recycling and on power loss distribution in single null discharges of TdeV during lower hybrid (LH) current drive and heating experiments are presented. The beneficial effects of negative biasing of the divertor plates, such as the ability to control power and particle fluxes in the SOL, remain effective in the presence of auxiliary heating and current drive. Up to 0.7 MW of auxiliary power were injected in these experiments. With a negative biasing of 150 V, and the ExB flow vector pointing towards the outer divertor chamber, a roughly 2 fold increase in the divertor pressure and the radiation from the divertor region is observed. ((orig.))

  8. Kinetic modeling of divertor heat load fluxes in the Alcator C-Mod and DIII-D tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Pankin, A. Y. [Tech-X Corporation, Boulder, Colorado 80303 (United States); Rafiq, T.; Kritz, A. H. [Department of Physics, Lehigh University, Bethlehem, Pennsylvania 18015 (United States); Park, G. Y. [National Fusion Research Institute, Daejeon, 305-333 (Korea, Republic of); Chang, C. S.; Ku, S. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Brunner, D.; Hughes, J. W.; LaBombard, B.; Terry, J. L. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Groebner, R. J. [General Atomics, San Diego, California 92121 (United States)

    2015-09-15

    The guiding-center kinetic neoclassical transport code, XGC0 [Chang et al., Phys. Plasmas 11, 2649 (2004)], is used to compute the heat fluxes and the heat-load width in the outer divertor plates of Alcator C-Mod and DIII-D tokamaks. The dependence of the width of heat-load fluxes on neoclassical effects, neutral collisions, and anomalous transport is investigated using the XGC0 code. The XGC0 code includes realistic X-point geometry, a neutral source model, the effects of collisions, and a diffusion model for anomalous transport. It is observed that the width of the XGC0 neoclassical heat-load is approximately inversely proportional to the total plasma current I{sub p.} The scaling of the width of the divertor heat-load with plasma current is examined for an Alcator C-Mod discharge and four DIII-D discharges. The scaling of the divertor heat-load width with plasma current is found to be weaker in the Alcator C-Mod discharge compared to scaling found in the DIII-D discharges. The effect of neutral collisions on the 1/I{sub p} scaling of heat-load width is shown not to be significant. Although inclusion of poloidally uniform anomalous transport results in a deviation from the 1/I{sub p} scaling, the inclusion of the anomalous transport that is driven by ballooning-type instabilities results in recovering the neoclassical 1/I{sub p} scaling. The Bohm or gyro-Bohm scalings of anomalous transport do not strongly affect the dependence of the heat-load width on plasma current. The inclusion of anomalous transport, in general, results in widening the width of neoclassical divertor heat-load and enhances the neoclassical heat-load fluxes on the divertor plates. Understanding heat transport in the tokamak scrape-off layer plasmas is important for strengthening the basis for predicting divertor conditions in ITER.

  9. Molecule-surface interaction processes of relevance to gas blanket type fusion device divertor design

    Energy Technology Data Exchange (ETDEWEB)

    Snowdon, K.J. [Newcastle Univ. (United Kingdom). Dept. of Physics; Tawara, H.

    1997-01-01

    The mechanisms which may lead to the departure of molecular species from surfaces exposed to low energy (0.1-100 eV) particle or photon and electron irradiation are reviewed. Where possible, the charge and electronic state, angular, translational and internal energy distributions of the departing molecules are described and the physical origin of the nature of those distributions identified. The consequences, for the departing molecules, of certain material choices become apparent from such an analysis. Such information may help guide the choice of appropriate materials for plasma facing components of gas-blanket type divertors such as that recently proposed for the International Thermonuclear Experimental Reactor (ITER). (author). 71 refs.

  10. Evidence for enhanced cross-field transport mechanisms in the TCV Snowflake divertor

    Science.gov (United States)

    Vijvers, Wouter

    2015-11-01

    TCV experiments demonstrate that cross-field plasma transport is enhanced in the Snowflake divertor (SFD) compared to a standard single-null divertor (SND). This enhanced cross-field transport spreads the exhaust power over a larger surface area than can be achieved by magnetic geometry alone and, thereby, reduces the peak heat flux. Comparison of the experiments with modelling identifies steepened radial gradients, ExB drift effects, and βp-driven instabilities as the responsible transport mechanisms. The uncovered physics is also relevant to the SND and may help improve predictive models for the target profiles in ITER and DEMO. In SFD variants with an X-point in the scrape-off layer (SOL), part of the heat flux profile is split off and redirected to an additional target. The resulting steepened radial gradients enhance cross-field diffusion. This is confirmed by EMC3-Eirene simulations, which show a factor two reduction of the parallel heat flux, even if diffusivities remain constant. Theoretical analysis predicts enhanced ExB drifts in the SFD by increased poloidal gradients of the temperature and density. The predictions are confirmed by target heat and particle flux measurements in dedicated experiments with both toroidal field directions. Cross-field convection by curvature-driven modes at high βp (``churning modes'') explains the large fluxes into the private flux region of the SFD. This activates the extra targets and reduces the peak power to the primary targets up to a factor four. This mechanism is expected to be most effective when the divertor conditions are most severe: near the separatrix of a narrow, high-pressure SOL of a large tokamak. These and other alternative divertor configurations thus provide potential solutions to the power exhaust challenge, as well as laboratories to study SOL transport, one of the most important topics in tokamak research. This project was carried out with financial support from NWO. The work was carried out within

  11. On the asymmetries of ELM divertor power deposition in JET and ASDEX Upgrade

    DEFF Research Database (Denmark)

    Eich, T.; Kallenbach, A.; Fundamenski, W.;

    2009-01-01

    An analytical expression was derived for describing the divertor target power during ELMs based on the model discussed in [W. Fundamenski, R.A. Pitts, Plasma Phys. Control. Fus. 48 (2006) 109] where the power load arises from a Maxwellian distribution of particles released into the SOL region......-streaming-particle (FSP) approach predicts a dependence of the ELM in/out energy balance of the pedestal Mach number as well as an inversion of the in/out balance by a change of the field line helicity as observed experimentally. From the FSP approach the value for EτIR (see text) is predicted to be 18–25% in good...

  12. Alternative schemes of power deposition with the ergodic divertor on Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Mank, G. E-mail: g.mank@fz-juelich.de; Ghendrih, Ph.; Grisolia, C.; Gunn, J.; Loarer, T.; Monier-Garbet, P.; Costanzo, L.; Finken, K.H.; Michelis, C. de; Reichle, R

    2001-03-01

    Two alternative schemes to distribute the energy flux over larger surfaces are proposed and tested at Tore Supra. (i) A good sharing of the energy flux to the pump limiter and the ergodic divertor (ED) is achieved at a reduced stochasticity. (ii) The operation at highest densities with the plasma leaning on the high field side is characterised by the same screening properties and achievable densities as for the standard operation. The results indicate that an efficient heat reduction and screening of impurities can be reached under ED operation. These experiments have partly been carried out in order to test special aspects of the dynamic ED (DED) at TEXTOR.

  13. VUV spectroscopic study of a localized impurity source in Tore Supra ergodic divertor plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Michelis, C. de; Monier-Garbet, P.; Becoulet, M.; Guirlet, R.; Hess, W.; Schunke, B.; Vallet, J.C. [Association EURATOM/CEA, CEA/DSM/DRFC, CEA-Cadarache, St. Paul lez Durance (France); Hogan, J. [Oak Ridge National Laboratory, Oak Ridge, TN (United States)

    2002-07-01

    A spectroscopic study of VUV emission from injected and intrinsic low-Z impurities has been carried out for Tore Supra ergodic divertor (ED) plasmas. Analysis of plasmas in which a nearby limiter effectively provides a spatially localized source of recycled impurities provides information illuminating the dynamical processes of impurity penetration in ED plasmas. The profile evolution behaviour is found to be consistent with an interpretation which identifies both ED-influenced edge confinement and scrape-off layer and edge phenomena provided by the localized source. Preferentially increased edge impurity transport is a beneficial aspect of the ED. (author)

  14. Alternative schemes of power deposition with the ergodic divertor on Tore Supra

    International Nuclear Information System (INIS)

    Two alternative schemes to distribute the energy flux over larger surfaces are proposed and tested at Tore Supra. (i) A good sharing of the energy flux to the pump limiter and the ergodic divertor (ED) is achieved at a reduced stochasticity. (ii) The operation at highest densities with the plasma leaning on the high field side is characterised by the same screening properties and achievable densities as for the standard operation. The results indicate that an efficient heat reduction and screening of impurities can be reached under ED operation. These experiments have partly been carried out in order to test special aspects of the dynamic ED (DED) at TEXTOR

  15. Particle collection and exhaust in ergodic divertor experiments on Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Loarer, T. E-mail: loarer@drfc.cad.cea.fr; Ghendrih, Ph.; Gunn, J.; Azeroual, A.; Costanzo, L.; Grisolia, C.; Guirlet, R.; Mank, G.; Monier-Garbet, P.; Pegourie, B

    2001-03-01

    The particle exhaust in ergodic divertor (ED) configuration has been enhanced by inserting the outboard pump limiter (OPL) in the ergodic layer. The influence on the edge parameters (electron density and temperature) induced by the insertion of the OPL at different positions in the ergodic layer is reported in this paper. The additional pumping delivered by the OPL is shown to improve by about 50% the plasma density control while low impurity (carbon) concentration, characteristic of the ED shots, is also obtained for high plasma density in spite of proximity of the OPL to the bulk plasma.

  16. Impurity radiation efficiency and retention in Tore Supra ergodic divertor experiments

    International Nuclear Information System (INIS)

    The screening effect of Ne and C in ergodic divertor plasmas is studied. Spectroscopic measurements show that the screening mechanism is not the same for the two impurities. A 2D model explains this difference by the longer penetration length of neutral Ne. 3D modelling of the plasma edge with the BBQ code confirms the brightness profile shape dependence on the edge Te. The 1D impurity transport code MIST coupled to BBQ interprets the screening effect as possibly due to strong impurity outfluxes coming out of the ergodic region. (author)

  17. Design of a Vacuum Pumping System for the Closed Helical Divertor for Steady State Operation in LHD

    OpenAIRE

    Shoji, Mamoru; MASUZAKI, Suguru; MORISAKI, Tomohiro; Kobayashi, Masahiro; TOKITANI, Masayuki; TAKEIRI, Yasuhiko

    2012-01-01

    A vacuum pumping system is installed in a Closed Helical Divertor (CHD) in the Large Helical Device (LHD) at the National Institute for Fusion Science for active control of the peripheral plasma density and impurity suppression in the core plasma. In the CHD configuration, the distance between the pumping system and the divertor plates (heat and particle source) is very short (only ∼0.1 m). One of the major issues in designing the pumping system is the reduction of heat load by radiation and ...

  18. Modifications of impurity transport and divertor sources by lithium wall conditioning in the National Spherical Torus Experiment

    Science.gov (United States)

    Scotti, Filippo

    In the National Spherical Torus Experiment (NSTX), lithium coatings are evaporated on graphite plasma facing components (PFCs) for wall conditioning. In lithium-conditioned H-mode discharges, carbon accumulation is observed with core concentrations ≤10%, leading to a lack of density control, while lithium ions have concentrations ≤0.1%. In this thesis, modifications of carbon and lithium divertor sources as well as scrape-off layer (SOL) and core transport due to lithium conditioning are studied. Spectroscopic impurity influxes (measured by filtered cameras) and 2D multi-fluid edge transport simulations via the UEDGE code are employed to study divertor impurity sources and SOL transport, respectively. Core transport of carbon and lithium is analyzed using the impurity transport code MIST and the neoclassical transport codes NEO and NCLASS. A reduction of the carbon sputtering yield in the lower divertor is observed with lithium evaporation. However, weaker divertor impurity retention resulting from reduced recycling (inferred from UEDGE simulations) and the possible importance of wall sources can counteract this reduction in divertor carbon influxes. The suppression of edge-localized-modes (ELMs) is the primary cause of the increased carbon inventories in lithium-conditioned discharges, leading to lack of density control. Deviations from neoclassical predictions for carbon transport are observed at the pedestal top in lithium-conditioned discharges, indicating the presence of anomalous outward convection. While the lithium sputtering yield from lithium-coated graphite in the divertor is consistent with physical and temperature-enhanced sputtering, a strong reduction in ionized lithium influxes is observed, possibly due to prompt re-deposition. The different poloidal source distribution and the stronger divertor retention for lithium (inferred from UEDGE simulations) contribute to a lower edge lithium source with respect to carbon. The latter is due to the

  19. Manufacturing and testing of a copper/CFC divertor mock-up for JET

    International Nuclear Information System (INIS)

    An actively cooled divertor is a possible option for future developments at The Joint European Torus (JET). A proof of principle actively cooled tile has been produced in order to qualify the relevant manufacturing technologies and the non destructive control processes. In this frame Ansaldo Ricerche (ARI) has been involved in the construction of a mock-up comprising 6 OFHC copper tubes for water cooling that are brazed to a plate made out of carbon fibre composite (CFC). The final objective was the high heat flux testing of the mock-up at JET in order to evaluate the general behaviour of the component under relevant operating conditions. The key point of the work was the realisation of a sound joint by adapting the expertise gained in ARI in previous R and D activities on brazing heterogeneous materials. Reliable methods for ultrasonic examinations of the pieces were also set up. For successful application to the JET pumped divertor a water-cooled CFC target plate must show surface temperatures of 2. Furthermore, global hydraulic considerations specific to JET limit the system pressure to 0.7 MPa. In such a design, critical heat flux is not the key limit, rather the reliability of the CFC-copper joint in terms of extent of wetting. First tests in the neutral beam test bed at JET show an adequate response for fluxes up to 15 MW/m2. (orig.)

  20. Production and fabrication of vanadium alloys for the radiative divertor program of DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, W.R.; Smith, J.P.

    1997-08-01

    V-4Cr-4Ti alloy has been selected for use in the manufacture of a portion of the DIII-D Radiative Divertor upgrade. The production of a 1200-kg ingot of V-4Cr-4Ti alloy, and processing into final sheet and rod product forms suitable for components of the DIII-D Radiative Divertor Program (RDP), has been completed by Wah Chang (formerly Teledyne Wah Chang) of Albany, Oregon (WCA). CVN impact tests on sheet material indicate that the material has properties comparable to other previously-processed V-4Cr-4Ti and V-5Cr-5Ti alloys. Joining of V-4Cr-4Ti alloy has been identified as the most critical fabrication issue for its use in the RDP, and research into several joining methods for fabrication of the RDP components, including resistance seam, friction, and electron beam welding, and explosive bonding is being pursued. Preliminary trials have been successful in the joining of V-alloy to itself by resistance, friction, and electron beam welding processes, and to Inconel 625 by friction welding. In addition, an effort to investigate the explosive bonding of V-4Cr-4Ti alloy to Inconel 625, in both tube-to-bar and sheet-to-sheet configurations, has been initiated, and results have been encouraging.

  1. Production and fabrication of vanadium alloys for the radiative divertor program of DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, W.R.; Smith, J.P.; Trester, P.W.

    1997-04-01

    V-4Cr-4Ti alloy has been selected for use in the manufacture of a portion of the DIII-D Radiative Divertor upgrade. The production of a 1200-kg ingot of V-4Cr-4Ti alloy, and processing into final sheet and rod product forms suitable for components of the DIII-D Radiative Divertor structure, has been completed at Wah Chang (formerly Teledyne Wah Chang) of Albany, Oregon (WCA). Joining of V-4Cr-4Ti alloy has been identified as the most critical fabrication issue for its use in the RD Program, and research into several joining methods for fabrication of the RD components, including resistance seam, friction, and electron beam welding, is continuing. Preliminary trials have been successful in the joining of V-alloy to itself by electron beam, resistance, and friction welding processes, and to Inconel 625 by friction welding. An effort to investigate the explosive bonding of V-4Cr-4Ti alloy to Inconel 625 has also been initiated, and results have been encouraging. In addition, preliminary tests have been completed to evaluate the susceptibility of V-4Cr-4Ti alloy to stress corrosion cracking in DIII-D cooling water, and the effects of exposure to DIII-D bakeout conditions on the tensile and fracture behavior of V-4Cr-4Ti alloy.

  2. Microstructure and inhomogeneous fuel trapping at divertor surfaces in the JET tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bergsåker, H., E-mail: henricb@kth.se [Department for Fusion Plasma Physics, Association EURATOM-VR, School of Electrical Engineering, Royal Institute of Technology, S-10405 Stockholm (Sweden); Bykov, I.; Petersson, P. [Department for Fusion Plasma Physics, Association EURATOM-VR, School of Electrical Engineering, Royal Institute of Technology, S-10405 Stockholm (Sweden); Possnert, G. [Uppsala Universitet, Tandem Laboratory, Association EURATOM-VR, S-75105 Uppsala (Sweden); Likonen, J.; Koivuranta, S.; Coad, J.P. [VTT, Association Euratom-Tekes, PO Box 1000, FI-02044 VTT (Finland); Widdowson, A.M. [JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)

    2014-08-01

    The plasma deposited layers at divertor surfaces in the JET tokamak with carbon wall have been studied post mortem, using micro ion beam analysis (μ-IBA) methods, optical microscopy and scanning electron microscopy (SEM). The layers were formed during plasma operations over different periods through 1998–2009. They frequently have a columnar structure. For μ-IBA a 3 MeV {sup 3}He beam was used, focused to about 5–15 μm size. Nuclear reaction analysis was used to measure D, Be and C. Elemental mapping was carried out both at the original surface and on polished layer cross sections. Trapped deuterium is predominantly found in remote areas on the horizontal bottom divertor tiles and in regions with locally enhanced deuterium concentration on the vertical tiles. Pockets with enhanced deuterium concentration are found in the carbon fibre composite (CFC) substrate. Areas with dimensions of about 100 μm with enhanced deuterium concentration are also found inside the deposited layers. The inhomogeneous fuel trapping is tentatively explained with co-deposition in partly protected pits in the substrate and by incorporation of dust particles in the growing layers.

  3. Divertor load footprint of ELMs in pellet triggering and pacing experiments at JET

    Energy Technology Data Exchange (ETDEWEB)

    Frigione, D., E-mail: domenico.frigione@frascati.enea.it [Unità Tecnica Fusione, ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Garzotti, L. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom); Lennholm, M. [EFDA CSU, Culham Science Centre, OX14 3DB (United Kingdom); Alper, B. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom); Artaserse, G. [Unità Tecnica Fusione, ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Bennett, P. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom); Giovannozzi, E. [Unità Tecnica Fusione, ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Eich, T. [Max Planck Institute for Plasma Physics, Garching (Germany); Kocsis, G. [WIGNER RCP RMI, POB 49, 1525 Budapest (Hungary); Lang, P.T. [Max Planck Institute for Plasma Physics, Garching (Germany); Maddaluno, G. [Unità Tecnica Fusione, ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Mooney, R. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom); Rack, M. [Institut für Energieforschung – Plasmaphysik, Forschungszentrum Jülich, 52425 Jülich (Germany); Sips, G. [EFDA CSU, Culham Science Centre, OX14 3DB (United Kingdom); Tvalashvili, G. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom); Viola, B. [Unità Tecnica Fusione, ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Wilkes, D. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom)

    2015-08-15

    An investigation of pellet pacing and triggering of Edge Localized Modes (ELMs) was carried out in the frame of ELM mitigation studies aimed at reducing their damaging effects on the plasma-facing components (PFCs). The divertor power load footprint of triggered ELMs was compared with gas puffing controlled ELMs. Small pellets, corresponding to a few per cent of the target plasma particle inventory, were used to minimize the fueling effect and the total particle throughput. There is no evidence that pellets can reduce the divertor power load with respect to gas fueling when operating at the same ELM frequency. The line average density and the energy confinement time remained constant when the gas was progressively substituted by pellets. The launch from the Vertical High Field Side (VHFS) confirmed to be more efficient in ELM triggering than from the Low Field Side (LFS) while the power load footprint remained the same both in time evolution and in spatial distribution when changing the injection geometry.

  4. Concept design of divertor remote handling system for the FAST machine

    Energy Technology Data Exchange (ETDEWEB)

    Di Gironimo, G., E-mail: giuseppe.digironimo@unina.it [Association Euratom/ENEA/CREATE, Università di Napoli Federico II, 80125 Napoli (Italy); Labate, C.; Renno, F. [Association Euratom/ENEA/CREATE, Università di Napoli Federico II, 80125 Napoli (Italy); Brolatti, G.; Crescenzi, F.; Crisanti, F. [CR ENEA Frascati, Via E. Fermi 27, Frascati (RM) (Italy); Lanzotti, A. [Association Euratom/ENEA/CREATE, Università di Napoli Federico II, 80125 Napoli (Italy); Lucca, F. [LT Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy); Siuko, M. [VTT Systems Engineering, Tekniikankatu 1, 33720 Tampere (Finland)

    2013-10-15

    The paper presents a concept design of a remote handling (RH) system oriented to maintenance operations on the divertor second cassette in FAST, a satellite of ITER tokamak. Starting from ITER configuration, a suitably scaled system, composed by a cassette multifunctional mover (CMM) connected to a second cassette end-effector (SCEE), can represent a very efficient solution for FAST machine. The presence of a further system able to open the divertor port, used for RH aims, and remove the first cassette, already aligned with the radial direction of the port, is presumed. Although an ITER-like system maintains essentially shape and proportions of its reference configuration, an appropriate arrangement with FAST environment is needed, taking into account new requirements due to different dimensions, weights and geometries. The use of virtual prototyping and the possibility to involve a great number of persons, not only mechanical designers but also physicist, plasma experts and personnel assigned to remote handling operations, made them to share the multiphysics design experience, according to a concurrent engineering approach. Nevertheless, according to the main objective of any satellite tokamak, such an approach benefits the study of enhancements to ITER RH system and the exploration of alternative solutions.

  5. An automated approach to magnetic divertor configuration design, using an efficient optimization methodology

    Energy Technology Data Exchange (ETDEWEB)

    Blommaert, Maarten; Reiter, Detlev [Institute of Energy and Climate Research (IEK-4), FZ Juelich GmbH, D-52425 Juelich (Germany); Heumann, Holger [Centre de Recherche INRIA Sophia Antipolis, BP 93 06902 Sophia Antipolis (France); Baelmans, Martine [KU Leuven, Department of Mechanical Engineering, 3001 Leuven (Belgium); Gauger, Nicolas Ralph [TU Kaiserslautern, Chair for Scientific Computing, 67663 Kaiserslautern (Germany)

    2015-05-01

    At present, several plasma boundary codes exist that attempt to describe the complex interactions in the divertor SOL (Scrape-Off Layer). The predictive capability of these edge codes is still very limited. Yet, in parallel to major efforts to mature edge codes, we face the design challenges for next step fusion devices. One of them is the design of the helium and heat exhaust system. In past automated design studies, results indicated large potential reductions in peak heat load by an increased magnetic flux divergence towards the target structures. In the present study, a free boundary magnetic equilibrium solver is included into the simulation chain to verify these tendencies. Additionally, we expanded the applicability of the automated design method by introducing advanced ''adjoint'' sensitivity computations. This method, inherited from airfoil shape optimization in aerodynamics, allows for a large number of design variables at no additional computational cost. Results are shown for a design application of the new WEST divertor.

  6. Production and fabrication of vanadium alloys for the radiative divertor program of DIII-D

    International Nuclear Information System (INIS)

    V-4Cr-4Ti alloy has been selected for use in the manufacture of a portion of the DIII-D Radiative Divertor upgrade. The production of a 1200-kg ingot of V-4Cr-4Ti alloy, and processing into final sheet and rod product forms suitable for components of the DIII-D Radiative Divertor Program (RDP), has been completed by Wah Chang (formerly Teledyne Wah Chang) of Albany, Oregon (WCA). CVN impact tests on sheet material indicate that the material has properties comparable to other previously-processed V-4Cr-4Ti and V-5Cr-5Ti alloys. Joining of V-4Cr-4Ti alloy has been identified as the most critical fabrication issue for its use in the RDP, and research into several joining methods for fabrication of the RDP components, including resistance seam, friction, and electron beam welding, and explosive bonding is being pursued. Preliminary trials have been successful in the joining of V-alloy to itself by resistance, friction, and electron beam welding processes, and to Inconel 625 by friction welding. In addition, an effort to investigate the explosive bonding of V-4Cr-4Ti alloy to Inconel 625, in both tube-to-bar and sheet-to-sheet configurations, has been initiated, and results have been encouraging

  7. Thermal analysis of an exposed tungsten edge in the JET divertor

    Energy Technology Data Exchange (ETDEWEB)

    Arnoux, G., E-mail: gilles.arnoux@ccfe.ac.uk [CCFE Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Coenen, J. [Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, 52425 Jülich (Germany); Bazylev, B. [Forshungzentrum Karlsruhe GmbH, P.O.Box 3640, D-76021 Karlsruhe (Germany); Corre, Y. [CEA/DSM/IRFM, CEA Cadarache, 13108 Saint Paul Lez Durance (France); Matthews, G.F.; Balboa, I. [CCFE Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Clever, M. [Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, 52425 Jülich (Germany); Dejarnac, R. [IPP.CR, Institute of Plasma Physics AS CR, Za Slovankou 3, 182 21 Praha 8 (Czech Republic); Devaux, S.; Eich, T. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Gauthier, E. [CEA/DSM/IRFM, CEA Cadarache, 13108 Saint Paul Lez Durance (France); Frassinetti, L. [Fusion Plasma Physics, EES, KTH, SE-10044 Stockholm (Sweden); Horacek, J. [IPP.CR, Institute of Plasma Physics AS CR, Za Slovankou 3, 182 21 Praha 8 (Czech Republic); Jachmich, S. [Laboratory for Plasma Physics Koninklijke Militaire School – Ecole Royale Militaire, Renaissancelaan, 30 Avenue de la Renaissance, B-1000 Brussels (Belgium); Kinna, D. [CCFE Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Marsen, S. [Max-Planck-Institut für Plasmaphysik, Teilinsitut Greifswald, D-17491 Greifswald (Germany); and others

    2015-08-15

    Highlights: • We provide experimental evidences that melting of the JET tungsten divertor is achieved by transients only. • The measurements show that less than half the parallel heat flux reaches the melted sample. • We propose ideas to investigate to explain the missing heat flux. - Abstract: In the recent melt experiments with the JET tungsten divertor, we observe that the heat flux impacting on a leading edge is 3–10 times lower than a geometrical projection would predict. The surface temperature, tungsten vaporisation rate and melt motion measured during these experiments is consistent with the simulations using the MEMOS code, only if one applies the heat flux reduction. This unexpected observation is the result of our efforts to demonstrate that the tungsten lamella was melted by ELM induced transient heat loads only. This paper describes in details the measurements and data analysis method that led us to this strong conclusion. The reason for the reduced heat flux are yet to be clearly established and we provide some ideas to explore. Explaining the physics of this heat flux reduction would allow to understand whether it can be extrapolated to ITER.

  8. ADX: a high field, high power density, Advanced Divertor test eXperiment

    Science.gov (United States)

    Vieira, R.; Labombard, B.; Marmar, E.; Irby, J.; Shiraiwa, S.; Terry, J.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; ADX Team

    2014-10-01

    The MIT PSFC and collaborators are proposing an advanced divertor experiment (ADX) - a tokamak specifically designed to address critical gaps in the world fusion research program on the pathway to FNSF/DEMO. This high field (6.5 tesla, 1.5 MA), high power density (P/S ~ 1.5 MW/m2) facility would utilize Alcator magnet technology to test innovative divertor concepts for next-step DT fusion devices (FNSF, DEMO) at reactor-level boundary plasma pressures and parallel heat flux densities while producing high performance core plasma conditions. The experimental platform would also test advanced lower hybrid current drive (LHCD) and ion-cyclotron range of frequency (ICRF) actuators and wave physics at the plasma densities and magnetic field strengths of a DEMO, with the unique ability to deploy launcher structures both on the low-magnetic-field side and the high-field side - a location where energetic plasma-material interactions can be controlled and wave physics is most favorable for efficient current drive, heating and flow drive. This innovative experiment would perform plasma science and technology R&D necessary to inform the conceptual development and accelerate the readiness-for-deployment of FNSF/DEMO - in a timely manner, on a cost-effective research platform. Supported by DE-FC02-99ER54512.

  9. 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.

  10. Improving concept design of divertor support system for FAST tokamak using TRIZ theory and AHP approach

    Energy Technology Data Exchange (ETDEWEB)

    Di Gironimo, G., E-mail: giuseppe.digironimo@unina.it [Association Euratom/ENEA/CREATE, Università di Napoli Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Carfora, D.; Esposito, G.; Labate, C.; Mozzillo, R.; Renno, F.; Lanzotti, A. [Association Euratom/ENEA/CREATE, Università di Napoli Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Siuko, M. [VTT Systems Engineering, Tekniikankatu 1, 33720 Tampere (Finland)

    2013-11-15

    Highlights: • Optimization of the RH system for the FAST divertor using TRIZ. • Participative design approach using virtual reality. • Comparison of product alternatives in an immersive virtual reality environment. • Prioritization of concept alternatives based on AHP. -- Abstract: The paper focuses on the application of the Theory of Inventive Problem Solving (TRIZ) to divertor Remote Handling (RH) issues in Fusion Advanced Studies Torus (FAST), a satellite tokamak acting as a test bed for the study and the development of innovative technologies oriented to ITER and DEMO programs. The objective of this study consists in generating concepts or solutions able to overcome design and technical weak points in the current maintenance procedure. Two different concepts are designed with the help of a parametric CAD software, CATIA V5, using a top-down modeling approach; kinematic simulations of the remote handling system are performed using Digital Mock-Up (DMU) capabilities of the software. The evaluation of the concepts is carried out involving a group of experts in a participative design approach using virtual reality, classifying the concepts with the help of the Analytical Hierarchy Process (AHP)

  11. Production and fabrication of vanadium alloys for the radiative divertor program of DIII-D

    International Nuclear Information System (INIS)

    V-4Cr-4Ti alloy has been selected for use in the manufacture of a portion of the DIII-D Radiative Divertor upgrade. The production of a 1200-kg ingot of V-4Cr-4Ti alloy, and processing into final sheet and rod product forms suitable for components of the DIII-D Radiative Divertor structure, has been completed at Wah Chang (formerly Teledyne Wah Chang) of Albany, Oregon (WCA). Joining of V-4Cr-4Ti alloy has been identified as the most critical fabrication issue for its use in the RD Program, and research into several joining methods for fabrication of the RD components, including resistance seam, friction, and electron beam welding, is continuing. Preliminary trials have been successful in the joining of V-alloy to itself by electron beam, resistance, and friction welding processes, and to Inconel 625 by friction welding. An effort to investigate the explosive bonding of V-4Cr-4Ti alloy to Inconel 625 has also been initiated, and results have been encouraging. In addition, preliminary tests have been completed to evaluate the susceptibility of V-4Cr-4Ti alloy to stress corrosion cracking in DIII-D cooling water, and the effects of exposure to DIII-D bakeout conditions on the tensile and fracture behavior of V-4Cr-4Ti alloy

  12. Development of ITER divertor Thomson scattering support structure design on the basis of engineering analyses

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

    Highlights: • The results of multiphysics simulations for TS support structure in consequent design iterations are presented and discussed. • The close to final design of front diagnostic rack developed on the basis of simulation results and suitable to sustain all main ITER loads is presented. • The distortion of mirrors surfaces is analyzed and possible problems are indicated. • The new design of the mirror mounting system is proposed. - Abstract: The support structure for divertor Thomson scattering equipment – the front diagnostic rack, which actually plays plugging role of the divertor port, should be designed to sustain the severe ITER conditions. Meeting the requirements of multifield analyses (which often contradict each other) results in an iterative design process. A number of design variants based on engineering analyses results were developed in 2011–2012. We study here the close to the final design of the diagnostic rack for consistency to electromagnetic, thermal and seismic loads. The specific ITER environment imposes a restricted list of materials and requires a careful design of optical elements to accommodate their thermal expansion. Special attention is focused on the mirror deformed shape under operating loading conditions and its effect on optical system performance, which is vital for all optical systems with mirrors specially designed for the ITER.

  13. First tests of diagnostic mirrors in a tokamak divertor: An overview of experiments in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Litnovsky, A. [Institut fuer Energieforschung - Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ, D-52425, Juelich (Germany)], E-mail: a.litnovsky@fz-juelich.de; Rudakov, D.L. [University of California, San Diego, La Jolla, CA 92093-0417 (United States); De Temmerman, G. [Institute of Physics, University of Basel, Klingelbergstrasse 82, CH-4056 Basel (Switzerland); Wienhold, P.; Philipps, V.; Samm, U. [Institut fuer Energieforschung - Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ, D-52425, Juelich (Germany); McLean, A.G. [University of Toronto, Institute for Aerospace Studies, Toronto, Ontario, Canada M3H 5T6 (Canada); West, W.P.; Wong, C.P.C.; Brooks, N.H. [General Atomics, San Diego, CA 92186-5608 (United States); Watkins, J.G.; Wampler, W.R. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Stangeby, P.C. [University of Toronto, Institute for Aerospace Studies, Toronto, Ontario, M3H 5T6 (Canada); Boedo, J.A.; Moyer, R.A. [University of California, San Diego, La Jolla, CA 92093-0417 (United States); Allen, S.L.; Fenstermacher, M.E.; Groth, M.; Lasnier, C.J. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Boivin, R.L. [General Atomics, San Diego, CA 92186-5608 (United States)] (and others)

    2008-01-15

    Mirrors will be used in ITER in all optical diagnostic systems observing the plasma radiation in the ultraviolet, visible and infrared ranges. Diagnostic mirrors in ITER will suffer from electromagnetic radiation, energetic particles and neutron irradiation. Erosion due to impact of fast neutrals from plasma and deposition of plasma impurities may significantly degrade optical and polarization characteristics of mirrors influencing the overall performance of the respective diagnostics. Therefore, maintaining the best possible performance of mirrors is of the crucial importance for the ITER optical diagnostics. Mirrors in ITER divertor are expected to suffer from deposition of impurities. The dedicated experiment in a tokamak divertor was needed to address this issue. Investigations with molybdenum diagnostic mirrors were made in DIII-D divertor. Mirror samples were exposed at different temperatures in the private flux region to a series of ELMy H-mode discharges with partially detached divertor plasmas. An increase of temperature of mirrors during the exposure generally led to the mitigation of carbon deposition, primarily due to temperature-enhanced chemical erosion of carbon layers by D atoms. Finally, for the mirrors exposed at the temperature of {approx}160 {sup o}C neither carbon deposition nor degradation of optical properties was detected.

  14. Numerical Study of High Heat Flux Performances of Flat-Tile Divertor Mock-ups with Hypervapotron Cooling Concept

    Science.gov (United States)

    Chen, Lei; Liu, Xiang; Lian, Youyun; Cai, Laizhong

    2015-09-01

    The hypervapotron (HV), as an enhanced heat transfer technique, will be used for ITER divertor components in the dome region as well as the enhanced heat flux first wall panels. W-Cu brazing technology has been developed at SWIP (Southwestern Institute of Physics), and one W/CuCrZr/316LN component of 450 mm×52 mm×166 mm with HV cooling channels will be fabricated for high heat flux (HHF) tests. Before that a relevant analysis was carried out to optimize the structure of divertor component elements. ANSYS-CFX was used in CFD analysis and ABAQUS was adopted for thermal-mechanical calculations. Commercial code FE-SAFE was adopted to compute the fatigue life of the component. The tile size, thickness of tungsten tiles and the slit width among tungsten tiles were optimized and its HHF performances under International Thermonuclear Experimental Reactor (ITER) loading conditions were simulated. One brand new tokamak HL-2M with advanced divertor configuration is under construction in SWIP, where ITER-like flat-tile divertor components are adopted. This optimized design is expected to supply valuable data for HL-2M tokamak. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2011GB110001 and 2011GB110004)

  15. Concept design of the DEMO divertor cassette-to-vacuum vessel locking system adopting a systems engineering approach

    Energy Technology Data Exchange (ETDEWEB)

    Di Gironimo, G., E-mail: giuseppe.digironimo@unina.it [Università degli Studi di Napoli “Federico II”, Dipartimento di Ingegneria Industriale, Piazzale Tecchio 80, 80135 Napoli (Italy); Carfora, D. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); VTT Technical Research Centre of Finland, Tekniikankatu 1, PO Box 1300, FI-33101 Tampere (Finland); Università degli Studi di Napoli “Federico II”, Dipartimento di Ingegneria Industriale, Piazzale Tecchio 80, 80135 Napoli (Italy); Esposito, G.; Lanzotti, A.; Marzullo, D. [Università degli Studi di Napoli “Federico II”, Dipartimento di Ingegneria Industriale, Piazzale Tecchio 80, 80135 Napoli (Italy); Siuko, M. [VTT Technical Research Centre of Finland, Tekniikankatu 1, PO Box 1300, FI-33101 Tampere (Finland)

    2015-05-15

    Highlights: • An iterative and incremental design process for cassette-to-VV locking system of DEMO divertor is presented. • Three different concepts have been developed with a systematic design approach. • The final concept has been selected with Fuzzy-Analytic Hierarchy Process in virtual reality. - Abstract: This paper deals with pre-concept studies of DEMO divertor cassette-to-vacuum vessel locking system under the work program WP13-DAS-07-T06: Divertor Remote Maintenance System pre-concept study. An iterative design process, consistent with Systems Engineering guidelines and named Iterative and Participative Axiomatic Design Process (IPADeP), is used in this paper to propose new innovative solutions for divertor locking system, which can overcome the difficulties in applying the ITER principles to DEMO. The solutions conceived have been analysed from the structural point of view using the software Ansys and, eventually, evaluated using the methodology known as Fuzzy-Analytic Hierarchy Process. Due to the lack and the uncertainty of the requirements in this early conceptual design stage, the aim is to cover a first iteration of an iterative and incremental process to propose an innovative design concept to be developed in more details as the information will be completed.

  16. Concept design of the DEMO divertor cassette-to-vacuum vessel locking system adopting a systems engineering approach

    International Nuclear Information System (INIS)

    Highlights: • An iterative and incremental design process for cassette-to-VV locking system of DEMO divertor is presented. • Three different concepts have been developed with a systematic design approach. • The final concept has been selected with Fuzzy-Analytic Hierarchy Process in virtual reality. - Abstract: This paper deals with pre-concept studies of DEMO divertor cassette-to-vacuum vessel locking system under the work program WP13-DAS-07-T06: Divertor Remote Maintenance System pre-concept study. An iterative design process, consistent with Systems Engineering guidelines and named Iterative and Participative Axiomatic Design Process (IPADeP), is used in this paper to propose new innovative solutions for divertor locking system, which can overcome the difficulties in applying the ITER principles to DEMO. The solutions conceived have been analysed from the structural point of view using the software Ansys and, eventually, evaluated using the methodology known as Fuzzy-Analytic Hierarchy Process. Due to the lack and the uncertainty of the requirements in this early conceptual design stage, the aim is to cover a first iteration of an iterative and incremental process to propose an innovative design concept to be developed in more details as the information will be completed

  17. Loss of beam ions to the inside of the PDX [Poloidal Divertor Experiment] tokamak during the fishbone instability

    International Nuclear Information System (INIS)

    Using data from two vertical charge-exchange detectors on the Poloidal Divertor Experiment (PDX), we have identified a set of conditions for which loss of beam ions inward in major radius is observed during the fishbone instability. Previously, it was reported that beam ions were lost only to the outside of the PDX tokamak

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

    Science.gov (United States)

    Schmitz, O.; Becoulet, M.; Cahyna, P.; 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.

    2016-06-01

    Results from three-dimensional modeling of plasma edge transport and plasma-wall interactions during application of resonant magnetic perturbation (RMP) fields for control of edge-localized modes in the ITER standard 15 MA Q  =  10 H-mode are presented. The full 3D plasma fluid and kinetic neutral transport code EMC3-EIRENE is used for the modeling. Four characteristic perturbed magnetic topologies are considered and discussed with reference to the axisymmetric case without RMP fields. Two perturbation field amplitudes at full and half of the ITER ELM control coil current capability using the vacuum approximation are compared to a case including a strongly screening plasma response. In addition, a vacuum field case at high q 95  =  4.2 featuring increased magnetic shear has been modeled. Formation of a three-dimensional plasma boundary is seen for all four perturbed magnetic topologies. The resonant field amplitudes and the effective radial magnetic field at the separatrix define the shape and extension of the 3D plasma boundary. Opening of the magnetic field lines from inside the separatrix establishes scrape-off layer-like channels of direct parallel particle and heat flux towards the divertor yielding a reduction of the main plasma thermal and particle confinement. This impact on confinement is most accentuated at full RMP current and is strongly reduced when screened RMP fields are considered, as well as for the reduced coil current cases. The divertor fluxes are redirected into a three-dimensional pattern of helical magnetic footprints on the divertor target tiles. At maximum perturbation strength, these fingers stretch out as far as 60 cm across the divertor targets, yielding heat flux spreading and the reduction of peak heat fluxes by 30%. However, at the same time substantial and highly localized heat fluxes reach divertor areas well outside of the axisymmetric heat flux decay profile. Reduced RMP amplitudes due to screening or reduced RMP

  19. Impact of nonlinear 3D equilibrium response on edge topology and divertor heat load in Wendelstein 7-X

    Science.gov (United States)

    Suzuki, Y.; Geiger, J.

    2016-06-01

    The impact of the 3D equilibrium response on the plasma edge topology is studied. In Wendelstein 7-X, the island divertor concept is used to assess scenarios for quasi-steady-state operation. However, the boundary islands necessary for the island divertor are strongly susceptible to plasma beta and toroidal current density effects because of the low magnetic shear. The edge magnetic topology for quasi-steady-state operation scenarios is calculated with the HINT-code to study the accompanying changes of the magnetic field structures. Two magnetic configurations have been selected, which had been investigated in self consistent neoclassical transport simulations for low bootstrap current but which use the alternative natural island chains to the standard iota value of 1 (ι b   =  5/5, periodicity), namely, at high-iota (ι b   =  5/4) and at low-iota (ι b   =  5/6). For the high-iota configuration, the boundary islands are robust but the stochasticity around them is enhanced with beta. The addition of toroidal current densities enhances the stochasticity further. The increased stochasticity changes the footprints on in-vessel components with a direct impact on the corresponding heat loads. In the low-iota configuration the boundary islands used for the island divertor are dislocated radially due to the low shear and even show healing effects, i.e. the island width vanishes. In the latter case the plasma changes from divertor to limiter operation. To realize the predicted high-performance quasi-steady-state operation of the transport simulations, further adjustments of the magnetic configuration may be necessary to achieve a proper divertor compatibility of the scenarios.

  20. Divertor Heat Flux Mitigation in High-Performance H-mode Discharges in the National Spherical Torus Experiment.

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V A; Maingi, R; Gates, D; Menard, J

    2008-12-31

    Experiments conducted in high-performance 1.0 MA and 1.2 MA 6 MW NBI-heated H-mode discharges with a high magnetic flux expansion radiative divertor in NSTX demonstrate that significant divertor peak heat flux reduction and access to detachment may be facilitated naturally in a highly-shaped spherical torus (ST) configuration. Improved plasma performance with high {beta}{sub t} = 15-25%, a high bootstrap current fraction f{sub BS} = 45-50%, longer plasma pulses, and an H-mode regime with smaller ELMs has been achieved in the strongly-shaped lower single null configuration with elongation {kappa} = 2.2-2.4 and triangularity {delta} = 0.6-0.8. Divertor peak heat fluxes were reduced from 6-12 MW/m{sup 2} to 0.5-2 MW/m{sup 2} in ELMy H-mode discharges using the inherently high magnetic flux expansion f{sub m} = 16-25 and the partial detachment of the outer strike point at several D{sub 2} injection rates. A good core confinement and pedestal characteristics were maintained, while the core carbon concentration and the associated Z{sub eff} were reduced. The partially detached divertor regime was characterized by an increase in divertor radiated power, a reduction of ion flux to the plate, and a large neutral compression ratio. Spectroscopic measurements indicated a formation of a high-density, low temperature region adjacent to the outer strike point, where substantial increases in the volume recombination rate and CII, CIII emission rates was measured.

  1. Benchmarking of a 1D Scrape-off layer code SOLF1D with SOLPS and its use in modelling long-legged divertors

    CERN Document Server

    Havlickova, E; Subba, F; Coster, D; Wischmeier, M; Fishpool, G

    2013-01-01

    A 1D code modelling SOL transport parallel to the magnetic field (SOLF1D) is benchmarked with 2D simulations of MAST-U SOL performed via the SOLPS code for two different collisionalities. Based on this comparison, SOLF1D is then used to model the effects of divertor leg stretching in 1D, in support of the planned Super-X divertor on MAST. The aim is to separate magnetic flux expansion from volumetric power losses due to recycling neutrals by stretching the divertor leg either vertically or radially.

  2. Preliminary comparison of the conventional and quasi-snowflake divertor configurations with the 2D code EDGE2D/EIRENE in the FAST tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Viola, B.; Maddaluno, G.; Pericoli Ridolfini, V. [EURATOM-ENEA Association, C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Rome) (Italy); Corrigan, G.; Harting, D. [Culham Centre of Fusion Energy, EURATOM-Association, Abingdon (United Kingdom); Mattia, M. [Dipartimento di Informatica, Sistemi e Produzione, Universita di Roma, Tor Vergata, Via del Politecnico, 00133 Roma (Italy); Zagorski, R. [Institute of Plasma Physics and Laser Microfusion-EURATOM Association, 01-497 Warsaw (Poland)

    2014-06-15

    The new magnetic configurations for tokamak divertors, snowflake and super-X, proposed to mitigate the problem of the power exhaust in reactors have clearly evidenced the need for an accurate and reliable modeling of the physics governing the interaction with the plates. The initial effort undertaken jointly by ENEA and IPPLM has been focused to exploit a simple and versatile modeling tool, namely the 2D TECXY code, to obtain preliminary comparison between the conventional and snowflake configurations for the proposed new device FAST that should realize an edge plasma with properties quite close to those of a reactor. The very interesting features found for the snowflake, namely a power load mitigation much larger than expected directly from the change of the magnetic topology, has further pushed us to check these results with the more sophisticated computational tool EDGE2D coupled with the neutral code module EIRENE. After a preparatory work that has been carried out in order to adapt this code combination to deal with non-conventional, single null equilibria and in particular with second order nulls in the poloidal field generated in the snowflake configuration, in this paper we describe the first activity to compare these codes and discuss the first results obtained for FAST. The outcome of these EDGE2D runs is in qualitative agreement with those of TECXY, confirming the potential benefit obtainable from a snowflake configuration. (copyright 2014 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  3. An FPGA-based bolometer for the MAST-U Super-X divertor

    Science.gov (United States)

    Lovell, Jack; Naylor, Graham; Field, Anthony; Drewelow, Peter; Sharples, Ray

    2016-11-01

    A new resistive bolometer system has been developed for MAST-Upgrade. It will measure radiated power in the new Super-X divertor, with millisecond time resolution, along 16 vertical and 16 horizontal lines of sight. The system uses a Xilinx Zynq-7000 series Field-Programmable Gate Array (FPGA) in the D-TACQ ACQ2106 carrier to perform real time data acquisition and signal processing. The FPGA enables AC-synchronous detection using high performance digital filtering to achieve a high signal-to-noise ratio and will be able to output processed data in real time with millisecond latency. The system has been installed on 8 previously unused channels of the JET vertical bolometer system. Initial results suggest good agreement with data from existing vertical channels but with higher bandwidth and signal-to-noise ratio.

  4. Calculations of Energy Losses due to Atomic Processes in Tokamaks with Applications to the ITER Divertor

    CERN Document Server

    Post, D; Clark, R E H; Putvinskaya, N

    1995-01-01

    Reduction of the peak heat loads on the plasma facing components is essential for the success of the next generation of high fusion power tokamaks such as the International Thermonuclear Experimental Reactor (ITER) 1 . Many present concepts for accomplishing this involve the use of atomic processes to transfer the heat from the plasma to the main chamber and divertor chamber walls and much of the experimental and theoretical physics research in the fusion program is directed toward this issue. The results of these experiments and calculations are the result of a complex interplay of many processes. In order to identify the key features of these experiments and calculations and the relative role of the primary atomic processes, simple quasi-analytic models and the latest atomic physics rate coefficients and cross sections have been used to assess the relative roles of central radiation losses through bremsstrahlung, impurity radiation losses from the plasma edge, charge exchange and hydrogen radiation losses f...

  5. Modelling of hydrogen isotope retention in the tungsten divertor of EAST during ELMy H-mode

    International Nuclear Information System (INIS)

    In this work, we study hydrogen isotopes (HI) inventory inside tungsten plasma-facing materials during high confinement mode discharges with repetitive edge localized modes (ELMy H-mode) based on the operating parameters of the EAST device, since tungsten is considered as the primary plasma-facing material and the ELMy H-mode is an important operation regime for EAST and future devices. The upgraded Hydrogen Isotope Inventory Processes Code (HIIPC) is applied with the incident depth profile provided by SRIM-2013 to make the study. The code is first verified by comparison with experimental measurements. The effects of the incident ion energy and ion flux on the retention are then studied. Finally, using the parameters obtained from EAST diagnostics, the HI retention inside the W divertor during ELMy H-mode is studied, which indicates the retained HI can be increased dramatically mainly due to ion-induced trap sites by ELMs

  6. Surface modifications of W divertor components for EAST during exposure to high heat loads with He

    Energy Technology Data Exchange (ETDEWEB)

    Li, C., E-mail: lichun10@mails.tsinghua.edu.cn [Laboratory of Advanced Materials, School of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China); Greuner, H. [Max Planck Institute for Plasma Physics, Boltzmannstr. 2, 85748 Garching (Germany); Yuan, Y. [School of Physics and Nuclear Energy Engineering, Beihang University, Beijing 100191 (China); Zhao, S.X.; Luo, G.N. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Böswirth, B. [Max Planck Institute for Plasma Physics, Boltzmannstr. 2, 85748 Garching (Germany); Fu, B.Q.; Jia, Y.Z. [Laboratory of Advanced Materials, School of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China); Liu, X. [Southwestern Institute of Physics, Chengdu, Sichuan 610041 (China); Liu, W., E-mail: liuw@mail.tsinghua.edu.cn [Laboratory of Advanced Materials, School of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China)

    2015-08-15

    Flat-type W/Cu plasma-facing components have been developed for the new generation divertor of the Chinese Experimental Advanced Superconducting Tokamak. Surface modifications of such actively water-cooled W components following short and long pulse high heat loading coupled with He particle loads with fluence of 3 × 10{sup 22} m{sup −2} have been investigated. An adiabatically loaded W block was investigated as a comparison and exposed to short pulse loads. Blistering was observed on all sample surfaces, but was less pronounced on the components than on the W block, due to the significant lower surface temperature caused by active cooling. For components, longer pulse loads gave rise to a rougher surface. Furthermore, most blisters on components are found to be less than 1 μm in diameter, with just a very few blisters larger than 1 μm, observed only in some near 〈1 1 1〉 grains.

  7. Evaluation of Nb-base alloys for the divertor structure in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Purdy, I.M. [Argonne National Laboratory, Upton, IL (United States)

    1996-04-01

    Niobium-base alloys are candidate materials for the divertor structure in fusion reactors. For this application, an alloy should resist aqueous corrosion, hydrogen embrittlement, and radiation damage and should have high thermal conductivity and low thermal expansion. Results of corrosion and embrittlement screening tests of several binary and ternary Nb alloys in high-temperature water indicated the Mb-1Zr, Nb-5MO-1Zr, and Nb-5V-1Z4 (wt %) showed sufficient promise for further investigation. These alloys, together with pure Nb and Zircaloy-4 have been exposed to high purity water containing a low concentration of dissolved oxygen (<12 ppb) at 170, 230, and 300{degrees}C for up to {approx}3200 h. Weight-change data, microstructural observations, and qualitative mechanical-property evaluation reveal that Nb-5V-1Zr is the most promising alloy at higher temperatures. Below {approx}200{degrees}C, the alloys exhibit similiar corrosion behavior.

  8. 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.

  9. Thermo-mechanical tests of a CFC divertor mock-up

    Science.gov (United States)

    Cardella, A.; Akiba, M.; Duwe, R.; Di Pietro, E.; Suzuki, S.; Satoh, K.; Reheis, N.

    1994-04-01

    Thermo-mechanical tests have been performed on a divertor mock-up consisting of a metallic tube armoured with five carbon fibre composite tiles. The tube is inserted inside the tiles and brazed with TiCuSil braze (monoblock concept). The tube material is TZM, a molybdenum alloy, and the armour material is SEP CARB N112, a high conductivity carbon-carbon composite. Using special surface preparation consisting of laser drilling, small (˜- 500 μm) holes in the composite have been made to increase the surface wetted by the braze and the resistance. The mock-up has been tested at the JAERI 400 kW electron beam test facility JEBIS. The aim of the test was to assess the performance of the mock-up in screening and thermal fatigue tests with particular attention to the behaviour of the armour to heat sink joint.

  10. L-Mode and Inter-ELM Divertor Particle and Heat Flux Width Scaling on MAST

    CERN Document Server

    Harrison, J R; Kirk, A

    2013-01-01

    The distribution of particles and power to plasma-facing components is of key importance in the design of next-generation fusion devices. Power and particle decay lengths have been measured in a number of MAST L-mode and H-mode discharges in order to determine their parametric dependencies, by fitting power and particle flux profiles measured by divertor Langmuir probes, to a convolution of an exponential decay and a Gaussian function. In all discharges analysed, it is found that exponential decay lengths mapped to the midplane are mostly dependent on separatrix electron density and plasma current (or parallel connection length). The widths of the convolved Gaussian functions have been used to derive an approximate diffusion coefficient, which is found to vary from 1m2/s to 7m2/s, and is systematically lower in H-mode compared with L-mode.

  11. Recycling studies in the ASDEX divertor with pellet or gas puff refuelling

    Science.gov (United States)

    Haas, G.; Kaufmann, M.; Lang, R. S.; ASDEX Team; Pellet Team; Mertens, V.; Niedermeyer, H.; Sandmann, W.; Becker, G.; Bosch, H. S.; Brocken, H.; Büchl, K.; Eberhagen, A.; Fussmann, G.; Gehre, O.; Gernhardt, J.; Gierke, G.; Glock, E.; Gruber, O.; Hofmann, J.; Izvozchikov, A.; Janeschitz, G.; Karger, F.; Keilhacker, M.; Klüber, O.; Kornherr, M.; Lackner, K.; Lenoci, M.; Lisitano, G.; Mast, F.; Mayer, H. M.; McCormick, K.; Meisel, D.; Müller, E. R.; Murmann, H.; Neuhauser, J.; Pietrzyk, Z. A.; Poschenrieder, W.; Rapp, H.; Riedler, H.; Röhr, H.; Roth, J.; Ryter, F.; Schneider, F.; Setzensack, C.; Siller, G.; Smeulders, P.; Söldner, F. X.; Speth, E.; Steuer, K.-H.; Tsois, N.; Vlases, G.; Vollmer, O.; Wagner, F.; Ugniewski, S.; Zasche, D.

    1987-02-01

    Discharges fuelled by stationary pellet injection (PI), gas puffing (GP) or a combination of the two methods are compared with respect to recycling in the divertor and particle confinement. Fuelling by PI yields much better global particle confinement than by GP. This has been found for both low and high recycling. In the low-recycling case this improvement is due to the deeper particle deposition for PI than for GP since the transport in the inner plasma is not reduced. For high recycling the improvement results from both the deeper deposition and a reduction in the transport. The best global particle confinement was found for phases with low or no GP. This, however, can be reached for short times only. Since with PI alone it is impossible to keep the recycling on a high level, GP is unavoidable for sustaining the favourable high-recycling condition.

  12. BBQ Modeling of Recycling from the Tore Supra Ergodic Divertor Neutraliser

    Science.gov (United States)

    Giannella, R.; Guirlet, R.; Demichelis, C.; Hogan, J.; Cherigier, L.

    1998-11-01

    Generation and recycling of carbon and hydrocarbon impurities, and recycling of neon at the Tore Supra pumped ergodic divertor have been analyzed using the BBQ 3-D scrape-off layer transport code. Code results are compared with spectroscopic observations from fibres located on the neutralizer plates, and background plasma conditions used in the code are constrained with data from langmuir probes embedded in the plates. The sensitivity of neon recycling to assumed reflection coefficients has been studied. A detailed 3-D geometry model for the neutralizer, including all 4 plates, and recycling from the notches between plates, has been prepared. A version of the code describing deuterium processes is being developed to study conditions during the onset of detachment at high density

  13. Characterisation of radiation and flux measurements on a neutraliser plate of the Tore Supra ergodic divertor

    Energy Technology Data Exchange (ETDEWEB)

    Corre, Y. E-mail: corre@drfc.cad.cea.fr; Giannella, R.; De Michelis, C.; Guirlet, R.; Azeroual, A.; Chareyre, E.; Costanzo, L.; Escarguel, A.; Gauthier, E.; Ghendrih, P.; Gunn, J.; Hogan, J.; Monier-Garbet, P.; Pegourie, B.; Pospieszczyk, A.; Tsitrone, E

    2001-03-01

    A recent extensive experimental study of impurity production and penetration for various density regimes is described. Deuterium and carbon emissions near a neutraliser plate (NP) of the Tore Supra Ergodic Divertor (ED) has been measured with an absolutely calibrated visible endoscope, for high- and low-density plasma regimes. From these radiation measurements, we have deduced an effective carbon flux, and at an order-of-magnitude estimate of the NP erosion: 50 m for the fall 1999 experimental ED campaign ({approx}500 shots). Combining the measured carbon and deuterium fluxes, we deduce a global experimental carbon sputtering yield for the NP in the range 2x10{sup -2}-3x10{sup -1} for ohmic pulses, showing evidence of the importance of carbon self-sputtering.

  14. Evaluation of copper alloys for fusion reactor divertor and first wall components

    DEFF Research Database (Denmark)

    Fabritsiev, S.A.; Zinkle, S.J.; Singh, B.N.

    1996-01-01

    This paper presents a critical analysis of the main factors of radiation damage limiting the possibility to use copper alloys in the ITER divertor and first wall structure. In copper alloys the most significant types of radiation damage in the proposed temperature-dose operation range are swelling......, creep, and low-temperature radiation embrittlement. Low-temperature radiation embrittlement at T-irr ... to similar to 0.1% after irradiation doses of 0.01 to 0.1 dpa. At irradiation temperatures above 300 degrees C, pronounced softening occurs in PH copper alloys due to radiation-enhanced precipitate coarsening and dislocation recovery and recrystallization processes. The DS copper alloys are relatively...

  15. Neutron diffraction stress determination in W-laminates for structural divertor applications

    Directory of Open Access Journals (Sweden)

    R. Coppola

    2015-07-01

    Full Text Available Neutron diffraction measurements have been carried out to develop a non-destructive experimental tool for characterizing the crystallographic structure and the internal stress field in W foil laminates for structural divertor applications in future fusion reactors. The model sample selected for this study had been prepared by brazing, at 1085 °C, 13 W foils with 12 Cu foils. A complete strain distribution measurement through the brazed multilayered specimen and determination of the corresponding stresses has been obtained, assuming zero stress in the through-thickness direction. The average stress determined from the technique across the specimen (over both ‘phases’ of W and Cu is close to zero at −17 ± 32 MPa, in accordance with the expectations.

  16. Effect of radiation power loss due to impurity gas puff on divertor plasma

    International Nuclear Information System (INIS)

    In order to examine the effect of radiation power loss due to impurity gas puff in divertor plasma, we calculated time evolution of impurity ion densities and electron temperature for several cases with Ne gas puff rates 1012 and 1013 cm-3s-1 during 0.01 s and initial electron temperature 10-300 eV for a plasma with electron density 1013 cm-3 with a one zone model. We found that the electron temperature decreases less than 1 eV for the cases with gas puff rate 1013 cm-3s-1, independent of initial electron temperature and in this case we expect plasma detachment. Detailed conditions for plasma detachment should also depend on time history of radiation power. (author)

  17. Effects of drifts and ballooning instability on the divertor in–out asymmetry in EAST tokamak

    International Nuclear Information System (INIS)

    The divertor in–out asymmetry (DIOA) of the disconnected double null (DDN) configuration with the active X-point at the bottom in EAST has been investigated using the SOLPS code package to explore its forming mechanism. Effects of classical drifts and ballooning-like transport (BLT) instability are evaluated individually and compared to the DIOA during H-mode. The results show that the BLT plays a much larger role in forming the DIOA. It is further found that the combined effects of the classical drifts and BLT can account for the formation of the DIOA observed experimentally in EAST. In addition, the results also imply that reversing the toroidal magnetic field may be used to mitigate the EAST DIOA under certain circumstances

  18. Effects of drifts and ballooning instability on the divertor in–out asymmetry in EAST tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Du, Hailong [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China); Sang, Chaofeng, E-mail: sang@dlut.edu.cn [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China); Wang, L., E-mail: lwang@ipp.ac.cn [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Bonnin, X. [LSPM-CNRS, Université Paris 13, Sorbonne Paris Cité, 99 avenue J.-B. Clément, Villetaneuse F-93430 (France); Guo, H.Y. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Sun, Jizhong [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China); Wang, Dezhen, E-mail: wangdez@dlut.edu.cn [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China)

    2015-08-15

    The divertor in–out asymmetry (DIOA) of the disconnected double null (DDN) configuration with the active X-point at the bottom in EAST has been investigated using the SOLPS code package to explore its forming mechanism. Effects of classical drifts and ballooning-like transport (BLT) instability are evaluated individually and compared to the DIOA during H-mode. The results show that the BLT plays a much larger role in forming the DIOA. It is further found that the combined effects of the classical drifts and BLT can account for the formation of the DIOA observed experimentally in EAST. In addition, the results also imply that reversing the toroidal magnetic field may be used to mitigate the EAST DIOA under certain circumstances.

  19. Electron microscopy characterization of some carbon based nanostructures with application in divertors coatings from fusion reactor

    Science.gov (United States)

    Ciupina, V.; Morjan, I.; Lungu, C. P.; Vladoiu, R.; Prodan, G.; Prodan, M.; Zarovschi, V.; Porosnicu, C.; Stanescu, I. M.; Contulov, M.; Mandes, A.; Dinca, V.; Sugiyama, K.

    2011-10-01

    Nanostructured carbon materials have increasingly attracted the interest of the scientific community, because of their fascinating physical properties and potential applications in high-tech devices. In the current ITER design, the tiles made of carbon fiber composites (CFCs) are foreseen for the strike point zone and tungsten (W) for other parts of the divertor region. This choice is a compromise based mainly on experience with individual materials in many different tokamaks. Also Beryllium is the candidate material for the First Wall in ITER. In order to prepare nanostructured carbon-tungsten nanocomposite for the divertor part in fusion applications, the original method thermionic vacuum arc (TVA) was used in two electronic guns configuration. One of the main advantages of this technology is the bombardment of the growing thin film just by the ions of the depositing film. The nanostructured C-W and C-Be films were characterized by Scanning Electron Microscopy (SEM), Transmission Electron Microscopy (TEM) and Atomic Force Microscopy (AFM). The C-W films were identified as a nanocrystals complex (5 nm average diameter) surrounded by amorphous structures with a strong graphitization tendency, allowing the creating of adherent and wear resistant films. The C-Be films are polycrystalline with mean grain size about 15 nm. The friction coefficients (0.15 - 0.35) of the C-W coatings was decreased more than 3-5 times in comparison with the uncoated substrates proving excellent tribological properties. C-W nanocomposites coatings were designed to have excellent tribological properties while the structure is composed by nanocrystals complex surrounded by amorphous structures with a strong graphitization tendency, allowing the creating of adherent and wear resistant films.&updat

  20. Application of carbon-aluminum nanostructures in divertor coatings from fusion reactor

    Science.gov (United States)

    Ciupina, V.; Lungu, C. P.; Vladoiu, R.; Epure, T. D.; Prodan, G.; Porosnicu, C.; Prodan, M.; Stanescu, I. M.; Contulov, M.; Mandes, A.; Dinca, V.; Zarovschi, V.

    2012-10-01

    Nanostructured carbon materials have increasingly attracted the interest of the scientific community, because of their fascinating physical properties and potential applications in high-tech devices. In the current ITER design, the tiles made of carbon fiber composites (CFCs) are foreseen for the strike point zone and tungsten (W) for other parts of the divertor region. This choice is a compromise based mainly on experience with individual materials in many different tokamaks. Also Carbon-Aluminum composites are the candidate material for the First Wall in ITER. In order to prepare nanostructured carbon-aluminum nanocomposite for the divertor part in fusion applications, the original method thermionic vacuum arc (TVA) was used in two electronic guns configuration. One of the main advantages of this technology is the bombardment of the growing thin film just by the ions of the depositing film. Moreover, the energy of ions can be controlled. Thermo-electrons emitted by an externally heated cathode and focused by a Wehnelt focusing cylinder are strongly accelerated towards the anode whose material is evaporated and bright plasma is ignited by a high voltage DC supply. The nanostructured C-Al films were characterized by Scanning Electron Microscopy (SEM), Transmission Electron Microscopy (TEM). Tribological properties in dry sliding were evaluated using a CSM ball-on-disc tribometer. The carbon - aluminum films were identified as a nanocrystals complex (from 2nm to 50 nm diameters) surrounded by amorphous structures with a strong graphitization tendency, allowing the creating of adherent and wear resistant films. The friction coefficients (0.1 - 0.2, 0.5) of the C-Al coatings was decreased more than 2-5 times in comparison with the uncoated substrates proving excellent tribological properties. C-Al nanocomposites coatings were designed to have excellent tribological properties while the structure is composed by nanocrystals complex surrounded by amorphous structures

  1. Relationship of edge localized mode burst times with divertor flux loop signal phase in JET

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, S. C., E-mail: S.C.Chapman@warwick.ac.uk [Centre for Fusion, Space and Astrophysics, Department of Physics, University of Warwick, Coventry (United Kingdom); Max Planck Institute for the Physics of Complex Systems, Dresden (Germany); Dendy, R. O. [Centre for Fusion, Space and Astrophysics, Department of Physics, University of Warwick, Coventry (United Kingdom); Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxfordshire (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, Oxfordshire (United Kingdom); Todd, T. N.; Webster, A. J.; Morris, J. [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxfordshire (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, Oxfordshire (United Kingdom); Watkins, N. W. [Centre for Fusion, Space and Astrophysics, Department of Physics, University of Warwick, Coventry (United Kingdom); Max Planck Institute for the Physics of Complex Systems, Dresden (Germany); Centre for the Analysis of Time Series, London School of Economics, London (United Kingdom); Department of Engineering and Innovation, Open University, Milton Keynes (United Kingdom); Calderon, F. A. [Centre for Fusion, Space and Astrophysics, Department of Physics, University of Warwick, Coventry (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, Oxfordshire (United Kingdom)

    2014-06-15

    A phase relationship is identified between sequential edge localized modes (ELMs) occurrence times in a set of H-mode tokamak plasmas to the voltage measured in full flux azimuthal loops in the divertor region. We focus on plasmas in the Joint European Torus where a steady H-mode is sustained over several seconds, during which ELMs are observed in the Be II emission at the divertor. The ELMs analysed arise from intrinsic ELMing, in that there is no deliberate intent to control the ELMing process by external means. We use ELM timings derived from the Be II signal to perform direct time domain analysis of the full flux loop VLD2 and VLD3 signals, which provide a high cadence global measurement proportional to the voltage induced by changes in poloidal magnetic flux. Specifically, we examine how the time interval between pairs of successive ELMs is linked to the time-evolving phase of the full flux loop signals. Each ELM produces a clear early pulse in the full flux loop signals, whose peak time is used to condition our analysis. The arrival time of the following ELM, relative to this pulse, is found to fall into one of two categories: (i) prompt ELMs, which are directly paced by the initial response seen in the flux loop signals; and (ii) all other ELMs, which occur after the initial response of the full flux loop signals has decayed in amplitude. The times at which ELMs in category (ii) occur, relative to the first ELM of the pair, are clustered at times when the instantaneous phase of the full flux loop signal is close to its value at the time of the first ELM.

  2. Modeling cross-field drifts and current with the B2 code for the CIT divertor

    Energy Technology Data Exchange (ETDEWEB)

    Rognlien, T.D.; Milovich, J.L.; Rensink, M.E.

    1990-10-12

    We have modified the B2 edge-plasma code to include the effects of classical fluid drifts across the magnetic field lines and plasma currents. This report presents preliminary results of these effects for the CIT parameter regime. The basic plasma model described by Braams involves solving the continuity equation, the parallel momentum balance equation, and separate energy balance equations for the ions and the electrons. If multiple ion species are present, they are all assumed to have a common temperature, but their densities and parallel velocities are solved for using additional continuity and parallel momentum balance equations for each species. Momentum and heat transport parallel to the magnetic field, B, are given by the classical collisional theory. On the other hand, transport perpendicular to B is represented by anomalous diffusion coefficients which are adjusted to agree with experimental measurements. These transport coefficients are generally taken to be constant in radius and poloidal angle, although this is not necessary. The goal of our work has been to include both the classical cross-field drift terms and the effects of parallel currents in the equations used in the B2 code. The motivation for including the cross-field terms comes from simple model calculations which indicate that the classical flows can contribute an important asymmetry which may help explain the transition from L-mode to H-mode confinement. Radial electric fields which arise near the separatrix cause E {times} B poloidal rotation which may also be related to the L-to-H mode transition through its effect on edge turbulence. Including the parallel currents is done to provide a tool for understanding the biased divertor experiments on DIII-D at General Atomics. Such biasing may provide an effective means of controlling the asymmetry of the power flow to different divertor plates.

  3. Fracture mechanical analysis of tungsten armor failure of a water-cooled divertor target

    International Nuclear Information System (INIS)

    Highlights: • The FEM-based VCE method and XFEM were employed for computing KI (or J-integral) and predicting progressive cracking, respectively. • The most probable pattern of crack formation is radial cracking in the tungsten armor block. • The most probable site of cracking is the upper interfacial region of the tungsten armor block adjacent to the top position of the copper interlayer. • The initiation of a major crack becomes likely, only when the strength of tungsten armor block is significantly reduced from its original strength. - Abstract: The inherent brittleness of tungsten at low temperature and the embrittlement by neutron irradiation are its most critical weaknesses for fusion applications. In the current design of the ITER and DEMO divertor, the high heat flux loads during the operation impose a strong constraint on the structure–mechanical performance of the divertor. Thus, the combination of brittleness and the thermally induced stress fields due to the high heat flux loads raises a serious reliability issue in terms of the structural integrity of tungsten armor. In this study, quantitative estimates of the vulnerability of the tungsten monoblock armor cracking under stationary high heat flux loads are presented. A comparative fracture mechanical investigation has been carried out by means of two different types of computational approaches, namely, the extended finite element method (XFEM) and the finite element method (FEM)-based virtual crack tip extension (VCE) method. The fracture analysis indicates that the most probable pattern of crack formation is radial cracking in the tungsten armor starting from the interface to tube and the most probable site of cracking is the upper interfacial region of the tungsten armor adjacent to the top position of the copper interlayer. The strength threshold for crack initiation and the high heat flux load threshold for crack propagation are evaluated based on XFEM simulations and computations of

  4. 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.

  5. Particle exhaust with vented structures: application to the ergodic divertor of Tore Supra; Pompage des particules dans les tokamaks au moyen d'une structure a events: le divertor ergodique de Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Azeroual, A

    2000-04-04

    In a thermonuclear reactor, one must continuously fuel the discharge and extract the ashes resulting from fusion reactions. To avoid the risk of discharge poisoning, {alpha}-particle concentration is limited to {approx} 10 %. To allow for steady-state conditions requires then to extract {>=}2 % of the helium out flux. In Tore Supra, the ergodic divertor is the main component managing the heat and particle fluxes at the edge. Its principle consists in generating a resonant perturbation able to destroy magnetic surfaces at the plasma periphery. In this region, the field lines are open and connected at both ends to neutralizers which are wetted by the major part of the heat and particle fluxes and are the structures through which a part of the plasma out flux is pumped for maintaining the discharge in steady-state conditions. This work describes the neutral recirculation around the ergodic divertor and is based on a data base of 56 discharges. One discuss the two processes allowing for particle exhaust: the ballistic collection of ions and that of neutrals backscattered by atomic reactions. These two processes are modelled accounting for a realistic description of the divertor geometry. A comparison between simulations and experiments is presented for measurements characterising the three main actors of plasma-wall interaction: the edge plasma, the D{sub {alpha}} light emission and the neutral pressure in the divertor plenum. Last, one question how such a system can be extrapolated to next step machines, for which one must account for technical constraints linked to the presence of the shield protecting the coils from the high neutron flux. (author)

  6. Parallel electron temperature and density gradients measured in the JET MkI divertor using thermal helium beams

    Energy Technology Data Exchange (ETDEWEB)

    Davies, S.J. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking; Morgan, P.D. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking; Ul`Haq, Y. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking]|[Imperial College of Science, Technology and Medicine, London SW7 2BZ (United Kingdom); Maggi, C.F. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking; Erents, S.K. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking]|[UKAEA Fusion, Culham, Abingdon, Oxon OX14 3EA (United Kingdom); Fundamenski, W. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking]|[Institute for Aerospace Studies, University of Toronto, Toronto (Canada); Horton, L.D. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking; Loarte, A. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking; Matthews, G.F. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking; Monk, R.D. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking; Stangeby, P.C. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking]|[Institute for Aerospace Studies, University of Toronto, Toronto (Canada)

    1997-02-01

    This paper describes the first application of a thermal helium beam diagnostic to a divertor. The helium beam is used to determine spectroscopically the electron temperature and density from the inner and outer strike points up to the X-point, using helium line ratios which are primarily sensitive to electron density and temperature, as reported by Schweer (1992). Measurement of the neutral helium line intensities in the outer divertor target were performed under attached, high recycling and detached plasma conditions in ohmic and L-mode discharges. An interpretative model has been developed using the DIVIMP code at JET which incorporates the helium injection point, the nozzle divergence and the viewing arrangement of the periscope for a particular equilibrium. (orig.).

  7. Deuterium retention in tungsten exposed to mixed D + N plasma at divertor relevant fluxes in Magnum-PSI

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H.T., E-mail: heunlee@wakate.frc.eng.osaka-u.ac.jp [Graduate School of Engineering, Osaka University, Suita, Osaka 565-0871 (Japan); De Temmerman, G. [Dutch Institute for Fundamental Energy Research, Nieuwegein, 3439 MN (Netherlands); Gao, L.; Schwarz-Selinger, T.; Meisl, G.; Höschen, T. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Ueda, Y. [Graduate School of Engineering, Osaka University, Suita, Osaka 565-0871 (Japan)

    2015-08-15

    Nitrogen (N) has been proposed as an extrinsic impurity species in the divertor to reduce the local power load onto tungsten (W) plasma-facing components. Laboratory studies at low incident fluxes have indicated N increases deuterium (D) retention in tungsten. Here we show that W exposed to D + N Magnum-PSI plasmas under divertor relevant particle fluxes (∼10{sup 24} D/m{sup 2} s), also results in an increase in D retention by enhanced near surface trapping up to 1100 K due either to N or Mo impurities, and increased retention in the bulk at T > 700 K. These results demonstrate that N or Mo surface impurities have the potential to alter the tritium inventory in tungsten plasma-facing components under diverter relevant particle fluxes by affecting surface and bulk retention processes.

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

    International Nuclear Information System (INIS)

    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.

  9. Transport in the plasma edge and specific connexion to the wall in the Tore Supra ergodic divertor experiments

    International Nuclear Information System (INIS)

    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. (orig.)

  10. Influence of the E  ×  B drift in high recycling divertors on target asymmetries

    Science.gov (United States)

    Chankin, A. V.; Corrigan, G.; Groth, M.; Stangeby, P. C.; contributors, JET

    2015-09-01

    Detailed analysis of convective fluxes caused by E  ×  B drifts is carried out in a realistic JET configuration, based on a series of EDGE2D-EIRENE runs. The EDGE2D-EIRENE code includes all guiding centre drifts, E  ×  B as well as ∇B and centrifugal drifts. Particle sources created by divergences of radial and poloidal components of the E  ×  B drift are separately calculated for each flux tube in the divertor. It is demonstrated that in high recycling divertor conditions radial E  ×  B drift creates particle sources in the common flux region (CFR) consistent with experimentally measured divertor and target asymmetries, with the poloidal E  ×  B drift creating sources of an opposite sign but smaller in absolute value. That is, the experimentally observed asymmetries in the CFR are the opposite to what poloidal E  ×  B drift by itself would cause. In the private flux region (PFR), the situation is reversed, with poloidal E  ×  B drift being dominant. In this region poloidal E  ×  B drift by itself contributes to experimentally observed asymmetries. Thus, in each region, the dominant component of the E  ×  B drift acts so as to create the density (and hence, also temperature) asymmetries that are observed both in experiment and in 2D edge fluid codes. Since the total number of charged particles is much greater in the CFR than in PFR, divertor asymmetries caused by the E  ×  B drift should be attributed primarily to particle sources in the CFR caused by radial E  ×  B drift.

  11. Sizing of the thermal and electrical systems for an FED bundle divertor design with MgO insulation

    International Nuclear Information System (INIS)

    The high-order dependence of toroidal ripple from a bundle divertor on the magnet shield thickness increases the desirability of a magnet technology with minimal shielding requirements. A jacketed conductor with MgO powder insulation has been used successfully in highly irradiated environments. Its properties and limitations are described. A thermal and electrical sizing code has been developed for magnet design with this technology. Two design examples for ETF and FED missions show reduced recirculating power from previously reported designs

  12. Relevance of collisionality in the transport model assumptions for divertor detachment multi-fluid modelling on JET

    DEFF Research Database (Denmark)

    Wiesen, S.; Fundamenski, W.; Wischmeier, M.;

    2011-01-01

    A revised formulation of the perpendicular diffusive transport model in 2D multi-fluid edge codes is proposed. Based on theoretical predictions and experimental observations a dependence on collisionality is introduced into the transport model of EDGE2D–EIRENE. The impact on time-dependent JET ga...... strongest for scenarios with strike points on vertical targets and vanishes in case of asymmetric divertor configurations....

  13. Predictions of VRF on a Langmuir Probe under the RF Heating Spiral on the Divertor Floor on NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Hosea, J C [PPPL; Perkins, R J [PPPL; Jaworski, M A [PPPL; Kramer, G J [PPPL; Ahn, J-W [ORNL

    2014-07-01

    RF heating deposition spirals are observed on the divertor plates on NSTX as shown in for a NB plus RF heating case. It has been shown that the RF spiral is tracked quite well by the spiral mapping of the strike points on the divertor plate of magnetic field lines passing in front of the high harmonic fast wave (HHFW) antenna on NSTX. Indeed, both current instrumented tiles and Langmuir probes respond to the spiral when it is positioned over them. In particular, a positive increment in tile current (collection of electrons) is obtained when the spiral is over the tile. This current can be due to RF rectification and/or RF heating of the scrape off layer (SOL) plasma along the magnetic field lines passing in front of the the HHFW antenna. It is important to determine quantitatively the relative contributions of these processes. Here we explore the properties of the characteristics of probes on the lower divertor plate to determine the likelyhood that the primary cause of the RF heat deposition is RF rectification.

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

    International Nuclear Information System (INIS)

    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

  15. Time-to-burnout data for a prototypical ITER divertor tube during a simulated loss of flow accident

    International Nuclear Information System (INIS)

    The Loss of Flow Accident (LOFA) is a serious safety concern for the International Thermonuclear Experimental Reactor (ITER) as it has been suggested that greater than 100 seconds are necessary to safely shutdown the plasma when ITER is operating at full power. In this experiment, the thermal response of a prototypical ITER divertor tube during a simulated LOFA was studied. The divertor tube was fabricated from oxygen-free high-conductivity copper to have a square geometry with a circular coolant channel. The coolant channel inner diameter was 0.77 cm, the heated length was 4.0 cm, and the heated width was 1.6 cm. The mockup did not feature any flow enhancement techniques, i.e., swirl tape, helical coils, or internal fins. One-sided surface heating of the mockup was accomplished through the use of the 30 kW Sandia Electron Beam Test System. After reaching steady state temperatures in the mockup, as determined by two Type-K thermocouples installed 0.5 mm beneath the heated surface, the coolant pump was manually tripped off and the coolant flow allowed to naturally coast down. Electron beam heating continued after the pump trip until the divertor tube's heated surface exhibited the high temperature transient normally indicative of rapidly approaching burnout. Experimental data showed that time-to-burnout increases proportionally with increasing inlet velocity and decreases proportionally with increasing incident heat flux

  16. Characterizations of power loads on divertor targets for type-I, compound and small ELMs in the EAST superconducting tokamak

    DEFF Research Database (Denmark)

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

    2013-01-01

    of the plasma stored energy occurs at the onset of type-I ELMs (∼8%) and compound ELMs (∼5%), while no noticeable change in the plasma stored energy is observed for the small ELMs, including both type-III ELMs and very small ELMs. The peak heat flux on divertor targets for type-I ELMs currently achieved in EAST...... is about 10 MW m−2, as determined from the divertor-embedded triple Langmuir probe system with high time resolution. As expected, type-III ELMs lead to much smaller divertor power loads with a peak heat flux of about 2 MW m−2. Peak power loads for compound ELMs are between those for type-I and type......-III ELMs. It is remarkable that the new very small ELMy H-modes exhibit even lower target power deposition than type-III ELMs, with the peak heat flux generally below 1 MW m−2. These very small ELMs are usually accompanied by broadband fluctuations with frequencies ranging from 20 to 50 kHz, which may...

  17. Power handling of the bulk tungsten divertor row at JET: First measurements and comparison to the GTM thermal model

    Energy Technology Data Exchange (ETDEWEB)

    Mertens, Ph., E-mail: ph.mertens@fz-juelich.de [JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Institute of Energy and Climate Research IEK-4 (Plasma Physics), Forschungszentrum Jülich GmbH, Association EURATOM-FZJ, D-52425 Jülich (Germany); Coenen, J.W. [JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Institute of Energy and Climate Research IEK-4 (Plasma Physics), Forschungszentrum Jülich GmbH, Association EURATOM-FZJ, D-52425 Jülich (Germany); Devaux, S. [JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Max-Planck-Institut für Plasmaphysik, EURATOM Association, D-85748 Garching (Germany); Jachmich, S. [JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Laboratory for Plasma Physics-ERM/KMS, Association EURATOM-Belgian State, B-1000 Brussels (Belgium); Balboa, I.; Matthews, G.F.; Riccardo, V. [JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Culham Centre for Fusion Energy, EURATOM-CCFE Association, Abingdon OX14 3DB (United Kingdom); Sieglin, B. [JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Max-Planck-Institut für Plasmaphysik, EURATOM Association, D-85748 Garching (Germany); Tanchuk, V. [JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); SINTEZ Scientific Technical Centre, D.V. Efremov SRIEA, RUS-196641, St. Petersburg (Russian Federation); Terra, A. [JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Institute of Energy and Climate Research IEK-4 (Plasma Physics), Forschungszentrum Jülich GmbH, Association EURATOM-FZJ, D-52425 Jülich (Germany); and others

    2013-10-15

    Highlights: • Experiments related to the design of bulk tungsten divertor tiles for JET were performed along with extensive modelling. • Temperatures measured in the torus are in good agreement with the model. Some characteristic times show stronger deviations. • Those deviations have no incidence on the highest temperatures reached. • The experimental behaviour of the bulk W tiles is close to design values in a wide range of operational parameters. • The conclusions apply up to deposited energy densities around 30 MJ/m{sup 2}. -- Abstract: The design of the tile assemblies of the bulk tungsten divertor row in JET was improved in the course of several experiments as far as the power and energy performances are concerned: many prototypes were exposed to high heat fluxes in several electron and ion beam facilities during the development phase. These experiments were carried out in parallel with extensive modelling of the complete tungsten tile assembly in the so-called Global Thermal Model (GTM). The goal was to understand the heat flow from the plasma-facing surface through the supporting structure down to the base plate of the JET MkII divertor sufficiently to be able to later interpret operational data from the torus. Temperatures measured in the torus are in good agreement (−10/+15%) with the model. Some characteristic times show stronger deviations, with no incidence on the highest temperature at all times.

  18. Power handling of the bulk tungsten divertor row at JET: First measurements and comparison to the GTM thermal model

    International Nuclear Information System (INIS)

    Highlights: • Experiments related to the design of bulk tungsten divertor tiles for JET were performed along with extensive modelling. • Temperatures measured in the torus are in good agreement with the model. Some characteristic times show stronger deviations. • Those deviations have no incidence on the highest temperatures reached. • The experimental behaviour of the bulk W tiles is close to design values in a wide range of operational parameters. • The conclusions apply up to deposited energy densities around 30 MJ/m2. -- Abstract: The design of the tile assemblies of the bulk tungsten divertor row in JET was improved in the course of several experiments as far as the power and energy performances are concerned: many prototypes were exposed to high heat fluxes in several electron and ion beam facilities during the development phase. These experiments were carried out in parallel with extensive modelling of the complete tungsten tile assembly in the so-called Global Thermal Model (GTM). The goal was to understand the heat flow from the plasma-facing surface through the supporting structure down to the base plate of the JET MkII divertor sufficiently to be able to later interpret operational data from the torus. Temperatures measured in the torus are in good agreement (−10/+15%) with the model. Some characteristic times show stronger deviations, with no incidence on the highest temperature at all times

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

    Energy Technology Data Exchange (ETDEWEB)

    Skovorodin, D. I., E-mail: dskovorodin@gmail.com; Arakcheev, A. S. [Budker Institute of Nuclear Physics, Novosibirsk 630090 (Russian Federation); Pshenov, A. A.; Eksaeva, E. A.; Marenkov, E. D.; Krasheninnikov, S. I. [National Research Nuclear University MEPhI, Moscow 115409 (Russian Federation)

    2016-02-15

    The erosion of divertor targets caused by high heat fluxes during transients is a serious threat to ITER operation, as it is going to be the main factor determining the divertor lifetime. Under the influence of extreme heat fluxes, the surface temperature of plasma facing components can reach some certain threshold, leading to an onset of intense material evaporation. The latter results in formation of cold dense vapor and secondary plasma cloud. This layer effectively absorbs the energy of the incident plasma flow, turning it into its own kinetic and internal energy and radiating it. This so called vapor shielding is a phenomenon that may help mitigating the erosion during transient events. In particular, the vapor shielding results in saturation of energy (per unit surface area) accumulated by the target during single pulse of heat load at some level E{sub max}. Matching this value is one of the possible tests to verify complicated numerical codes, developed to calculate the erosion rate during abnormal events in tokamaks. The paper presents three very different models of vapor shielding, demonstrating that E{sub max} depends strongly on the heat pulse duration, thermodynamic properties, and evaporation energy of the irradiated target material. While its dependence on the other shielding details such as radiation capabilities of material and dynamics of the vapor cloud is logarithmically weak. The reason for this is a strong (exponential) dependence of the target material evaporation rate, and therefore the “strength” of vapor shield on the target surface temperature. As a result, the influence of the vapor shielding phenomena details, such as radiation transport in the vapor cloud and evaporated material dynamics, on the E{sub max} is virtually completely masked by the strong dependence of the evaporation rate on the target surface temperature. However, the very same details define the amount of evaporated particles, needed to provide an effective shielding

  20. Integration of a Radiative Divertor for Heat Load Control into JET Operational Scenarios

    International Nuclear Information System (INIS)

    Full text: The ITER-like Wall (ILW) project in JET is replacing its Plasma Facing Components currently made of Carbon-Fibre Composite with bulk beryllium main-chamber limiters with a full tungsten divertor: the material selection for the DT phase in ITER. The first concern that arises for JET plasma operation is to reduce PFC steady-state power loads to keep within the engineering limits for the ILW. The second is to limit the impurity production from the new PFC to maintain plasma performance, in terms of dilution (Be) and core radiation (W) by reducing the target plasma temperature. Both issues can be addressed by increasing divertor recycling and radiation. The latest JET experiments investigate the achievements of these aims by fuelling and/or impurity seeding of Ne and N2, and assesses the impact on plasma performance on three main standard scenarios the ELMy H-mode, the Steady-State scenario and the Hybrid scenario. In the case of the ELMy H-mode, it was found that the incident power on the outer target can be reduced from 25% of the 16 MW input power, to 11% by D2 fuelling alone with an acceptable degradation of the confinement, H98(y,2) ∼ 0.95, and a decrease of Zeff by 0.2 from the reference value of 1.7. The power reaching the outer target can be reduced further down to 5%. of the input power, with the additional seeding of Ne and N2 for a modest confinement loss, H98(y,2) ∼ 0.9. The Zeff is moderately increased by 0.3 and 0.2 for Ne and N2 respectively. It was found that by D2 fuelling alone, the target temperature does not decrease below 30 eV, with additional Ne not below 20 eV, while in the N2-seeded cases it can be reduced to 10 - 15 eV at the highest fuelling and seeding rates. The results of the ELMy H-mode and AT-like case studies will be extrapolated to the future carbon-free machine with the help of interpretative EDGE2D-EIRENE. It will be assessed whether there is any possibility to operate with D2 fuelling alone at the present rate of

  1. Flow instabilities in non-uniformly heated helium jet arrays used for divertor PFCs

    International Nuclear Information System (INIS)

    In this study, due to a lack of prototypical experimental data, little is known about the off-normal behavior of recently proposed divertor jet cooling concepts. This article describes a computational fluid dynamics (CFD) study on two jet array designs to investigate their susceptibility to parallel flow instabilities induced by non-uniform heating and large increases in the helium outlet temperature. The study compared a single 25-jet helium-cooled modular divertor (HEMJ) thimble and a micro-jet array with 116 jets. Both have pure tungsten armor and a total mass flow rate of 10 g/s at a 600 °C inlet temperature. We investigated flow perturbations caused by a 30 MW/m2 off-normal heat flux applied over a 25 mm2 area in addition to the nominal 5 MW/m2 applied over a 75 mm2 portion of the face. The micro-jet array exhibited lower temperatures and a more uniform surface temperature distribution than the HEMJ thimble. We also investigated the response of a manifolded nine-finger HEMJ assembly using the nominal heat flux and a 274 mm2 heated area. For the 30 MW/m2 case, the micro-jet array absorbed 750 W in the helium with a maximum armor surface temperature of 1280 °C and a fluid/solid interface temperature of 801 °C. The HEMJ absorbed 750 W with a maximum armor surface temperature of 1411 °C and a fluid/solid interface temperature of 844 °C. For comparison, both the single HEMJ finger and the micro-jet array used 5-mm-thick tungsten armor. The ratio of maximum to average temperature and variations in the local heat transfer coefficient were lower for the micro-jet array compared to the HEMJ device. Although high heat flux testing is required to validate the results obtained in these simulations, the results provide important guidance in jet design and manifolding to increase heat removal while providing more even temperature distribution and minimizing non-uniformity in the gas flow and thermal stresses at the armor joint

  2. Fluid-particle hybrid simulation on the transports of plasma, recycling neutrals, and carbon impurities in the Korea Superconducting Tokamak Advanced Research divertor region

    Science.gov (United States)

    Kim, Deok-Kyu; Hong, Sang Hee

    2005-06-01

    A two-dimensional simulation modeling that has been performed in a self-consistent way for analysis on the fully coupled transports of plasma, recycling neutrals, and intrinsic carbon impurities in the divertor domain of tokamaks is presented. The numerical model coupling the three major species transports in the tokamak edge is based on a fluid-particle hybrid approach where the plasma is described as a single magnetohydrodynamic fluid while the neutrals and impurities are treated as kinetic particles using the Monte Carlo technique. This simulation code is applied to the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak [G. S. Lee, J. Kim, S. M. Hwang et al., Nucl. Fusion 40, 575 (2000)] to calculate the peak heat flux on the divertor plate and to explore the divertor plasma behavior depending on the upstream conditions in its base line operation mode for various values of input heating power and separatrix plasma density. The numerical modeling for the KSTAR tokamak shows that its full-powered operation is subject to the peak heat loads on the divertor plate exceeding an engineering limit, and reveals that the recycling zone is formed in front of the divertor by increasing plasma density and by reducing power flow into the scrape-off layer. Compared with other researchers' work, the present hybrid simulation more rigorously reproduces severe electron pressure losses along field lines by the presence of recycling zone accounting for the transitions between the sheath limited and the detached divertor regimes. The substantial profile changes in carbon impurity population and ionic composition also represent the key features of this divertor regime transition.

  3. Modeling effects of local surface properties on heat flux deposition in the JET divertor

    Science.gov (United States)

    Corre, Y.; Hogan, J.; Gauthier, E.; Andrew, P.; Eich, T.; Jachmich, S.; Loarer, T.; Matthews, G.; Monier-Garbet, P.

    2003-10-01

    Understanding heat flux deposition from ELMs is an essential issue for a next step fusion device. A high time resolution infrared system is used in JET to measure the surface temperature distribution and its evolution on the divertor target plates. Previously, an empirical technique was developed, based on a flexible 1D model calculation, to assess possible complications due to surface layer properties, such as poorly adhered a-C:D layers [1]. The model validation used data from JET DOC-L discharges (DOC-L: inner and outer strike points positioned for optimized infrared measurements) with programmed constant L-mode power steps. The effect of layers was identified for the inner tile surface. In this paper we compare the 1D model for surface temperature evolution with results of 3-D modeling with the CASTEM-2000 thermal code, for these DOC-L power step cases. The 1-D values are shown to approach the 3-D results as the model power deposition width increases, showing that there is a absolute 30% accuracy for the 1D model along with a well-supported validation for its use in scaling studies. Additional modeling describing the role of layers and also of small localized heat sinks (dust), as is suggested for similar cases [2], will be presented. [1] Y. Corre et al, EPS 2003, St. Peterburg [2] E. Delchambre et al, J Nucl Mater 2003

  4. Nonlinear heat flux estimation in the JET divertor with the ITER like wall

    International Nuclear Information System (INIS)

    The present paper deals with a nonlinear unsteady heat flux calculation in the case when measurements are provided by only one thermocouple (TC) embedded in the material and the spatial shape of the unknown surface heat flux is given. This inverse problem is solved with the Conjugate Gradient Method (CGM) combined with the adjoint state, the direct problem being solved with the finite element method. This heat flux estimation technique is illustrated in the case of the plasma facing components located in the JET tokamak divertor that can be exposed to several MW m-2 during more than 10 s. In those tiles, few embedded thermocouples (TC) located 1 cm below the tile surface are used to measure the bulk temperatures of the Carbon Fiber Composite (CFC) composite tiles (which are coated with 14 mm of tungsten for the International Thermonuclear Experimental Reactor (ITER) like wall). A numerical study is first presented in order to validate the heat flux calculation and to study the accuracy of the method. Then experimental data from a recent shot with the ITER-like wall configuration are used in the heat flux calculation presented here. Results are compared with those obtained with the deconvolution technique in the linear case, on a simplified geometry of the tiles. (authors)

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

    International Nuclear Information System (INIS)

    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

  6. The influence of the dynamic ergodic divertor on the radial electric field at the Tokamak TEXTOR

    International Nuclear Information System (INIS)

    In this work the influence of external Resonant Magnetic Perturbations (RMPs) on the radial electric field Er in magnetically confined plasmas is investigated by Charge Exchange Recombination Spectroscopy (CXRS) at the Tokamak TEXTOR. Here, the RMPs are produced with the Dynamic Ergodic Divertor (DED), a set of 16 helical perturbation coils located at the high field side of TEXTOR. Within this work, the base mode number of perturbations has been m/n=6/2. We have first investigated the influence of external torque from neutral heating beams on plasma rotation and Er. The ergodic zone causes an electron loss, and subsequently a vector j x vector B force driven by the compensating ion return current. In addition, the DED changes the global confinement properties. Depending on the edge safety factor (''field line twist'') qa, either increased or decreased particle confinement is observed. In case of the increased particle confinement (IPC) the increase in density (40%) and particle confinement time τp (30%) is correlated to the connection of field lines at the q=5/2 surface to the DED target, locally changing the transport properties and the Er. Transport is reduced and the Er shear is increased locally at q=5/2 up to 1.5 . 105s-1, while the Er becomes more positive. (orig.)

  7. Design of a water cooled monoblock divertor for DEMO using Eurofer as structural material

    Energy Technology Data Exchange (ETDEWEB)

    Richou, Marianne, E-mail: marianne.richou@cea.fr [CEA, IRFM, F-13108 Saint Paul Lez Durance (France); Li-Puma, Antonella [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Visca, Eliseo [Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, IT-00044 Frascati (Italy)

    2014-10-15

    The performed investigation focus on a monoblock type design for a water cooled DEMO divertor using Eurofer as structural material. In 2013, a study case of such a concept was presented. It was shown that basic concepts using Eurofer as structural material are limited to an incident heat flux of 8 MW m{sup −2}. Since, the EFDA agency issued new specifications. In this study, the conceptual design is reassessed with regard to specifications. Then, steady state thermal analyses and thermo-mechanical elastic analyses have been performed to define an upgrade of the geometry taking into account new specifications, design criteria and the maximum heat flux requirement of 10 MW m{sup −2}. An analysis of the influence of each adjustable geometrical parameter on thermo-mechanical design criteria was performed. As a consequence, geometrical parameters were set in order to fit to materials requirements. For defined hydraulic conditions taken in the most favourable configuration, the limit of this design is estimated to an incident heat flux of 10 MW m{sup −2}. Margin to critical heat flux and rules against progressive deformation/ratcheting in structural material limit the design.

  8. The influence of the dynamic ergodic divertor on the radial electric field at the Tokamak TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Coenen, Jan Willem

    2009-11-06

    In this work the influence of external Resonant Magnetic Perturbations (RMPs) on the radial electric field Er in magnetically confined plasmas is investigated by Charge Exchange Recombination Spectroscopy (CXRS) at the Tokamak TEXTOR. Here, the RMPs are produced with the Dynamic Ergodic Divertor (DED), a set of 16 helical perturbation coils located at the high field side of TEXTOR. Within this work, the base mode number of perturbations has been m/n=6/2. We have first investigated the influence of external torque from neutral heating beams on plasma rotation and E{sub r}. The ergodic zone causes an electron loss, and subsequently a (vector)j x (vector)B force driven by the compensating ion return current. In addition, the DED changes the global confinement properties. Depending on the edge safety factor (''field line twist'') q{sub a}, either increased or decreased particle confinement is observed. In case of the increased particle confinement (IPC) the increase in density (40%) and particle confinement time {tau}{sub p} (30%) is correlated to the connection of field lines at the q=5/2 surface to the DED target, locally changing the transport properties and the E{sub r}. Transport is reduced and the E{sub r} shear is increased locally at q=5/2 up to 1.5 . 10{sup 5}s{sup -1}, while the E{sub r} becomes more positive. (orig.)

  9. Plasma flow and carbon production and circulation with the ergodic divertor of Tore Supra

    Science.gov (United States)

    Corre, Y.; Gunn, J.; Pégourié, B.; Guirlet, R.; DeMichelis, C.; Giannella, R.; Ghendrih, P.; Hogan, J.; Monier-Garbet, P.; Azéroual, A.; Escarguel, A.; Gauthier, E.

    2007-02-01

    This paper presents a detailed study of carbon production and transport from the ergodic divertor (ED) target plates to the plasma core in the Tore Supra tokamak. Adapted experimental and numerical modelling techniques have been used to describe each of the main phenomena in play. Edge electron density and temperature are measured with Langmuir probes. The C II, C III and Hα emission is measured with optical fibres and cameras. The background plasma flow is calculated consistently with the observed recycling pattern by the neutral transport code EDCOLL for the two magnetic connection schemes of interest (short or long connection lengths). 3D Monte-Carlo modelling of carbon near the neutralizer plate (BBQ code) shows that the transport of carbon ions is governed by the friction force in addition to the electric field. Finally, a simplified 3D test particle model is used to estimate the core penetration fraction of carbon. A high value is found for the carbon screening efficiency (fraction of particles that does not penetrate in the plasma core), in the range 95-97% depending on the edge plasma conditions. This value, combined with the calculated carbon influxes, yields the first quantitative estimate of the carbon core contamination during ED operation. The paper shows that the screening of carbon and core contamination are mainly dependent on the carbon source (partially controlled with the ED) and the plasma flow distribution in the laminar region (magnetic topology and particle drifts).

  10. Particle transport studies for single-null divertor discharges in DIII-D

    International Nuclear Information System (INIS)

    In this paper an investigation of the particle confinement for beam-heated single-null discharges in the open divertor configuration of Doublet III-D (DIII-D) [E. J. Doyle et al., Phys. Fluids B 3, 2300 (1991)] is described. Results are based on a Monte Carlo neutral transport model with a relatively simple plasma model that utilizes experimental data on density, temperature, and heat flux profiles in the edge plasma. For a typical discharge, it is found that the particle confinement time in the quiescent H-mode phase is only about a factor of 2 larger than during the L-mode phase, an increase comparable to the energy confinement time increase. For both H-mode and L-mode phases the particle confinement time is about a factor of 4 larger than the energy confinement time. It is also found that the core plasma fueling rate is higher in the H mode due to the increased transparency of a thinner scrape-off layer. The longer particle confinement time and the increased fueling rate both contribute to the observed density rise during the quiescent period following the L--H transition. Flux surface-averaged transport modeling of the time evolution for the core plasma density profile during H mode suggests that a strong inward particle pinch is necessary near the separatrix

  11. Analysis of energy flux deposition and sheath transmission factors during ergodic divertor operation on Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Costanzo, L. E-mail: costanzo@drfc.cad.cea.fr; Gunn, J.P.; Loarer, T.; Colas, L.; Corre, Y.; Ghendrih, Ph.; Grisolia, C.; Grosman, A.; Guilhem, D.; Monier-Garbet, P.; Reichle, R.; Roche, H.; Vallet, J.C

    2001-03-01

    The magnetic deflection of field lines to dedicated wall components in the ergodic divertor of Tore Supra generates complex patterns of power deposition. In this paper, we analyze the energy flux deposition on neutralizer plates as measured by infrared cameras and Langmuir probes. Three important features will be discussed: (1) The energy deposition during helium shots is as much as twice that for deuterium shots, for a given input power level. (2) The sheath heat transmission factor, deduced experimentally by comparison between probes and infrared measurements, increases with input power independently of the working gas from {approx}7.5 (P{sub TOT}=1 MW) to {approx}10-11 for P{sub TOT}=5 MW). In ohmic discharges, the standard value of 7 is recovered except specific cases in helium where {gamma} can decrease to 2 or 3. (3) These anomalous values put in doubt the validity of edge temperature measurements by Langmuir probes in detached plasmas and have led to the development of a promising 'infrared' degree of detachment (Dod)

  12. Comparison of transient and stationary neutral pressure response in the DIII-D advanced divertor

    International Nuclear Information System (INIS)

    The DIII-D divertor baffle system was designed to facilitate density control in long pulse H-mode discharges by removing a particle flux equal to the neutral beam fueling rate (∼20 Torr-1/s) with a ∼1mTorr neutral pressure under the baffle (p0). Initial measurements of the baffle pressure indicated that p0∼ 10 mTorr (without pumping or biasing), a value much in excess of that required for long pulse density control. Radial sweeps of the X-point position have been employed to determine the maximum p0, as well as to establish the dependence of this pressure on geometry. An estimate of the particle equilibration time for the baffle system has been made by studying the baffle pressure response to ''giant'' ELM effects. ''Steady state'' experiments in which the X-point position was fixed for ∼2.5s have also been carried out and steady baffle pressures were observed. The scaling of baffle pressure with plasma parameters has been found to be similar under transient and ''steady state'' conditions. Detailed modeling of these experiments with the B2, DEGAS, and WDIFFUSE (wall model) codes has been made

  13. Tearing mode physics studies applying the dynamic ergodic divertor on TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Koslowski, H R [Forschungszentrum Juelich GmbH, Institut fuer Plasmaphysik, Association EURATOM-FZ Juelich, D-52425 Juelich (Germany); Westerhof, E [FOM-Institute for Plasmaphysics ' Rijnhuizen' , Association EURATOM-FOM, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Bock, M de [FOM-Institute for Plasmaphysics ' Rijnhuizen' , Association EURATOM-FOM, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Classen, I [FOM-Institute for Plasmaphysics ' Rijnhuizen' , Association EURATOM-FOM, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Jaspers, R [FOM-Institute for Plasmaphysics ' Rijnhuizen' , Association EURATOM-FOM, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Kikuchi, Y [Forschungszentrum Juelich GmbH, Institut fuer Plasmaphysik, Association EURATOM-FZ Juelich, D-52425 Juelich (Germany); Kraemer-Flecken, A [Forschungszentrum Juelich GmbH, Institut fuer Plasmaphysik, Association EURATOM-FZ Juelich, D-52425 Juelich (Germany); Lazaros, A [FOM-Institute for Plasmaphysics ' Rijnhuizen' , Association EURATOM-FOM, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Liang, Y [Forschungszentrum Juelich GmbH, Institut fuer Plasmaphysik, Association EURATOM-FZ Juelich, D-52425 Juelich (Germany); Loewenbrueck, K [Forschungszentrum Juelich GmbH, Institut fuer Plasmaphysik, Association EURATOM-FZ Juelich, D-52425 Juelich (Germany); Varshney, S [FOM-Institute for Plasmaphysics ' Rijnhuizen' , Association EURATOM-FOM, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Hellermann, M von [FOM-Institute for Plasmaphysics ' Rijnhuizen' , Association EURATOM-FOM, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Wolf, R [Forschungszentrum Juelich GmbH, Institut fuer Plasmaphysik, Association EURATOM-FZ Juelich, D-52425 Juelich (Germany); Zimmermann, O [Forschungszentrum Juelich GmbH, Institut fuer Plasmaphysik, Association EURATOM-FZ Juelich, D-52425 Juelich (Germany)

    2006-12-15

    The dynamic ergodic divertor (DED) on the TEXTOR tokamak allows for the reproducible destabilization of the m/n = 2/1 tearing mode which is phase locked to the external static or rotating perturbation field. In combination with its flexible heating systems (co- and counter-neutral beam injection, ion cyclotron resonance heating, electron cyclotron resonance heating (ECRH) with steerable launcher) dedicated experiments to study the mode onset, properties of large islands and mode stabilization can be performed. The dependence of the mode excitation threshold (field penetration) on the plasma rotation shows a resonance character, with minimum threshold when the external perturbation frequency matches the MHD frequency of the 2/1 mode. Mode stabilization by ECRH heating shows that for the TEXTOR plasma heating is more effective than the current drive in O-point. Extrapolation to ITER yields a significant contribution to the mode suppression originating from the temperature increase within the island. Alfven-like modes, which have been previously identified in the vicinity of large islands on FTU (Buratti et al 2005 Nuclear Fusion 45 1446), are found to be created already before island formation above a certain threshold of the externally applied perturbation field.

  14. Development of remote pipe cutting tool for divertor cassettes in JT-60SA

    International Nuclear Information System (INIS)

    Remote pipe cutting tool accessing from inside pipe has been newly developed for JT-60SA. The tool head equips a disk-shaped cutter blade and four rollers which are subjected to the reaction force. The tool pushes out the cutter blade by decreasing the distance between two cams. The tool cuts a cooling pipe by both pushing out the cutter blade and rotating the tool head itself. The roller holder is not pushed out anymore after touching the inner wall of the pipe. In other words, only cutter blade is pushed out after bringing the tool axis into the pipe axis. Outer diameter of the cutting tool head is 44 mm. The cutting tool is able to push out the cutter blade up to 32.5 mm in radius, i.e. 65 mm in diameter, which is enough to cut the pipe having an outer diameter of 59.8 mm. The thickness and material of the cooling pipe are 2.8 mm and SUS316L, respectively. The length of the cutting tool head is about 1 m. The tool is able to cut a pipe locates about 480 mm in depth from the mounting surface on the divertor cassette. The pipe cutting system equips two cutting heads and they are able to cut two pipes at the same time in order to remove the inner target plate. Reproducibility of the cross-sectional shape of the cut pipe is required for re-welding. The degree of reproducibility is inside 0.1 mm except for burr at outside of the pipe, which is enough to re-weld the cut pipe. Some swarf is generated during cutting the double-layered pipe assuming a plug located on the top of the pipe. The swarf is deposited on the bottom of the plug and collected by pulling out the plug in the actual equipment

  15. Interfacial metallurgy study of brazed joints between tungsten and fusion related materials for divertor design

    International Nuclear Information System (INIS)

    Highlights: • We created brazed joints between tungsten and EUROFER 97, Cu and SS316L with Au80Cu19Fe1 filler. • No elemental transitions were detected between the W and the AuCuFe filler in either direction. • Transition regions between filler to EUROFER97/316L showed similar elastic modulus and hardness to the filler. • Smooth elemental and mechanical properties transition were detected between the filler and Cu. - Abstract: In the developing DEMO divertor, the design of joints between tungsten to other fusion related materials is a significant challenge as a result of the dissimilar physical metallurgy of the materials to be joined. This paper focuses on the design and fabrication of dissimilar brazed joints between tungsten and fusion relevant materials such as EUROFER 97, oxygen-free high thermal conductivity (OFHC) Cu and SS316L using a gold based brazing foil. The main objectives are to develop acceptable brazing procedures for dissimilar joining of tungsten to other fusion compliant materials and to advance the metallurgical understanding within the interfacial region of the brazed joint. Four different butt-type brazed joints were created and characterised, each of which were joined with the aid of a thin brazing foil (Au80Cu19Fe1, in wt.%). Microstructural characterisation and elemental mapping in the transition region of the joint was undertaken and, thereafter, the results were analysed as was the interfacial diffusion characteristics of each material combination produced. Nano-indentation tests are performed at the joint regions and correlated with element composition information in order to understand the effects of diffused elements on mechanical properties. The experimental procedures of specimen fabrication and material characterisation methods are presented. The results of elemental transitions after brazing are reported. Elastic modulus and nano-hardness of each brazed joints are reported

  16. Deep drawing of tungsten plates for structural divertor applications in future fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Reiser, J., E-mail: jens.reiser@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM-AWP) (Germany); Rieth, M.; Dafferner, B.; Baumgaertner, S.; Ziegler, R. [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM-AWP) (Germany); Hoffmann, A. [PLANSEE Metall GmbH, Reutte (Austria)

    2011-12-15

    Highlights: Black-Right-Pointing-Pointer Successful deep drawing of a 1 mm tungsten plate. Black-Right-Pointing-Pointer The grains follow the contour - grain boundary alignment. Black-Right-Pointing-Pointer Deep drawing fits to the needs of a mass production. Black-Right-Pointing-Pointer Charpy tests of 1 mm tungsten plate material prove the anisotropy material behavior. - Abstract: The reference design of a helium cooled divertor for future fusion reactors makes use of hundreds of thousands of finger units consisting of a pressurized structural part called a thimble. Due to the high number of parts needed, the thimble has to be fabricated by mass production techniques like deep drawing. As the thimble is a pressurized part exposed to an internal pressure of 100 bar, the demands for the material are high, which means that it requires the best available tungsten material. Former work has shown that pure tungsten material has the best impact properties and has to be preferred over other commercially available tungsten materials, such as that doped with potassium or strengthened with oxides like lanthanum oxide. Furthermore the inherent weakness of the grain boundaries has to be taken into account, which requires the need for grains that are aligned to the contour of the part (grain boundary alignment). This paper describes the successful deep drawing of a 1 mm tungsten plate in high vacuum at 600 Degree-Sign C. In doing this, a thimble can be machined with grains that follow the contour. Furthermore the characterization of a 1 mm tungsten plate is conducted by tensile tests at room temperature and at 600 Degree-Sign C, as well as by Charpy tests taking into account the anisotropic material behaviour.

  17. 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.

  18. Assessment of the integration of a He-cooled divertor system in the power conversion system for the dual-coolant blanket concept (TW2-TRP-PPCS12D8)

    Energy Technology Data Exchange (ETDEWEB)

    Norajitra, P.; Kruessmann, R.; Malang, S.; Reimann, G.

    2002-12-01

    Application of a helium-cooled divertor together with the dual-coolant blanket concept is considered favourable for achieving a high thermal efficiency of the power plant due to its relatively high coolant outlet temperature. A new FZK He-cooled modular divertor concept with integrated pin arrays (HEMP) is introduced. Its main features and function are described in detail. The result of the thermalhydraulic analysis shows that the HEMP divertor concept has the potential of resisting, a heat flow density of at least 10-15 MW/m{sup 2} at a reachable heat transfer coefficient of approx. 60 kW/m{sup 2}K and a reasonable pumping power. Integration of this divertor concept into the power conversion system using a closed Brayton gas turbine system with three-stage compression leads to a net efficiency of the blanket/divertor cycle of about 43%. (orig.)

  19. In situ spectral calibration method for the impurity influx monitor (divertor) for ITER using angled physical contact fibers.

    Science.gov (United States)

    Iwamae, A; Ogawa, H; Sugie, T; Kusama, Y

    2011-03-01

    The in situ calibration method for the impurity influx monitor (divertor) is experimentally examined. The total reflectance of the optical path from the focal point of the Cassegrain telescope to the first mirror is derived using a micro retroreflector array. An optical fiber with angled physical contact (APC) connectors reduces the return edge reflection. APC fibers and a multimode coupler increase the signal-to-noise ratio by about one order compared to that of triple-branched fibers and enable measurement of the wavelength dependence of the total reflectance of the optical system even after potential deterioration of mirror surfaces reduces reflectance.

  20. Regime of Improved Confinement and High Beta in Neutral-Beam-Heated Divertor Discharges of the ASDEX Tokamak

    Science.gov (United States)

    Wagner, F.; Becker, G.; Behringer, K.; Campbell, D.; Eberhagen, A.; Engelhardt, W.; Fussmann, G.; Gehre, O.; Gernhardt, J.; Gierke, G. V.; Haas, G.; Huang, M.; Karger, F.; Keilhacker, M.; Klüber, O.; Kornherr, M.; Lackner, K.; Lisitano, G.; Lister, G. G.; Mayer, H. M.; Meisel, D.; Müller, E. R.; Murmann, H.; Niedermeyer, H.; Poschenrieder, W.; Rapp, H.; Röhr, H.; Schneider, F.; Siller, G.; Speth, E.; Stäbler, A.; Steuer, K. H.; Venus, G.; Vollmer, O.; Yü, Z.

    1982-11-01

    A new operational regime has been observed in neutral-injection-heated ASDEX divertor discharges. This regime is characterized by high βp values comparable to the aspect ratio A (βp=1.9 MW, a mean density n¯e>=3×1013 cm-3, and a q(a) value >=2.6. Beyond these limits or in discharges with material limiter, low βp values and reduced particle and energy confinement times are obtained compared to the Ohmic heating phase.

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

    International Nuclear Information System (INIS)

    The MCI transport model was used to compare chemical and physical sputtering for a DIII-D divertor plasma near detachment. With physical sputtering alone the integrated carbon influx was 8.4 x 1019 neutral/s while physical plus chemical sputtering produced an integrated carbon influx of 1.7 x 1021 neutrals/s. The average carbon concentration in the computational volume increased from 0.012% with only physical sputtering to 0.182% with both chemical and physical sputtering. This increase in the carbon inventory produced more radiated power which is in better agreement with experimental measurements

  2. Modelling of spatial structure of divertor footprints caused by edge-localized modes mitigated by magnetic perturbations

    OpenAIRE

    Cahyna, Pavel; Becoulet, Marina; Huijsmans, Guido T. A.; Orain, Francois; Morales, Jorge; Kirk, Andrew; Thornton, Andrew J.; Pamela, Stanislas; Panek, Radomir; Hoelzl, Matthias

    2016-01-01

    Resonant magnetic perturbations (RMPs) can mitigate the edge-localized modes (ELMs), i.e. cause a change of the ELM character towards smaller energy loss and higher frequency. During mitigation a change of the spatial structure of ELM loads on divertor was observed on DIII-D and MAST: the power is deposited predominantly in the footprint structures formed by the magnetic perturbation. In the present contribution we develop a theory explaining this effect, based on the idea that part of the EL...

  3. Abelian modules

    OpenAIRE

    S. Halıcıoğlu; Harmanci, A.; GÜNGÖROĞLU, G.; N. Agayev

    2009-01-01

    In this note, we introduce abelian modules as a generalization of abelian rings. Let R be an arbitrary ring with identity. A module M is called abelian if, for any m Î M and any a Î R, any idempotent e Î R, mae=mea. We prove that every reduced module, every symmetric module, every semicommutative module and every Armendariz module is abelian. For an abelian ring R, we show that the module MR is abelian iff M[x]R[x] is abelian. We produce an example to show that M[x, α] need not be abe...

  4. Production and fabrication of vanadium alloys for the radiative divertor program of DIII-D - Annual report input for 1996

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, W.R.; Smith, J.P.; Stambaugh, R.D.

    1996-10-01

    V-4Cr-4Ti alloy has been selected for use in the manufacture of a portion of the DIII-D Radiative Divertor (RD) upgrade. The production of a 1200-kg ingot of V-4Cr-4Ti alloy has been completed at Teledyne Wah Chang of Albany, Oregon (TWCA) to provide {approximately}800-kg of applicable product forms, and two billets have been extruded from the ingot. Chemical compositions of the ingot and both extruded billets were acceptable. Material from these billets will be converted into product forms suitable for components of the DIII-D Radiative Divertor structure. Joining of V-4Cr-4Ti alloy has been identified as the most critical fabrication issue for its use in the RD Program, and research into several joining methods for fabrication of the RD components, including resistance seam, friction, and electron beam welding, is continuing. Preliminary trials have been successful in the joining of V-alloy to itself by electron beam, resistance, and friction welding processes and to Inconel 625 by friction welding.

  5. Material deposition on inner divertor quartz-micro balances during ITER-like wall operation in JET

    Science.gov (United States)

    Esser, H. G.; Philipps, V.; Freisinger, M.; Widdowson, A.; Heinola, K.; Kirschner, A.; Möller, S.; Petersson, P.; Brezinsek, S.; Huber, A.; Matthews, G. F.; Rubel, M.; Sergienko, G.

    2015-08-01

    The migration of beryllium, tungsten and carbon to remote areas of the inner JET-ILW divertor and the accompanying co-deposition of deuterium has been investigated using post-mortem analysis of the housings of quartz-micro balances (QMBs) and their quartz crystals. The analysis of the deposition provides that the rate of beryllium atoms is significantly reduced compared to the analogue deposition rate of carbon during the carbon wall conditions (JET-C) at the same locations of the QMBs. A reduction factor of 50 was found at the entrance gap to the cryo-pumps while it was 14 under tile 5, the semi-horizontal target plate. The deposits consist of C/Be atomic ratios of typically 0.1-0.5 showing an enrichment of carbon in remote areas compared to directly exposed areas with less carbon. The deuterium retention fraction D/Be is between 0.3 and 1 at these unheated locations in the divertor.

  6. Production and fabrication of vanadium alloys for the radiative divertor program of DIII-D - Annual report input for 1996

    International Nuclear Information System (INIS)

    V-4Cr-4Ti alloy has been selected for use in the manufacture of a portion of the DIII-D Radiative Divertor (RD) upgrade. The production of a 1200-kg ingot of V-4Cr-4Ti alloy has been completed at Teledyne Wah Chang of Albany, Oregon (TWCA) to provide ∼800-kg of applicable product forms, and two billets have been extruded from the ingot. Chemical compositions of the ingot and both extruded billets were acceptable. Material from these billets will be converted into product forms suitable for components of the DIII-D Radiative Divertor structure. Joining of V-4Cr-4Ti alloy has been identified as the most critical fabrication issue for its use in the RD Program, and research into several joining methods for fabrication of the RD components, including resistance seam, friction, and electron beam welding, is continuing. Preliminary trials have been successful in the joining of V-alloy to itself by electron beam, resistance, and friction welding processes and to Inconel 625 by friction welding

  7. Potential and limits of water cooled divertor concepts based on monoblock design as possible candidates for a DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Li-Puma, Antonella, E-mail: antonella.lipuma@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Richou, Marianne; Magaud, Philippe; Missirlian, Marc [CEA, IRFM, F-13108 Saint Paul Lez Durance (France); Visca, Eliseo [Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, IT-00044 Frascati (Italy); Ridolfini, Vincenzo Pericoli [EFDA-CSU Garching, PPPT department, D-85748 Garching bei München (Germany)

    2013-10-15

    In this paper water-cooled divertor concepts based on tungsten monoblock design identified in previous studies as candidate for fusion power plant have been reviewed to assess their potential and limits as possible candidates for a DEMO concept deliverable in a short to medium term (“conservative baseline design”). The rationale and technology development assumptions that have led to their selection are revisited taking into account present factual information on reactor parameters, materials properties and manufacturing technologies. For that purpose, main parameters impacting the divertor design are identified and their relevance discussed. The state of the art knowledge on materials and relevant manufacturing techniques is reviewed. Particular attention is paid to material properties change after irradiation; phenomenon thresholds (if any) and possible operating ranges are identified (in terms of temperature and damage dose). The suitability of various proposed heat sink/structural and sacrificial layer materials, as proposed in the past, are re-assessed (e.g. with regard to the possibility of reducing peak heat flux and/or neutron radiation damages). As a result, potential and limits of various proposed concepts are highlighted, ranges in which they could operate (if any) defined and possible improvements are proposed. Identified missing point in materials database and/or manufacturing techniques knowledge that should be uppermost investigated in future R and D activities are reported. This work has been carried out in the frame of EFDA PPPT Work Programme activities.

  8. DiMES Studies of Temperature Dependence of Carbon Erosion and Re-Deposition in the DIII-D Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Rudakov, D; Jacob, W; Krieger, K; Litnovsky, A; Philipps, V; West, W; Wong, C; Allen, S; Bastasz, R; Boedo, J; Brooks, N; Boivin, R; De Temmerman, G; Fenstermacher, M; Groth, M; Hollmann, E; Lasnier, C; McLean, A; Moyer, R; Stangeby, P; Wampler, W; Watkins, J; Wienhold, P; Whaley, J

    2006-10-02

    A strong effect of a moderately elevated surface temperature on net carbon deposition and deuterium co-deposition in the DIII-D divertor was observed under detached conditions. A DiMES sample with a gap 2 mm wide and 18 mm deep was exposed to lower-single-null (LSN) L-mode plasmas first at room temperature, and then at 200 C. At the elevated temperature, deuterium co-deposition in the gap was reduced by an order of magnitude. At the plasma-facing surface of the heated sample net carbon erosion was measured at a rate of 3 nm/s, whereas without heating net deposition is normally observed under detachment. In a related experiment three sets of molybdenum mirrors recessed 2 cm below the divertor floor were exposed to identical LSN ELMy H-mode discharges. The first set of mirrors exposed at ambient temperature exhibited net carbon deposition at a rate of up to 3.7 nm/s and suffered a significant drop in reflectivity. In contrast, two other mirror sets exposed at elevated temperatures between 90 C and 175 C exhibited virtually no carbon deposition.

  9. DiMES Studies of Temperature Dependence of Carbon Erosion and Re-Deposition in the DIII-D Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Rudakov, D L; Jacob, W; Krieger, K; Litnovsky, A; Philipps, V; West, W P; Wong, C C; Allen, S L; Bastasz, R J; Boedo, J A; Brooks, N H; Boivin, R L; De Temmerman, G; Fenstermacher, M E; Groth, M; Hollmann, E M; Lasnier, C J; McLean, A G; Moyer, R A; Stangeby, P C; Wampler, W R; Watkins, J G; Wienhold, P; Whaley, J

    2007-03-15

    A strong effect of a moderately elevated surface temperature on net carbon deposition and deuterium co-deposition in the DIII-D divertor was observed under detached conditions. A DiMES sample with a gap 2 mm wide and 18 mm deep was exposed to lower-single-null (LSN) L-mode plasmas first at room temperature, and then at 200 C. At the elevated temperature, deuterium co-deposition in the gap was reduced by an order of magnitude. At the plasma-facing surface of the heated sample net carbon erosion was measured at a rate of 3 nm/s, whereas without heating net deposition is normally observed under detachment. In a related experiment three sets of molybdenum mirrors recessed 2 cm below the divertor floor were exposed to identical LSN ELMy H-mode discharges. The first set of mirrors exposed at ambient temperature exhibited net carbon deposition at a rate of up to 3.7 nm/s and suffered a significant drop in reflectivity. In contrast, two other mirror sets exposed at elevated temperatures between 90 C and 175 C exhibited practically no carbon deposition.

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

    Energy Technology Data Exchange (ETDEWEB)

    Tamain, P., E-mail: patrick.tamain@cea.fr [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Joffrin, E.; Bufferand, H. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Järvinen, A. [Aalto University, Espoo (Finland); Brezinsek, S. [Institut für Energie- und Klimaforschung IEK-4, FZJ, TEC, 52425 Jülich (Germany); Ciraolo, G. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Delabie, E. [FOM Institute for Plasma Physics Rijnhuizen, Postbus 1207, 3430 BE Nieuwegein (Netherlands); Frassinetti, L. [Royal Institute of Technology KTH, SE-10691 Stockholm (Sweden); Giroud, C. [Culham Centre for Fusion Energy, OX14 3DB Abingdon (United Kingdom); Groth, M. [Aalto University, Espoo (Finland); Lipschultz, B. [Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Lomas, P. [Culham Centre for Fusion Energy, OX14 3DB Abingdon (United Kingdom); Marsen, S. [Max-Planck-Institut for Plasma Physics, Greifswald (Germany); Menmuir, S. [Royal Institute of Technology KTH, SE-10691 Stockholm (Sweden); Oberkofler, M. [Max-Planck-Institut für Plasmaphysik, 85748 Garching (Germany); Stamp, M. [Culham Centre for Fusion Energy, OX14 3DB Abingdon (United Kingdom); Wiesen, S. [Institut für Energie- und Klimaforschung IEK-4, FZJ, TEC, 52425 Jülich (Germany); JET EFDA contributors [JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)

    2015-08-15

    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 essentially unchanged. Large variations of the global plasma confinement have been observed, the H{sub 98} factor changing from typically 0.7 when the outer strike-point is on the vertical or horizontal targets to 0.9 when it is located in the pump duct entrance. Profiles are mainly impacted in the pedestal but core gradient lengths, especially for the density, are also modified. Although substantial differences are observed in the divertor conditions, none seem to correlate directly with the confinement. Modelling with the EDGE2D-EIRENE and SOLEDGE2D-EIRENE transport codes exhibits differences in the energy losses due to neutrals inside the separatrix, but orders of magnitude are too low to explain simply the impact on the confinement.

  11. Potential and limits of water cooled divertor concepts based on monoblock design as possible candidates for a DEMO reactor

    International Nuclear Information System (INIS)

    In this paper water-cooled divertor concepts based on tungsten monoblock design identified in previous studies as candidate for fusion power plant have been reviewed to assess their potential and limits as possible candidates for a DEMO concept deliverable in a short to medium term (“conservative baseline design”). The rationale and technology development assumptions that have led to their selection are revisited taking into account present factual information on reactor parameters, materials properties and manufacturing technologies. For that purpose, main parameters impacting the divertor design are identified and their relevance discussed. The state of the art knowledge on materials and relevant manufacturing techniques is reviewed. Particular attention is paid to material properties change after irradiation; phenomenon thresholds (if any) and possible operating ranges are identified (in terms of temperature and damage dose). The suitability of various proposed heat sink/structural and sacrificial layer materials, as proposed in the past, are re-assessed (e.g. with regard to the possibility of reducing peak heat flux and/or neutron radiation damages). As a result, potential and limits of various proposed concepts are highlighted, ranges in which they could operate (if any) defined and possible improvements are proposed. Identified missing point in materials database and/or manufacturing techniques knowledge that should be uppermost investigated in future R and D activities are reported. This work has been carried out in the frame of EFDA PPPT Work Programme activities

  12. Metallurgical Bonding Development of V-4Cr-4Ti Alloy for the DIII-D Radiative Divertor Program

    International Nuclear Information System (INIS)

    General Atomics (GA), in conjunction with the Department of Energy's (DOE) DIII-D Program, is carrying out a plan to utilize a vanadium alloy in the DIII-D tokamak as part of the DIII-D Radiative Divertor (RD) upgrade. The V-4Cr-4Ti alloy has been selected in the U.S. as the leading candidate vanadium alloy for fusion applications. This alloy will be used for the divertor fabrication. Manufacturing development with the V-4Cr-4Ti alloy is a focus of the DIII-D RD Program. The RD structure, part of which will be fabricated from V-4Cr-4Ti alloy, will require many product forms and types of metal/metal bonded joints. Metallurgical bonding methods development on this vanadium alloy is therefore a key area of study by GA. Several solid state (non-fusion weld) and fusion weld joining methods are being investigated. To date, GA has been successful in producing ductile, high strength, vacuum leak tight joints by all of the methods under investigation. The solid state joining was accomplished in air, i.e., without the need for a vacuum or inert gas environment to prevent interstitial impurity contamination of the V-4Cr-4Ti alloy

  13. Studies of high-{delta} (baffled) and low-{delta} (open) pumped divertor operation on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Allen, S.L.; Fenstermacher, M.E. [Lawrence Livermore National Lab., CA (United States); Greenfield, C.M. [General Atomics, San Diego, CA (United States)] [and others

    1998-08-01

    The authors report new experimental results with the RDP-OB (Radiative Divertor Project-outer baffle) and cryopump in both upper single-null (USN) and double-null (DN) ELMing H-mode discharges. The baffled divertor reduced the core ionization ({approximately}2--2.5{times}), in reasonable agreement with predictions from UEDGE/DEGAS modeling ({approximately}3.75{times}). The upper cryopump achieved density control of n{sub e}/n{sub gw} {approximately} 0.22 (line density/Greenwald density) with Z{sub eff} {approximately} 2 in high-{delta} plasmas. The measured exhaust is comparable to the lower pump, except at lower core electron densities (n{sub e} < 5 {times} 10{sup 19} m{sup {minus}3}). Efficient impurity exhaust was obtained with deuterium SOL flow. Preliminary experiments with DN operation has shown that the particle exhaust to the upper pump depends on the up/down magnetic balance. Preliminary experiments indicate that the DN exhaust is roughly 40--50% of the USN exhaust at n{sub e} {approximately} 4 {times} 10{sup 19} m{sup {minus}3}.

  14. Plasma-wall interaction study in the open divertor of Globus-M spherical tokamak

    International Nuclear Information System (INIS)

    Globus-M is the first Russian spherical tokamak, which was built at the A.F. Ioffe Institute in 1999. Currently about 90% of in vessel area, which is faced to plasma, is covered by protection tiles. Tiles are constructed from RGTi graphite, doped by 2 at.% Ti and 0.3-0.7 at.% Si. Gradual increase in plasma parameters was recorded as the wall area protected by tiles constantly grew up. The plasma-facing surface including graphite armor was covered by a-B/C:H layers periodically. Globus-M is usually operating in the density range n ∼ (3-10) x 1019 m-3 with plasma currents 0.2-0.25 MA and toroidal magnetic field of 0.4 T. The main working gas is deuterium. NB injection and IC resonance heating at hydrogen minority fundamental frequency are used as auxiliary heating methods. Specific power deposition is high up to value of several MW/m. Power density deposited at the first wall is also high. This is due to 'spherical' geometry of plasma column in which the first wall area is small with respect to plasma volume and plasma is close to the first wall position. The 'focusing' of power fluxes along separatrix strike points could increase power density up to 10 MW/m2 at the divertor target. RGTi diverter tiles analysis was performed after irradiation by plasma during big number of shots (10000 shots in average). Composition and morphology of the surface layers were examined by different diagnostic tools (electron probe microanalysis, scanning electron microscope, Rutherford backscattering, nuclear resonance reactions, thermal desorption spectroscopy). The most of tiles were covered with deposited mixed layers. Deposits exist even in high flux regions (separatrix strike points). The mixed layers are composed from the elements used in the vessel construction and in conditioning technology processes. The concentration and absorbed deuterium depth profiles in tiles are analyzed and conditions of deuterium desorption are studied. Important result is that deuterium is absorbed

  15. Assessment of a water hydraulics joint for RH operations in the divertor region

    International Nuclear Information System (INIS)

    Due to the high level of radiations, all the nominal maintenance in the divertor region of ITER will be carried out with help of robotic means. In reduced volumes, hydraulic applications can provide powerful actuators. They become an interesting technology to build a heavy duty manipulator for operations in space constrained areas. Oil hydraulics can not ensure the cleanliness level required for all maintenance operations in the vacuum vessel. Therefore, pure water hydraulics proposes a good alternative to oil and developments are today focusing on that direction. Although basic hydraulic elements like pumps, on-off valves, filters running with pure water are already available on the market, actuators are not so many and generally limited to linear motions. Fine control of the joint is achieved with help of servovalves. Today's off the shelf products are only adaptations from standard oil servovalves and are not specifically designed for water use. Operational experience for these products shows short lifetime expectancy and could not last a complete shutdown. Starting from the oil hydraulic version CEA with help of Cybernetix redesigned for water applications the elbow vane actuator of a Maestro arm, a six-degrees-of-freedom hydraulic manipulator used in decommissioning activities. In parallel with help of In-LHC, CEA developed a servovalve for water hydraulic applications that fits the space constraints of a Maestro manipulator. This prototype is a pressure-control valve. To a current input this servovalve supplies a very accurate pressure difference output instead of a flow rate in the case of flow control servovalve that are generally used in that kind of applications. The advantage is the improvement of the performances and stability of the force control loop. This paper presents the performances of the modified vane actuator and its servovalve. Both static and dynamic responses of the servovalve prototype with and without actuator are presented. Position and

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

    International Nuclear Information System (INIS)

    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

  17. Conceptual design of a He-cooled divertor with integrated flow and heat transfer promoters (PPCS subtask TW3-TRP-001-D2). Pt. 1. Summary

    Energy Technology Data Exchange (ETDEWEB)

    Norajitra, P. (ed.); Kruessmann, R. (comp.)

    2004-04-01

    Within the framework of the EU power plant conceptual study (PPCS), helium-cooled modular divertor concepts with a flow promoter (HEMP as a pin array and HEMS as slot array version) have been investigated at the Forschungszentrum Karlsruhe since 2002. The design goal is to achieve a high heat flux performance of 15 MW/m{sup 2}. In this summary of the detailed report, research areas related to the development of a helium-cooled divertor shall be addressed. Latest changes in thermohydraulic layout as well as current results of simulation calculations shall be presented exemplarily for the slot concept HEMS which has the crucial advantage of being easier to manufacture. The divertor construction resulting from the requirements as well as the design-related issues shall be discussed. Possible manufacturing processes for divertor components of tungsten are assessed. Chapters 7 and 8 have been completely revised comprising the latest results of the thermohydraulic layout and thermomechanical analyses. Calculation results have to be verified by experiments. For this purpose, a helium loop will be built at the Efremov Institute, St. Petersburg, Russia, in 2004. An outlook on an alternative multi-jet design (HEMJ) will be given at the end of this report. (orig.)

  18. Particle and impurity transport in the Axial Symmetric Divertor Experiment Upgrade and the Joint European Torus, experimental observations and theoretical understanding

    DEFF Research Database (Denmark)

    Angioni, C.; Carraro, L.; Dannert, T.;

    2007-01-01

    Experimental observations on core particle and impurity transport from the Axial Symmetric Divertor Experiment Upgrade [O. Gruber, H.-S. Bosch, S. Gunter , Nucl Fusion 39, 1321 (1999)] and the Joint European Torus [J. Pamela, E. R. Solano, and JET EFDA Contributors, Nucl. Fusion 43, 1540 (2003...

  19. 3-D Finite Element Electromagnetic and Stress Analyses of the JET LB-SRP Divertor Element (Tungsten Lamella Design)

    International Nuclear Information System (INIS)

    Within the ITER-like wall project at the JET, the original plasma facing divertor tiles made of tungsten coated carbon fibre composite (CFC) are to be replaced by bulk tungsten. The design concept should comply with the power and energy handling requirements, the electromagnetic (EM) forces and the mechanical constraints of the existing remote handling system. Through a number of intermediate design options the '' lamella '' option has been developed. Each divertor block consists of three main parts: the plasma facing tiles, the inconel wedge holding the tiles and the inconel interface plate attaching the wedge to the JET CFC base plate. In order to minimize eddy currents the wedge is equipped with slits and the lamellae are isolated from each other. Defined electrical contact from lamellae via wedge to the base plate is required for defined path of halo currents. Eight tungsten lamella stacks are attached to the wedge. The individual lamellae are isolated from each other by means of insulated spacers. Tie rods keep the stack of tungsten lamellae and ceramic coated spacers together. The aim of this study is verification of the divertor block design with the load bearing septum replacement plate (LB-SRP) with respect to electromagnetic loads in the block components by means of essentially 3-D Finite Element (FE) electromagnetic and stress analyses. The following problems have been simulated and studied: · 3-D FE modeling of eddy and halo currents distribution for different cases of plasma current ramp down · Calculation of EM loads arising in the structure components due to interaction of the currents with external electromagnetic fields for different possible directions of magnetic fields · Selection of the worst load combination cases performed during post-processing of results of EM FE analysis · 3-D multi-contact non-linear stress analysis for the worst load combinations with paying attention to the system integrity at the elements separation planes. As a

  20. Study on the heat flux reconstruction with the infrared thermography for the divertor target plates in the KSTAR tokamak

    Science.gov (United States)

    Kang, C. S.; Lee, H. H.; Oh, S.; Lee, S. G.; Wi, H. M.; Kim, Y. S.; Kim, H. S.

    2016-08-01

    An infrared (IR) thermography is the preferred diagnostic that can quantify heat flux by measuring the surface temperature distributions of the divertor plates. The IR thermography is successfully instrumented on Korea Superconducting Tokamak Advanced Research (KSTAR). In this study, finite volume method is considered to solve the heat conduction equations. 1D-, 2D-, and 3D models are developed and compared with various calculation algorithms, such as Duhamel's theorem and THEODOR. These comparisons show good agreement. In order to acquire more efficient and reliable calculation results, we consider two numerical analysis schemes, influence of temperature on thermal properties and image stabilization. Recently, this reconstruction code is successfully applied to the KSTAR IR thermography.

  1. Modification of the internal electric field by biasing of the divertor plates in the Tokamak de Varennes (TdeV)

    Energy Technology Data Exchange (ETDEWEB)

    Lafrance, D.; Huang, R.; Stansfield, B.L.; Haddad, E.; Lachambre, J. [Centre Canadien de Fusion Magnetique, Varennes, Quebec J3X 1S1 (Canada)

    1997-10-01

    The radial electric field inside the separatrix has been deduced from spectroscopic measurements of impurities on TdeV (Tokamak de Varennes), using the reduced radial momentum balance and two neoclassical models [R. D. Hazeltine, Phys. Fluids {bold 17}, 961 (1974) and Y. B. Kim, P. H. Diamond, and R. J. Groebner, Phys. Fluids B {bold 3}, 2050 (1991)]. The results from all three models are in fair agreement. Furthermore, the electric field has been deduced using the same models both with and without biasing the divertor plates relative to the machine wall, showing an inward propagation of the effect of the biasing created in the scrape-off layer (SOL). Undeniably, the electric field has been modified well inside the separatrix (0.6{approx_lt}r/a{approx_lt}0.9), revealing the possibility of modifying the internal electric field by external means. {copyright} {ital 1997 American Institute of Physics.}

  2. Modification of the internal electric field by biasing of the divertor plates in the Tokamak de Varennes (TdeV)

    International Nuclear Information System (INIS)

    The radial electric field inside the separatrix has been deduced from spectroscopic measurements of impurities on TdeV (Tokamak de Varennes), using the reduced radial momentum balance and two neoclassical models [R. D. Hazeltine, Phys. Fluids 17, 961 (1974) and Y. B. Kim, P. H. Diamond, and R. J. Groebner, Phys. Fluids B 3, 2050 (1991)]. The results from all three models are in fair agreement. Furthermore, the electric field has been deduced using the same models both with and without biasing the divertor plates relative to the machine wall, showing an inward propagation of the effect of the biasing created in the scrape-off layer (SOL). Undeniably, the electric field has been modified well inside the separatrix (0.6 approx-lt r/a approx-lt 0.9), revealing the possibility of modifying the internal electric field by external means. copyright 1997 American Institute of Physics

  3. Study on the heat flux reconstruction with the infrared thermography for the divertor target plates in the KSTAR tokamak.

    Science.gov (United States)

    Kang, C S; Lee, H H; Oh, S; Lee, S G; Wi, H M; Kim, Y S; Kim, H S

    2016-08-01

    An infrared (IR) thermography is the preferred diagnostic that can quantify heat flux by measuring the surface temperature distributions of the divertor plates. The IR thermography is successfully instrumented on Korea Superconducting Tokamak Advanced Research (KSTAR). In this study, finite volume method is considered to solve the heat conduction equations. 1D-, 2D-, and 3D models are developed and compared with various calculation algorithms, such as Duhamel's theorem and THEODOR. These comparisons show good agreement. In order to acquire more efficient and reliable calculation results, we consider two numerical analysis schemes, influence of temperature on thermal properties and image stabilization. Recently, this reconstruction code is successfully applied to the KSTAR IR thermography. PMID:27587124

  4. Power deposition onto plasma facing components in poloidal divertor tokamaks during type-I ELMs and disruptions

    Energy Technology Data Exchange (ETDEWEB)

    Eich, T. [Max-Planck-Institut fuer Plasmaphysik, IPP-EURATOM Association, Boltzmann str. 2, D-85748 Garching (Germany)]. E-mail: thomas.eich@ipp.mpg.de; Herrmann, A.; Pautasso, G.; Fuchs, J.C.; Gruber, O. [Max-Planck-Institut fuer Plasmaphysik, IPP-EURATOM Association, Boltzmann str. 2, D-85748 Garching (Germany); Andrew, P. [EURATOM-UKAEA Fusion Association, Culham Science Center, Abingdon, Oxon OX 14 3DB (United Kingdom); Asakura, N. [Naka Fusion Research Establishment, JAERI (Japan); Boedo, J.A. [University of California, San Diego, La Jolla, CA 92093 (United States); Corre, Y. [Association EURATOM CEA, Cadarache, 13108 St. Paul-lez-Durance (France); Fenstermacher, M.E. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Fundamenski, W. [EURATOM-UKAEA Fusion Association, Culham Science Center, Abingdon, Oxon OX 14 3DB (United Kingdom); Federici, G. [ITER JWS, Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Gauthier, E. [Association EURATOM CEA, Cadarache, 13108 St. Paul-lez-Durance (France); Goncalves, B. [Associacao EURATOM/IST, Instituto Superior Tecnico (Portugal); Kirk, A. [EURATOM-UKAEA Fusion Association, Culham Science Center, Abingdon, Oxon OX 14 3DB (United Kingdom); Leonard, A.W. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Loarte, A. [CSU-EFDA, Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Matthews, G.F. [EURATOM-UKAEA Fusion Association, Culham Science Center, Abingdon, Oxon OX 14 3DB (United Kingdom); Neuhauser, J. [Max-Planck-Institut fuer Plasmaphysik, IPP-EURATOM Association, Boltzmann str. 2, D-85748 Garching (Germany); Pitts, R.A. [Association EURATOM, CRPP-EPFL, 1015 Lausanne (Switzerland); Riccardo, V. [EURATOM-UKAEA Fusion Association, Culham Science Center, Abingdon, Oxon OX 14 3DB (United Kingdom); Silva, C. [Associacao EURATOM/IST, Instituto Superior Tecnico (Portugal)

    2005-03-01

    A comparative analysis of the spatial and temporal characteristics of transient energy loads (ELMs and disruptions) on plasma facing components (PFCs) in present tokamak devices and their extrapolation to next step devices is presented. Type I ELMs lead to the expulsion of energy by the plasma in helical structures with ballooning-like features and toroidal numbers in the range n = 10-15. The plasma energy is transported towards the divertor and the main chamber PFCs leading to significant transient energy loads at these two locations on small wetted area. The largest transient energy fluxes onto PFCs in tokamaks are measured during the thermal quench of disruptions. These fluxes do not exceed greatly those of large Type I ELMs, due to the much larger wetted area for energy flux during the thermal quench compared to Type I ELMs. The implications of these findings for the next step devices are discussed.

  5. Real-time mass measurement of dust particles deposited on vessel wall in a divertor simulator using quartz crystal microbalances

    Energy Technology Data Exchange (ETDEWEB)

    Tateishi, Mizuki [Faculty of Information Science and Electrical Engineering, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395 (Japan); Koga, Kazunori, E-mail: koga@ed.kyushu-u.ac.jp [Faculty of Information Science and Electrical Engineering, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395 (Japan); Katayama, Ryu; Yamashita, Daisuke [Faculty of Information Science and Electrical Engineering, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395 (Japan); Kamataki, Kunihiro [Faculty of Arts and Science, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395 (Japan); Seo, Hyunwoong [Faculty of Information Science and Electrical Engineering, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395 (Japan); Itagaki, Naho [Faculty of Information Science and Electrical Engineering, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395 (Japan); PRESTO JST, 5 Sanban-cho, Chiyoda-ku, Tokyo 102-0075 (Japan); Shiratani, Masaharu [Faculty of Information Science and Electrical Engineering, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395 (Japan); Ashikawa, Naoko; Masuzaki, Suguru; Nishimura, Kiyohiko; Sagara, Akio [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki-city, Gifu 509-5292 (Japan)

    2015-08-15

    We are developing a dust monitoring method using quartz crystal microbalances (QCMs) equipped with a dust eliminating filter. Here we report a dust eliminating ratio of the filter and first measurement results of the QCMs in a divertor simulator. The volume of spherical dust in unit area on the filter and QCM under the filter were 2.09 × 10{sup −9} and 1.22 × 10{sup −10} m{sup 3} m{sup −2}, respectively. Thus, the dust eliminating ratio of the filter is 94.2%. The QCM without the filter gives deposition rate due to radicals and dust particles, whereas the QCM with the filter gives deposition rate predominantly due to radicals. From the results, we deduce information of mass fraction of dust particles in deposits.

  6. Effects of background plasma characteristics on tungsten impurity transport in the SOL/divertor region using IMPGYRO code

    Energy Technology Data Exchange (ETDEWEB)

    Yamoto, S., E-mail: yamoto@ppl.appi.keio.ac.jp [Faculty of Science and Technology, Keio University, Yokohama 223-8522 (Japan); Homma, Y. [Faculty of Science and Technology, Keio University, Yokohama 223-8522 (Japan); Hoshino, K. [Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan); Sawada, Y. [Faculty of Science and Technology, Keio University, Yokohama 223-8522 (Japan); Bonnin, X. [LSPM – CNRS, Université Paris 13, Sorbonne Paris Cité, 99 Avenue Jean-Baptiste Clément, F-93430 Villetaneuse (France); Coster, D. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, D-85748 Garching (Germany); Schneider, R. [Ernst-Moritz-Arndt University Greifswald, Felix-Hausdorff Str. 6, D-17487 Greifswald (Germany); Hatayama, A. [Faculty of Science and Technology, Keio University, Yokohama 223-8522 (Japan)

    2015-08-15

    The difference of tungsten impurity transport characteristics between a high recycling regime and a partially detached regime has been studied with the IMPGYRO code. Background plasma profiles from a JT-60U model geometry, computed from SOLPS, have been used. To obtain such characteristic regimes, we have changed electron (Q{sub e}) and ion (Q{sub i}) input powers at the core boundary. In the high-recycling regime, the tungsten impurities are transported toward the upstream of the SOL. On the other hand, in the partially detached regime, most tungsten impurities are localized near the inner and outer divertors. These features are mainly related to the background plasma temperature and ion flow.

  7. Real-time mass measurement of dust particles deposited on vessel wall in a divertor simulator using quartz crystal microbalances

    International Nuclear Information System (INIS)

    We are developing a dust monitoring method using quartz crystal microbalances (QCMs) equipped with a dust eliminating filter. Here we report a dust eliminating ratio of the filter and first measurement results of the QCMs in a divertor simulator. The volume of spherical dust in unit area on the filter and QCM under the filter were 2.09 × 10−9 and 1.22 × 10−10 m3 m−2, respectively. Thus, the dust eliminating ratio of the filter is 94.2%. The QCM without the filter gives deposition rate due to radicals and dust particles, whereas the QCM with the filter gives deposition rate predominantly due to radicals. From the results, we deduce information of mass fraction of dust particles in deposits

  8. Development of a mirror-based endoscope for divertor spectroscopy on JET with the new ITER-like wall (invited).

    Science.gov (United States)

    Huber, A; Brezinsek, S; Mertens, Ph; Schweer, B; Sergienko, G; Terra, A; Arnoux, G; Balshaw, N; Clever, M; Edlingdon, T; Egner, S; Farthing, J; Hartl, M; Horton, L; Kampf, D; Klammer, J; Lambertz, H T; Matthews, G F; Morlock, C; Murari, A; Reindl, M; Riccardo, V; Samm, U; Sanders, S; Stamp, M; Williams, J; Zastrow, K D; Zauner, C

    2012-10-01

    A new endoscope with optimised divertor view has been developed in order to survey and monitor the emission of specific impurities such as tungsten and the remaining carbon as well as beryllium in the tungsten divertor of JET after the implementation of the ITER-like wall in 2011. The endoscope is a prototype for testing an ITER relevant design concept based on reflective optics only. It may be subject to high neutron fluxes as expected in ITER. The operating wavelength range, from 390 nm to 2500 nm, allows the measurements of the emission of all expected impurities (W I, Be II, C I, C II, C III) with high optical transmittance (≥ 30% in the designed wavelength range) as well as high spatial resolution that is ≤ 2 mm at the object plane and ≤ 3 mm for the full depth of field (± 0.7 m). The new optical design includes options for in situ calibration of the endoscope transmittance during the experimental campaign, which allows the continuous tracing of possible transmittance degradation with time due to impurity deposition and erosion by fast neutral particles. In parallel to the new optical design, a new type of possibly ITER relevant shutter system based on pneumatic techniques has been developed and integrated into the endoscope head. The endoscope is equipped with four digital CCD cameras, each combined with two filter wheels for narrow band interference and neutral density filters. Additionally, two protection cameras in the λ > 0.95 μm range have been integrated in the optical design for the real time wall protection during the plasma operation of JET.

  9. Liquid Lithium Divertor and Scrape-Off-Layer Interactions on the National Spherical Torus Experiment: 2010 ? 2013 Progress Report

    Energy Technology Data Exchange (ETDEWEB)

    None

    2013-08-27

    The implementation of the liquid Lithium Divertor (LLD) in NSTX presented a unique opportunity in plasma-material interactions studies. A high density Langmuir Probe (HDLP) array utilizing a dense pack of triple Langmuir probes was built at PPPL and the electronics designed and built by UIUC. It was shown that the HDLP array could be used to characterize the modification of the EEDF during lithium experiments on NSTX as well as characterize the transient particle loads during lithium experiments as a means to study ELMs. With NSTX being upgraded and a new divertor being installed, the HDLP array will not be used in NSTX-U. However UIUC is currently helping to develop two new systems for depositing lithium into NSTX-U, a Liquid Lithium Pellet Dripper (LLPD) for use with the granular injector for ELM mitigation and control studies as well as an Upward-Facing Lithium Evaporator (U-LITER) based on a flash evaporation system using an electron beam. Currently UIUC has Daniel Andruczyk Stationed at PPPL and is developing these systems as well as being involved in preparing the Materials Analysis Particle Probe (MAPP) for use in LTX and NSTX-U. To date the MAPP preparations have been completed. New sample holders were designed by UIUC?s Research Engineer at PPPL and manufactured at PPPL and installed. MAPP is currently being used on LTX to do calibration and initial studies. The LLPD has demonstrated that it can produce pellets. There is still some adjustments needed to control the frequency and particle size. Equipment for the U-LITER has arrived and initial test are being made of the electron beam and design of the U-LITER in progress. It is expected to have these ready for the first run campaign of NSTX-U.

  10. Heat pulse propagation studies around magnetic islands induced by the Dynamic Ergodic Divertor in TEXTOR

    NARCIS (Netherlands)

    Spakman, G. W.; Hogeweij, G. M. D.; Jaspers, R. J. E.; Schüller, F. C.; Westerhof, E.; Boom, J. E.; Classen, I.G.J.; Delabie, E.; Domier, C.; Donne, A. J. H.; Kantor, M. Y.; Kramer-Flecken, A.; Liang, Y.; N C Luhmann Jr.,; Park, H. K.; van de Pol, M.J.; Schmitz, O.; Oosterbeek, J. W.

    2008-01-01

    Since the efficiency of the tearing mode suppression by heating depends on the electron heat diffusivity it is important to know if the electron heat transport coefficients inside the island are reduced compared with the ambient plasma. With that aim, modulated ECRH has been employed for heat pulse

  11. Test of divertor materials under simulated ITER plasma disruption conditions using the hot plasma stream of the 2MK-200 facility

    Energy Technology Data Exchange (ETDEWEB)

    Arkhipov, N.I.; Bakhtin, V.; Konkashbaev, I. [Troitsk Inst. for Innovation and Fusion Research (Russian Federation)] [and others

    1994-12-31

    The high divertor heat load during 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 and material erosion are experimentally investigated at the 2MK-200 facility under conditions simulating the thermal quench phase of ITER tokamak plasma disruptions. Material samples from graphite, tungsten, boron nitrite and quartz were exposed to deuterium plasma streams with the following parameters: density < 10{sup 16}cm{sup {minus}3}, temperature T{sub e}+T{sub i} < 0.8 keV, plasma beta 0.25, plasma flow width 2 cm, power density 10 MW/cm{sup 2} and time duration of the pulse 20 {mu}s.

  12. Supplement to 'ASDEX Upgrade, definition of a tokamak experiment with a reactor-compatible poloidal divertor' (IPP-report 1/197, March 1982)

    International Nuclear Information System (INIS)

    Since March 1982 the better understanding of the divertor physics, both by theory and experiments, and the development of the ASDEX Upgrade concept have considerably improved and simplified the ASDEX Upgrade design. Single null poloidal divertor configurations were calculated, which can well compete with elongated limiter configurations in reduced poloidal field effort. The role of recycling and its limitation set by the available energy flux, observed experimentally and explained by a plasma boundary flow model, led to a refined formulation of the line density requirements. Finally, a discussion of the attainable temperature and densities allowed clearly to distinguish between ASDEX and ASDEX Upgrade and pointed out the dominant role of the plasma current. The ASDEX Upgrade basic data are summarized as presented to the EURATOM advisory board. (orig.)

  13. Theoretical and experimental substantiation of the fusion reactor divertor and first wall shield concept by means of lithium capillary porous systems

    International Nuclear Information System (INIS)

    Basic properties of the lithium capillary porous systems (LCPS) for protection of divertor and the first wall shield TNR are described. The LCPS behaviour under conditions of impact of high stationary and pulse heat loads are studied. Lithium ions and neutrons migration in the tokamak reactor with liquid lithium is studied numerically and lithium flow in the crossed magnetic field experimentally. The LCPS behaviour is studied also experimentally under the conditions of the tokamak plasma impact

  14. Task II: ECRH and transport modeling in tandem mirrors and divertor physics. Annual progress report on fusion plasma theory, January 1, 1983-December 31, 1983

    International Nuclear Information System (INIS)

    The research performed under Task II of this contract has focused on (1) the coupling of an ECRH ray tracing and absorption code to a tandem mirror transport code in order to self-consistently model the temporal and spatial evolution of the plasma, and (2) the further development of a semi-analytical kinetic model for plasma flow in divertors and pumped limiters. Work on these topics is briefly summarized in this progress report

  15. Simulation experiment of interaction of plasma facing materials and transient heat loads in ITER divertor by use of magnetized coaxial plasma gun

    Science.gov (United States)

    Nakatsuka, M.; Ando, K.; Higashi, T.; Kikuchi, Y.; Fukumoto, N.; Nagata, M.

    2009-11-01

    Interaction of plasma facing materials and transient head loads such as type I ELMs is one of the critical issues in ITER divertor. The heat load to the ITER divertor during type I ELMs is estimated to be 0.5-3 MJ/m^2 with a pulse length of 0.1-0.5 ms. We have developed a magnetized coaxial plasma gun (MCPG) for the simulation experiment of transient heat load during type I ELMs in ITER divertor. The MCPG has inner and outer electrodes made of stainless steel 304. In addition, the inner electrode is covered with molybdenum so as to suppress the release of impurities from the electrode during the discharge. The diameters of inner and outer electrodes are 0.06 m and 0.14 m, respectively. The power supply for the MCPG is a capacitor bank (7 kV, 1 mF, 25 kJ). The plasma velocity estimated by the time of flight measurement of the magnetic fields was about 50 km/s, corresponding to the ion energy of 15 eV (H) or 30 eV (D). The absorbed energy density of the plasma stream was measured a calorimeter made of graphite. It was found that the absorbed energy density was 0.9 MJ/m^2 with a pulse width of 0.5 ms at the distance of 100 mm from the inner electrode. In the conference, experimental results of plasma exposure on the plasma facing materials in ITER divertor will be shown.

  16. Scrape-off layer ion temperature measurements at the divertor target during type III and type I ELMs in MAST measured by RFEA

    Science.gov (United States)

    Elmore, S.; Allan, S. Y.; Fishpool, G.; Kirk, A.; Thornton, A. J.; Walkden, N. R.; Harrison, J. R.; the MAST Team

    2016-06-01

    In future nuclear fusion reactors high heat load events, such as edge-localised modes (ELMs), can potentially damage divertor materials and release impurities into the main plasma, limiting plasma performance. The most difficult to handle are type I ELMs since they carry the largest fraction of energy from the plasma and therefore deposit the largest heat flux at the target and on first wall materials. Knowing the temperature of the ions released from ELM events is important since it determines the potential sputtering they would cause from plasma facing materials. To make measurements of T i by retarding field energy analyser (RFEA) during type I ELMs a new operational technique has been used to allow faster measurements to be made; this is called the fast swept technique (FST). The FST method allows measurements to be made within the time of the ELM event which has previously not been possible with T i measurements. This new technique has been validated by comparing it with a slower average measurement previously used to make ion temperature measurements of ELMs. Presented here are the first T i measurements during Type I ELMs made at a tokamak divertor. Temperatures as high as 20 eV are measured more than 15 cm from the peak heat flux of an ELM, in a region where no inter-ELM current is measured by the RFEA; showing that ELM events cause hot ions to reach the divertor target far into the scrape off layer. Fast camera imaging has been used to investigate the type of ELM filaments that have been measured by the divertor RFEA. It is postulated that most of the ion temperatures measured in type I ELMs are from secondary ELM filaments which have not been previously identified in MAST plasmas.

  17. Characterization of deuterium retention and co-deposition of fuel with lithium on the divertor tile of EAST using laser induced breakdown spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Li, Cong, E-mail: cli@mail.dlut.edu.cn [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams, Chinese Ministry of Education, School of Physics and Optical Electronic Technology, Dalian University of Technology, Dalian 116024 (China); Zhao, Dongye [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams, Chinese Ministry of Education, School of Physics and Optical Electronic Technology, Dalian University of Technology, Dalian 116024 (China); Hu, Zhenhua [Institute of Plasma Physics, Chinese Academy of Sciences, PO Box 1126, Hefei 230031 (China); Wu, Xingwei [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams, Chinese Ministry of Education, School of Physics and Optical Electronic Technology, Dalian University of Technology, Dalian 116024 (China); Luo, Guang-Nan; Hu, Jiansheng [Institute of Plasma Physics, Chinese Academy of Sciences, PO Box 1126, Hefei 230031 (China); Ding, Hongbin, E-mail: hding@dlut.edu.cn [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams, Chinese Ministry of Education, School of Physics and Optical Electronic Technology, Dalian University of Technology, Dalian 116024 (China)

    2015-08-15

    A laser induced breakdown spectroscopy (LIBS) system has been developed to measure and monitor the composition evolution on plasma facing materials (PFMs) of Experimental Advanced Superconducting Tokamak (EAST). As a necessity and important proof of principle experiment, LIBS analysis has been performed for lithium–deuterium co-deposition layer diagnosis of EAST divertor tiles in lab experiments. The distribution of deuterium retention has been obtained from the depth of 0.5–4 μm in the divertor tiles. The deuterium/hydrogen concentration ratio was estimated as 0.17 ± 0.02 in lithium–deuterium co-deposition layer. Moreover, the depth profile behaviors of lithium and deuterium indicate that the deuterium retention in divertor tile came from lithium–deuterium co-deposition processes during deuterium discharge in EAST. This work would improve the understanding of deuterium retention and lithium–deuterium co-deposition mechanism and give a guidance to optimize the LIBS system which will be a unique and useful diagnostic approach in EAST 2014-campaign.

  18. Control of 3D edge radiation structure with resonant magnetic perturbation fields applied to the stochastic layer and stabilization of radiative divertor plasma in LHD

    International Nuclear Information System (INIS)

    It is found that resonant magnetic perturbation (RMP) fields have a stabilizing effect on the radiating edge plasma, realizing stable sustainment of radiative divertor (RD) operation in the Large Helical Device (LHD). Without RMP, thermal instability leads to radiative collapse. Divertor power load is reduced by a factor of 3 ∼ 10 during the RMP assisted RD phase, while maintaining relatively good core plasma confinement with confinement enhancement factor τEexp / fren τEISS04 ∼ 0.96. It has also been demonstrated that the RMP field itself can initiate transition to RD operation by increasing perturbation strength. The results show a possibility of a new control knob for divertor power load in a 3D magnetic field configuration. It is also found that after the transition to RD, the energy confinement enhancement factor based on ISS04 scaling increases by a factor of 1.4 compared to the attached phase. The operation range of the RMP assisted RD is identified in terms of RMP strength and radial location of resonance layer of the RMP. (author)

  19. Characterization of deuterium retention and co-deposition of fuel with lithium on the divertor tile of EAST using laser induced breakdown spectroscopy

    International Nuclear Information System (INIS)

    A laser induced breakdown spectroscopy (LIBS) system has been developed to measure and monitor the composition evolution on plasma facing materials (PFMs) of Experimental Advanced Superconducting Tokamak (EAST). As a necessity and important proof of principle experiment, LIBS analysis has been performed for lithium–deuterium co-deposition layer diagnosis of EAST divertor tiles in lab experiments. The distribution of deuterium retention has been obtained from the depth of 0.5–4 μm in the divertor tiles. The deuterium/hydrogen concentration ratio was estimated as 0.17 ± 0.02 in lithium–deuterium co-deposition layer. Moreover, the depth profile behaviors of lithium and deuterium indicate that the deuterium retention in divertor tile came from lithium–deuterium co-deposition processes during deuterium discharge in EAST. This work would improve the understanding of deuterium retention and lithium–deuterium co-deposition mechanism and give a guidance to optimize the LIBS system which will be a unique and useful diagnostic approach in EAST 2014-campaign

  20. Experimental Investigation MHD Instabilities of Free Surface Jet and Film flow for Liquid Metal Divertor/Limiter Option

    International Nuclear Information System (INIS)

    The magneto-hydrodynamics (MHD) stability of free surface jet and film flow is one of the key remaining issues for liquid metal free surface divertor/limiter system. Recently, several laboratories have observed some different MHD instabilities phenomena of free surface jet and film under non-uniform or a gradient transverse magnetic field. In this paper, the results are presented of the simple modeling of jet flow MHD stability and experimental measured. The modeling can give out the changes of velocity, cross section and three dimensional skeleton map of jet in its path during the free surface jet flowing in a gradient transverse magnetic field. The experiments were performed in three cases. Three insulator nozzles was used in First case (nozzle diameter of 6 mm), One insulator nozzle (nozzle diameter of 6 mm) was used in second case, One electric conductivity nozzle (nozzle diameter of 12 mm) was used in third case. The gradient transverse magnetic field is 0.2 ∼ 1.95 Tesla in 30 cm path. The velocity of jet at the end of nozzle was 2.9 m/s, 3.24 m/s and 4.10 m/s, respectively. Furthermore, for third case, a 55 mm wide of film flow was observed. Under these experimental conditions, the experimental results indicated that transverse gradient magnetic field strongly shorten the jet flow range, and the shape of cross section of jet flow deformed from round to elliptical, and finally bowed down shape in jet flow downstream. The distinct difference MHD effect is not observed for different jet nozzles and in one-nozzle or multi-nozzles cases. Two type different MHD phenomena was observed for film flow, one is in retardarce flow and another is in rivulets. The film rivulet phenomena is firstly observed in present experiment. The simple modeling expected values are good agreement with jet flow experimental results. Combining jet flow and film flow (tentative call jet-film flow) maybe is a good chosen for liquid metal divertor/limiter system. (author)

  1. Measurement of LHCD edge power deposition through modulation techniques on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Faust, I. C.; Brunner, D.; LaBombard, B.; Parker, R. R.; Baek, S. G.; Chilenksi, M. A.; Hubbard, A.; Hughes, J. W.; Terry, J. L.; Shiraiwa, S.; Walk, J. R.; Wallace, G. M.; Whyte, D. G. [MIT Plasma Science and Fusion Center, Cambridge, MA USA (United States); Edlund, E. [Princeton Plasma Physics Laboratory, Princeton, NJ USA (United States)

    2015-12-10

    The efficiency of LHCD on Alcator C-Mod drops exponentially with line average density. At reactor relevant densities (> 1 · 1020 [m{sup −3}]) no measurable current is driven. While a number of causes have been suggested, no specific mechanism has been shown to be responsible for the loss of current drive at high density. Fast modulation of the LH power was used to isolate and quantify the LHCD deposition within the plasma. Measurements from these plasmas provide unique evidence for determining a root cause. Modulation of LH power in steady plasmas exhibited no correlated change in the core temperature. A correlated, prompt response in the edge suggests that the loss in efficiency is related to a edge absorption mechanism. This follows previous results which found the generation of n{sub ||}-independent SOL currents. Multiple Langmuir probe array measurements of the conducted heat conclude that the lost power is deposited near the last closed flux surface. The heat flux induced by LH waves onto the outer divertor is calculated. Changes in the neutral pressure, ionization and hard X-ray emission at high density highlight the importance of the active divertor in the loss of efficiency. Results of this study implicate a mechanism which may occur over multiple passes, leading to power absorption near the LCFS.

  2. Conceptual design of a He-cooled divertor with integrated flow and heat transfer Promoters (PPCS subtask TW3-TRP-001-D2). Pt. 2. Detailed version

    Energy Technology Data Exchange (ETDEWEB)

    Norajitra, P. (ed.); Kruessmann, R. (comp.)

    2004-04-01

    This report represents a summary of our knowledge after little more than one year of development of a helium-cooled divertor. The design goal is to reach at least 10 MW/m{sup 2} at a reasonable pumping power for a fusion power plant operating under DEMO conditions. In the first part, design requirements for the divertor are given and the current design using low-activation materials is described. In the second part, materials choice and promising tungsten alloy materials are pointed out. In view of the operation temperature window defined, materials choice for the divertor components is limited, i.e. tungsten for the thermal shield in the form of small tiles, W-1%La{sub 2}O{sub 3} for the thimble, and high-temperature ODS for the back bone structure. To broaden the operating temperature window of the divertor for obtaining a larger safety margin in the design, further development of tungsten alloys as thimble material is required. Promising methods (EDM, ECM and PIM) are identified for the fabrication of pin and slot arrays from tungsten, which need to be further developed. In the third part, computational fluid dynamics (CFD) analyses and thermomechanical finite element (FE) simulation calculations are covered. Comparisons of the pressure loss calculated by the CFD programs with first results of the pressure loss measurement performed at EFREMOV are made, the results are discussed. FE simulations revealed opportunities for the improvement of the design. The last part deals with the planning of experimental devices to confirm the theoretical findings. To validate the CFD programs, helium experiments are planned to be performed in the helium blanket test loop HEBLO at FZK/IMF III in the middle of 2004 using a single finger test mock-up of 10:1 in scale. For the high-heat-flux tests, a large helium loop is planned to be constructed at the EFREMOV Institute in St. Petersburg, Russia. Planning and specification of the experiment programmes are under way. The overall

  3. High-flux deuterium plasma exposure tests of actively-cooled divertor plate units in PISCES-B

    International Nuclear Information System (INIS)

    An actively-cooled divertor plate mock-up with three kinds of carbon-based armor tiles (IG430U, MFC-1, and CX2002U) designed and fabricated by JAERI was bombarded with steady-state and high-flux deuterium plasmas produced in UCLA PISCES-B. The plasma densities, electron temperatures, and ion fluxes were measured from 1 to 3x1019 m-3, from 4 to 12 eV, and from 1.2 to 2.1x1023 ions/m2s, respectively. The total ion fluence was of the order of 1026 ions/m2. Interesting surface morphologies have been observed for the plasma-bombarded surfaces, having relatively large agglomerated carbon particles with diameters up to 100 micrometer. The plasma heat flux was measured with a calorimeter embedded in a graphite (IG430U) to range from 1.1 to 4.4 MW/m2, which is in good agreement with the calculated value with a simple sheath theory. (author)

  4. Bulk tungsten in the JET divertor: Potential influence of the exhaustion of ductility and grain growth on the lifetime

    Science.gov (United States)

    Mertens, Ph.; Thompson, V.; Matthews, G. F.; Nicolai, D.; Pintsuk, G.; Riccardo, V.; Devaux, S.; Sieglin, B.; JET-EFDA contributors

    2013-07-01

    The divertor of the ITER-like Wall in JET currently includes a solid tungsten row for the outer strike point. The use of plasma-facing tungsten in fusion devices is limited by its brittleness in the low temperature domain (arbitrarily ˜TW 1200 °C). In the absence of active cooling, an extreme case of thermal cycling is represented by the situation in JET: the plasma-facing surface of the bulk tungsten tile experiences cyclic excursions from 200 °C to about 2000 °C. Thermal fatigue for impact factors of 11-24 MW m-2 s0.5 is investigated with a Manson-Coffin model; tungsten properties come from production samples. Recrystallization is studied in metallographic cuts of tungsten lamellae identical to those installed in the torus which were exposed in the MARION facility to JET relevant heat fluxes for >300 pulses (Pdep ⩽ 9 MW/m2, angle of attack 6°). The calculations suggest that the number of high temperature cycles should be limited with appropriate budgeting, especially if the grain growth degrades material properties. Values for JET range from 150 to thousands of pulses depending on the temperatures reached.

  5. Modelling of spatial structure of divertor footprints caused by edge-localized modes mitigated by magnetic perturbations

    CERN Document Server

    Cahyna, Pavel; Huijsmans, Guido T A; Orain, Francois; Morales, Jorge; Kirk, Andrew; Thornton, Andrew J; Pamela, Stanislas; Panek, Radomir; Hoelzl, Matthias

    2016-01-01

    Resonant magnetic perturbations (RMPs) can mitigate the edge-localized modes (ELMs), i.e. cause a change of the ELM character towards smaller energy loss and higher frequency. During mitigation a change of the spatial structure of ELM loads on divertor was observed on DIII-D and MAST: the power is deposited predominantly in the footprint structures formed by the magnetic perturbation. In the present contribution we develop a theory explaining this effect, based on the idea that part of the ELM loss is caused by parallel transport in the homoclinic tangle formed by the magnetic perturbation of the ELM. The modified tangle resulting from the combination of the ELM perturbation and the applied RMP has the expected property of bringing open field lines in the same areas as the tangle from the RMP alone. We show that this explanation is consistent with features of the mitigated ELMs on MAST. We in addition validated our theory by an analysis of simulations of mitigated ELMs using the code JOREK. We produced detail...

  6. Modeling The Effect of Drifts on the Edge, Scrape-Off Layer, and Divertor Plasma in DIII-D

    International Nuclear Information System (INIS)

    Simulations of plasmas with a DIII-D shape indicate plasma drifts are important at power levels near the L- to H-mode plasma transition. In addition to enhancing plasma flows in the divertor region, drifts are found to play a key role in establishing highly sheared radial electric fields in the edge of the confined plasma, for the physics of the high confinement operating mode (H-mode). Measurements of the plasma structure in the vicinity of the X-point of DIII-D indicate the importance of drifts on plasma flow between the scrape-off layer (SOL) and closed field lines. The large electric fields provide large flows around the X-point, and these are conjectured to play a role in the transition from L- to H-mode confinement. These results indicate the relevance of modeling the edge and SOL plasmas of present tokamak devices using models which include E x B, (del)B, and pressure gradient drifts. The results of simulation of specific DIII-D discharges is reported in this paper. They start with discussion of the simulation of an Ohmic discharge in Section 2, including a study of the effect of varying several operational parameters. Simulation of a higher triangularity L-mode discharge is discussed in Section, and a summary is given in Section 4

  7. Irreducible Specht modules are signed Young modules

    OpenAIRE

    Hemmer, David J.

    2005-01-01

    Recently Donkin defined signed Young modules as a simultaneous generalization of Young and twisted Young modules for the symmetric group. We show that in odd characteristic, if a Specht module $S^\\lambda$ is irreducible, then $S^\\lambda$ is a signed Young module. Thus the set of irreducible Specht modules coincides with the set of irreducible signed Young modules. This provides evidence for our conjecture that the signed Young modules are precisely the class of indecomposable self-dual module...

  8. *-Modules, co-*-modules and cotilting modules over Noetherian rings

    Institute of Scientific and Technical Information of China (English)

    汪明义; 许永华

    1996-01-01

    Let R be a Noetherian ring. The projectivity and injectivity of modules over R are discussed. The concept of modules is introduced and the descriptions for co-*-modules over R are given. At last, cotilting modules over R are characterized by means of co-*-modules.

  9. Study of Scrape-Off-Layer Width in Ohmic and Lower Hybrid Wave Heated Double-Null Divertor Plasma in EAST%Study of Scrape-Off-Layer Width in Ohmic and Lower Hybrid Wave Heated Double-Null Divertor Plasma in EAST

    Institute of Scientific and Technical Information of China (English)

    王亮; 刘鹏; 蒋敏; 熊豪; 万宝年; 高翔; 郭后扬; 胡立群; 吴振伟; 朱思铮; 罗广南; 徐国盛; 常加峰; 张炜; 颜宁; 丁斯晔; 刘少承; 明廷凤; 汪惠乾

    2011-01-01

    Edge profiles in Ohmic and lower hybrid (LH) wave heated discharges in EAST are presented. A comparison of the measured profiles is made with those from the theoretical prediction for the scrape-off-layer (SOL) width. The edge plasma parameters are diagnosed by a triple probe divertor diagnostic system and fast reciprocating probes at the outer mid-plane. The experimental results show that the SOL width of double-null (DN) divertor plasmas in EAST appears to exhibit a negative dependence on the power crossing the separatrix, which is consistent with the collisional SOL scalings of JET and Alcator C-Mod. This will provide useful information for extrapolation to the ITER SOL width scaling for power deposition.

  10. Signed Young Modules and Simple Specht Modules

    OpenAIRE

    Danz, Susanne; Lim, Kay Jin

    2015-01-01

    By a result of Hemmer, every simple Specht module of a finite symmetric group over a field of odd characteristic is a signed Young module. While Specht modules are parametrized by partitions, indecomposable signed Young modules are parametrized by certain pairs of partitions. The main result of this article establishes the signed Young module labels of simple Specht modules. Along the way we prove a number of results concerning indecomposable signed Young modules that are of independent inter...

  11. Optical emission measurements of H 2 and D 2 molecules in the divertor region of ASDEX Upgrade

    Science.gov (United States)

    Fantz, U.; Behringer, K.; Gafert, J.; Coster, D.; ASDEX Upgrade Team

    A spectroscopic method has been developed for measuring molecular influxes and particle densities in fusion edge plasmas, which is based on the H 2 and D 2 Fulcher emission bands around 600 nm wavelength. A first application to the ASDEX Upgrade divertor plasma is described. The influx of hydrogen molecules was determined from the population of the upper Fulcher state using the theoretical number of ionization and dissociation events per Fulcher photon ( Seff + Deff)/XB Ful, as calculated by a collisional-radiative model. These results were compared with expectations on the basis of the atomic hydrogen fluxes and a typical molecule/atom ratio. Measurements and calculations agree in their time dependence, but the experimental values are somewhat lower, which may be within the error margin or of more significance. The Fulcher radiation was also compared directly to B2-EIRENE predictions, resulting in a higher discrepancy. In addition, the vibrational population of the ground state molecules was determined from that of the excited state using a method based on Franck-Condon factors. It can be characterized by a Tvib between 3000 and 9000 K, inversely correlated with electron temperature. This variation is predicted by the collisional-radiative code and even allows an estimate of Te. Vibrational excitation increases ionization and dissociation rate coefficients, as clearly demonstrated by the code calculations. It is therefore very likely that the observed discrepancy in molecular intensity is mainly caused by the omission of vibrational excitation in the present version of B2-EIRENE. The described flux measurements are expected to be accurate above Te=5 eV, but are more difficult at lower temperatures due to the strong Te dependence of ( Seff + Deff)/XB Ful in that region.

  12. Fabrication of a 1200 kg Ingot of V-4Cr-4Ti for the DIII-D Radiative Divertor Program

    International Nuclear Information System (INIS)

    Vanadium chromium titanium alloys are attractive materials for fusion reactors because of their high temperature capability and their potential for low neutron active and rapid activation decay. A V-4Cr-4Ti alloy has been selected in the U.S. as the current leading candidate vanadium alloy for future use in fusion reactor structural applications. General Atomics (GA), in conjunction with the Department of Energy's (DOE) DIII-D Program, is carrying out a plan for the utilization of this vanadium alloy in the DIII-D tokamak. The plan will culminate in the fabrication, installation, and operation of a V-4Ti alloy structure in the DIII-D Radiative Divertor (RD) upgrade. The deployment of vanadium alloy will provide a meaningful step in the development and technology acceptance of this advanced material for future fusion power devices. Under a GA contract and material specification, an industrial scale 1200 kg heat (ingot) of a V-4Cr-4Ti alloy has been produced and converted into product forms by Wah Chang of Albany, Oregon (WCA). To assure the proper control of minor and trace impurities which affect the mechanical and activation behavior of this vanadium alloy, selected lots of raw vanadium base metal were processed by aluminothermic reduction of high purity vanadium oxide, and were then electron beam melted into two high purity vanadium ingots. The ingots were then consolidated with high purity Cr and Ti, and double vacuum-arc melted to obtain a 1200 kg V-4Cr-4Ti alloy ingot. Several billets were extruded from the ingot, and were then fabricated into plate, sheet, and rod at WCA. Tubing was subsequently processed from plate material. The chemistry and fabrication procedures for the product forms were specified on the basis of experience and knowledge gained from DOE Fusion Materials Program studies on previous laboratory scale heats and a large scale ingot (500 kg)

  13. Experimental Study of the Atomic and Molecular Processes Related to Plasma Detachment in Steady-State Divertor Simulator MAP-II

    Institute of Scientific and Technical Information of China (English)

    S.Kado; S.Kajita; Y.Iida; B.Xiao; T.Shikama; D.Yamasaki; T.Oishi; S.Tanaka

    2004-01-01

    Atomic and molecular processes relevant to the volumetric recombination phenomena were investigated in a linear divertor plasma simulator MAP-II. Volumetric recombination is induced in He plasma by puffing of He or H2. In the He puffing case, the reduction of the ion flux is dominated by the electron-ion recombination. In the H2 puffing case, however, it is dominated by the molecule-assisted recombination (MAR), which is characterized by the disappearance of the Helium Rydberg spectra and by the existence of the hydrogen negative ions. Current achievement and the future prospect are described.

  14. Application of optical emission spectroscopy for He I considering the spatial structure of radiation trapping in MAP-II divertor simulator

    International Nuclear Information System (INIS)

    The He I optical emission spectroscopy that considers the spatial structure of radiation trapping was proposed by us and was applied to a MAP-II divertor simulator. The spatial distribution of the optical escape factor was calculated from the n 1P (n≥3) state profiles measured by visible spectroscopy. The profile of 2 1P, which is immeasurable by visible spectroscopy, needs to be broader than that of the 3 1P state. The sensitivity of the 2 1P profile to the Te value estimated by He I spectroscopy is investigated.

  15. Deuterium Balmer/Stark spectroscopy and impurity profiles: first results from mirror-link divertor spectroscopy system on the JET ITER-like wall

    CERN Document Server

    Meigs, A G; Clever, M; Huber, A; Marsen, S; Nicholas, C; Stamp, M; Zastrow, K-D; Contributors, JET EFDA

    2013-01-01

    For the ITER-like wall, the JET mirror link divertor spectroscopy system was redesigned to fully cover the tungsten horizontal strike plate with faster time resolution and improved near-UV performance. Since the ITER-like wall project involves a change in JET from a carbon dominated machine to a beryllium and tungsten dominated machine with residual carbon, the aim of the system is to provide the recycling flux, equivalent, to the impinging deuterium ion flux, the impurity fluxes (C, Be, O) and tungsten sputtering fluxes and hence give information on the tungsten divertor source. In order to do this self-consistently, the system also needs to provide plasma characterization through the deuterium Balmer spectra measurements of electron density and temperature during high density. L-Mode results at the density limit from Stark broadening/line ratio analysis will be presented and compared to Langmuir probe profiles and 2D-tomography of low-n Balmer emission [1]. Comparison with other diagnostics will be vital fo...

  16. Photovoltaic module and module arrays

    Science.gov (United States)

    Botkin, Jonathan; Graves, Simon; Lenox, Carl J. S.; Culligan, Matthew; Danning, Matt

    2012-07-17

    A photovoltaic (PV) module including a PV device and a frame. The PV device has a PV laminate defining a perimeter and a major plane. The frame is assembled to and encases the laminate perimeter, and includes leading, trailing, and side frame members, and an arm that forms a support face opposite the laminate. The support face is adapted for placement against a horizontal installation surface, to support and orient the laminate in a non-parallel or tilted arrangement. Upon final assembly, the laminate and the frame combine to define a unitary structure. The frame can orient the laminate at an angle in the range of 3.degree.-7.degree. from horizontal, and can be entirely formed of a polymeric material. Optionally, the arm incorporates integral feature(s) that facilitate interconnection with corresponding features of a second, identically formed PV module.

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

    International Nuclear Information System (INIS)

    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/m2 to the erosion model already subjected to the normal condition loads for 10,000 s, the plasma shaping of the

  18. Simulations of Material Damage and High Energy Fluxes to ITER Divertor and First Wall during Transients and Runaway Electron Loads

    International Nuclear Information System (INIS)

    Full text: The anticipated regime of the tokamak ITER is the H-mode in which the repetitive outbreaks of the edge-localized mode (ELM) produce plasma fluxes which determine the erosion rate and the lifetime of PFCs. The disruptions also reduce the PFC lifetime, despite of mitigation measures such as the massive gas injection (MGI), in particular because of high heat fluxes by runaway electrons and the radiation flush. The lost plasma dumped mainly into the scrape-off layer (SOL) produces surface erosion by sputtering, melting, splashing, cracking and vaporization. The expected transient heat fluxes on the PFCs are: Type I ELM 0.5 - 4 MJ/m2 on the timescale 0.3 - 0.6 ms, thermal quench flux 2 - 13 MJ/m2 in 1 - 3 ms. Mitigated disruption radiative flux 0.1 - 2 MJ/m2 in 2 - 5 ms, and the runaway flux more than 10 MJ/m2 on the timescale 10 - 100 ms. In ITER the CFC and tungsten macrobrush armour as PFCs for the divertor and the dome, and beryllium macrobrushes for the first wall (FW) are foreseen. The fluid motion in a thin molten layer of W and Be during transients may produce melt splashing and thus dust emission by droplets. The expected erosion of ITER PFCs can be properly estimated by numerical simulations validated against erosion experiments at the plasma gun facilities QSPA-T, MK-200UG and QSPA-Kh50. The measured material erosion was used to validate the melt dynamics code MEMOS and the thermomechanic code PEGASUS that were then applied to model the erosion of ITER PFCs under the anticipated transient loads. The results of experiments carried out at QSPA-T allowed validation of numerical model for the melt splashing based on Kelvin-Helmholtz-and Rayleigh-Taylor instabilities. The crack formation at W surface was modeled using the code PEGASUS and validated against the experiments carried out at QSPA-Kh50. The models were applied for simulations of PFCs damage under expected ITER-like scenarios. Numerical simulations under radiation and runaway electron impact

  19. The contribution of radio-frequency rectification to field-aligned losses of high-harmonic fast wave power to the divertor in the National Spherical Torus eXperiment

    Energy Technology Data Exchange (ETDEWEB)

    Perkins, R. J., E-mail: rperkins@pppl.gov; Hosea, J. C.; Jaworski, M. A.; Diallo, A.; Bell, R. E.; Bertelli, N.; Gerhardt, S.; Kramer, G. J.; LeBlanc, B. P.; Phillips, C. K.; Podestà, M.; Roquemore, L.; Taylor, G.; Wilson, J. R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Ahn, J.-W.; Gray, T. K. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); McLean, A. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Sabbagh, S. [Department of Applied Physics and Applied Mathematics, Columbia University, New York, New York 10027 (United States)

    2015-04-15

    The National Spherical Torus eXperiment (NSTX) can exhibit a major loss of high-harmonic fast wave (HHFW) power along scrape-off layer (SOL) field lines passing in front of the antenna, resulting in bright and hot spirals on both the upper and lower divertor regions. One possible mechanism for this loss is RF sheaths forming at the divertors. Here, we demonstrate that swept-voltage Langmuir probe characteristics for probes under the spiral are shifted relative to those not under the spiral in a manner consistent with RF rectification. We estimate both the magnitude of the RF voltage across the sheath and the sheath heat flux transmission coefficient in the presence of the RF field. Although precise comparison between the computed heat flux and infrared (IR) thermography cannot yet be made, the computed heat deposition compares favorably with the projections from IR camera measurements. The RF sheath losses are significant and contribute substantially to the total SOL losses of HHFW power to the divertor for the cases studied. This work will guide future experimentation on NSTX-U, where a wide-angle IR camera and a dedicated set of coaxial Langmuir probes for measuring the RF sheath voltage directly will quantify the contribution of RF sheath rectification to the heat deposition from the SOL to the divertor.

  20. The contribution of radio-frequency rectification to field-aligned losses of high-harmonic fast wave power to the divertor in the National Spherical Torus eXperiment

    International Nuclear Information System (INIS)

    The National Spherical Torus eXperiment (NSTX) can exhibit a major loss of high-harmonic fast wave (HHFW) power along scrape-off layer (SOL) field lines passing in front of the antenna, resulting in bright and hot spirals on both the upper and lower divertor regions. One possible mechanism for this loss is RF sheaths forming at the divertors. Here, we demonstrate that swept-voltage Langmuir probe characteristics for probes under the spiral are shifted relative to those not under the spiral in a manner consistent with RF rectification. We estimate both the magnitude of the RF voltage across the sheath and the sheath heat flux transmission coefficient in the presence of the RF field. Although precise comparison between the computed heat flux and infrared (IR) thermography cannot yet be made, the computed heat deposition compares favorably with the projections from IR camera measurements. The RF sheath losses are significant and contribute substantially to the total SOL losses of HHFW power to the divertor for the cases studied. This work will guide future experimentation on NSTX-U, where a wide-angle IR camera and a dedicated set of coaxial Langmuir probes for measuring the RF sheath voltage directly will quantify the contribution of RF sheath rectification to the heat deposition from the SOL to the divertor

  1. Scrape-off layer ion temperature measurements at the divertor target during type III and type I ELMs in MAST measured by RFEA

    CERN Document Server

    Elmore, S; Fishpool, G; Kirk, A; Thornton, A J; Walkden, N R; Harrison, J R

    2016-01-01

    In future nuclear fusion reactors high heat load events, such as edge-localised modes (ELMs), can potentially damage divertor materials and release impurities into the main plasma, limiting plasma performance. The most difficult to handle are type I ELMs since they carry the largest fraction of energy from the plasma and therefore deposit the largest heat flux at the target and on first wall materials. Knowing the temperature of the ions released from ELM events is important since it determines the potential sputtering they would cause from plasma facing materials. To make measurements of Ti by retarding field energy analyser (RFEA) during type I ELMs a new operational technique has been used to allow faster measurements to be made; this is called the fast swept technique (FST).

  2. Micro- and nano-scale damage on the surface of W divertor component during exposure to high heat flux loads with He

    Science.gov (United States)

    Li, C.; Greuner, H.; Zhao, S. X.; Böswirth, B.; Luo, G. N.; Zhou, X.; Jia, Y. Z.; Liu, X.; Liu, W.

    2015-11-01

    Micro- and nano-scale surface damage on a W divertor component sample exposed to high heat flux loads generated with He atoms has been investigated through SEM, EBSD, AFM and FIB-SEM. The component sample was supplied by the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) and AT&M company, China, and the loading experiment was performed in the GLADIS facility at IPP Garching, Germany. Two typical damage structures were observed on the surface: the first one is characterized by obvious blisters and some grooves formed from ruptured blisters, and the other one is a kind of porous structure accompanying with at least ∼25 nm surface material loss. As the grain orientation is further away from , the damage morphology gradually changes from the former structure to the latter. The possible damage mechanism is discussed.

  3. Response to high heat fluxes and metallurgical examination of a brazed carbon-fiber-composite/refractory-metal divertor mock-up

    International Nuclear Information System (INIS)

    As a feasibility-study an actively cooled divertor mock-up has been subjected to high heat flux loading in electron beam simulation. The divertor design concept is based on a carbon-fiber-composite material (Aerolor 05) brazed onto a TZM/Mo41Re heat sink. The plasma facing carbon armor is divided in seven tiles to allow variable loading parameters - and repeated destructive tests. The mock-up has survived high heat flux loading up to about 12 MW/m2 surface heat flux in steady-state conditions. One armor tile showed no change in the thermal response even after 500 s at ∝14 MW/m2. To estimate the general thermal response of the mock-up design, numerical methods were applied. The predicted behavior was confirmed by the experimental results. The loading experiments were followed by a detailed metallurgical investigation of the loaded sample regions and the braze joints. The typical damages after high heat flux testing and cycling were failure (i.e. detachment) in the Zr brazed carbon/TZM joint, and failure in the CuPd bonded TZM/TZM joint due to an excess of the melting temperature of the brazes. The microstructural changes in the braze regions and the recrystallization behavior of the refractory alloys are discussed. Only in one case the loaded surface of the carbon armor shows considerable erosion, caused by a partial detachment along a braze joint and thus loss of the good thermal contact during the last applied loading shots. The thermal analyses and high heat flux performance of the Aerolor-05 armored mock-up are compared to the thermal response of a previously tested mock-up of corresponding geometry with armor tiles of isotropic graphite. (orig.)

  4. Module theory, extending modules and generalizations

    CERN Document Server

    Tercan, Adnan

    2016-01-01

    The main focus of this monograph is to offer a comprehensive presentation of known and new results on various generalizations of CS-modules and CS-rings. Extending (or CS) modules are generalizations of injective (and also semisimple or uniform) modules. While the theory of CS-modules is well documented in monographs and textbooks, results on generalized forms of the CS property as well as dual notions are far less present in the literature. With their work the authors provide a solid background to module theory, accessible to anyone familiar with basic abstract algebra. The focus of the book is on direct sums of CS-modules and classes of modules related to CS-modules, such as relative (injective) ejective modules, (quasi) continuous modules, and lifting modules. In particular, matrix CS-rings are studied and clear proofs of fundamental decomposition results on CS-modules over commutative domains are given, thus complementing existing monographs in this area. Open problems round out the work and establish the...

  5. Ballasted photovoltaic module and module arrays

    Science.gov (United States)

    Botkin, Jonathan; Graves, Simon; Danning, Matt

    2011-11-29

    A photovoltaic (PV) module assembly including a PV module and a ballast tray. The PV module includes a PV device and a frame. A PV laminate is assembled to the frame, and the frame includes an arm. The ballast tray is adapted for containing ballast and is removably associated with the PV module in a ballasting state where the tray is vertically under the PV laminate and vertically over the arm to impede overt displacement of the PV module. The PV module assembly can be installed to a flat commercial rooftop, with the PV module and the ballast tray both resting upon the rooftop. In some embodiments, the ballasting state includes corresponding surfaces of the arm and the tray being spaced from one another under normal (low or no wind) conditions, such that the frame is not continuously subjected to a weight of the tray.

  6. Simulation study on detachment operation of snowflake divertor for CFETR%中国聚变工程实验堆雪花偏滤器脱靶运行的SOLPS模拟

    Institute of Scientific and Technical Information of China (English)

    吴昊声; 毛世峰; 陈彬; 张传家; 罗正平; 郭勇; 彭学兵; 叶民友

    2015-01-01

    Background:In the conceptual design of China Fusion Engineering Test Reactor (CFETR), two additional poloidal coils, with respect to International Thermonuclear Experimental Reactor (ITER), are used to generate snowflake divertor configuration proposed recently for the purpose of exploring effective way for reducing heat loads onto divertor targets. Heat flux onto divertor targets was dramatically reduced in detached regime, while the performance of impurity screening would also be reduced due to the decrease of divertor temperature.Purpose: This study aims to simulate the detachment operation of snowflake divertor for CFETR.Methods:The detachment operational status was investigated by numerical simulation based on the edge plasma simulation software SOLPS (Scrape-off Layer Plasma Simulation). A D2 gas puffing in the main chamber was used to change plasma density. Results:When the gas puffing rate was sufficiently high, snowflake divertor of CFETR was completely detached, and the ion flux and heat loads onto the targets significantly decreased. However, the plasma temperature in the divertor region was too low and the impurities could easily pass through the X-point to core plasma, which implied a risk of radiation instability.Conclusion:Therefore, a proper operational status for the snowflake divertor in CFETR should be partial detachment.%在中国聚变工程实验堆(China Fusion Engineering Test Reactor, CFETR)的工程概念设计中,为探索有效降低偏滤器靶板热负荷的途径,相对于国际热核聚变实验堆(International Thermonuclear Experimental Reactor, ITER)特别增加了两个极向场线圈用于产生近年来新提出的雪花偏滤器位形。脱靶运行状态下,偏滤器靶板上的热负荷显著降低,但同时由于偏滤器温度的降低,杂质约束性能会变差,因此需要对CFETR雪花偏滤器的脱靶运行状态进行研究。基于边界等离子体物理模拟软件SOLPS (Scrape-off Layer Plasma

  7. 中国聚变工程实验堆雪花偏滤器脱靶运行的SOLPS模拟%Simulation study on detachment operation of snowflake divertor for CFETR

    Institute of Scientific and Technical Information of China (English)

    吴昊声; 毛世峰; 陈彬; 张传家; 罗正平; 郭勇; 彭学兵; 叶民友

    2015-01-01

    在中国聚变工程实验堆(China Fusion Engineering Test Reactor, CFETR)的工程概念设计中,为探索有效降低偏滤器靶板热负荷的途径,相对于国际热核聚变实验堆(International Thermonuclear Experimental Reactor, ITER)特别增加了两个极向场线圈用于产生近年来新提出的雪花偏滤器位形。脱靶运行状态下,偏滤器靶板上的热负荷显著降低,但同时由于偏滤器温度的降低,杂质约束性能会变差,因此需要对CFETR雪花偏滤器的脱靶运行状态进行研究。基于边界等离子体物理模拟软件SOLPS (Scrape-off Layer Plasma Simulation),通过数值模拟研究了CFETR雪花偏滤器的脱靶运行状态。模拟中通过主等离子体室内的D2充气改变等离子体密度。当充气速度足够高时,CFETR雪花偏滤器实现完全脱靶,靶板上的离子流和热负荷都显著降低。但此时偏滤器区域等离子体温度已经非常低,杂质将容易通过X点进入芯部,有产生辐射不稳定性的风险。因此,对于CFETR雪花偏滤器较为合适的工作状态应当是部分脱靶运行。%Background:In the conceptual design of China Fusion Engineering Test Reactor (CFETR), two additional poloidal coils, with respect to International Thermonuclear Experimental Reactor (ITER), are used to generate snowflake divertor configuration proposed recently for the purpose of exploring effective way for reducing heat loads onto divertor targets. Heat flux onto divertor targets was dramatically reduced in detached regime, while the performance of impurity screening would also be reduced due to the decrease of divertor temperature.Purpose: This study aims to simulate the detachment operation of snowflake divertor for CFETR.Methods:The detachment operational status was investigated by numerical simulation based on the edge plasma simulation software SOLPS (Scrape-off Layer Plasma Simulation). A D2 gas puffing in the main chamber was used to

  8. Intensity modulated proton therapy

    OpenAIRE

    Kooy, H. M.; Grassberger, C

    2015-01-01

    Intensity modulated proton therapy (IMPT) implies the electromagnetic spatial control of well-circumscribed “pencil beams” of protons of variable energy and intensity. Proton pencil beams take advantage of the charged-particle Bragg peak—the characteristic peak of dose at the end of range—combined with the modulation of pencil beam variables to create target-local modulations in dose that achieves the dose objectives. IMPT improves on X-ray intensity modulated beams (intensity modulated radio...

  9. MHD analysis and heat transfer characteristics of liquid metal thin film flows in quasi-coplanar magnetic field for Tokamak liquid metal divertor

    International Nuclear Information System (INIS)

    Numerical analysis of an open-channel liquid metal thin film with a quasi-coplanar strong applied magnetic field is carried out for a liquid metal divertor of tokamak device. The wall conductance ratio and the magnetic field inclinded angle appear to be the most important parameters to explain flow characteristics. As the flow rate increases, the velocity distribution with applied magnetic field is flat in the core region of flow and has jets at free surface of liquid metal film flow. In case of conductive walls, that effect is larger than insulated walls since open-channel, induced current circuits are constructed through walls, which causes a large magnetohydro-dynamic (MHD) drag in that region. In case with inclined magnetic field, as the flow rate increases, the film height increases and the flow experiences three regimes whether wall is conductive ro not. Regime 1 is dominant by the viscous force, regime 2 by the film height direction component of magnetic field (y component), and regime 3 by the channel width direction component of magnetic field (z component). Characteristics and limits of each regime are examined. Using calculated velocity distributions, heat transfer at the free surface is examined. In case of ordinary hydrodynamic flow, the heat removal characteristic is superior to the MHD case

  10. Fourier-spectral element approximation of the ion-electron Braginskii system with application to tokamak edge plasma in divertor configuration

    Science.gov (United States)

    Minjeaud, Sebastian; Pasquetti, Richard

    2016-09-01

    Due to the extreme conditions required to produce energy by nuclear fusion in tokamaks, simulating the plasma behavior is an important but challenging task. We focus on the edge part of the plasma, where fluid approaches are probably the best suited, and our approach relies on the Braginskii ion-electron model. Assuming that the electric field is electrostatic, this yields a set of 10 strongly coupled and non-linear conservation equations that exhibit multiscale and anisotropy features. The computational domain is a torus of complex geometrical section, that corresponds to the divertor configuration, i.e. with an "X-point" in the magnetic surfaces. To capture the complex physics that is involved, high order methods are used: The time-discretization is based on a Strang splitting, that combines implicit and explicit high order Runge-Kutta schemes, and the space discretization makes use of the spectral element method in the poloidal plane together with Fourier expansions in the toroidal direction. The paper thoroughly describes the algorithms that have been developed, provides some numerical validations of the key algorithms and exhibits the results of preliminary numerical experiments. In particular, we point out that the highest frequency of the system is intermediate between the ion and electron cyclotron frequencies.

  11. Koszul differential graded modules

    Institute of Scientific and Technical Information of China (English)

    HE JiWei; WU QuanShui

    2009-01-01

    The concept of Koszulity for differential graded (DG, for short) modules is introduced. It is shown that any bounded below DG module with bounded Ext-group to the trivial module over a Koszul DG algebra has a Koszul DG submodule (up to a shift and truncation), moreover such a DG module can be approximated by Koszul DG modules (Theorem 3.6). Let A be a Koszul DG algebra, and Dc (A) be the full triangulated subcategory of the derived category of DG A-modules generated by the object AA. If the trivial DG module kA lies in Dc(A), then the heart of the standard t-structure on Dc(A) is anti-equivalent to the category of finitely generated modules over some finite dimensional algebra. As a corollary, Dc(A) is equivalent to the bounded derived category of its heart as triangulated categories.

  12. Koszul differential graded modules

    Institute of Scientific and Technical Information of China (English)

    2009-01-01

    The concept of Koszulity for differential graded (DG, for short) modules is introduced. It is shown that any bounded below DG module with bounded Ext-group to the trivial module over a Koszul DG algebra has a Koszul DG submodule (up to a shift and truncation), moreover such a DG module can be approximated by Koszul DG modules (Theorem 3.6). Let A be a Koszul DG algebra, and Dc(A) be the full triangulated subcategory of the derived category of DG A-modules generated by the object AA. If the trivial DG module kA lies in Dc(A), then the heart of the standard t-structure on Dc(A) is anti-equivalent to the category of finitely generated modules over some finite dimensional algebra. As a corollary, Dc(A) is equivalent to the bounded derived category of its heart as triangulated categories.

  13. Modulating lignin in plants

    Science.gov (United States)

    Apuya, Nestor; Bobzin, Steven Craig; Okamuro, Jack; Zhang, Ke

    2013-01-29

    Materials and methods for modulating (e.g., increasing or decreasing) lignin content in plants are disclosed. For example, nucleic acids encoding lignin-modulating polypeptides are disclosed as well as methods for using such nucleic acids to generate transgenic plants having a modulated lignin content.

  14. CS-Rickart modules

    OpenAIRE

    Abyzov, A. N.; Nhan, T. H. N.

    2014-01-01

    In this paper, we introduce and study the concept of CS-Rickart modules, that is a module analogue of the concept of ACS rings. A ring $R$ is called a right weakly semihereditary ring if every its finitly generated right ideal is of the form $P\\oplus S,$ where $P_R$ is a projective module and $S_R$ is a singular module. We describe the ring $R$ over which $\\mathrm{Mat}_n (R)$ is a right ACS ring for any $n \\in \\mathbb {N}$. We show that every finitely generated projective right $R$-module wil...

  15. Reduced Multiplication Modules

    Indian Academy of Sciences (India)

    Karim Samei

    2011-05-01

    An -module is called a multiplication module if for each submodule of , = for some ideal of . As defined for a commutative ring , an -module is said to be reduced if the intersection of prime submodules of is zero. The prime spectrum and minimal prime submodules of the reduced module are studied. Essential submodules of are characterized via a topological property. It is shown that the Goldie dimension of is equal to the Souslin number of Spec (). Also a finitely generated module is a Baer module if and only if Spec () is an extremally disconnected space; if and only if it is a -module. It is proved that a prime submodule is minimal in if and only if for each $x\\in N,\\mathrm{Ann}(x)\

  16. Solar cell module. Taiyo denchi module

    Energy Technology Data Exchange (ETDEWEB)

    Nakano, Akihiko.

    1990-01-24

    This invention concerns a module frame of solar cell and a solar cell module using this frame. In particular, it concerns a frame and a module useful for the CdS/CdTe or CdS/CuInSe {sub 2} based cell. In the existing solar cell module, sealant is packed in between the edges of a glass substrate, a resin layer and a back protective thin film, etc. and a grooved frame of U-shaped section. For the sealant, silicon based resin and butyl rubber based resin are used many times, but either resin has defects such as their overflow from the module structure. In order to solve these defects, this invention proposes to provide stair-shaped protrusions along the four sides of the bottom of the box frame (herein after called the lower frame) of the module and at the same time, provide a groove for pooling the sealant at the portion where such protrusion meets the side wall, furthermore to provide depressions for pooling the sealant at the upper edge inside the side wall of the lower frame or to punch holes at the corners of the bottom of the lower frame. 9 figs.

  17. Directed network modules

    CERN Document Server

    Pálla, G; Farkas, I J; Pollner, P; Vicsek, T; Derenyi, Imre; Farkas, Illes J.; Palla, Gergely; Pollner, Peter; Vicsek, Tamas

    2007-01-01

    A search technique locating network modules, i.e., internally densely connected groups of nodes in directed networks is introduced by extending the Clique Percolation Method originally proposed for undirected networks. After giving a suitable definition for directed modules we investigate their percolation transition in the Erdos-Renyi graph both analytically and numerically. We also analyse four real-world directed networks, including Google's own webpages, an email network, a word association graph and the transcriptional regulatory network of the yeast Saccharomyces cerevisiae. The obtained directed modules are validated by additional information available for the nodes. We find that directed modules of real-world graphs inherently overlap and the investigated networks can be classified into two major groups in terms of the overlaps between the modules. Accordingly, in the word-association network and among Google's webpages the overlaps are likely to contain in-hubs, whereas the modules in the email and t...

  18. A photovoltaic module

    OpenAIRE

    KREBS Frederik C.; Sommer-Larsen, Peter

    2013-01-01

    The present invention relates to a photovoltaic module comprising a carrier substrate, said carrier substrate carrying a purely printed structure comprising printed positive and negative module terminals, a plurality of printed photovoltaic cell units each comprising one or more printed photovoltaic cells, wherein the plurality of printed photovoltaic cell units are electrically connected in series between the positive and the negative module terminals such that any two neighbouring photovolt...

  19. Model theory and modules

    CERN Document Server

    Prest, M

    1988-01-01

    In recent years the interplay between model theory and other branches of mathematics has led to many deep and intriguing results. In this, the first book on the topic, the theme is the interplay between model theory and the theory of modules. The book is intended to be a self-contained introduction to the subject and introduces the requisite model theory and module theory as it is needed. Dr Prest develops the basic ideas concerning what can be said about modules using the information which may be expressed in a first-order language. Later chapters discuss stability-theoretic aspects of module

  20. Delphi Accounts Receivable Module

    Data.gov (United States)

    Department of Transportation — Delphi accounts receivable module contains the following data elements, but are not limited to customer information, cash receipts, line of accounting details, bill...

  1. Cosmetology. Computerized Learning Modules.

    Science.gov (United States)

    Finnerty, Kathy, Ed.

    Intended to help reading-limited students meet course objectives, these 11 modules are based on instructional materials in cosmetology that have a higher readability equivalent. Modules cover bacteriology, chemical waving, scalp and hair massage, chemistry, hair shaping, hairstyling, chemical hair relaxing, hair coloring, skin and scalp,…

  2. Modulation gamma resonance spectroscopy

    International Nuclear Information System (INIS)

    Possibility to control dynamic processes in a matter through gamma-resonance modulation by high-frequency external variable fields in excess of inverse lifetimes of the Moessbauer nuclei excited states, that is, within the megahertz frequency range lies in the heart of the modulation gamma-resonance spectroscopy. Through the use of the gamma-resonance process theoretical analysis methods and of the equation solution method for the density matrix with the secondary quantization of gamma-radiation field one attacks the problems dealing with the effect of both variable fields and relaxation on gamma-resonance. One has studied the gamma-radiation ultrasound modulation stages. One points out a peculiar role of the gamma-magnetic resonance effect in modulation gamma resonance spectroscopy formation. One forecasts development of the modulation gamma-resonance spectroscopy into the nonlinear gamma-resonance spectroscopy

  3. Rigidity of tilting modules

    DEFF Research Database (Denmark)

    Haahr Andersen, Henning; Kaneda, Masaharu

    ) modules for $U_q$ are rigid, i.e., have identical radical and socle filtrations. Moreover, we obtain the same for a large class of Weyl modules for $U_q$. On the other hand, we give examples of non-rigid indecomposable tilting modules as well as non-rigid Weyl modules. These examples are for type $B_2......$ and in this case as well as for type $A_2$ we calculate explicitly the Loewy structure for all regular Weyl modules. We also demonstrate that these results carry over to the modular case when the highest weights in question are in the so-called Jantzen region. At the same time we show by examples that as soon...

  4. Solar energy modulator

    Science.gov (United States)

    Hale, R. R. (Inventor); Mcdougal, A. R.

    1984-01-01

    A module is described with a receiver having a solar energy acceptance opening and supported by a mounting ring along the optic axis of a parabolic mirror in coaxial alignment for receiving solar energy from the mirror, and a solar flux modulator plate for varying the quantity of solar energy flux received by the acceptance opening of the module. The modulator plate is characterized by an annular, plate-like body, the internal diameter of which is equal to or slightly greater than the diameter of the solar energy acceptance opening of the receiver. Slave cylinders are connected to the modulator plate for supporting the plate for axial displacement along the axis of the mirror, therby shading the opening with respect to solar energy flux reflected from the surface of the mirror to the solar energy acceptance opening.

  5. Photovoltaic module and interlocked stack of photovoltaic modules

    Science.gov (United States)

    Wares, Brian S.

    2014-09-02

    One embodiment relates to an arrangement of photovoltaic modules configured for transportation. The arrangement includes a plurality of photovoltaic modules, each photovoltaic module including a frame. A plurality of individual male alignment features and a plurality of individual female alignment features are included on each frame. Adjacent photovoltaic modules are interlocked by multiple individual male alignment features on a first module of the adjacent photovoltaic modules fitting into and being surrounded by corresponding individual female alignment features on a second module of the adjacent photovoltaic modules. Other embodiments, features and aspects are also disclosed.

  6. On Modules Whose Singular Subgenerated Modules Are Weakly Injective

    Institute of Scientific and Technical Information of China (English)

    S. Dhompongsa; J. Sanwong; S.Plubtieng; H.Tansee

    2001-01-01

    Rings over which every singular right module is injective (briefly,right SI-rings) were introduced and investigated by Goodearl. Weakly injective modules, as a generalization of injective modules, were introduced by Jain and modules are weakly injective, which we call SwI-rings. This concept is extended to SwI-modules, i.e., modules whose singular subgenerated modules are weakly injective. Several characterizations and properties of SwI-rings and SwI-modules are obtained which generalize some earlier known results on SI-rings and weakly semisimple rings.

  7. Autonomous cotton module forming system

    Science.gov (United States)

    Cotton producers often have difficulty finding adequate labor during harvest. Module builder operators are often inexperienced and may build poorly shaped modules. Equipment manufacturers have recently introduced harvesters with on-board module building capabilities to reduce labor requirements; h...

  8. Optical modulator including grapene

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Ming; Yin, Xiaobo; Zhang, Xiang

    2016-06-07

    The present invention provides for a one or more layer graphene optical modulator. In a first exemplary embodiment the optical modulator includes an optical waveguide, a nanoscale oxide spacer adjacent to a working region of the waveguide, and a monolayer graphene sheet adjacent to the spacer. In a second exemplary embodiment, the optical modulator includes at least one pair of active media, where the pair includes an oxide spacer, a first monolayer graphene sheet adjacent to a first side of the spacer, and a second monolayer graphene sheet adjacent to a second side of the spacer, and at least one optical waveguide adjacent to the pair.

  9. The ANTARES optical module

    International Nuclear Information System (INIS)

    The ANTARES collaboration is building a deep sea neutrino telescope in the Mediterranean Sea. This detector will cover a sensitive area of typically 0.1 km2 and will be equipped with about 1000 optical modules. Each of these optical modules consists of a large area photomultiplier and its associated electronics housed in a pressure resistant glass sphere. The design of the ANTARES optical module, which is a key element of the detector, has been finalized following extensive R and D studies and is reviewed here in detail

  10. Microlensing modulation by binaries

    CERN Document Server

    Dubath, F; Durrer, R; Dubath, Florian; Gasparini, Maria Alice; Durrer, Ruth

    2006-01-01

    We compute the effect of the lens quadrupole on microlensing. The time dependence of the quadrupole can lead to specific modulations of the amplification signal. We study especially binary system lenses in our galaxy. The modulation is observable if the rotation period of the system is smaller than the time over which the amplification is significant and if the impact parameter of the passing light ray is sufficiently close to the Einstein radius so that the amplification is very large. Observations of this modulation can reveal important information on the quadrupole and thus on the gravitational radiation emitted by the lens.

  11. Solar cell module. Taiyo denchi module

    Energy Technology Data Exchange (ETDEWEB)

    Nakano, Akihiko; Matsumoto, Hitoshi; Komatsu, Yasumitsu; Shirai, Sadaharu.

    1989-09-29

    In the solar cell module of this invention, such junctions as CdS/CdTe or CdS/CuInSe {sub 2} are contained as a photoelectromotive force part coexists with air in a closed space which consists of glass, metal parts and a bonding resin layer; the photoelectromotive force part is coated either with a fluorine resin or a silicone resin. The fluorine resin contains a fundamental skeleton of an alternative copolymer of fluoroolefin and a hydrocarbon-based vinyl monomer; the silicone resin has three types, i.e., addition-reacted, condensated or UV-curing type, and the released oxygen is sealed in the closed space. The resin layer which adheres the glass and the metal plate is a thermoplastic resin which is polyethylene modified by copolymerization of acid anhydride. By this, the reliability of the solar cell module was enhanced. 3 figs.

  12. Module bay with directed flow

    Science.gov (United States)

    Torczynski, John R.

    2001-02-27

    A module bay requires less cleanroom airflow. A shaped gas inlet passage can allow cleanroom air into the module bay with flow velocity preferentially directed toward contaminant rich portions of a processing module in the module bay. Preferential gas flow direction can more efficiently purge contaminants from appropriate portions of the module bay, allowing a reduced cleanroom air flow rate for contaminant removal. A shelf extending from an air inlet slit in one wall of a module bay can direct air flowing therethrough toward contaminant-rich portions of the module bay, such as a junction between a lid and base of a processing module.

  13. A photovoltaic module

    DEFF Research Database (Denmark)

    2013-01-01

    The present invention relates to a photovoltaic module comprising a carrier substrate, said carrier substrate carrying a purely printed structure comprising printed positive and negative module terminals, a plurality of printed photovoltaic cell units each comprising one or more printed...... photovoltaic cells, wherein the plurality of printed photovoltaic cell units are electrically connected in series between the positive and the negative module terminals such that any two neighbouring photovoltaic cell units are electrically connected by a printed interconnecting electrical conductor....... The carrier substrate comprises a foil and the total thickness of the photovoltaic module is below 500 [mu]m. Moreover, the nominal voltage level between the positive and the negative terminals is at least 5 kV DC....

  14. Ultrasound Modulated Bioluminescence Tomography

    CERN Document Server

    Bal, Guillaume

    2013-01-01

    We propose a method to reconstruct the density of a luminescent source in a highly-scattering medium from ultrasound modulated optical measurements. Our approach is based on the solution to a hybrid inverse source problem for the diffusion equation.

  15. GREET Pretreatment Module

    Energy Technology Data Exchange (ETDEWEB)

    Adom, Felix K.; Dunn, Jennifer B.; Han, Jeongwoo

    2014-09-01

    A wide range of biofuels and biochemicals can be produced from biomass via different pretreatment technologies that yield sugars. This report documents the material and energy flows that occur when fermentable sugars from four lignocellulosic feedstocks (corn stover, miscanthus, switchgrass, and poplar) are produced via dilute acid pretreatment and ammonia fiber expansion. These flows are documented for inclusion in the pretreatment module of the Greenhouses Gases, Regulated Emissions, and Energy Use in Transportation (GREET) model. Process simulations of each pretreatment technology were developed in Aspen Plus. Material and energy consumption data from Aspen Plus were then compiled in the GREET pretreatment module. The module estimates the cradle-to-gate fossil energy consumption (FEC) and greenhouse gas (GHG) emissions associated with producing fermentable sugars. This report documents the data and methodology used to develop this module and the cradle-to-gate FEC and GHG emissions that result from producing fermentable sugars.

  16. Top Local Cohomology Modules

    Institute of Scientific and Technical Information of China (English)

    Mohammad T. Dibaei; Siamak Yassemi

    2007-01-01

    For a finitely generated module M over a commutative Noetherian local ring (R, m), it is shown that there exist only a finite number of non-isomorphic top localeohomology modules Hdim(M) (M) for all ideals a of RIt is also shown that for a giveninteger r ≥ 0, if Hra(R/p) is zero for all p in Supp(M), then Hia(M)=0 for all I ≥ r.

  17. Electrical modulation of emissivity

    OpenAIRE

    Vassant, Simon; Moldovan Doyen, Ioana Cristina; Marquier, François; Pardo, F.; Gennser, Ulf; Cavanna, Antonella; Pelouard, Jean-Luc; Greffet, Jean-Jacques

    2013-01-01

    We demonstrate that it is possible to modulate the thermal emission through an electrical modulation of the emissivity. The basic idea is to design a device where absorption is due to a resonant phenomenon. If the resonance can be electrically controlled, then absorption and, therefore, thermal emission can be controlled. We demonstrate this general concept using THz resonant absorption by surface phonon polaritons coupled through a gold grating. In our device, absorption is mostly due to a s...

  18. Groups, rings, modules

    CERN Document Server

    Auslander, Maurice

    2014-01-01

    This classic monograph is geared toward advanced undergraduates and graduate students. The treatment presupposes some familiarity with sets, groups, rings, and vector spaces. The four-part approach begins with examinations of sets and maps, monoids and groups, categories, and rings. The second part explores unique factorization domains, general module theory, semisimple rings and modules, and Artinian rings. Part three's topics include localization and tensor products, principal ideal domains, and applications of fundamental theorem. The fourth and final part covers algebraic field extensions

  19. Photovoltaic module reliability workshop

    Energy Technology Data Exchange (ETDEWEB)

    Mrig, L. (ed.)

    1990-01-01

    The paper and presentations compiled in this volume form the Proceedings of the fourth in a series of Workshops sponsored by Solar Energy Research Institute (SERI/DOE) under the general theme of photovoltaic module reliability during the period 1986--1990. The reliability Photo Voltaic (PV) modules/systems is exceedingly important along with the initial cost and efficiency of modules if the PV technology has to make a major impact in the power generation market, and for it to compete with the conventional electricity producing technologies. The reliability of photovoltaic modules has progressed significantly in the last few years as evidenced by warranties available on commercial modules of as long as 12 years. However, there is still need for substantial research and testing required to improve module field reliability to levels of 30 years or more. Several small groups of researchers are involved in this research, development, and monitoring activity around the world. In the US, PV manufacturers, DOE laboratories, electric utilities and others are engaged in the photovoltaic reliability research and testing. This group of researchers and others interested in this field were brought together under SERI/DOE sponsorship to exchange the technical knowledge and field experience as related to current information in this important field. The papers presented here reflect this effort.

  20. Directed network modules

    International Nuclear Information System (INIS)

    A search technique locating network modules, i.e. internally densely connected groups of nodes in directed networks is introduced by extending the clique percolation method originally proposed for undirected networks. After giving a suitable definition for directed modules we investigate their percolation transition in the Erdos-Renyi graph both analytically and numerically. We also analyse four real-world directed networks, including Google's own web-pages, an email network, a word association graph and the transcriptional regulatory network of the yeast Saccharomyces cerevisiae. The obtained directed modules are validated by additional information available for the nodes. We find that directed modules of real-world graphs inherently overlap and the investigated networks can be classified into two major groups in terms of the overlaps between the modules. Accordingly, in the word-association network and Google's web-pages, overlaps are likely to contain in-hubs, whereas the modules in the email and transcriptional regulatory network tend to overlap via out-hubs