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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. Module of lithium divertor for KTM tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  5. Simulations of NSTX with a liquid lithium divertor module

    International Nuclear Information System (INIS)

    A strategy to develop self-consistent simulations of the behavior of lithium in the Liquid Lithium Divertor (LLD) module to be installed in NSTX is described. In this initial stage of the plan, the UEDGE edge plasma transport code is used to simulate an existing NSTX shot, with UEDGE's transport coefficients set using midplane and divertor diagnostic data. The LLD is incorporated into the simulations as a reduction in the recycling coefficient over the outer divertor. Heat transfer calculations performed using the resulting heat flux profiles provide preliminary estimates on operating limits for the LLD as well as input data for subsequent steps in the LLD modeling effort (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  6. 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)

  7. Manufacturing and testing of a Be/OFHCCu divertor module

    Science.gov (United States)

    Araki, M.; Youchison, D. L.; Akiba, M.; Watson, R. D.; Sato, K.; Suzuki, S.

    1996-10-01

    Beryllium, carbon-based materials and tungsten are considered as plasma facing materials for the next generation of fusion machines such as the international thermonuclear experimental reactor (ITER). Beryllium is one of the primary candidate materials because of its low atomic number and lack of tritium codeposition. However, joining of a beryllium armor to a copper heat sink remains a critical problem due to the formation of brittle intermetallics at the interface. To address this concern, the Japan Atomic Energy Research Institute manufactured a beryllium/Cu divertor module with Cr and Ni diffusion barriers. This Be/Cu module was tested in the electron beam test system of Sandia National Laboratories in the framework of the US—Japan Fusion Collaboration. The divertor module consisted of four beryllium tiles, 25 mm × 25 mm, and a square copper heat sink with convolutions like a screw nut inside the coolant channel. To evaluate the integrity of the brazed bonds under various heat fluxes, beryllium tiles of two different thicknesses, 2 and 10 mm, were bonded to the copper heat sink. Cooling conditions of 10 m/s water flow velocity at 1 MPa, and a water inlet temperature of 20°C were selected based on the thermal analysis. During high heat flux testing the 10 mm thick Be tiles detached at an absorbed heat flux around 5 MW/m 2 for several shots due to flaws at the braze joint confirmed by optical observation after manufacturing. One of the 2 mm thick Be tiles failed after 550 cycles at the steady state heat flux of 6.5 MW/m 2. Most likely the failure was caused by brittleness at the interface caused by the presence of BeCu intermetallics.

  8. Manufacturing and testing of a HETS module for DEMO divertor

    Energy Technology Data Exchange (ETDEWEB)

    Visca, Eliseo, E-mail: eliseo.visca@enea.it [ENEA - Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, 00044 Frascati (Italy); Agostini, Pietro; Crescenzi, Fabio; Malavasi, A.; Pizzuto, Aldo; Rossi, Paolo; Storai, Sandro; Utili, Marco [ENEA - Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, 00044 Frascati (Italy)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer The HETS (high-efficiency thermal shield) concept, initially developed by ENEA for water, has been adapted for use with He as coolant. Black-Right-Pointing-Pointer This DEMO divertor concept is based on elements joined in series and protected by a hemispheric dome. Black-Right-Pointing-Pointer It has been calculated to be capable of sustaining an incident heat flux of 10 MW/m{sup 2} when operating at 10 MPa, an inlet He temperature of 600 Degree-Sign C, and an outlet temperature of 800 Degree-Sign C. Black-Right-Pointing-Pointer The activity is focused on the manufacturing of a single HETS module with W armor and on its thermal-hydraulic testing. Black-Right-Pointing-Pointer A CFD analysis by ANSYS-CFX was performed in order to predict the thermal-mechanical behavior of the module and a final comparison with the experimental data is required to validate the CFD results. - Abstract: The development of a divertor concept for fusion power plants that is able to grant efficient recovery and conversion of the considerable fraction ({approx}15%) of the total fusion thermal power incident is deemed to be an urgent task to meet in the EU Fast Track scenario. The He-cooled conceptual divertor design is one of the possible candidates. Helium cooling offers several advantages including chemical and neutronic inertness and the ability to operate at higher temperatures and lower pressures than those required for water cooling. The HETS (high-efficiency thermal shield) concept, initially developed by ENEA for water, has been adapted for use with He as coolant. This DEMO divertor concept is based on elements joined in series and protected by a hemispheric dome; it allows an increase of thermal exchange coefficient both for high speed of gas and for 'jet impingement' effects of gas coming out from the internal side of hemispheric dome. It has been calculated to be capable of sustaining an incident heat flux of 10 MW/m{sup 2} when

  9. Status of design and experimental activity on module of lithium divertor for KTM tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lyublinski, Igor E., E-mail: lyublinski@yandex.ru [JSC “Red Star”, Moscow (Russian Federation); Vertkov, Alexey V.; Zharkov, Mikhail Yu.; Semenov, Vladimir V. [JSC “Red Star”, Moscow (Russian Federation); Mirnov, Sergey V.; Lazarev, Vladimir B. [GSC RF TRINITI, Troitsk, Moscow Region (Russian Federation); Tazhibayeva, Irina L.; Shapovalov, Gennadiy V.; Kulsartov, Timur V.; D’yachenko, Alexandr V. [IAE of National Nuclear Center, Kurchatov (Kazakhstan); Mazzitelli, Giuseppe [Associazione EURATOM-ENEA sulla Fusione, C.R. ENEA Frascati, Rome (Italy); Agostini, Pietro [ENEA Brasimone, Camugnano, BO (Italy)

    2013-10-15

    Highlights: • Lithium divertor module based on capillary-porous system is created for KTM tokamak. • The hydraulic tests of lithium divertor module were conducted. • The results were compared with the calculation data. • The analysis of results’ discrepancies was conducted. • The lithium divertor module is ready for tests on KTM tokamak. -- Abstract: The projects of ITER and DEMO reactors showed that there are serious difficulties with solving the issues of plasma facing elements (PFE) based on the solid materials. Problems of PFE can be overcome by the use of liquid lithium. Application of lithium will allow to create a self-renewal and MHD stable liquid metal surface of the in-vessel devices possessing practically unlimited service life. Realization of these advantages is based on use of so-called lithium capillary-porous system (CPS) – new material, in which liquid lithium fills a solid matrix from porous material. The progress in development of lithium technology and also lithium experiments in the tokamaks TFTR, T-11M, T-10, FTU, NSTX, LTX, HT-7 and stellarator TJ II is a good basis for development of the project of steady-state operating lithium divertor module for Kazakhstan tokamak. At present the lithium divertor module for KTM tokamak is development and manufacturing. The paper describes main design features of the module of lithium divertor (MLD). The first step of the hydraulic tests of MLD with fully assembled external thermo-stabilization system, which was connected to in-vessel lithium unit, were performed using ethanol as a model heat transfer media. Test results of MLD have shown that operating parameters of designed and manufactured system for thermo-stabilization are sufficient for proper operation; basic hydraulic characteristics of the system are close to expected values.

  10. 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.

  11. 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

  12. 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.

  13. Thermal cycling tests of RG-Ti/Mo monoblock divertor modules

    International Nuclear Information System (INIS)

    Test results for actively cooled, graphite RG-Ti type brazed to molybdenum tube divertor mock-ups are presented. Particular attention was paid to finding a reliable, non-destructive method of investigating brazed joints. The mock-ups were ultrasonically tested and the results correlated with the microstructure of the joints. Two types of tests were carried out: screening tests with a stepped increase of the surface heat flux from shot to shot and thermal cycling tests at a fixed power density level. The heat loads were varied in the range from 6 MW/m2 to 12 MW/m2. About 150 cycles of heat loads with a power density of 10 MW/m2 were realized and no macrodamage was found. The heat flux experiments were accompanied by detailed metallographic investigations. ((orig.))

  14. 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.

  15. 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

  16. 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)

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. The jet divertor coils

    International Nuclear Information System (INIS)

    This paper reports on the JET Tokamak which is to be modified to incorporate a divertor. A coil system in the vacuum vessel has been developed, which can produce a range of different divertor plasmas. The divertor coils are of conventional construction and are contained in this Inconel cases. They will be assembled in the vacuum vessel, welded into their cases and impregnated with epoxy resin

  3. 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.

  4. 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.

  5. '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

  6. 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

  7. 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)

  8. 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.

  9. Study of the radiation in divertor plasmas

    International Nuclear Information System (INIS)

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

  10. 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)

  11. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Laugier, F

    2000-10-19

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

  13. 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

  14. 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.

  15. 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)

  16. Twin tori for a new bundle divertor

    International Nuclear Information System (INIS)

    A new bundle divertor system using the straight stagnation axis in toroidal field together with the uniform field along the axis is discussed in detail. We call this type of divertor as the ''muffler divertor'' because of its shape. (author)

  17. 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)

  18. 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.

  19. PDX divertor operation

    International Nuclear Information System (INIS)

    PDX was brought into operation in January 1980 as a diverted tokamak with typical parameters of Bsub(T) = 15-20 kG, a = 38 cm, R0 = 123-159 cm, Isub(p) = 180-300 kA, q approx. equal to 3.7, anti nsub(e) = 1-3.8 x 1013 cm-3, anti Z = 1.1-3, tausub(E)sub(e) approx. equal to 25 ms, and pulse lengths up to 0.7 s. Internal vacuum components that were exposed to the plasma (such as limiters, shields, microwave horns, etc.) were fabricated from 99% pure titanium. Glow discharge cleaning with 3 x 10-2 Torr H2 and pulse discharge cleaning were used to condition the vessel for high power discharges. For the divertor studies, work has concentrated on obtaining long, high current stable discharges. Radial position, plasma current, and gas injection control systems have been used to facilitate this effort. Discharges of inside-D, square, and inverse-D cross-section have been produced. Microwave interferometers, spectroscopy, an X-ray pulse height analyzer system, scanning and fixed bolometers, and thermocouple array have been used to determine plasma and impurity densities, temperature, radiation, and power loss to the divertor. A comparison of diverted and undiverted discharges is presented. (orig.)

  20. Development of microwave interferometer system for divertor simulation experiments in GAMMA 10/PDX

    Science.gov (United States)

    Kohagura, J.; Wang, X.; Kanno, S.; Yoshikawa, M.; Kuwahara, D.; Nagayama, Y.; Shima, Y.; Chikatsu, M.; Nojiri, K.; Sakamoto, M.; Imai, T.; Nakashima, Y.; Mase, A.

    2015-12-01

    Microwave interferometer has newly been installed on GAMMA 10/PDX for divertor simulation study. A divertor simulation experimental module (D-module) is used to investigate the physics of divertor in the end-cell of GAMMA 10/PDX where an open magnetic field configuration is formed. D-module has a rectangular chamber with an inlet aperture. Two tungsten target plates are mounted in V-shape inside the chamber. In order to develop understandings of divertor simulation experiments the microwave interferometer using heterodyne scheme and a 1D horn-antenna mixer array (HMA) is applied to obtain electron density and density distribution inside the V-shaped target plates. Line-averaged electron density distributions inside D-module are first observed in H2 gas injection experiments.

  1. 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.))

  2. High temperature divertor plasma operation

    International Nuclear Information System (INIS)

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

  3. The divertor remote maintenance project

    International Nuclear Information System (INIS)

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

  4. 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

  5. Edge plasma in snowflake divertor

    International Nuclear Information System (INIS)

    The snowflake divertor (Ryutov 2007, Phys. Plasmas 14, 064502) uses a 2nd order null of the poloidal magnetic field instead of the 1st order null used in the standard divertor. This leads to a number of interesting geometric properties such as stronger fanning of the poloidal flux, stronger magnetic shear in the edge region, larger radiating volume, and larger connection length in the scrape-off layer. These can potentially lead to new ways for alleviating heat loads on the divertor target plates. Discussion of properties of snowflake is presented, along with results of numerical modeling. Divertor leg volume is larger in snowflake than in the standard x-point configuration, which leads to larger fraction of radiated power in the divertor. This allows the snowflake to transition to a strongly detached plasma regime more easily than for the standard x-point. Besides, stronger shearing of the magnetic field in snowflake may be beneficial for controlling magneto-hydrodynamic instabilities in the edge (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  6. 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.

  7. Tokamak Physics Experiment divertor design

    International Nuclear Information System (INIS)

    The Tokamak Physics Experiment (TPX) tokamak requires a symmetric up/down double-null divertor capable of operation with steady-state heat flux as high as 7.5 MW/m2. The divertor is designed to operate in the radiative mode and employs a deep slot configuration with gas puffing lines to enhance radiative divertor operation. Pumping is provided by cryopumps that pump through eight vertical ports in the floor and ceiling of the vessel. The plasma facing surface is made of carbon-carbon composite blocks (macroblocks) bonded to multiple parallel copper tubes oriented vertically. Water flowing at 6 m/s is used, with the critical heat flux (CHF) margin improved by the use of enhanced heat transfer surfaces. In order to extend the operating period where hands on maintenance is allowed and to also reduce dismantling and disposal costs, the TPX design emphasizes the use of low activation materials. The primary materials used in the divertor are titanium, copper, and carbon-carbon composite. The low activation material selection and the planned physics operation will allow personnel access into the vacuum vessel for the first 2 years of operation. The remote handling system requires that all plasma facing components (PFCs) are configured as modular components of restricted dimensions with special provisions for lifting, alignment, mounting, attachment, and connection of cooling lines, and instrumentation and diagnostics services

  8. 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.

  9. 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

  10. Progress in ergodic divertor operation on Tore Supra

    International Nuclear Information System (INIS)

    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 ∼ 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)

  11. 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)

  12. High heat flux tests of small-scale Be/Cu mock-ups for ITER

    International Nuclear Information System (INIS)

    Several kinds of Be/Cu joints have been made by hot isostatic press (HIP) in China in order to develop the ITER-FW blanket fabrication technology. At the first stage, high temperature HIP technology was investigated, and both Ti film and PVD (physical vapor deposition)-coating were adopted as intermediate layers between high purity beryllium made by HIP and CuCrZr alloy. The average bonding strength of Be/CuCrZr joints is larger than 60 Mpa and a good metallurgical bonding was formed. The Be/CuCrZr joints at optimized technology can sustain 1000 cycles under an absorbed power density of about 2.5 MW·m-2, which shows relatively good thermal fatigue properties. Temperature and stress distributions were also calculated by 2D ANSYS, showing a good accord with experimental results. Low temperature HIP joining is being developed and the heat load evaluation is also under way. (author)

  13. 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

  14. 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.

  15. Kinetic Modeling of Divertor Plasma

    Science.gov (United States)

    Ishiguro, Seiji; Hasegawa, Hiroki; Pianpanit, Theerasarn

    2015-11-01

    Particle-in-Cell (PIC) simulation with the Monte Carlo collisions and the cumulative scattering angle coulomb collision can present kinetic dynamics of divertor plasmas. We are developing two types of PIC codes. The first one is the three dimensional bounded PIC code where three dimensional kinetic dynamics of blob is studied and current flow structures related to sheath formation are unveiled. The second one is the one spatial three velocity space dimensional (1D3V) PIC code with the Monte Carlo collisions where formation of detach plasma is studied. First target of our research is to construct self-consistent full kinetic simulation modeling of the linear divertor simulation experiments. This work is performed with the support and under the auspices of NIFS Collaboration Research program (NIFS15KNSS059, NIFS14KNXN279, and NIFS13KNSS038) and the Research Cooperation Program on Hierarchy and Holism in Natural Science at NINS.

  16. Development of divertor remote maintenance system

    International Nuclear Information System (INIS)

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

  17. 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.

  18. SOL–divertor plasma simulations introducing anisotropic temperature with virtual divertor model

    International Nuclear Information System (INIS)

    A 1D SOL–divertor plasma simulation code by introducing the anisotropic ion temperature with virtual divertor model has been developed. By introducing the anisotropic ion temperature directly, the second-order derivative parallel ion viscosity term in the momentum transport equation can be excluded and the boundary condition at the divertor plate will not be required in the simulation. In order to express the effects of the divertor plate and accompanying sheath implicitly, a virtual divertor model which has artificial sinks for the particle, momentum and energy has been introduced. Periodic boundary condition becomes available by the use of the virtual divertor model. By using this model, SOL–divertor plasmas which satisfy the Bohm condition has been successfully obtained. The dependence of the ion temperature anisotropy on the normalized mean free path of ion and the validity of the parallel ion viscous flux for the Braginskii expression and the limited one are also investigated

  19. 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)

  20. 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

  1. Moving Divertor Plates in a Tokamak

    International Nuclear Information System (INIS)

    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

  2. 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

  3. 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.

  4. 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)

  5. Electron yield from Be-Cu induced by highly charged Xe q+ ions

    Czech Academy of Sciences Publication Activity Database

    Krása, Josef; Láska, Leoš; Stöckli, M. P.; Fehrenbach, C. W.

    2002-01-01

    Roč. 196, - (2002), s. 61-67. ISSN 0168-583X R&D Projects: GA AV ČR IAA1010105; GA MŠk LN00A100 Institutional research plan: CEZ:AV0Z1010921 Keywords : highly charged ion-induced electron emission * angle impact effect * Be-Cu Subject RIV: BH - Optics, Masers, Lasers Impact factor: 1.158, year: 2002

  6. 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

  7. Divertor and gas blanket impurity control study

    International Nuclear Information System (INIS)

    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

  8. Activation analysis of coolant water in ITER blanket and divertor

    International Nuclear Information System (INIS)

    Coolant water in blankets and divertor cassettes will be activated by neutrons during ITER operation. 16N and 17N are determined to be the most important activation products in the coolant water in terms of their impact on ITER design and performance. In this study, the geometry of cooling channels in blanket module 4 was described precisely in the ITER neutronics model ‘Alite-4’ based on the latest CAD model converted using MCAM developed by FDS Team. The 16N and 17N concentration distribution in the blanket, divertor cassette and their primary heat transport systems were calculated by MCNP with data library FENDL2.1. The activation of cooling pipes induced 17N decay neutrons was analyzed and compared with that induced by fusion neutrons, using FISPACT-2007 with data library EAF-2007. The outlet concentration of blanket and divertor cooling systems was 1.37 × 1010 nuclide/cm3 and 1.05 × 1010 nuclide/cm3 of 16N, 8.93 × 106 nuclide/cm3 and 0.33 × 105 nuclide/cm3 of 17N. The decay gamma-rays from 16N in activated water could be a problem for cryogenic equipments inside the cryostat. Near the cryostat, the activation of pipes from 17N decay neutrons was much lower than that from fusion neutrons.

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

    International Nuclear Information System (INIS)

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

  10. Development and characterization of Be/Cu joint obtained by hot isostatic pressing

    International Nuclear Information System (INIS)

    Beryllium is one of the possible candidate for Plasma Facing Components (PFC) such as divertor or first wall in the ITER project. Due to its high thermal conductivity, copper alloys are used as heat sink material. So, in one way or another, beryllium will have to be fixed onto copper alloys. In this study, the joining of beryllium onto copper is achieved by Hot Isostatic Pressing (Solid HIP). This joining technique allows an homogeneous bonding. But, as direct bonding between Be and Cu induces intermetallics which are deleterious to the joint (interlayers are needed to avoid reaction between Be and Cu. This paper gives a description and the role of different associated interlayers used as diffusion barriers between copper and beryllium for a low in-service working temperature. Moreover, a mock-up was fabricated. Shear resistance of the junction was measured from test specimens machined from the mock-up. (authors)

  11. 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.)

  12. 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.

  13. 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

  14. Manufacture and installation of JET MKII divertor support structure

    International Nuclear Information System (INIS)

    The water cooled support structure, comprising twenty-four modules is the main component of the JET MKII divertor system. It is to be installed in the vacuum vessel with high accuracy with respect to the magnetic center and the other in-vessel components. The paper describes the design and manufacturing cycle including the required tolerances, the assembly and installation method and the material production process required to ensure the accuracy and reliability of the MKII support structure system. The water cooling holes, machined into the support structure require the procurement of special material to prevent risks of leaks inside the vacuum vessel

  15. Dust divertor for a tokamak fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tang, X Z [Los Alamos National Laboratory; Delzanno, G L [Los Alamos National Laboratory

    2009-01-01

    Micron-size tungsten particulates find equilibrium position in the magnetized plasma sheath in the normal direction of the divertor surface, but are convected poloidally and toroidally by the sonic-ion-flow drag parallel to the divertor surface. The natural circulation of dust particles in the magnetized plasma sheath can be used to set up a flowing dust shield that absorbs and exhausts most of the tokamak heat flux to the divertor. The size of the particulates and the choice of materials offer substantial room for optimization.

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

    International Nuclear Information System (INIS)

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

  17. 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.

  18. Stability, divertors and innovative concepts

    International Nuclear Information System (INIS)

    This paper contains a short resume of the sections on 'Stability, Divertors and Innovative Concepts' presented at the 19th IAEA Fusion Energy Conference. The main conclusions are: (1) the problem of type I ELMs in tokamaks seems to be not so dramatic; (2) it was demonstrated that the working pulse length of large thermonuclear devices can achieve 100 s and more; (3) the problem of tritium retention seems to be not so dramatic now; probable approaches of its solution are visible; (4) active methods of plasma instabilities suppression (NTM, RWM, sawteeth, external MHD) work successfully; (5) new methods of mitigation of the disruption consequences were offered. New technological ideas and new ideas on magnetic confinement were presented. (author)

  19. 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

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

    International Nuclear Information System (INIS)

    Based on the design requirements for the plasma facing components in ITER/FER, we have performed design studies on the conductively cooled first wall armor and the divertor plate with sliding supports. The full-scale armor tiles were fabricated for heat load tests, and good thermal performances were obtained in heat load tests of 0.2-0.4 MW/m2. It is shown by the thermomechanical analysis on the divertor plate that thermal stresses and bending deformation are reduced significantly by using the sliding supports. The divertor test module with the sliding supports has been fabricated to investigate its fabricability and to verify the functions of the sliding supports during a high heat load of about 10 MW/m2. (orig.)

  1. Radiative divertor and SOL experiments in open and baffled divertors on DIII-D

    International Nuclear Information System (INIS)

    We present recent progress towards an understanding of the physical processes in the divertor and scrape-off-layer (SOL) plasmas in DIII-D. This has been made possible by a combination of new diagnostics, improved computational models, and changes in divertor geometry. We have focused primarily on ELMing H-mode discharges. The physics of Partially Detached Divertor (PDD) plasmas, with divertor heat flux reduction by divertor radiation enhancement using D2 puffing, has been studied in 2-D, and a model of the heat and particle transport has been developed that includes conduction, convection, ionization, recombination, and flows. Plasma and impurity particle flows have been measured with Mach probes and spectroscopy and these flows have been compared with the UEDGE model. The model now includes self-consistent calculations of carbon impurities. Impurity radiation has been increased in the divertor and SOL with 'puff and pump' techniques using SOL D2 puffing, divertor cryopumping, and argon puffing. The important physical processes in plasma-wall interactions have been examined with a DiMES probe, plasma characterization near the divertor plate, and the REDEP code. Experiments comparing single-null (SN) plasma operation in baffled and open divertors have demonstrated a change in the edge plasma profiles. These results are consistent with a reduction in the core ionization source calculated with UEDGE. Divertor particle control in ELMing H-mode with pumping and baffling has resulted in reduction in H-mode core densities to ne/ngw=0.25. Divertor particle exhaust and heat flux has been studied as the plasma shape was varied from a lower SN, to a balanced double null (DN), and finally to an upper SN. (author)

  2. Conceptual design of pebble drop divertor

    International Nuclear Information System (INIS)

    A pebble drop divertor concept is proposed for future fusion reactor. The marked feature of this system is the use of multi-layer pebbles that consists of a central kernel and some coating layers, as a divertor surface component. By using multi-layer pebbles, pebble drop divertor have the advantages such as steady state wall pumping with low bulk tritium retention. The performance of whole divertor system depends on the characteristics of the multi-layer pebble. Particularly the maximum heat load of the system is determined by the dimensions, the layer structure and the material of a kernel. A kernel also has an important role to determine surface temperature, which affects the wall pumping efficiency. This paper presents the numerical results of the maximum allowable heat load and the surface temperature of the divertor pebble. From the numerical estimation of thermal stress and surface temperature, it is found that the radius of divertor pebble with ceramic kernel should be 0.5 - 1 mm. (author)

  3. Conceptual design of pebble drop divertor

    International Nuclear Information System (INIS)

    A pebble drop divertor concept is proposed for future fusion reactor. The marked feature of this system is the use of multi-layer pebbles that consists of a central kernel and some coating layers, as a divertor surface component. By using multi-layer pebbles, pebble drop divertor have the advantages such as steady state wall pumping with low bulk tritium retention. The performance of whole divertor system depends on the characteristics of the multi-layer pebble. Particularly the maximum heat load of the system is determined by the dimensions, the layer structure and the material of a kernel. A kernel also has an important role to determine surface temperature, which affects the wall pumping efficiency. This paper presents the numerical results of the maximum allowable heat load and the surface temperature of the divertor pebble. From the numerical estimation of thermal stress and surface temperature, it is found that the radius of divertor pebble with ceramic kernel should be 0.5-1 mm. (author)

  4. 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.)

  5. 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.

  6. Recent DIII-D divertor research

    International Nuclear Information System (INIS)

    DIII-D currently operates with a single- or double-null open divertor and graphite walls. Active particle control with a divertor cryopump has demonstrated density control, efficient helium exhaust, and reduction of the inventory of particles in the wall. Gas puffing of D2 and impurities has demonstrated reduction of the peak divertor beat flux by factors of 3--5 by radiation. A combination of active cryopumping and feedback-controlled D2 gas puffing has produced similar divertor heat flux reduction with density control. Experiments with neon puffing have shown that the radiation is equally-divided between a localized zone near the X-point and a mantle around the plasma core. The density in these experiments has also been controlled with cryopumping. These experimental results combined with modeling were used to develop the new Radiative Divertor for DIII-D. This is a double-null slot divertor with four cryopumps to provide particle control and neutral shielding for high-triangularity advanced tokamak discharges. UEDGE and DEGAS simulations, benchmarked to experimental data, have been used to optimize the design

  7. 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.

  8. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

  10. 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.

  11. Integrated core-edge-divertor modeling studies

    International Nuclear Information System (INIS)

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

  12. 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)

  13. 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.

  14. 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

  15. Thermal fatigue cycling of Be/Cu joining mock-ups

    International Nuclear Information System (INIS)

    To evaluate beryllium-to-copper joining techniques for potential use by US manufacturers in making first wall components for International Thermonuclear Experimental Reactor (ITER), we tested two mock-ups with S65C beryllium (Be) tiles Hot Isostatic Pressing (HIP) bonded to CuCrZr heat sinks. Under the aegis of the US ITER Project Office, Sandia prepared the mock-ups working with industrial vendors and performed high heat flux testing at Sandia's Plasma Material Test Facility (PMTF) to ascertain the robustness of the Be/Cu joints to 1000 thermal fatigue cycles at a heat flux level of 1.5 MW/m2. Thermal stress analysis provided insight into choosing the heat flux and flow conditions required for accelerated fatigue testing at 1000 cycles and 1.5 MW/m2 that is comparable to the 12,000 cycles and 0.875 MW/m2 required for the ITER First Wall Qualification Mock-ups. Each mock-up had three Be tiles, 35.5 mm square and 10 mm thick, bonded to a CuCrZr heat sink 134.5 mm x 36 mm x 25 mm with a single bored 12.7 mm (dia.) cooling channel. The bonding techniques included various interlayer metallizations and HIPping at 100 MPa pressure and temperature of 580 or 560 deg. C for 2 h. Each tile had a thermocouple (TC) in the center 1 mm below the Be/Cu interface. The test arrangement allowed for both mock-ups to be tested at the same time with alternate heating and cooling cycles of equal duration of 30 s. A total power of 12.7 kW was absorbed by the heated area of 4000 mm2 during the on-cycle. The mock-up was cooled by water at 2.3 m/s (0.27 kg/s), 1 MPa and 20 deg. C inlet temperature. These operating conditions did not permit the mock-ups to cool down to their initial temperature state during the off-cycle. Both mock-ups survived 1000 cycles with no significant changes. The temperature of the top surface on each reached 254 deg. C; while the center TCs reached 136 and 139 deg. C, respectively. Despite localized changes observed in the surface emissivity, the corrected

  16. 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.

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

    International Nuclear Information System (INIS)

    The use of lithium-coated plasma facing components for plasma density control is studied in the National Spherical Torus Experiment (NSTX). A recently installed liquid lithium divertor (LLD) module has a porous molybdenum surface, separated by a stainless steel liner from a heated copper substrate. Lithium is deposited on the LLD from two evaporators. Two new spectroscopic diagnostics are installed to study the plasma surface interactions on the LLD: (1) A 20-element absolute extreme ultraviolet (AXUV) diode array with a 6 nm bandpass filter centered at 121.6 nm (the Lyman-α transition) for spatially resolved divertor recycling rate measurements in the highly reflective LLD environment, and (2) an ultraviolet-visible-near infrared R=0.67 m imaging Czerny-Turner spectrometer for spatially resolved divertor D I, Li I-II, C I-IV, Mo I, D2, LiD, CD emission and ion temperature on and around the LLD module. The use of photometrically calibrated measurements together with atomic physics factors enables studies of recycling and impurity particle fluxes as functions of LLD temperature, ion flux, and divertor geometry.

  18. EU R and D on divertor components

    International Nuclear Information System (INIS)

    Since the last SOFT conference held in Helsinki in 2002, substantial progress has been made in the EU R and D on the divertor components. A number of activities have been completed and new ones have been launched. The present paper gives an update of the works carried out by the EU Participating Team in support of the development of the divertor, which is one of the most challenging components of the next-step ITER machine. The following topics are covered: (1) the further development and consolidation of suitable technologies for the production of high heat-flux components, which culminated with the successful manufacturing and testing of a full-scale vertical target prototype; (2) the completion of the post-irradiation testing of divertor mock-ups and samples; (3) the preparation for the hydraulic and assembly tests of a complete set of full-scale divertor components; (4) the on-going R and D on the definition of workable acceptance criteria for the procurement of ITER high heat-flux components; (5) the activities in support of the divertor design

  19. 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

  20. 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.

  1. 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

  2. Divertor E X B Plasma Convection in DIII-D

    International Nuclear Information System (INIS)

    Extensive two-dimensional measurements of plasma potential in the DIII-D tokamak divertor region are reported for standard (ion VBT drift toward divertor X-point) and reversed BT directions; for low (L) and high (H) confinement modes; and for partially detached divertor mode. The data are consistent with recent computational modeling identifying E x BT circulation, due to potentials sustained by plasma gradients, as the main cause of divertor plasma sensitivity to BT direction

  3. Impact of divertor geometry on radiative divertor performance in JET H-mode plasmas

    Science.gov (United States)

    Jaervinen, A. E.; Brezinsek, S.; Giroud, C.; Groth, M.; Guillemaut, C.; Belo, P.; Brix, M.; Corrigan, G.; Drewelow, P.; Harting, D.; Huber, A.; Lawson, K. D.; Lipschultz, B.; Maggi, C. F.; Matthews, G. F.; Meigs, A. G.; Moulton, D.; Stamp, M. F.; Wiesen, S.; Contributors, JET

    2016-04-01

    Radiative divertor operation in JET high confinement mode plasmas with the ITER-like wall has been experimentally investigated and simulated with EDGE2D-EIRENE in horizontal and vertical low field side (LFS) divertor configurations. The simulations show that the LFS divertor heat fluxes are reduced with N2-injection in similar fashion in both configurations, qualitatively consistent with experimental observations. The simulations show no substantial difference between the two configurations in the reduction of the peak LFS heat flux as a function of divertor radiation, nitrogen concentration, or pedestal Zeff. Consistently, experiments show similar divertor radiation and nitrogen injection levels for similar LFS peak heat flux reduction in both configurations. Nevertheless, the LFS strike point is predicted to detach at 20% lower separatrix density in the vertical than in the horizontal configuration. However, since the peak LFS heat flux in partial detachment in the vertical configurations is shifted towards the far scrape-off layer (SOL), the simulations predict no benefit in the reduction of LFS peak heat flux for a given upstream density in the vertical configuration relative to a horizontal one. A factor of 2 reduction of deuterium ionization source inside the separatrix is observed in the simulations when changing to the vertical configuration. The simulations capture the experimentally observed particle and heat flux reduction at the LFS divertor plate in both configurations, when adjusting the impurity injection rate to reproduce the measured divertor radiation. However, the divertor D α -emissions are underestimated by a factor of 2-5, indicating a short-fall in radiation by the fuel species. In the vertical configuration, detachment is experimentally measured and predicted to start next to the strike point, extending towards the far SOL with increasing degree of detachment. In contrast, in the horizontal configuration, the entire divertor particle flux

  4. 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)

  5. Thermal effects of runaway electrons in an armoured divertor

    International Nuclear Information System (INIS)

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

  6. Liquid metal cooled divertor for ARIES

    International Nuclear Information System (INIS)

    A liquid metal, Ga-cooled divertor design was completed for the double null ARIES-II divertor design. The design analysis indicated a surface heat flux removal capability of up to 15 MW/m2, and its relative easy maintenance. Design issues of configuration, thermal hydraulics, thermal stresses, liquid metal loop and safety effects were evaluated. For coolant flow control, it was found that it is necessary to use some part of the blanket cooling ducts for the draining of liquid metal from the top divertor. In order to minimize the inventory of Ga, it was recommended that the liquid metal loop equipment should be located as close to the torus as possible. More detailed analysis of transient conditions especially under accident conditions was identified as an issue that will need to be addressed

  7. 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.

  8. Conceptual Design for a Bulk Tungsten Divertor Tile in JET

    International Nuclear Information System (INIS)

    With ITER on the verge of being build, the ITER-like Wall project (ILW) for JET aims at providing the plasma chamber of the tokamak with an environment of mixed materials which will be relevant to the support of decisions to the first wall construction and, from the point of view of plasma physics, to the corresponding investigations of possible plasma configuration and plasma-wall interaction. In both respects, tungsten plays a key role in the divertor cladding whereas beryllium will be used for the vessel's first wall. For the central tile, also called LB-SRP for '' Load-Bearing Septum Replacement Plate '', resort to bulk tungsten is envisaged in order to cope with the high loads expected (up to 10 MW/m2 for about 10 s). This is indeed the preferred plasma-facing component for positioning the outer strike-point in the divertor. Forschungszentrum Juelich has developed a conceptual design for this tile, based on an assembly of tungsten blades or lamellae. It was selected in the frame of an extensive R-and-D study in search of a suitable, inertially cooled component(T. Hirai et al., R-and-D on full tungsten divertor and beryllium wall for JET ITER-like Wall Project: this conference). As reported elsewhere, the design is actually driven by electromagnetic considerations in the first place(S. Sadakov et al., Detailed electromagnetic analysis for optimisation of a tungsten divertor plate for JET: this conference). The lamellae are grouped in four stacks per tile which are independently attached to an equally re-designed supporting structure. A so-called adapter plate, also a new design, takes care of an appropriate interface to the base carrier of JET, onto which modules of two tiles are positioned and screwed by remote handling (RH) procedures. The compatibility of the design on the whole with RH requirements is another essential ingredient which was duly taken into account throughout. The concept and the underlying philosophy will be presented along with important

  9. 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...

  10. Design Integration of Liquid Surface Divertors

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-11-13

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

  11. Control of divertor configuration in JT-60

    International Nuclear Information System (INIS)

    The control algorithm of JT-60 divertor configuration is presented. JT-60 has five types of poloidal magnetic field coil with each power supply in order to regulate the control objectives mentioned above. However, if one controls each objective by each coil current independently, there must inevitably occur large interaction between control objectives. Because the relation between control objectives and coil currents is complicated. This situation may be the same with a fusion reactor device. For making it possible to control each objective independently without causing large interaction, the authors adopt the noninteracting control algorithm. Hence, this report demonstrates the availability of this method to the control of JT-60 divertor configuration

  12. Development of heat sink concept for near-term fusion power plant divertor

    International Nuclear Information System (INIS)

    The development of the efficient divertor concept is an important task to meet in the scenario of the future fusion power plant (DEMO). The divertor has to discharge the considerable fraction ∼15% of the total fusion thermal power incident on the divertor, therefore it has to survive very high thermal loads (∼10 MW/m2). In the present study, a new high efficient divertor heat sink (HEDHS) concept is proposed for the future post ITER tokamak called as 'DEMO'. The first wall of the diverter made-up of several modules to overcome the stresses caused by high heat flux, in the present design. Thermal hydraulic performance of one such HEDHS module is numerically investigated using the Fluent software. The effects of critical thermal hydraulic and geometric parameters on the heat transfer characteristics of HEDHS are presented with the Reynolds number (Re) range of 1.2 × 104 - 3.0 × 104. The stresses induced in the HEDHS by the thermal and pressure loads are an important factor that limits the performance and life of the divertor. Therefore, heat transfer coefficient received from the computational fluid dynamics (CFD) analysis is used to perform the thermo-mechanical analysis through finite element based approach. The result revealed that, the proposed design is capable to accommodate the design loads at the acceptable pumping power ratio, and stresses are well within the allowable limits. In addition, detailed of fluid flow and heat transfer mechanism associated with geometric variation have also been studied for the HEDHS to enhance the thermal performance. (author)

  13. 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).

  14. 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)

  15. 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...

  16. First study of EAST divertor by impurities puffing

    International Nuclear Information System (INIS)

    A series of experiments has recently been carried out in EAST under different plasma conditions to investigate the basic divertor performance, divertor and SOL screening efficiency and radiative divertor effect. Detached divertor plasmas have been achieved by density ramp-up. It is found that CIII emission from the low-field side (LFS) exhibits a strong dependence on poloidal locations and plasma operation regimes from methane (CH4) puffing experiments. In addition, the radiative divertor experiments by injection of mixed Ar (5.7% Ar in D2) into the outer divertor chamber reduce the peak heat flux by 50% at the outer target plate, which also reduce the divertor plasma temperature. The in-out heat flux distribution asymmetry is improved.

  17. 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.

  18. ELM induced divertor heat loads on TCV

    Czech Academy of Sciences Publication Activity Database

    Marki, J.; Pitts, R. A.; Horáček, Jan; Turri, G.; Tskhakaya, D.; TCV, team.

    Basel: Swiss Physical Society, 2008. s. 75-75. ISBN N. [Annual meeting of the Swiss physical society 2008. 26.03.208-27.03.2008, Geneva] Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak TCV * divertor heat load * ELM Subject RIV: BL - Plasma and Gas Discharge Physics http://www.sps.ch/uploads/media/SPS2008_Plasma.pdf

  19. 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

  20. Divertor Materials Evaluation System (DiMES)

    International Nuclear Information System (INIS)

    The mission of the Divertor Materials Evaluation System (DiMES) in DIII-D is to establish an integrated data base from measurements in the divertor of a tokamak in order to address some of the ITER and fusion power reactor plasma material interaction issues. Carbon and metal coatings of Be, W, V, and Mo were exposed to the steady-state outer strike point on DIII-D for 4-18 s. These short exposure times ensure controlled exposure conditions, and the extensive arrays of DIII-D divertor diagnostics provide a well-characterized plasma for modeling efforts. Postexposure analysis provides a direct measure of surface material erosion rates and the amount of retained deuterium. For carbon, these results match closely with the results of accumulated carbon deposition and erosion, and the corresponding deuterium retention of long term exposure tiles in DIII-D. Deuterium retention of different materials was measured using the 3He(d,p) 4He nuclear reaction. For carbon, these measurements showed peak deuterium areal density of about 8 x 10 18 D/cm2 in a co-deposited layer about 6 microm deep, mainly at the usually detached inboard divertor leg. That layer of carbon near the inner divertor strike point has an atomic saturation concentration of D/C ∼ 0.25, which is not significantly lower than the laboratory-measured saturation retention of 0.4. Under the carbon contaminated background plasma of DIII-D, metal coatings of Be, V, Mo, and W were exposed to the steady state outer strike point under ELMing and ELM-free H-mode discharges. The rate of material erosion and tritium retention were measured. As expected, W shows the lowest erosion rate at 0.1 nm/s and the lowest deuterium uptake

  1. 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

  2. Fabrication and high heat flux test of divertor cooling elements

    International Nuclear Information System (INIS)

    The plasma facing components in ITER are subjected to a high heat flux from fusion plasma. The heat flux is not only higher than that of existing tokamaks but also has a longer pulse duration (burn time). To minimize a peaking of the heat flux, the thermal deformation towards the plasma should be restrained. One-meter-long monoblock divertor modules with a sliding support structure were fabricated and tested at JAERI. Two kinds of support mechanisms were provided to minimize the thermal deformation of the modules in the upward and downward directions ; one is a pin type sliding structure and the other is a rail type support structure. Both modules were tested on the electron beam HHF test facility, JEBIS (JAERI Electron Beam Irradiation System), in JAERI. The steady-state heat flux of 15 MW/m2 was applied to the surface of the modules to simulate the design condition of ITER CDA. As a result of the HHF test, the performance of the sliding support structures was successfully demonstrated. Three dimensional elastic stress analyses were conducted using a finite element method. The result shows that the relatively high thermal stress is observed at the cooling tube ; and that the maximum thermal stress at the cooling tube exceeds its yield strength. It is necessary to perform the lifetime evaluation of the copper cooling tube against cyclic thermal stresses. (author)

  3. 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.

  4. ITER tungsten divertor design development and qualification program

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  5. ITER tungsten divertor design development and qualification program

    International Nuclear Information System (INIS)

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

  6. 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.

  7. Development of a radiative divertor for DIII-D

    International Nuclear Information System (INIS)

    We have used experiments and modeling to develop a new radiative divertor configuration for DIII-D. Gas puffing experiments with the existing open divertor have shown the creation of a localized (∼10 cm diameter) radiation zone which results in substantial reduction (3--10) in the divertor heat flux while δE remains ∼2 times ITER-89P scaling. However, ne increases with D2 puffing, and Zeff increases with neon puffing. Divertor structures are required to minimize the effects on the core plasma. The UEDGE fluid code, benchmarked with DIII-D data, and the DEGAS neutrals transport code are used to estimate the effectiveness of divertor configurations; slots reduce the core ionization more than baffles. The overall divertor shape is set by confinement studies which indicate that high triangularity (δ ∼0.8) is important for high τE VH-modes. Results from engineering feasibility studies, including diagnostic access, will be presented

  8. Plasma density control with ergodic divertor on Tore Supra

    International Nuclear Information System (INIS)

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

  9. 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.

  10. 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

  11. ELM induced divertor heat loads on TCV

    Czech Academy of Sciences Publication Activity Database

    Marki, J.; Pitts, R. A.; Horáček, Jan; Turri, G.; Tskhakaya, D.; TCV Team, T.

    Madrid: Centro de Investigaciones Energética Medioambiental y Tecnológica (CIEMAT), 2008. P1-41-P1-41. ISBN N. [International Conference on Plasma Surface Interactions ,PSI 18/18th./. 26.05.2008-30.05.2008, Toledo] Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak TCV * divertor heat load * ELM Subject RIV: BL - Plasma and Gas Discharge Physics http://psi2008.ciemat.es/documents/Book_of_abstracts3.pdf

  12. ELM induced divertor heat loads on TCV

    Czech Academy of Sciences Publication Activity Database

    Marki, J.; Pitts, R. A.; Horáček, Jan; Tskhakaya, D.; TCV, team.

    309-391, - (2009), s. 801-805. ISSN 0022-3115. [International Conference on Plasma-Surface Interactions in Controlled Fusion Devices/18th./. Toledo, 26.05.2008-30.5.2008] Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak TCV * divertor heat load * ELM * EVOLUTION * JET Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.933, year: 2009 http://dx.doi.org/10.1016/j.jnucmat.2009.01.212

  13. Influence of helium puff on divertor asymmetry in experimental advanced superconducting tokamak

    DEFF Research Database (Denmark)

    Liu, S. C.; Guo, H. Y.; Xu, G. S.; Wang, L.; Wang, H. Q.; Ding, R.; Duan, Y. M.; Gan, K. F.; Shao, L. M.; Chen, L.; Yan, Ning; Zhang, W.; Chen, R.; Xiong, H.; Ding, S.; Hu, G. H.; Liu, Y. L.; Zhao, N.; Li, Y. L.; Gao, X.

    2014-01-01

    Divertor asymmetries with helium puffing are investigated in various divertor configurations on Experimental Advanced Superconducting Tokamak (EAST). The outer divertor electron temperature decreases significantly during the gas injection at the outer midplane. As soon as the gas is injected into...

  14. 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.

  15. Plasma diagnostics for the DIII-D divertor upgrade

    International Nuclear Information System (INIS)

    The DIII-D tokamak is being upgraded to allow for divertor biasing, baffling, and pumping experiments. This paper gives an overview of the new diagnostics added to DIII-D as part of this Advanced Divertor Program. They include tile current monitors, fast reciprocating Langmuir probes, a fixed probe array in the divertor, fast neutral pressure gauges, and Hα measurements with TV cameras and fiber optics coupled to a high resolution spectrometer. 9 refs

  16. 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.

  17. Divertor materials evaluation system (DiMES)

    International Nuclear Information System (INIS)

    The mission of the Divertor Materials Evaluation System (DiMES) in DIII-D is to establish an integrated data base from measurements in the divertor of a tokamak in order to address some of the ITER and fusion power reactor plasma material interaction issues. Carbon and metal coatings of Be, W, V, and Mo were exposed to the steady-state outer strike point on DIII-D for 4--18 s. These short exposure times ensure controlled exposure conditions, and the extensive arrays of DIII-D divertor diagnostics provide a well-characterized plasma for modeling efforts. Post-exposure analysis provides a direct measure of surface material erosion rates and the amount of retained deuterium. For carbon, these results match closely with the results of accumulated carbon deposition and erosion, and the corresponding deuterium retention of long term exposure tiles in DIII-D. Under the carbon-contaminated background plasma of DIII-D, metal coatings of Be, V, Mo, and W were exposed to the steady-state outer strike point under ELMing and ELM-free H-mode discharges. The rate of material erosion and deuterium retention were measured. As expected, W shows the lowest erosion rate at 0.1 mm/s and the lowest deuterium uptake of 2 x 1020/m2

  18. Divertor geometry optimization for ASDEX Upgrade

    International Nuclear Information System (INIS)

    One of the critical questions to be solved for ITER (or any other reactor) is the power exhaust problem (compatible with particle exhaust). Optimized divertors have to be tested in existing geometries based mainly on the idea of closing them very efficiently to the main chamber and, by the choice of the plate and baffle geometry, positively influencing the flow pattern of hydrogen favoring good impurity entrainment. Also, for ASDEX Upgrade there is an experimental necessity for an improved divertor due to the increased heating power (24 MW will be available from 1997 on compared to the present 18 MW). We present the optimization strategy for the divertor II of ASDEX Upgrade, using elaborate numerical models and codes (B2-Eirene) as well as simple models. We start with the choice of a proper target plate geometry, and then further discuss how main chamber and private flux baffling will be done, and how this affects neutral recirculation pattern and pumping properties. For the final configuration the impurity entrainment properties are analyzed. (orig.)

  19. UEDGE modeling of the effect of divertor modifications on divertor performance

    International Nuclear Information System (INIS)

    The DIII'D upper divertor will be modified in late 1999 by installing a continuous dome in the private flux region with an independent pumping capability for the inner strike point. A ''bump'' on the inner cylinder also has been considered to enhance impurity and neutral control. Using the UEDGE code, we have examined the effect of this dome and the ''bump'' on core ionization and core impurity content. For typical parameters, results indicate that the planned divertor modifications enable detachment at higher heating power and the ''fueling efficiency'' (ratio of core neutral ionization rate to total divertor ion current) decreases, however, the core carbon content increases. The inner ''bump'' does enhance ''fueling efficiency'' compared to the private flux dome alone, but it does not reduce the increased core impurity content

  20. 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

  1. Towards a physics-integrated view on divertor pumping

    International Nuclear Information System (INIS)

    Highlights: • Physics-integrated design approaches are to be preferred over approaches based on simple requirement lists. • A physics-integrated assessment is presented for the divertor vacuum pumping system based on detachment onset conditions for the divertor. • This approach considers density dependent pump albedo to reflect the effects of gas recycling at the divertor and the changes in flow regime with density. • A comparison with DEMO indicates that the divertor pumping system for a pulsed DEMO scales less than linearly with fusion power. - Abstract: One key requirement to design the inner fuel cycle of a divertor tokamak is defined by the torus vessel gas throughput and composition, and the sub-divertor neutral pressure at which the exhaust gas has to be pumped. This paper illustrates how divertor physics aspects can be translated to requirements on the divertor vacuum pumping system. An example workflow is presented that links the realization of detachment conditions with the sub-divertor neutral gas flow patterns in order to determine the appropriate number of torus vacuum pumps. For the example case of a fusion DEMO size machine, it was found that 7 actively pumping cryopumps (ITER-type) are necessary to handle the gas throughput that is needed to manage the heat flux and densities related to detachment onset

  2. 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)

  3. 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.

  4. Comparative study of divertor and limiter concepts in FER

    International Nuclear Information System (INIS)

    Comparative study of engineering features is carried out for divertor and pumped limiter reactor concepts. These concepts are: double null divertor as a reference concept of FER, single null divertor, single pumped limiter with medium edge temperature and single pumped limiter with low edge temperature. Plasma parameters of these concepts are determined by maintaining plasma confinement performance. It is found that the double null divertor is the least favorable; and the medium edge temperature limiter is the most favorable from most of the engineering standpoints. (author)

  5. 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.

  6. 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.

  7. Plasma flow in the DIII-D divertor

    International Nuclear Information System (INIS)

    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

  8. Divertor target shape optimization in realistic edge plasma geometry

    International Nuclear Information System (INIS)

    Tokamak divertor design for next-step fusion reactors heavily relies on numerical simulations of the plasma edge. Currently, the design process is mainly done in a forward approach, where the designer is strongly guided by his experience and physical intuition in proposing divertor shapes, which are then thoroughly assessed by numerical computations. On the other hand, automated design methods based on optimization have proven very successful in the related field of aerodynamic design. By recasting design objectives and constraints into the framework of a mathematical optimization problem, efficient forward-adjoint based algorithms can be used to automatically compute the divertor shape which performs the best with respect to the selected edge plasma model and design criteria. In the past years, we have extended these methods to automated divertor target shape design, using somewhat simplified edge plasma models and geometries. In this paper, we build on and extend previous work to apply these shape optimization methods for the first time in more realistic, single null edge plasma and divertor geometry, as commonly used in current divertor design studies. In a case study with JET-like parameters, we show that the so-called one-shot method is very effective is solving divertor target design problems. Furthermore, by detailed shape sensitivity analysis we demonstrate that the development of the method already at the present state provides physically plausible trends, allowing to achieve a divertor design with an almost perfectly uniform power load for our particular choice of edge plasma model and design criteria. (paper)

  9. 2D modelling and assessment of divertor performance for ITER

    International Nuclear Information System (INIS)

    The results of the ITER divertor modelling performed during the EDA are summarised in the paper. Studies on the operating window and optimisation of the divertor geometry are presented together with preliminary results on the start-up limiter performance. The issue of model validation against the experimental data which is crucial for extrapolation to ITER is also addressed. (author)

  10. 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

  11. He-cooled demo divertor: Design verification testing against mechanical impact loads

    International Nuclear Information System (INIS)

    A modular He-cooled divertor concept for DEMO has been developed at Karlsruhe Institute of Technology. The design goal is to achieve a DEMO-relevant high heat flux of 10 MW/m2. The reference design HEMJ (He-cooled modular divertor with multiple-jet cooling) uses small finger modules, which consist of a tungsten tile and a thimble made of tungsten alloy. Both components are connected by soldering. They are cooled by helium gas (10 MPa, 634 °C) impinging directly onto the inner heated surface of the thimble. One of the most difficult to predict incident events is the disruption that may damage the divertor structure by an extraordinary impact loading. This danger is particularly acute by the brittle property of the tungsten material, which is generally characterized by the DBTT. In this paper an estimate of the mechanical impact loading induced by electro-magnetic forces during a disruption and an appropriate experimental setup are outlined, and the test results discussed.

  12. Recent results from divertor and sol studies at JET

    International Nuclear Information System (INIS)

    Recent progress in the study of divertor and scrape-off layer plasma (SOL) phenomena in JET is reviewed. Up to the present time, three pumped divertors (Mark I, Mark IIA/AP and Mark IIGB) have been installed and exploited under reactor relevant conditions. With increased divertor closure, it is found that the particle exhaust rate has increased and neutral compression factors of >100 are obtained with the Mark IIGB divertor. Helium enrichment factors of >0.2 are measured under a wide range of conditions and satisfy the minimum requirements for ITER. Fast infrared camera measurements show broad deposition profiles during type I ELMs and energy densities of ∼0.12MJm-2. During the recent D-T experiments, the codeposition of tritium on cold shadowed surfaces in the inner divertor has been identified as an important form of long-term tritium retention. This has serious implications for the divertor design and tritium inventory in a next-step tokamak. Core plasma purity has not improved with enhanced divertor closure or decreased main chamber neutral pressure. Studies of the chemical sputtering yield have shown a dependence on surface temperature and hydrogen isotope. This accounts for the observation of increased impurity production and lower disruptive density limits in Mark II (at 500K) compared to Mark I (at 300K). Significant progress has been made in the study of divertor detachment, and volume recombination has been spectroscopically identified. With increasing isotope mass, detachment and the disruptive density limit occur at lower main plasma density as predicted by the EDGE2D/NIMBUS codes. Using differential gas fuelling in the Mark IIGB divertor, it has been possible to modify the in-out asymmetry of the divertor plasma for the first time. (author)

  13. Recent results from divertor and SOL studies at JET

    International Nuclear Information System (INIS)

    Recent progress in the study of divertor and scrape-off layer plasma (SOL) phenomena in JET is reviewed. Up to the present time, three pumped divertors (Mark I, Mark IIA/AP and Mark IIGB) have been installed and exploited under reactor relevant conditions. With increased divertor closure, it is found that the particle exhaust rate has increased and neutral compression factors of >100 are obtained with the Mark IIGB divertor. Helium enrichment factors of >0.2 are measured under a wide range of conditions and satisfy the minimum requirements for ITER. Fast infra-red camera measurements show broad deposition profiles during type I ELMs and energy densities of ∼0.12MJm-2. During the recent D-T experiments, the codeposition of tritium on cold shadowed surfaces in the inner divertor has been identified as an important form of long-term tritium retention. This has serious implications for the divertor design and tritium inventory in a next-step tokamak. Core plasma purity has not improved with enhanced divertor closure or decreased main chamber neutral pressure. Studies of the chemical sputtering yield have shown a dependence on surface temperature and hydrogen isotope. This accounts for the observation of increased impurity production and lower disruptive density limits in Mark II (at 500K) compared to Mark I (at 300K). Significant progress has been made in the study of divertor detachment, and volume recombination has been spectroscopically identified. With increasing isotope mass, detachment and the disruptive density limit occur at lower main plasma density as predicted by the EDGE2D/NIMBUS codes. Using differential gas fuelling in the Mark IIGB divertor, it has been possible to modify the in-out asymmetry of the divertor plasma for the first time. (author)

  14. 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...

  15. First measurements of electron temperature and density with divertor Thomson Scattering in radiative divertor discharges on DIII-D

    International Nuclear Information System (INIS)

    We have obtained the first measurements of ne and Te in the DIII-D divertor region with a multi-pulse (20 Hz) Divertor Thomson Scattering (DTS) system. Eight measurement locations are distributed vertically up to 21 cm above the divertor plate. Two-dimensional distributions have been obtained by sweeping the divertor plasma across the DTS measurement location. Several operating modes have been studied, including ohmic, L-mode, Elming H-mode, and Radiative Divertor operation with puffing of D2 and impurities. Mapping of the data to either the (Lpol, φ) or (R, Z) planes with the EFIT equilibrium is used to analyze the 2D profiles. We find that in ELMing H-mode: ne, Te, and Pe are relatively constant along field lines from the X-point to the divertor plate, especially near the separatrix field line. With D2 puffing, the DTS profiles indicate that Te in a large part of divertor region below the X-point is dramatically reduced from ∼30-40 eV in ELMing H-mode to 1-2 eV. This results in a fairly uniform low-Te divertor, with an increased electron density in the range of 2 to 4 x 1020 m-3. Detailed comparisons of the spatial profiles of ne, Te, and electron pressure Pe, are presented for several operating modes. In addition, these data are compared with initial calculations from the UEDGE fluid code

  16. Results from Radiating Divertor Experiments with RMP ELM Suppression

    International Nuclear Information System (INIS)

    Full text: The successful integration of ELM suppression using resonant magnetic perturbations (RMPs) with puff-and-pump radiating divertor operation is demonstrated. In addition, because higher gas injection rates are needed to maintain plasma density after the RMP coils have been activated, a radiating divertor with resonant magnetic perturbation (RMP) produces considerably higher levels of radiated power from the divertor and scrape-off layer (SOL)/edge plasma regions than comparable non-RMP discharges at the same density. The radiating divertor has long been proposed as a reliable way to moderate steady heat flux at the divertor targets. Studies at DIII-D have demonstrated that RMPs are effective in suppressing ELMs and thus might be an attractive way to deal with the transient ELM-related heat flux problems expected for ITER. However, it was not clear as to whether RMP-based ELM suppression and radiating divertor scenarios were compatible. Recent DIII-D experiments comprise our first attempts to assess this compatibility by directly comparing the behaviors of injected 'seed' argon impurities in RMP and non-RMP puff-and-pump environments and by identifying issues that might limit the use of RMP ELM suppression with a radiating divertor approach. In the puff-and-pump scenarios used, argon was injected in the private flux region of a single-null magnetic configuration near the outer divertor target, while plasma flows into the divertor were enhanced by a combination of particle pumping near the outer divertor target and deuterium gas puffing upstream of the divertor targets. Differences in argon accumulation in the main plasma between RMP ELM-suppressed and similar non-RMP ELMing H-mode plasmas were relatively small, typically less than 20%. The core concentration of argon decreased as the deuterium gas puff rate was raised in RMP and non-RMP cases, suggesting that the detailed UEDGE analysis reported previously [1] for non-RMP divertor and SOL radiating divertor

  17. 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.

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

    International Nuclear Information System (INIS)

    In this paper, detailed characteristics and controllability of plasmas emitted from the end-cell of the GAMMA 10 tandem mirror are described from the viewpoint of divertor simulation studies. The energy analysis of ion flux by using end-loss ion energy analyzer (ELIEA) proved that the obtained high ion temperature (100 - 400 eV) was comparable to SOL plasma parameters in toroidal devices and was controlled by changing the ICRF power. Parallel ion temperature Ti∥ determined from the probe and calorimeter shows a linear relationship with the ICRF power in the central-cell and agrees with the results of ELIEA. Additional ICRF heating revealed a significant enhancement of particle flux, which indicated an effectiveness of additional plasma heating in adjacent cells toward the improvement of the performance. Superimposing the ECH pulse of 380 kW, 5 ms attained the maximum heat-flux more than 10 MW/m2 on axis. This value comes up to the heat-load of the divertor plate of ITER, which gives a clear prospect of generating the required heat density for divertor studies by building up heating systems to the end-mirror cell. Initial results of plasma irradiation experiment and construction of new divertor module are also described. (author)

  19. 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.

  20. 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.

  1. A super-cusp divertor configuration for tokamaks

    International Nuclear Information System (INIS)

    Our study demonstrates a remarkable flexibility of advanced divertor configurations created with the remote poloidal field coils. The emphasis here is on the configurations with three poloidal field nulls in the divertor area. We are seeking the structures where all three nulls lie on the same separatrix, thereby creating two zones of a very strong flux expansion, as envisaged in the concept of Takase's cusp divertor. It turns out that the set of remote coils can produce a cusp divertor, with additional advantages of: (i) a large stand-off distance between the divertor and the coils and (ii) a thorough control that these coils exert over the fine features of the configuration. In reference to these additional favourable properties acquired by the cusp divertor, the resulting configuration could be called 'a super-cusp'. General geometrical features of the three-null configurations produced by remote coils are described. Furthermore, issues on the way to practical applications include the need for a more sophisticated control system and possible constraints related to excessively high currents in the divertor coils

  2. 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

  3. 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

  4. A review of ELMs in divertor tokamaks

    International Nuclear Information System (INIS)

    This paper reviews what is known about edge localized modes (ELMs), with an emphasis on their effect on the scrape-off layer and divertor plasmas. ELM effects have been measured in the ASDEX-U, C-Mod, COMPASS-D, DIII-D, JET, JFT-2M,JT-60U, and TCV tokamaks and are reported here. At least three types of ELMs have been identified and their salient features determined. Type-1 giant ELMs can cause the sudden loss of up to 10-15% of the plasma stored energy but their amplitude (ΔW/W) does not increase with increasing power. Type- 3 ELMs are observed near the H-mode power threshold and produce small energy dumps (1-3% of the stored energy). All ELMs increase the scrape- off layer plasma and produce particle fluxes on the divertor targets which are as much as ten times larger that the quiescent phase between ELMs. The divertor heat pulse is largest on the inner target, unlike that of L-Mode or quiescent H-mode; some tokamaks report radial structure in the heat flux profile which is suggestive of islands or helical structures. The power scaling of Type-1 ELM amplitude and frequency have been measured in several tokamaks and has recently been applied to predictions of the ELM Size in ITER. Concern over the expected ELM amplitude has led to a number of experiments aimed at demonstrating active control of ELMs. Impurity gas injection with feedback control on the radiation loss in ASDEX-U suggests that a promising mode of operation (the CDH-mode) with a very small type-3 ELMs can be maintained with heating power sell above the H-mode threshold, where giant type-1 ELMs can be maintained with heating power well above the H-mode threshold, where Giant type-1 ELMs are normally observed. While ELMs have many potential negative effects, the beneficial effect of ELMs in providing density control and limiting the core plasma impurity content in high confinement H- mode discharges should not be overlooked

  5. 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.)

  6. 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.

  7. 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

  8. Two-dimensional divertor modeling and scaling laws

    International Nuclear Information System (INIS)

    Two-dimensional numerical models of divertors contain large numbers of dimensionless parameters that must be varied to investigate all operating regimes of interest. To simplify the task and gain insight into divertor operation, we employ similarity techniques to investigate whether model systems of equations plus boundary conditions in the steady state admit scaling transformations that lead to useful divertor similarity scaling laws. A short mean free path neutral-plasma model of the divertor region below the x-point is adopted in which all perpendicular transport is due to the neutrals. We illustrate how the results can be used to benchmark large computer simulations by employing a modified version of UEDGE which contains a neutral fluid model. (orig.)

  9. 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.)

  10. Compatibility of detached divertor operation with robust edge pedestal performance

    Energy Technology Data Exchange (ETDEWEB)

    Leonard, A.W., E-mail: leonard@fusion.gat.com [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Makowski, M.A.; McLean, A.G. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Osborne, T.H.; Snyder, P.B. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States)

    2015-08-15

    The compatibility of detached radiative divertor operation with a robust H-mode pedestal is examined in DIII-D. A density scan produced low temperature plasmas at the divertor target, T{sub e} ⩽ 2 eV, with high radiation leading to a factor of ⩾4 drop in peak divertor heat flux. The cold radiative plasma was confined to the divertor and did not extend across the separatrix in X-point region. A robust H-mode pedestal was maintained with a small degradation in pedestal pressure at the highest densities. The response of the pedestal pressure to increasing density is reproduced by the EPED pedestal model. However, agreement of the EPED model with experiment at high density requires an assumption of reduced diamagnetic stabilization of edge Peeling–Ballooning modes.

  11. Numerical analysis of divertor plasma for demo-CREST

    International Nuclear Information System (INIS)

    The numerical analysis of the demonstration fusion reactor Demo-CREST has been carried out; this analysis focuses on impurity seeding. Several design activities for DEMO have been carried out; however, its detailed divertor plasma analysis remains to be carried out. Therefore, in this study, we discuss the possibility of neon puffing in demo-CREST to decrease the power load to the divertor plate by using the B2-EIRENE code. It has been shown that the radiation power loss by neon increases with upstream plasma density and that the peak power load to the divertor plate comes close to the allowable level by using the preliminary divertor configuration (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  12. Plasma detachment with molecular processes in divertor plasmas

    International Nuclear Information System (INIS)

    Molecular processes in detached recombining plasmas are briefly reviewed. Several reactions with vibrationally excited hydrogen molecule related to recombination processes are described. Experimental evidence of molecular activated recombination observed in a linear divertor plasma simulator is also shown. (author)

  13. 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.

  14. Transport and divertor studies in the FM-1 spherator

    International Nuclear Information System (INIS)

    Fundamental problems of toroidal fusion devices have been investigated in the FM-1 Spherator. These subjects include the transport due to drift wave turbulence in the trapped electron regime, poloidal divertor and impurities, and lower hybrid heating. (auth)

  15. 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.

  16. 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

  17. 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.

  18. 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

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

    International Nuclear Information System (INIS)

    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 λ < 450 nm) to prevent the transmission loss in fiber and in the diagnostic room (for λ ≥ 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 γ-ray irradiation effect on windows, a calibration method, an alignment method, a remote handling method and a data acquisition method are considered. (author)

  20. 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)

  1. 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)

  2. Diagnostics for the DIII-D radiative divertor

    International Nuclear Information System (INIS)

    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

  3. Nuclear analysis of the ITER full-tungsten divertor

    Energy Technology Data Exchange (ETDEWEB)

    Villari, R., E-mail: rosaria.villari@enea.it [ENEA Fusion Division, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Barabash, V.; Escourbiac, F.; Ferrand, L.; Hirai, T.; Komarov, V.; Loughlin, M.; Merola, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Moro, F. [ENEA Fusion Division, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Petrizzi, L. [IAEA Representative at OECD Nuclear Energy Agency, 92130 Issy-les-Moulinaux (France); Podda, S. [ENEA Fusion Division, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Polunovsky, E. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Brolatti, G. [ENEA Fusion Division, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy)

    2013-10-15

    Highlights: • 3D nuclear analysis of the last design of ITER full-W divertor. • Calculate nuclear heating, damage and he-production in divertor components. • Evaluate impact on design and maintenance of the system. -- Abstract: This paper presents the nuclear analysis performed for the ITER full-tungsten divertor using the MCNP-5 Monte Carlo Code in a 3-D geometry. A detailed model of the geometry of the divertor based on the last design specifications has been integrated into the latest ITER MCNP model. Nuclear heating, damage and helium production have been calculated. The presented results are consistent with recent analysis performed with ATTILA code and with the previous ones with MCNP. The shielding capabilities of the last design are reduced in comparison with the design of 2004 but a negligible impact is expected. The reweldability of radial pipes of the cassette body (CB) is not a concern even assuming irradiation during the whole ITER lifetime. The reweldability of pipes for the refurbishment of the CB with new plasma facing components depends on maintenance scenario. It is not expected that last 2012 divertor design leads to an appreciable increase of the nuclear loads on vacuum vessel and toroidal field coils, since the relative contribution of the divertor region remains low.

  4. Multi-fluid modeling of low-recycling divertor regimes

    International Nuclear Information System (INIS)

    The low-recycling regimes of divertor operation in a single-null NSTX magnetic configuration are studied using computer simulations with the edge plasma transport code UEDGE. The edge plasma transport properties pertinent to the low-recycling regimes are demonstrated. These include the flux-limited character of the parallel heat transport and the high plasma temperatures with the flattened profiles in the scrape-off-layer. It is shown that to maintain the balance of particle fluxes at the core interface the deuterium gas puffing rate should increase as the divertor recycling coefficient decreases. The radial profiles of the heat load to the outer divertor plate, the upstream radial plasma profiles, and the effects of the cross-field plasma transport in the low-recycling regimes are discussed. It is also shown that recycling of lithium impurities evaporating from the divertor plate at high surface temperatures can reverse the low-recycling divertor operational regime to the high-recycling one and may cause thermal instability of the divertor plate (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  5. 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.

  6. 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

  7. Application of best-fit survey techniques throughout design, manufacturing and installation of the MKII divertor at JET

    International Nuclear Information System (INIS)

    The precise installation and alignment of large components in an activated and beryllium contaminated fusion device is a problem which must be faced in JET as well as future devices such as ITER. To guarantee the successful alignment of the MKII Divertor in JET it was essential that, early in the design phase, realistic manufacturing and installation tolerances and restrictions were identified and considered. The main components of the MKII divertor structure are an inner and outer ring mounted on a base plate. The overall diameter of the assembly is 6m and is dismantled into 24 sub-assemblies for installation. The structure must be installed very accurately whilst wearing full pressurized suits. As the other major in-vessel components remain unchanged it is important that the new divertor be installed to the same center as these components. Major considerations in the design process were the installation accuracy required, the installation method and restrictions imposed by the existing in-vessel structure. Joints between modules could only be made from one side due to access restrictions. Design of the support system had to be such that minimal modification to the existing in-vessel structure would be required. The tight tolerances necessary to ensure the mechanical integrity of the module joints were compromised by the necessity to have realistic assembly tolerances

  8. 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.

  9. 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...

  10. 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.

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

    International Nuclear Information System (INIS)

    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 (Ip=170 kA, ne=3 x 1019 m-3, BT=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

  12. 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

  13. Initial performance results of the DIII-D Divertor 2000

    International Nuclear Information System (INIS)

    A major upgrade of the DIII-D divertor, with the goal of enhancing impurity and density control and increasing the thermal pulse length limit of advanced tokamak (AT) plasmas has been successfully completed and commissioned. The integrated system that includes independent cryopumps at both the inner and the outer legs of the divertor, private flux region and outboard baffles, and improved graphite divertor armor, has been successfully applied to a variety of plasma conditions. Comparison of similar discharges before and after the upgrades show that with the new divertor the core plasma neutral source and carbon content are lower by as much as 50%. Calculations supported by preliminary infra-red (IR) camera measurements show that the new graphite armor design increases the limit on the discharge duration, due to temperature of the tile edges reaching sublimation point, by an order of magnitude. With the new system we have been able to control the density of high confinement H-mode plasmas to less than 1/3 of the Greenwald limit. It is observed that with divertor pumping during the current ramp phase the wall particle inventory and consequently the density rise after the H-mode transition can be significantly reduced

  14. 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.)

  15. Analysis of sweeping heat loads on divertor plate materials

    International Nuclear Information System (INIS)

    The heat flux on the divertor plate of a fusion reactor is probably one of the most limiting constraints on its lifetime. The current heat flux profile on the outer divertor plate of a device like ITER is highly peaked with narrow profile. The peak heat flux can be as high as 30--40 MW/m2 with full width at half maximum (FWHM) is in the order of a few centimeters. Sweeping the separatrix along the divertor plate is one of the options proposed to reduce the thermomechanical effects of this highly peaked narrow profile distribution. The effectiveness of the sweeping process is investigated parametrically for various design values. The optimum sweeping parameters of a particular heat load will depend on the design of the divertor plate as well as on the profile of such a heat load. In general, moving a highly peaked heat load results in substantial reduction of the thermomechanical effects on the divertor plate. 3 refs., 8 figs

  16. Initial Development of the NSTX-U Snowflake Divertor Control

    Science.gov (United States)

    Vail, Patrick; Kolemen, Egemen; Welander, Anders; Lanctot, Matthew

    2015-11-01

    A feedback control system has been implemented at NSTX-U for real-time detection and manipulation of snowflake divertor (SFD) magnetic configurations. The SFD is an alternative magnetic divertor concept that is characterized by a second-order null formed by two x-points in close proximity. The SFD is an attractive option for heat flux mitigation for NSTX-U in which unmitigated peak heat fluxes in standard divertor operation near 20 MW/m2 may compromise plasma-facing components. The real-time control system at NSTX-U is capable of simultaneous control of multiple SFD parameters, such as the separation between the two x-points in the divertor region and their orientation. Control of SFD configurations in NSTX-U has been simulated in TOKSYS using the upgraded sets of poloidal field coils in both the upper and lower divertor regions. Performance of the real-time control system and its effect on plasma performance will be assessed experimentally as an initial step toward the development of the SFD concept at NSTX-U. Supported by the US DOE under DE-AC02-09CH11466.

  17. A novel approach to magnetic divertor configuration design

    International Nuclear Information System (INIS)

    Divertor exhaust system design and analysis tools are crucial to evolve from experimental fusion reactors towards commercial power plants. In addition to material research and dedicated vessel geometry design, improved magnetic configurations can contribute to sustaining the diverted heat loads. Yet, computational design of the magnetic divertor is a challenging process involving a magnetic equilibrium solver, a plasma edge grid generator and a computationally demanding plasma edge simulation. In this paper, an integrated approach to efficient sensitivity calculations is discussed and applied to a set of slightly reduced divertor models. Sensitivities of target heat load performance to the shaping coil currents are directly evaluated. Using adjoint methods, the cost for a sensitivity evaluation is reduced to about two times the simulation cost of one specific configuration. Further, the use of these sensitivities in an optimal design framework is illustrated by a case with realistic Joint European Torus (JET) configurational parameters

  18. 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.

  19. 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.)

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

    International Nuclear Information System (INIS)

    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

  1. 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

  2. A novel approach to magnetic divertor configuration design

    Energy Technology Data Exchange (ETDEWEB)

    Blommaert, M., E-mail: m.blommaert@fz-juelich.de [Institute of Energy and Climate Research, Plasma Physics (IEK-4), FZ Jülich GmbH, D-52425 Jülich (Germany); Baelmans, M., E-mail: martine.baelmans@kuleuven.be [KU Leuven, Department of Mechanical Engineering, 3001 Leuven (Belgium); Dekeyser, W. [Institute of Energy and Climate Research, Plasma Physics (IEK-4), FZ Jülich GmbH, D-52425 Jülich (Germany); KU Leuven, Department of Mechanical Engineering, 3001 Leuven (Belgium); Gauger, N.R. [RWTH Aachen, Department of Mathematics and Center for Computational Engineering Science, D-52062 Aachen (Germany); Reiter, D. [Institute of Energy and Climate Research, Plasma Physics (IEK-4), FZ Jülich GmbH, D-52425 Jülich (Germany)

    2015-08-15

    Divertor exhaust system design and analysis tools are crucial to evolve from experimental fusion reactors towards commercial power plants. In addition to material research and dedicated vessel geometry design, improved magnetic configurations can contribute to sustaining the diverted heat loads. Yet, computational design of the magnetic divertor is a challenging process involving a magnetic equilibrium solver, a plasma edge grid generator and a computationally demanding plasma edge simulation. In this paper, an integrated approach to efficient sensitivity calculations is discussed and applied to a set of slightly reduced divertor models. Sensitivities of target heat load performance to the shaping coil currents are directly evaluated. Using adjoint methods, the cost for a sensitivity evaluation is reduced to about two times the simulation cost of one specific configuration. Further, the use of these sensitivities in an optimal design framework is illustrated by a case with realistic Joint European Torus (JET) configurational parameters.

  3. 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

  4. 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.

  5. Tungsten Divertor Target Technology and Test Facilities Development

    International Nuclear Information System (INIS)

    Full text: Tungsten divertor target technology development is in progress at IPR for water-cooled divertors of ITER-like tokamak. Test mock-ups are fabricated using tungsten materials in macro-brush as well as mono-block fashion. Vacuum brazing technique is used for macro-brush fabrication whereas high pressure high temperature diffusion bonding technique is used for mono-block fabrication. Experimental facilities are also being set-up at IPR for Non-destructive testing and high heat flux testing of divertor targets. Present paper describes recent results on high heat flux testing of the test mock-ups and briefly mention about some of the experimental test facilities being set-up at IPR. (author)

  6. 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

  7. 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)

  8. ITER full tungsten divertor qualification program and progress

    International Nuclear Information System (INIS)

    The full tungsten divertor qualification program was defined for the R and D activity in domestic agencies. The qualification program consists of two steps: (i) technology development and validation and (ii) a full-scale demonstration. Small-scale mock-ups were manufactured in Japanese and European industries and delivered to the ITER divertor test facility in Russia for high heat flux testing. In parallel activity to the qualification program, both domestic agencies demonstrated that W monoblock technologies withstanding up to 20 MW m−2 were available. (paper)

  9. Divertor development for a future fusion power plant

    International Nuclear Information System (INIS)

    Nuclear fusion is considered as a future source of sustainable energy supply. In the first chapter, the physical principle of magnetic plasma confinement, and the function of a tokamak are described. Since the discovery of the H-mode in ASDEX experiment ''Divertor I'' in 1982, the divertor has been an integral part of all modern tokamaks and stellarators, not least the ITER machine. The goal of this work is to develop a feasible divertor design for a fusion power plant to be built after ITER. This task is particularly challenging because a fusion power plant formulates much greater demands on the structural material and the design than ITER in terms of neutron wall load and radiation. First several divertor concepts proposed in the literature e.g. the Power Plant Conceptual Study (PPCS) using different coolants are reviewed and analyzed with respect to their performance. As a result helium cooled divertor concept exhibited the best potential to come up to the highest safety requirements and therefore has been chosen for the design process. From the third chapter the necessary steps towards this goal are described. First, the boundary conditions for the arrangement of a divertor with respect to the fusion plasma are discussed, as this determines the main thermal and neutronic load parameters. Based on the loads material selection criteria are inherently formulated. In the next step, the reference design is defined in accordance with the established functional design specifications. The developed concept is of modular nature and consists of cooling fingers of tungsten using an impingement cooling in order to achieve a heat dissipation of 10 MW/m2. In the next step, the design was subjected to the thermal-hydraulic and thermo-mechanical calculations in order to analyze and improve the performance and the manufacturing technologies. Based on these results, a prototype was produced and experimentally tested on their cooling capacity, their thermo-cyclic loading behavior

  10. An analytic model for flow reversal in divertor plasmas

    International Nuclear Information System (INIS)

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

  11. PARAMETRIC SENSITIVITY STUDY OF A FUSION REACTOR DIVERTOR COOLING FINGER

    OpenAIRE

    Martin, Oliver; SIMONOVSKI IGOR

    2012-01-01

    In this paper the results of coupled thermal-mechanical Finite Element (FE) analysis on the design of a fusion reactor divertor cooling finger are presented. Beside its main purpose, to remove alpha particles, helium and other impurities from the plasma stream, a divertor has to remove approximately 15% of the total thermal power of the fusion reactor. The aim of the analysis is to assess the influence of a number of physical properties of the brazing layer (BL) of the cooling finger on the o...

  12. 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

  13. 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.

  14. 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)

  15. 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.

  16. 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.

  17. Three-dimensional neutronics and shielding analyses for the ITER divertor

    International Nuclear Information System (INIS)

    3-D neutronics and shielding analyses have been performed for the divertor region of the ITER interim design. The peak neutron wall loading in the divertor region is 0.6 MW/m2 at the divertor cassette dome. The total nuclear heating in the 60 divertor cassettes is 102.4 MW. The peak helium production in the VV behind the pumping ducts is 0.5 He appm/FPY implying that rewelding might be feasible. The total nuclear heating in the parts of the TF coils in the divertor region is only 2.1 kW. 5 refs., 4 figs., 5 tabs

  18. 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

  19. 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.)

  20. 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.).

  1. Energy and particle control characteristics of the ASDEX Upgrade 'LYRA' divertor

    International Nuclear Information System (INIS)

    In 1997 the new 'LYRA' divertor went into operation at ASDEX Upgrade and the neutral beam heating power was increased to 20 MW by installation of a second injector. This leads to the relatively high value of P/R of 12 MW/m. It has been shown that the ASDEX Upgrade LYRA divertor is capable of handling such high heating powers. Measurements presented in this paper reveal a reduction of the maximum heat flux in the LYRA divertor by more than a factor of two compared to the open Divertor I. This reduction is caused by radiative losses inside the divertor region. Carbon radiation cools the divertor plasma down to a few eV where hydrogen radiation losses become significant. They are increased due to an effective reflection of neutrals into the hot separatrix region. B2-Eirene modelling of the performed experiments supports the experimental findings and refines the understanding of loss processes in the divertor region. (and others)

  2. 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. PMID:26931846

  3. Design study on divertor plates of Large Helical Device (LHD)

    International Nuclear Information System (INIS)

    A conceptual design has been completed for the divertor plates of the Large Helical Device (LHD, R = 3.9 m, ap = 50 ∼ 60 cm, Bh = 3 ∼ 4T/ superconducting coils of NbTi) and the detailed technical design is now in progress. The design concept and the status of research and development (R and D) programs are described. (author)

  4. Edge plasma control by local island divertor in LHD

    International Nuclear Information System (INIS)

    In the Large Helical Device (LHD) program, one of the key research issues is to enhance helical plasma performance through the edge plasma control. For the first time in the LHD program, the edge plasma control was performed with a local island divertor (LID) that is a closed divertor, utilizing an m/n 1/1 island generated externally by 20 small perturbation coils, and fundamental LID functions were demonstrated experimentally. It was found that the outward heat and particle fluxes crossing the island separatrix flow along the field lines to the backside of the divertor head, where carbon plates are placed to receive the heat and particle loads. Accordingly high efficient pumping was demonstrated, which is considered to be the key in realizing high temperature divertor operation, resulting in an improvement of energy confinement. In the present experiment, a factor of ∼1.2 improvement of the energy confinement time, τE, was observed at a magnetic axis position, Rax, of 3.75 m over the International Stellarator Scaling 95. Results of edge modeling are also presented by using the EMC3-EIRENE code. (author)

  5. 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

  6. 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)

  7. 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.

  8. 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

  9. 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

  10. Edge plasma control by local island divertor in LHD

    International Nuclear Information System (INIS)

    In the Large Helical Device (LHD) program, one of the key research issues is to enhance helical plasma performance through the edge plasma control. For the first time in the LHD program, the edge plasma control was performed with a local island divertor (LID) that is a closed divertor, utilizing an m/n=1/1 island generated externally by 20 small perturbation coils, and fundamental LID functions were demonstrated experimentally. It was found that the outward heat and particle fluxes crossing the island separatrix flow along the field lines to the backside of the divertor head, where carbon plates are placed to receive the heat and particle loads. Accordingly high efficient pumping was demonstrated, which is considered to be the key in realizing high temperature divertor operation, resulting in an improvement of energy confinement. In the present experiment, relatively good energy confinement is achieved in the high density regime at a magnetic axis position, Rax, of 3.75 m. Results of edge modelling are also presented by using the EMC3-EIRENE code. (author)

  11. 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.

  12. Spectroscopic study of JT-60U divertor plasma

    International Nuclear Information System (INIS)

    Particle behavior in the JT-60U divertor plasmas has been studied spectroscopically. Doppler profiles of the Dα line have been investigated for understanding of atomic and molecular processes in deuterium particle recycling and Dα line emission. Near the divertor plates, dissociative excitation from deuterium molecules and molecular ions plays an important role for the line emission. By investigation of spectral profiles of the He I line (667.8 nm), Doppler broadening due to elastic scattering by protons has been found. It is estimated that the penetration probability of the helium atoms from the divertor plates to the main plasma and the helium atom flux to the gap for pumping increase by 30% due to the elastic scattering. Intensity distribution of the CD band (around 430.5 nm) has been compared between the W-shaped divertor with a dome in the private flux region and the previous open one. The dome prevents the upstream transport of hydrocarbon impurity produced by chemical sputtering. (author)

  13. 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

  14. The control of divertor carbon erosion/redeposition in the DIII-D tokamak

    International Nuclear Information System (INIS)

    The DIII-D tokamak has demonstrated an operational scenario where the graphite-covered divertor is free of net erosion. Reduction of divertor carbon erosion is accomplished using a low temperature (detached) divertor plasma that eliminates physical sputtering. Likewise, the carbon source rate arising from chemical erosion is found to be very low in the detached divertor. Near strikepoint regions, the rate of carbon deposition is ∼3 cm/burn-year, with a corresponding hydrogenic codeposition rate >1kg/m2/burn-year; rates both problematic for steady-state fusion reactors. The carbon net deposition rate in the divertor is consistent with carbon arriving from the core plasma region. Carbon influx from the main wall is measured to be relatively large in the high-density detached regime and is of sufficient magnitude to account for the deposition rate in the divertor. Divertor redeposition is therefore determined by non-divertor erosion and transport. Despite the success in reducing divertor erosion on DIII-D with detachment, no significant reduction is found in the core plasma carbon density, illustrating the importance of non-divertor erosion and the complex coupling between erosion/redeposition and impurity plasma transport. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Carfora, D., E-mail: dario.carfora@gmail.com [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); ENEA/CREATE/University of Naples Federico II, 80125 Naples (Italy); Di Gironimo, G. [ENEA/CREATE/University of Naples Federico II, 80125 Naples (Italy); Järvenpää, J. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Huhtala, K. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T.; Siuko, M. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland)

    2014-11-15

    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.

  16. 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

  17. 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).

  18. 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).

  19. 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).

  20. Latest results and future plans from JT-60U and JFT-2M divertor research

    International Nuclear Information System (INIS)

    The JT-60U divertor research program is designed to address key issues for ITER physics R and D. Radiative cooling divertor is investigated to reduce the heat load on the divertor plates in L- and ELMy H-mode discharges. The mechanism of radiation losses is spectroscopically studied with visible spectroscopy. The peak heat flux density reaches 300-400 MW/m2 and most of the power is deposited within a few milliseconds due to ELMs. In high density plasmas, the physically sputtered impurities from target plates are reduced and chemical sputtering by neutral particles which strike the divertor plates in the private region becomes dominant. Helium exhaust from the core plasma is observed due to wall pumping caused by Solid Target Boronization. A new W-shaped pumped divertor in JT-60U providing a dense and cold divertor plasma is the basis to develop the integrated performance in quasi-steady state. Divertor configuration was modified an open divertor to a closed divertor and initial results on the closed divertor has been obtained in JFT-2M. (author)

  1. Modeling of divertor geometry effects in China fusion engineering testing reactor by SOLPS/B2-Eirene

    International Nuclear Information System (INIS)

    The China Fusion Engineering Testing Reactor (CFETR) is currently under design. The SOLPS/B2-Eirene code package is utilized for the design and optimization of the divertor geometry for CFETR. Detailed modeling is carried out for an ITER-like divertor configuration and one with relatively open inner divertor structure, to assess, in particular, peak power loading on the divertor target, which is a key issue for the operation of a next-step fusion machine, such as ITER and CFETR. As expected, the divertor peak heat flux greatly exceeds the maximum steady-state heat load of 10 MW/m2, which is a limit dictated by engineering, for both divertor configurations with a wide range of edge plasma conditions. Ar puffing is effective at reducing divertor peak heat fluxes below 10 MW/m2 even at relatively low densities for both cases, favoring the divertor configuration with more open inner divertor structure

  2. Investigation of tokamak solid-divertor target options

    International Nuclear Information System (INIS)

    Analysis of survival constraints on the design of solid targets for tokamak bundle divertors is presented. Previous target design efforts are reviewed. Considerations of heat removal, surface erosion, and fatigue life are included in a generalized design window methodology which facilitates target selection. Using subcooled water as coolant, eight possible target materials are evaluated for use in tubular and plate targets as substrates, coatings, and claddings. Subject to the severe environment of the tokamak plasma, the most promising conventional designs are identified. A thermally bonded, mechanically unbonded laminated design is proposed and evaluated as a target design well suited to the divertor target environment. Due to fatigue and sputtering erosion this configuration has limited life, but appears to constitute an upper bound for the capabilities of a solid target design. Needs for experimental work are identified

  3. 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.

  4. Alignment systems for pumped divertor installation at JET

    International Nuclear Information System (INIS)

    The installation of the JET Pumped Divertor, designed to study impurity control, has recently been completed. The main components are four magnetic coils, forty eight divertor plate assemblies, one toroidal cryopump, eight ICRH antennae, sixteen inner wall guard limiters and twelve poloidal limiters. Due to the high thermal loads, accurate positioning of plasma facing components to the magnetic centre of the machine was a major requirement. Typically alignment within ± 2 mm was required, with steps between tiles on a component being controlled to ± 0.25 mm. In some cases a set of components was required to be concentric, while also lying within a narrow band defined by the position of some other components. A typical example of this was the positioning of the poloidal limiters, which perform the dual function of limiting the plasma and also protecting the antennae. Clearly, a measuring system accurate to better than ± 0.5 mm was required. (author) 4 refs.; 3 figs

  5. Alignment systems for pumped divertor installation at JET

    International Nuclear Information System (INIS)

    The installation of the JET Pumped Divertor, designed to study impurity control, has recently been completed. The main components are four magnetic coils, forty eight divertor plate assemblies, one toroidal cryopump, eight ICRH antennae, sixteen inner wall guard limiters and twelve poloidal limiters. Due to the high thermal loads, accurate positioning of plasma facing components to the magnetic centre of the machine was a major requirement. Typically alignment within ± 2 mm was required, with steps between tiles on a component being controlled to ± 0.25 mm. In some cases a set of components was required to be concentric, while also lying within a narrow band defined by the position of some other components. A typical example of this was the positioning of the poloidal limiters, which perform the dual function of limiting the plasma and also protecting the antennae. Clearly, a measuring system accurate to better than ± 0.5 mm was required. (orig.)

  6. Measurements of divertor impurity concentrations on DIII-D

    International Nuclear Information System (INIS)

    Carbon emissions in the DIII-D divertor during partial detachment have been measured, and the deduced radiated power and the temporal behavior of the impurity emissions from spectroscopy are in good agreement with bolometer measurements. Effective electron temperatures from line ratios for CIV (9-11 eV) and CIII (6-8 eV) are correlated with DTS measured electron temperatures to determine the spatial location of the carbon radiation zone. During PDD operation, the bulk of the divertor radiation is emitted from CIV near the X- point while deuterium radiation is strongest near the outer strikepoint. The carbon ion concentrations are in the range of 1% - 4% of the electron density

  7. 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

  8. Results of the H-mode experiments with JT-60 outer and lower divertors

    International Nuclear Information System (INIS)

    In JT-60, hydrogen H-mode experiments with outer and lower divertors were performed. In the outer divertor, H-mode were obtained, similar to the ones observed in the other lower/upper divertors. Its threshold absorbed power and electron density were 16 MW and 1.8 x 1019m-3. In the two combined heatings with NB+ICRF and NB+LHRF, H-mode discharges are also obtained. Moreover, in new configuration of lower divertor, H-mode phases without and with ELM are obtained. Typical results of the lower divertor are shown to compare the H-mode characteristics between the two configurations. Improvement of the energy confinement time in the two divertors was limited to 10 %. Analyses on ballooning/interchange instabilities were carried out with precise equlibria of JT-60. These results showed that the both modes were enough stable. (author)

  9. The control of convection by fuelling and pumping in the JET pumped divertor

    Energy Technology Data Exchange (ETDEWEB)

    Harbour, P.J.; Andrew, P.; Campbell, D.; Clement, S.; Davies, S.; Ehrenberg, J.; Erents, S.K.; Gondhalekar, A.; Gadeberg, M.; Gottardi, N.; Von Hellermann, M.; Horton, L.; Loarte, A.; Lowry, C.; Maggi, C.; McCormick, K.; O`Brien, D.; Reichle, R.; Saibene, G.; Simonini, R.; Spence, J.; Stamp, M.; Stork, D.; Taroni, A.; Vlases, G. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

    1994-07-01

    Convection from the scrape-off layer (SOL) to the divertor will control core impurities, if it retains them in a cold, dense, divertor plasma. This implies a high impurity concentration in the divertor, low at its entrance. Particle flux into the divertor entrance can be varied systematically in JET, using the new fuelling and pumping systems. The convection ratio has been estimated for various conditions of operation. Particle convection into the divertor should increase thermal convection, decreasing thermal conduction, and temperature and density gradients along the magnetic field, hence increasing the frictional force and decreasing the thermal force on impurities. Changes in convection in the SOL, caused by gaseous fuelling, have been studied, both experimentally in the JET Mk I divertor and with EDGE2/NIMBUS. 1 ref., 4 figs., 1 tab.

  10. Active control of divertor heat and particle fluxes in EAST towards advanced steady state operations

    International Nuclear Information System (INIS)

    Significant progress has been made in EAST towards advanced steady state operations by active control of divertor heat and particle fluxes. Many innovative techniques have been developed to mitigate transient ELM and stationary heat fluxes on the divertor target plates. It has been found that lower hybrid current drive (LHCD) can lead to edge plasma ergodization, striation of the stationary heat flux and lower ELM transient heat and particle fluxes. With multi-pulse supersonic molecular beam injection (SMBI) to quantitatively regulate the divertor particle flux, the divertor power footprint pattern can be actively modified. H-modes have been extended over 30 s in EAST with the divertor peak heat flux and the target temperature being controlled well below 2 MW/m2 and 250 °C, respectively, by integrating these new methods, coupled with advanced lithium wall conditioning and internal divertor pumping, along with an edge coherent mode to provide continuous particle and power exhaust

  11. Divertor heat and particle control experiments on the DIII-D tokamak

    International Nuclear Information System (INIS)

    In this paper we present a summary of recent DIII-D divertor physics activity and plans for future divertor upgrades. During the past year, DIII-D experimental effort was focused on areas of active heat and particle control and divertor target erosion studies. Using the DIII-D Advanced Divertor system we have succeeded for the first time to control the plasma density and demonstrate helium exhaust in H-mode plasmas. Divertor heat flux control by means of D2 gas puffing and impurity injection were studied separately and in, both cases up to a factor of five reduction of the divertor peak heat flux was observed. Using the DiMES sample transfer system we have obtained erosion data on various material samples in well diagnosed plasmas and compared the results with predictions of numerical models

  12. Active control of divertor heat and particle fluxes in EAST towards advanced steady state operations

    Energy Technology Data Exchange (ETDEWEB)

    Wang, L., E-mail: lwang@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Dalian University of Technology, Dalian 116024 (China); Guo, H.Y. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); General Atomics, P. O. Box 85608, San Diego, CA 92186 (United States); Li, J.; Wan, B.N.; Gong, X.Z.; Zhang, X.D.; Hu, J.S. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Liang, Y. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Association EURATOM-FZJ, D-52425 Jülich (Germany); Xu, G.S. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Zou, X.L. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Loarte, A. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Maingi, R.; Menard, J.E. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Luo, G.N.; Gao, X.; Hu, L.Q.; Gan, K.F.; Liu, S.C.; Wang, H.Q.; Chen, R. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); and others

    2015-08-15

    Significant progress has been made in EAST towards advanced steady state operations by active control of divertor heat and particle fluxes. Many innovative techniques have been developed to mitigate transient ELM and stationary heat fluxes on the divertor target plates. It has been found that lower hybrid current drive (LHCD) can lead to edge plasma ergodization, striation of the stationary heat flux and lower ELM transient heat and particle fluxes. With multi-pulse supersonic molecular beam injection (SMBI) to quantitatively regulate the divertor particle flux, the divertor power footprint pattern can be actively modified. H-modes have been extended over 30 s in EAST with the divertor peak heat flux and the target temperature being controlled well below 2 MW/m{sup 2} and 250 °C, respectively, by integrating these new methods, coupled with advanced lithium wall conditioning and internal divertor pumping, along with an edge coherent mode to provide continuous particle and power exhaust.

  13. Effect of localized gas puffing on divertor plasma behavior in EAST

    International Nuclear Information System (INIS)

    Localized D2 puffing from various divertor locations has been carried out under double null (DN) and lower single null (LSN) divertor configurations to investigate the effect of gas puff locations on the divertor behavior in ohmic L-mode discharges in EAST. Localized gas puffing from the dome has a higher fueling efficiency than that from the inner and outer targets for both DN and LSN configurations. Under the DN configuration, gas puffing from the inner target exhibits a much better fueling efficiency than that from the outer target. In contrast, the gas fueling efficiency shows little difference between the inner and outer divertor gas puff locations in the LSN configuration. In LSN, localized gas puffing from the outer divertor target tends to promote detachment at the outer target. This will be employed as a means to control heat fluxes to the outer divertor target plates for high power long pulse operations.

  14. Multiscale study on hydrogen mobility in metallic fusion divertor material

    OpenAIRE

    Heinola, Kalle

    2010-01-01

    For achieving efficient fusion energy production, the plasma-facing wall materials of the fusion reactor should ensure long time operation. In the next step fusion device, ITER, the first wall region facing the highest heat and particle load, i.e. the divertor area, will mainly consist of tiles based on tungsten. During the reactor operation, the tungsten material is slowly but inevitably saturated with tritium. Tritium is the relatively short-lived hydrogen isotope used in the fusion reactio...

  15. 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...

  16. Method of plasma impurity control without magnetic divertor

    International Nuclear Information System (INIS)

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

  17. 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

  18. 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...

  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 fueling for both deuterium and impurities. A feedback algorithm based on real time Langmuir probe measurements has been developed to monitor detachment and avoid disruptions. (authors)

  20. 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)

  1. 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

  2. Analytical calculations for impurity seeded partially detached divertor conditions

    Science.gov (United States)

    Kallenbach, A.; Bernert, M.; Dux, R.; Reimold, F.; Wischmeier, M.; ASDEX Upgrade Team

    2016-04-01

    A simple analytical formula for the impurity seeded partially detached divertor operational point has been developed using 1D modelling. The inclusion of charge exchange momentum loss terms improves the 1D modelling for ASDEX Upgrade conditions and its extrapolation to larger devices. The investigations are concentrated around a partially detached divertor working point of low heat flux and an electron temperature around 2.5 eV at the target which are required to maintain low sputtering rates at a tungsten target plate. An experimental formula for the onset of detachment by nitrogen seeding in ASDEX Upgrade is well reproduced, and predictions are given for N, Ne and Ar seeding for variable device size. Moderate deviations from a linear {{P}\\text{sep}}/R size dependence of the detachment threshold are seen in the modelling caused by upstream radiation at longer field line lengths. The presented formula allows the prediction of the neutral gas or seed impurity pressure which is required to achieve partial detachment for a given {{P}\\text{sep}} in devices with a closed divertor similar to the geometry in ASDEX Upgrade.

  3. Parametric study of FER first wall and divertor plate performance

    International Nuclear Information System (INIS)

    Thermal, mechanical, and lifetime performance of various first wall and divertor plate materials were examined over a broad range of conditions, representative of those considered for next-generation tokamaks such as FER. Candidate plasma side materials include beryllium, graphite, silicon carbide, molybdenum, tantalum, and tungsten. Copper, copper alloy C17510, austenitic stainless steel (316SS), ferritic stainless steel (HT-9), vanadium alloy V-15Cr-5Ti, and molybdenum alloy TZM were considered as candidate heat sink/structural materials. Performance was examined at heat fluxes ranging from 0.05 MW/m2 for the first wall up to 5.0 MW/m2 for the divertor plate. Ion flux, plasma edge temperature, burn time per pulse, and number of operating cycles were the other major parameters varied in this study. The analysis model used for these studies includes: (1) a thermal model; (2) a thermal stress model; (3) a disruption erosion model; (4) a sputtering erosion model; and (5) a fatique lifetime model. Results show that recommended first wall and divertor plate designs perform adequately over most of the range of conditions considered for FER design options. Thermal shock of the plasma facing material during intense disruption heating and radiation damage and temperature limitations for graphite are identified as major concerns reguiring experimental investigation. (author)

  4. 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.)

  5. Non-linear effects on neutral gas transport in divertors

    International Nuclear Information System (INIS)

    The effects of neutral particles on the condition of the plasma edge play a key role in divertor and limiter physics. In computational models they are usually treated in the linear test particle approximation. However, in some divertor concepts a large neutral gas pressure is required in the divertor chamber to provide sufficient neutral-plasma interaction in the plasma fan (momentum removal and energy dissipation) and to permit adequate pumping performance. In such regimes viscous effects in the neutral gas may become relevant. We have extended the EIRENE code to solve the Boltzmann equation with a non-linear BGK-model collision term added to its standard linear collision integrals. The linear in-elastic collision integrals are reconsidered with respect to volume recombination and momentum removal efficiency from the plasma. The numerical procedure in the EIRENE Monte Carlo code is outlined. A simple test application (Couette flow) shows that the procedure works properly. First numerical studies have been carried out and the results are discussed. (orig.)

  6. Observation of the heteroclinic tangles in the heat flux pattern of the ergodic divertor at TEXTOR

    International Nuclear Information System (INIS)

    A fine structure of open chaotic field lines, namely, a heteroclinic tangle, in the ergodic divertor has been observed by measurements of heat deposition pattern on the divertor plates at TEXTOR. Calculations show that magnetic footprints on the divertor plates are formed by open field lines coming from the plasma along narrow stripe regions called fingers. The latter are determined by the structure of stable and unstable manifolds of the outermost resonant magnetic island. This fact is confirmed by observations of the bifurcations of the heat flux pattern on the divertor plates with changing edge safety factor

  7. 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 ...

  8. Experimental studies on an axisymmetric divertor in DIVA(JFT-2a)

    International Nuclear Information System (INIS)

    DIVA(JFT-2a) is the first tokamak with an axisymmetric divertor in the world. Objectives of the experiments were i) Plasma production and confinement in a tokamak with a separatrix magnetic surface, and ii) divertor effects on radiation loss and plasma confinement. The results so far are as follows: i) The equilibrium with a separatrix magnetic surface is stable during the discharge. ii) There is an ergodic region near the separatrix magnetic surface due to non-axisymmetric magnetic perturbations. iii) The divertor reduces radiation loss and increases energy confinement time. iv) The divertor does not affect the transport process in the main plasma. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

  11. 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.

  12. 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

  13. 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)

  14. 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.

  15. Specimen alignment in an axial tensile test of thin films using direct imaging and its influence on the mechanical properties of BeCu

    International Nuclear Information System (INIS)

    This paper proposes a new system for verification of the alignment of loading fixtures and test specimens during tensile testing of thin film with a micrometer size through direct imaging. The novel and reliable image recognition system to evaluate the misalignment between the load train and the specimen axes during tensile test of thin film was developed using digital image processing technology with CCD. The decision of whether alignment of the tensile specimen is acceptable or not is based on a probabilistic analysis through the edge feature extraction of digital imaging. In order to verify the performance of the proposed system and investigate the effect of the misalignment of the specimen on tensile properties, the tensile tests were performed as displacement control in air and at room temperature for metal thin film, the beryllium copper (BeCu) alloys. In the case of the metal thin films, bending stresses caused by misalignment are insignificant because the films are easily bent during tensile tests to eliminate the bending stresses. And it was observed that little effects and scatters on tensile properties occur by stress gradient caused by twisting at in-plane misalignment, and the effects and scatters on tensile properties are insignificant at out-of-plane misalignment, in the case of the BeCu thin film.

  16. 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.

  17. Installation, features, and capabilities of the DIII-D advanced tokamak radiative divertors

    Energy Technology Data Exchange (ETDEWEB)

    Friend, M.E. E-mail: friend@fusion.gat.com; Bozek, A.S.; Baxi, C.B.; O' Neill, R.C.; Reis, E.E.; Mahdavi, M.A

    2001-10-01

    The DIII-D program has completed a series of density control and plasma core confinement experiments this past year. These experiments were designed to investigate the performance of baffled and open divertors with single-null plasmas and particle control in double-null plasmas. The experiments utilized all three of the DIII-D divertor assemblies located in the lower outer corner, the upper outer corner, and the upper inner corner of the vessel, which were installed last year. Each divertor consists of a liquid helium cryopump, a shielded protective ring, and a gas puff system. The divertors were designed to optimize pumping performance and to withstand the electromagnetic loads from both halo and toroidal, induced currents. With theoretical pumping speeds varying from 15,000 to 32,000 l/s, the cryopumps, combined with the baffle structures, collect particles and prevent them from recirculating back into the plasma core. The intent of the gas puff systems is to inject neutral gases in and around the divertors to minimize the heat flux on the divertors, minimizing the impurities generated by the excessive heating of the divertor graphite tiles. This hardware permits either single- or double-null plasma experiments and enables continued research of well confined high beta divertor plasmas with noninductive current drive, which is one of the primary research goals of DIII-D.

  18. Near-infrared spectroscopy for divertor plasma diagnosis and control in DIII-D tokamak

    International Nuclear Information System (INIS)

    New near infrared (NIR) spectroscopic measurements performed in the DIII-D tokamak divertor plasma suggest new viable diagnostic applications: divertor recycling and low-Z impurity flux measurements, a spectral survey for divertor Thomson scattering (DTS) diagnostic, and Te monitoring for divertor detachment control. A commercial 0.3 m spectrometer coupled to an imaging lens via optical fiber and a InGaAs 1024 pixel array detector enabled deuterium and impurity emission measurements in the range 800–2300 nm. The first full NIR survey identified D, He, B, Li, C, N, O, Ne lines and provided plasma Te, ne estimates from deuterium Paschen and Brackett series intensity and Stark line broadening analysis. The range 1.000–1.060 mm was surveyed in high-density and neon seeded divertor plasmas for spectral background emission studies for λ = 1.064 μm laser-based DTS development. The ratio of adjacent deuterium Paschen-α and Brackett Br9 lines in recombining divertor plasmas is studied for divertor Te monitoring aimed at divertor detachment real-time feedback control

  19. Near-infrared spectroscopy for divertor plasma diagnosis and control in DIII-D tokamak.

    Science.gov (United States)

    Soukhanovskii, V A; McLean, A G; Allen, S L

    2014-11-01

    New near infrared (NIR) spectroscopic measurements performed in the DIII-D tokamak divertor plasma suggest new viable diagnostic applications: divertor recycling and low-Z impurity flux measurements, a spectral survey for divertor Thomson scattering (DTS) diagnostic, and Te monitoring for divertor detachment control. A commercial 0.3 m spectrometer coupled to an imaging lens via optical fiber and a InGaAs 1024 pixel array detector enabled deuterium and impurity emission measurements in the range 800-2300 nm. The first full NIR survey identified D, He, B, Li, C, N, O, Ne lines and provided plasma Te, ne estimates from deuterium Paschen and Brackett series intensity and Stark line broadening analysis. The range 1.000-1.060 mm was surveyed in high-density and neon seeded divertor plasmas for spectral background emission studies for λ = 1.064 μm laser-based DTS development. The ratio of adjacent deuterium Paschen-α and Brackett Br9 lines in recombining divertor plasmas is studied for divertor Te monitoring aimed at divertor detachment real-time feedback control. PMID:25430325

  20. 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)

  1. 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

    The effects of edge-localized modes (ELMs) on divertor particle and heat fluxes were investigated for the first time in the Experimental Advanced Superconducting Tokamak (EAST). The experiments were carried out with both double null and lower single null divertor configurations, and comparisons w...

  2. 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.

  3. The installation of the JET Mki divertor - features and achievements

    International Nuclear Information System (INIS)

    The installation of the MkI divertor involved the removal of 25 tonnes and installing nearly 80 tonnes of equipment into the JET vacuum vessel. The work was carried out over a twenty-two and a half month shutdown period requiring over 50,000 man hours of in-vessel effort. JET's requirements were successfully achieved in terms of component installation, dimensional accuracy and timescale. The work was organised into three stages: Stage One dealt with the stripout of all components and the removal from the vessel wall of many welded bosses. Two external toroidal coils were also replaced. This work was carried out in a Beryllium and Tritium contaminated environment with a high radiation level. An extensive cleaning and decontamination exercise completed the stage. Stage Two involved the fabrication of the four divertor coils at the end of which another full vessel cleaning was carried out. Stage Three was the installation of the pumped divertor and the associated wall and roof components. This paper reviews the overall concept of the shutdown and looks at the organisations, training and logistical support required. Certain key features will be described. These include the cleaning and decontamination of the vessel, without this full pressurised suits would have been needed for phases two and three, thus extending the programme to an unacceptable length. Material handling in the vessel which ranged from temporarily supporting the 24 tonne weight of the four coils from the roof of the vessel to an in-vessel crane and mobile personnel carriers. Finally, the installation and correct alignment of the target plates on which the success of the shutdown largely depended. (orig.)

  4. 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

  5. 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.)

  6. 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

  7. 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)

  8. 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)

  9. 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.

  10. 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

  11. Comparison of observed divertor heat flux and modeling results at LHD

    International Nuclear Information System (INIS)

    The divertor strike line pattern on the helical divertor of LHD was observed with an infra red camera. The derived heat flux pattern show multiple distinct strike lines depending on the equilibrium magnetic configuration. Predictions of such divertor heat loads thus require a modeling of the magnetic configuration and the heat transport in the magnetic edge. Equilibrium magnetic topologies were analyzed with HINT2, while the plasma fluid model code EMC3 was used to simulate the energy transport in the edge. The measured multi peak structure of the divertor heat flux is correlated to the intersection points of elongated loop shaped flux tubes of long LC field lines. But the fluid model could not recreate the total energy load and the multiple heat flux peaks on the divertor. A Variation in the plasma density ne as a transport parameter in order to fit the simulated heat flux to the measured one shows a contradicting tendency. (author)

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

    International Nuclear Information System (INIS)

    Net erosion and deposition of tungsten (W) in the ASDEX Upgrade divertor were determined after the 2007 campaign by using thin W marker stripes. ASDEX Upgrade had full-W plasma-facing components during this campaign. The inner divertor and the roof baffle were net W deposition areas with a maximum deposition of about 1x1018 W-atoms cm-2 in the private flux region below the inner strike point. Net erosion of W was observed in the whole outer divertor, with the largest erosion close to the outer strike point. Only a small fraction of the W eroded in the main chamber and in the outer divertor was found in redeposits in the inner divertor, while a large fraction was either redeposited at unidentified places in the main chamber or has formed dust.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-12-15

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

  14. HHF Test with 35x35x3 Be/Cu Mockups for Verifying the HIP Joining Technology of the ITER Blanket First Wall

    International Nuclear Information System (INIS)

    Since the high heat flux (HHF) test is essential for verifying the joint integrity of the ITER blanket first wall with the similar heat flux like the ITER operation conditions, several HHF tests were performed like the previous tests with the Cu/SS, Be/Cu, Be/Cu/SS mockups. In the present study, the HHF tests with Be/Cu mockups were introduced, which have three 35 mm x 35 mm Be tiles and each material depths were kept to be the same as the ITER blanket. Six mockups were fabricated with three kinds of interlayers such as 10μmCr/10μmCu, 1μmCr/10μmCu, 1μmTi/0.5μmCr/10μmCu, 5μmTi/10μmCu. Ten mockups were fabricated with the above conditions but one mockup (10μmCr/10μmCu interlayer) was failed during fabrication process. So, the first four mockups were excluded in the HHF test. Other six mockups were used in the present HHF test. According to the test conditions determined by the preliminary analysis with ANSYS code in the case of 1.5 and 1.0 MW/m2 heat fluxes, the tests were performed. The coolant conditions of the test facility, KoHLT-1 were considered in the simulation and used in the HHF test. Before the HHF test, shear test and non-destructive test with ultrasonic probe were performed with the fabricated mockups. One mockup showed a Be tile delamination during the screening test but the other survived up to 862 cycles under 1.0 MW/m2 heat flux. Other four mockups survived up to 1,100 cycles under the same heat flux without any delamination or Be tile damage, therefore, it shows that the joint integrity has no problem even with the loaded certain heat. And more, three interlayers show the their applicability as a Be to Cu joining one but it needs more attention at the interlayer coating for the reproducibility

  15. Development of modular helium-cooled divertor for demo based on the multi-jet impingement (HEMJ) concept: experimental validation of thermal performance

    International Nuclear Information System (INIS)

    A modular helium-cooled divertor design for the ''post-ITER'' demonstration reactor (DEMO) based on the multi-jet impingement concept (HEMJ) has been developed at the Forschungszentrum Karlsruhe. The design goal is to accommodate a surface heat flux of at least 10 MW/m2 at an acceptable pumping power. This paper describes the thermalhydraulic analyses and validation experiments performed in support of the HEMJ divertor design. Both thermal-hydraulic and thermo-mechanical simulations were performed to support the original design optimization process. The thermal-hydraulic analyses were performed using the FLUENT CFD software package; they showed that the HEMJ design can remove a heat load of up to 12 MW/m2 at an acceptable pumping power. Extremely high heat transfer coefficients were predicted (∝30 kW/m2-K). This experimental investigation has been undertaken to validate the results of the numerical simulations. A one-to-one scale test module that closely matches the reference geometry of the HEMJ design has been constructed and tested. Initial experiments have been performed using air as the coolant at different Reynolds numbers spanning the value for the actual helium-cooled HEMJ design. The experiments have been performed at heat fluxes of up to 1.0 MW/m2. The temperature distributions and local heat transfer coefficients have been measured over a wide range of operational conditions. The experimental data have been compared with the results of a-priori analyses performed using the FLUENT CFD package with the same model options used in the original HEMJ divertor design calculations. Comparison between the model predictions and experimental data provides the means for assessing the suitability of the numerical model to the design of the HEMJ divertor, as well as other gas-cooled high heat flux components at fusion reactor operating conditions. (orig.)

  16. 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

  17. 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.

  18. First EMC3-Eirene simulations of the TCV snowflake divertor

    International Nuclear Information System (INIS)

    One of the approaches to solve the heat load problem in a divertor tokamak is the so called ‘snowflake’ (SF) configuration, a magnetic equilibrium with two nearby x-points and two additional divertor legs. Here we report on the first EMC3-Eirene simulations of plasma- and neutral particle transport in the scrape-off layer of a series of TCV SF equilibria with different values of σ, the distance between the x-points normalized to the minor radius of the plasma. The constant cross-field transport coefficients were chosen such that the power- and particle deposition profiles at the primary inner strike point (SP) match the Langmuir probe measurements for the σ = 0.1 case. At the secondary SP on the floor, however, a significantly larger power flux than that predicted by the simulation was measured by the probes, indicating an enhanced transport across the primary separatrix. As the ideal SF configuration (σ = 0) is approached, the density as well as the radiation maximum are predicted to move from the target plates upward to the x-point by the simulation. (paper)

  19. Axisymmetric curvature-driven instability in a model divertor geometry

    International Nuclear Information System (INIS)

    A model problem is presented which qualitatively describes a pressure-driven instability which can occur near the null-point in the divertor region of a tokamak where the poloidal field becomes small. The model problem is described by a horizontal slot with a vertical magnetic field which plays the role of the poloidal field. Line-tying boundary conditions are applied at the planes defining the slot. A toroidal field lying parallel to the planes is assumed to be very strong, thereby constraining the possible structure of the perturbations. Axisymmetric perturbations which leave the toroidal field unperturbed are analyzed. Ideal magnetohydrodynamics is used, and the instability threshold is determined by the energy principle. Because of the boundary conditions, the Euler equation is, in general, non-separable except at marginal stability. This problem may be useful in understanding the source of heat transport into the private flux region in a snowflake divertor which possesses a large region of small poloidal field, and for code benchmarking as it yields simple analytic results in an interesting geometry

  20. 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)

  1. 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.

  2. 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

  3. 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...

  4. Direct measurement of divertor exhaust neo enrichment in DIII-D

    International Nuclear Information System (INIS)

    We report first direct measurements of divertor exhaust gas impurity enrichment, ηexh=(exhaust impurity concentration)divided-by(core impurity concentration), for both unpumped and D2 puff-with-divertor-pump conditions. The experiment was performed with neutral beam heated, ELMing H-mode, single-null diverted deuterium plasmas with matched core and exhaust parameters in the DIII-D tokamak. Neon gas impurity was puffed into the divertor. Neon density was measured in the exhaust by a specially modified Penning gauge and in the core by absolute charge exchange recombination spectroscopy. Neon particle accounting indicates that much of the puffed neon entered a temporary unmeasured reservoir, inferred to be the graphite divertor target, which makes direct measurements necessary to calculate divertor enrichments. D2 puff into the SOL (scrape-off layer) with pumping increased ηexh threefold over either unpumped conditions or D2 puff directly into the divertor with pumping. These results show that SOL flow plays an important role in divertor exhaust impurity enrichment

  5. A comprehensive 2-D divertor data set from DIII-D for edge theory validation

    International Nuclear Information System (INIS)

    A comprehensive set of experiments has been carried out on the DIII-D tokamak to measure the 2-D (R,Z) structure of the divertor plasma in a systematic way using new diagnostics. Measurements cover the divertor radially from inside the X-point to the outer target plate and vertically from the target plate to above the X-point. Identical, repeatable shots were made, each having radial sweeps of the X-point and divertor strike points, to allow complete plasma and radiation profile measurements. Data have been obtained in ohmic, L-mode, ELMing H-mode, and reversed BT operation (∇B drift away from the X-point). In addition, complete measurements were made of radiative divertor plasmas with a Partially Detached Divertor (PDD) induced by D2 injection and with a Radiating Mantle induced by Impurity injection (RMI) using neon and nitrogen. The data set includes first observations of the radial and poloidal profiles of the X-point, inner and outer leg plasmas in PDD and RMI radiative divertor operation. Preliminary data analysis shows that intrinsic impurities play a critical role in determining the SOL and divertor conditions

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

    Science.gov (United States)

    Sizyuk, V.; Hassanein, A.

    2015-01-01

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

  7. 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.

  8. 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.

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

    International Nuclear Information System (INIS)

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

  10. 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

  11. Structure of the temperature and density fields in the boundary layer during ergodic divertor operation in Tore Supra

    International Nuclear Information System (INIS)

    Active control of the edge plasma (in view of minimizing the impurity influx) is one of the crucial issues to be solved for tokamaks. The ergodic divertor (ED) is being studied in the Tore Supra tokamak. A theoretical analysis of transport in presence of the ED, including the effect of the wall, shows that one can expect two domains in the edge layer: a laminar region, where the field line connection lengths are short (and therefore there is little stochasticity), and the proper ergodic region, where the connection lengths to the wall are large. As a result of the transport properties imposed by the presence of different flux tube lengths and by the existence of localized sources and sinks, one expects some modulation of the electron temperature and density. These effects are studied. (K.A.) 11 refs.; 4 figs

  12. Structure of the temperature and density fields in the boundary layer during ergodic divertor operation in Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    DeMichelis, C.; Ghendrih, Ph.; Guirlet, R.; Hess, W.; Monier-Garbet, P.; Capes, H.; Grosman, A.; Guilhem, D.; Nguyen, F.; Vallet, J.C.; Valter, J. [Association Euratom-CEA, Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee

    1995-12-31

    Active control of the edge plasma (in view of minimizing the impurity influx) is one of the crucial issues to be solved for tokamaks. The ergodic divertor (ED) is being studied in the Tore Supra tokamak. A theoretical analysis of transport in presence of the ED, including the effect of the wall, shows that one can expect two domains in the edge layer: a laminar region, where the field line connection lengths are short (and therefore there is little stochasticity), and the proper ergodic region, where the connection lengths to the wall are large. As a result of the transport properties imposed by the presence of different flux tube lengths and by the existence of localized sources and sinks, one expects some modulation of the electron temperature and density. These effects are studied. (K.A.) 11 refs.; 4 figs.

  13. 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.

  14. Development and qualification of a bulk tungsten divertor row for JET

    Science.gov (United States)

    Mertens, Ph.; Altmann, H.; Hirai, T.; Philipps, V.; Pintsuk, G.; Rapp, J.; Riccardo, V.; Schweer, B.; Uytdenhouwen, I.; Samm, U.

    2009-06-01

    A bulk tungsten divertor row has been developed in the frame of the ITER-like Wall project at JET. It consists of 96 tiles grouped in 48 modules around the torus. The outer strike point is located on those tiles for most of the ITER-relevant, high triangularity plasmas. High power loads (locally up to 10-20 MW/m 2) and erosion rates are expected, even a risk of melting, especially with the transients or ELM loads. These are demanding conditions for an inertially cooled design as prescribed. A lamella design has been selected for the tungsten, arranged to control the eddy and halo current flows. The lamellae must also withstand high temperature gradients (2200 to 220 °C over 40 mm height), without overheating the supporting carrier (600-700 °C maximum). As a consequence of the tungsten emissivity, the radiative cooling drops appreciably in comparison with the current CFC tiles, calling for interleaved plasma scenarios in terms of performance. The compromise between shadowing and power handling is discussed, as well as the consequences for operation. Prototypes have been exposed in TEXTOR and in an electron beam facility (JUDITH-2) to the nominal power density of 7 MW/m 2 for 10 s and, in addition, to higher loads leading to surface temperatures above 2000 °C.

  15. Magnetic divertor design for the compact reversed-field pinch reactor

    International Nuclear Information System (INIS)

    A recently completed design of a pumped-limiter-based Compact Reversed-Field Pinch Reactor is used to estimate for the first time the impact of magnetic divertors. A range of divertor options for the low-toroidal-field RFP is examined, and a design selection is made constrained by consideration of field ripple (magnetic island), blanket displacement, recirculating power, cost, heat flux, and access. Design choices based on diversion of minority (toroidal) field lead to a preference for (poloidally) symmetric or bundle divertor geometries

  16. Feasibility study of inside automatic welding system of cooling pipe of divertors for FER

    International Nuclear Information System (INIS)

    In order to replace divertors for FER, cooling pipes of divertors should be cut and welded since they are too long to be replaced with divertors via horizontal maintenance ports. An inside cutting and welding system is also required because of an accessibility to pipes. A combination of an inside disc-cutting machine and an inside TIG-welding machine has been proposed as a candidate of the systems. We have made tests to confirm possibility to weld pipes which were cut with the disc-cutting machine. Possibility of welding has been proven. The tests result is described in the paper. (orig.)

  17. Present status of linear plasma devices and issues on DEMO divertor design

    International Nuclear Information System (INIS)

    Construction of ITER has inspired discussion on the design of DEMO reactor. In the country, a team for constructing technological base of the fusion prototype reactor (a joint core team) has been organized in the nuclear science and technology committee thermonuclear-research working group of MEXT. In IAEA, the DEMO program workshop has been held. With these situations, we consider design issues of divertor in DEMO reactor and the role of linear plasma devices. First, the status of researches in linear plasma devices is reviewed. Next we explain R and D issues in the design of DEMO divertors and in the modeling and numerical simulation of divertors. (author)

  18. DESIGN OPTIMIZATION OF A FUSION REACTOR DIVERTOR COOLING FINGER DESIGN FOR DECREASED THERMO-MECHANICAL STRESSES

    OpenAIRE

    MATTEOLI CAMILLA; Martin, Oliver; SIMONOVSKI IGOR

    2012-01-01

    In this paper a design of a divertor cooling finger for a fusion reactor is looked at with the aim of reducing the thermo-mechanical stresses. The function of a divertor in a fusion reactor is to reduce the dilution of the plasma by removing alpha particles, helium and other impurities. In addition, this component has to remove approximately 15% of the total thermal power. The divertor is therefore exposed to a significant thermal load, and during the operation it has to be actively cooled, w...

  19. 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

  20. Erosion/redeposition analysis of the DIII-D divertor

    International Nuclear Information System (INIS)

    Carbon and tungsten sputtering and transport in the DIII-D divertor is analyzed with the impurity transport codes REDEP and WBC. Analysis is carried out for a recent DiMES experiment in which a carbon sample with a tungsten marker in the center was exposed to six well controlled ELM-free plasma discharges. WBC analysis predicts a high rate of ionization of tungsten neutrals within the sheath and subsequent redeposition on the DiMES sample. Qualitative comparison of the tungsten redeposited flux agrees well with measurements. REDEP analysis of net carbon erosion shows a factor of 2-3 agreement with measured data on the outboard side of DiMES and poor agreement on the inboard side

  1. Performance of JT-60SA divertor Thomson scattering diagnostics

    International Nuclear Information System (INIS)

    For the satellite tokamak JT-60 Super Advanced (JT-60SA), a divertor Thomson scattering measurement system is planning to be installed. In this study, we improved the design of the collection optics based on the previous one, in which it was found that the solid angle of the collection optics became very small, mainly because of poor accessibility to the measurement region. By improvement, the solid angle was increased by up to approximately five times. To accurately assess the measurement performance, background noise was assessed using the plasma parameters in two typical discharges in JT-60SA calculated from the SONIC code. Moreover, the influence of the reflection of bremsstrahlung radiation by the wall is simulated by using a ray tracing simulation. The errors in the temperature and the density are assessed based on the simulation results for three typical field of views

  2. Performance of JT-60SA divertor Thomson scattering diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Kajita, Shin, E-mail: kajita.shin@nagoya-u.jp [Nagoya University, Nagoya 464-8603 (Japan); Hatae, Takaki; Tojo, Hiroshi; Hamano, Takashi; Shimizu, Katsuhiro; Kawashima, Hisato [Japan Atomic Energy Agency, Naka, Ibaraki 801-1 (Japan); Enokuchi, Akito [Genesia Co., Mitaka, Tokyo 181-0013 (Japan)

    2015-08-15

    For the satellite tokamak JT-60 Super Advanced (JT-60SA), a divertor Thomson scattering measurement system is planning to be installed. In this study, we improved the design of the collection optics based on the previous one, in which it was found that the solid angle of the collection optics became very small, mainly because of poor accessibility to the measurement region. By improvement, the solid angle was increased by up to approximately five times. To accurately assess the measurement performance, background noise was assessed using the plasma parameters in two typical discharges in JT-60SA calculated from the SONIC code. Moreover, the influence of the reflection of bremsstrahlung radiation by the wall is simulated by using a ray tracing simulation. The errors in the temperature and the density are assessed based on the simulation results for three typical field of views.

  3. Simulations of carbon sputtering in fusion reactor divertor plates

    International Nuclear Information System (INIS)

    The interaction of edge plasma with material surfaces raises key issues for the viability of the International Thermonuclear Reactor (ITER) and future fusion reactors, including heat-flux limits, net material erosion, and impurity production. After exposure of the graphite divertor plate to the plasma in a fusion device, an amorphous C/H layer forms. This layer contains 20-30 atomic percent D/T bonded to C. Subsequent D/T impingement on this layer produces a variety of hydrocarbons that are sputtered back into the sheath region. We present molecular dynamics (MD) simulations of D/T impacts on amorphous carbon layer as a function of ion energy and orientation, using the AIREBO potential. In particular, energies are varied between 10 and 150 eV to transition from chemical to physical sputtering. These results are used to quantify yield, hydrocarbon composition and eventual plasma contamination

  4. 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

  5. Correlation between hydrogen isotope profiles and surface structure of divertor tiles in JT-60U

    International Nuclear Information System (INIS)

    The present paper is devoted to depth profiles by secondary ion mass spectroscopy (SIMS) of hydrogen/deuterium in tiles taken from the dome unit area of JT-60U. This information is correlated with surface features, particularly from the aspect of erosion and deposition, determined by scanning electron microscope (SEM) and X-ray photoelectron spectroscopy (XPS). The outer divertor-facing surface was mostly covered by re-deposited layers a maximum of 10 μm thick, while the inner divertor-facing side was eroded. The deposition profile is opposite to the observation for the divertor area in most tokamaks that the outer divertor side is eroded, while the inner deposited. However, H + D retention was higher for the deposited layers than that for the eroded area. Nevertheless, hydrogen retention seems very small and showed no appreciable effects on C1s spectra of XPS compared to the constituent elements boron and oxygen

  6. Erosion and deposition on JET divertor and limiter tiles during the experimental campaigns 2005–2009

    Energy Technology Data Exchange (ETDEWEB)

    Krat, S., E-mail: stepan.krat@gmail.com [National Research Nuclear University “MEPhI”, Kashirskoe Road 31, 115409 Moscow (Russian Federation); Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany); Coad, J.P. [Culham Science Centre, EURATOM/UKAEA – Fusion Association, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Gasparyan, Yu. [National Research Nuclear University “MEPhI”, Kashirskoe Road 31, 115409 Moscow (Russian Federation); Hakola, A.; Likonen, J. [Association EURATOM-Tekes, Technical Research Centre of Finland, PO Box 1000, FI-02044 VTT (Finland); Mayer, M. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany); Pisarev, A. [National Research Nuclear University “MEPhI”, Kashirskoe Road 31, 115409 Moscow (Russian Federation); Widdowson, A. [Culham Science Centre, EURATOM/UKAEA – Fusion Association, Abingdon, Oxfordshire OX14 3DB (United Kingdom)

    2013-07-15

    Erosion from and deposition on JET divertor tiles used during the 2007–2009 campaign and on inner wall guard limiter (IWGL) tiles used during 2005–2009 are studied. The tungsten coating on the divertor tiles was mostly intact with the largest erosion ∼30% in a small local area. Locally high erosion areas were observed on the load bearing divertor tile 5 and on the horizontal surface of the divertor tile 8. The IWGL tiles show a complicated distribution of erosion and deposition areas. The total amount of carbon deposited on the all IWGL tiles during the campaign 2005–2009 is estimated to be 65 g. The density of carbon deposits is estimated to be 0.67–0.83 g/cm{sup 3}.

  7. Comparison of JET main chamber erosion with dust collected in the divertor

    Energy Technology Data Exchange (ETDEWEB)

    Widdowson, A., E-mail: anna.widdowson@ccfe.ac.uk [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Ayres, C.F.; Booth, S. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Coad, J.P.; Hakola, A. [Association EURATOM-TEKES, VTT, PO Box 1000, 02044 VTT, Espoo (Finland); Heinola, K. [Association EURATOM-TEKES, University of Helsinki, PO Box 64, 00560 Helsinki (Finland); Ivanova, D. [Laboratory, Royal Institute of Technology, Association EURATOM-VR, 100 44 Stockholm (Sweden); Koivuranta, S.; Likonen, J. [Association EURATOM-TEKES, VTT, PO Box 1000, 02044 VTT, Espoo (Finland); Mayer, M. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Stamp, M. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)

    2013-07-15

    A complete global balance for carbon in JET requires knowledge of the net erosion in the main chamber, net deposition in the divertor and the amount of dust and flakes collecting in the divertor region. This paper describes a number of measurements on aspects of this global picture. Profiler measurements and cross section microscopy on tiles that were removed in the 2009 JET intervention are used to evaluate the net erosion in the main chamber and net deposition in the divertor. In addition the mass of dust and flakes collected from the JET divertor during the same intervention is also reported and included as part of the balance. Spectroscopic measurements of carbon erosion from the main chamber are presented and compared with the erosion measurements for the main chamber.

  8. A full tungsten divertor for ITER: Physics issues and design status

    Energy Technology Data Exchange (ETDEWEB)

    Pitts, R.A., E-mail: richard.pitts@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Carpentier, S.; Escourbiac, F.; Hirai, T.; Komarov, V.; Lisgo, S.; Kukushkin, A.S.; Loarte, A.; Merola, M.; Sashala Naik, A.; Mitteau, R.; Sugihara, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Bazylev, B. [Karlsruhe Institute of Technology, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany); Stangeby, P.C. [University of Toronto Institute Aerospace Studies, Ontario, Canada M3H 5T6 (Canada)

    2013-07-15

    Budget restrictions have forced the ITER Organization to reconsider the baseline divertor strategy, in which operations would begin with carbon (C) in the high heat flux regions, changing out to a full-tungsten (W) variant before the first nuclear campaigns. Substantial cost reductions can be achieved if one of these two divertors is eliminated. The new strategy implies not only that ITER would start-up on a full-W divertor, but that this component should survive until well into the nuclear phase. This paper considers the risks engendered by such an approach with regard to known W plasma-material interaction issues and briefly presents the current status of a possible full-W divertor design.

  9. A full tungsten divertor for ITER: Physics issues and design status

    International Nuclear Information System (INIS)

    Budget restrictions have forced the ITER Organization to reconsider the baseline divertor strategy, in which operations would begin with carbon (C) in the high heat flux regions, changing out to a full-tungsten (W) variant before the first nuclear campaigns. Substantial cost reductions can be achieved if one of these two divertors is eliminated. The new strategy implies not only that ITER would start-up on a full-W divertor, but that this component should survive until well into the nuclear phase. This paper considers the risks engendered by such an approach with regard to known W plasma-material interaction issues and briefly presents the current status of a possible full-W divertor design

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

    Energy Technology Data Exchange (ETDEWEB)

    Klepper, C Christopher [ORNL; Biewer, T. M. [Oak Ridge National Laboratory (ORNL); Graves, Van B [ORNL; Andrew, P. [EURATOM / UKAEA, UK; Lukens, P. C. [United States ITER Project Office; Marcus, Chris [ORNL; Shimada, M. [ITER Organization, Saint Paul Lez Durance, France; Hughes, S. [ITER Organization, Cadarache, France; Boussier, B. [ITER Organization, Saint Paul Lez Durance, France; Johnson, D. W. [Princeton Plasma Physics Laboratory (PPPL); Gardner, W. L. [United States ITER Project Office; Hillis, D. L. [Oak Ridge National Laboratory (ORNL); Vayakis, G. [ITER Organization, Cadarache, France; Vayakis, G. [ITER Organization, Saint Paul Lez Durance, France; Walsh, M. [ITER Organization, Saint Paul Lez Durance, France

    2015-01-01

    One of the ITER diagnostics having reached an advanced design stage is a diagnostic RGA for the divertor, i.e. residual gas analysis system for the ITER divertor, which is intended to sample the divertor pumping duct region during the plasma pulse and to have a response time compatible with plasma particle and impurity lifetimes in the divertor region. Main emphasis is placed on helium (He) concentration in the ducts, as well as the relative concentration between the hydrogen isotopes (H2, D2, T2). Measurement of the concentration of radiative gases, such as neon (Ne) and nitrogen (N2), is also intended. Numerical modeling of the gas flow from the sampled region to the cluster of analysis sensors, through a long (~8m long, ~110mm diameter) sampling pipe terminating in a pressure reducing orifice, confirm that the desired response time (~1s for He or D2) is achieved with the present design.

  11. 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

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

    International Nuclear Information System (INIS)

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

  13. Comparison of JET main chamber erosion with dust collected in the divertor

    CERN Document Server

    Widdowson, A; Booth, S; Coad, J P; Hakola, A; Heinola, K; Ivanova, S; Koivuranta, S; Likonen, J; Mayer, M; Stamp, M; Contributors, JET-EFDA

    2013-01-01

    A complete global balance for carbon in JET requires knowledge of the net erosion in the main chamber, net deposition in the divertor and the amount of dust and flakes collecting in the divertor region. This paper describes a number of measurements on aspects of this global picture. Profiler measurements and cross section microscopy on tiles that were removed in the 2009 JET intervention are used to evaluate the net erosion in the main chamber and net deposition in the divertor. In addition the mass of dust and flakes collected from the JET divertor during the same intervention is also reported and included as part of the balance. Spectroscopic measurements of carbon erosion from the main chamber are presented and compared with the erosion measurements for the main chamber.

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

    International Nuclear Information System (INIS)

    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 C3+ 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-α (Dα) 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 Dα line spectral profile in the divertor region has been measured for the first time in this study. The observed results were compared with the calculated one by a neutral particle

  15. 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

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

    International Nuclear Information System (INIS)

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

  17. 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)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Mattas, R.F.; Smith, D.L. [Argonne National Lab., IL (United States); Wu, C.H. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany). NET Team; Koroda, T. [Japan Atomic Energy Research Inst., Ibaraki-ken (Japan); Shatalov, G. [Kurchatov Inst. of Atomic Energy, Moscow (USSR)

    1992-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Mattas, R.F.; Smith, D.L. (Argonne National Lab., IL (United States)); Wu, C.H. (Max-Planck-Institut fuer Plasmaphysik, Garching (Germany). NET Team); Koroda, T. (Japan Atomic Energy Research Inst., Ibaraki-ken (Japan)); Shatalov, G. (Kurchatov Inst. of Atomic Energy, Moscow (USSR))

    1992-01-01

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

  1. 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)

  2. 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

  3. Possible divertor solutions for a fusion reactor. Pt. I. Physical aspects based on present day divertor operation

    International Nuclear Information System (INIS)

    For pt.II see ibid., p.109-117 (1997). With an anticipated power flux across the separatrix of up to 300 MW of an ITER-like fusion reactor, conventional measures of power spread lead to a peak power load at the target plates in the order of 30 MW m-2, far beyond the technically feasible limit for stationary operation. Radiative cooling by seed impurities appears to be the most promising plasma-physical option to reduce the target power load, but extrapolations of present experiments predict an only marginally tolerable increase of the plasma effective charge Zeff. Key points will be the achievement of very high electron densities, leading to more effective radiative cooling by δPrad/δZeff∝ne2 while keeping the edge temperature within its optimum range. This range is bounded from below by the H→L mode temperature threshold due to confinement requirements, whereas the upper boundary is given by the ideal ballooning stability limit which is connected to type-I ELM activity which may cause non-tolerable divertor heat loads. The completely detached H-mode (CDH) in ASDEX Upgrade demonstrates radiative H-mode operation within this operational range exhibiting high-frequent type-III ELMs and target power load in the order of 10% of the heating power. At present, open questions on high density reactor operation are related to radiative instabilities as well as edge transport enhancement and H-mode impairment observed in several tokamaks under high density conditions. Measures to overcome these detrimental effects will be investigated with improved divertor concepts in the near future. The possible problems connected to high density reactor operation can be relaxed, if the design of plasma facing components with higher heat flux endurance is successful. (orig.)

  4. 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.

  5. 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

  6. 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. }

  7. Linear peeling–ballooning mode simulations in snowflake-like divertor configuration using BOUT++ code

    International Nuclear Information System (INIS)

    We present linear characteristics of peeling–ballooning (P–B) modes in the pedestal region of DIII-D tokamak with snowflake (SF) plus divertor configuration using edge two-fluid code BOUT++. A set of reduced magnetohydrodynamics (MHD) equations is found to simulate the linear P–B mode in both snowflake plus and standard (STD) single-null divertor configurations. Further analysis shows that the implementation of snowflake geometry changes the local magnetic shear in the pedestal region, which leads to different linear behaviours of the P–B mode in STD and SF divertor configuration. Primary linear simulation results are the following. (1) The growth rate of the coupled P–B mode in SF-plus divertor geometry is larger than that in STD divertor geometry. (2) The global linear mode structures are more radially extended yet less poloidally extended in SF-plus divertor geometry, especially for moderate and high toroidal mode numbers. (3) The current-gradient drive (the kink term) dominates the P–B mode for low n, while the pressure gradient drive (ballooning) dominates for n > 25. In addition, constraints on poloidal field and central solenoid coils for snowflake geometry are briefly discussed based on conclusions in this paper. (paper)

  8. 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

  9. 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.)

  10. 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

  11. Sensitivity analysis of upstream plasma condition for SST-1 X-divertor configuration with SOLPS

    International Nuclear Information System (INIS)

    The extensive power exhausts and target heat loads are anticipated in reactor grade fusion devices. Prototyping of an X-Divertor based power exhaust scheme is being attempted by means of simulations of Scrape-off Layer plasma transport in the diverted plasma equilibria of SST-1 tokamak using SOLPS5.1. Evaluation of the relative advantages of an X-Divertor configuration involves simulating the SST-1 standard divertor scheme plasma transport for the reference and then achieving equivalent upstream plasma conditions in the X-divertor equilibrium to ensure an equivalent core plasma in both the cases. The first optimization is to be achieved by simulating effects of an external gas puff in the SOL region for controlling separatrix density in the X-divertor configuration with visible modifications in the downstream plasma conditions. The present work analyzes sensitivity of the upstream SOL plasma conditions to the gas puff intensity and its effect on the plasma neutral transport in the divertor region. (author)

  12. 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.

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

    International Nuclear Information System (INIS)

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

  14. The impact of ELMs on the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Leonard, A.W.; Osborne, T.H.; Suttrop, W. [General Atomics, San Diego, CA (United States); Hermann, A. [Max Planck Inst. fuer Plasmaphysics, Garching (Germany); Itami, K. [Japan Atomic Energy Research Inst. (Japan); Lingertat, J. [JET Joint Undertaking, Abingdon (United Kingdom); Loarte, A. [Next European Torus, Garching (Germany)

    1998-07-01

    Edge-Localized-Modes (ELMs) are expected to present a significant transient flux of energy and particles to the ITER divertor. The threshold for ablation of the graphite target will be reached if the ELM transient exceeds Q/t{sup 1/2} {approximately} 45 MJ-m{sup {minus}2}-s{sup {minus}1/2} where Q is the ELM deposition energy density and t is the ELM deposition time. The ablation parameter in ITER can be determined by scaling four factors from present experiments: the ELM energy loss from the core plasma, the fraction of ELM energy deposited on the divertor target, the area of the ELM profile onto the target, and finally the time for the ELM deposition. Review of the ELM energy loss of Type 1 ELM data suggests an ITER ELM energy loss of 2--6% of the stored energy or 25--80 MJ. The fraction of heating power crossing the separatrix due to ELMs is nearly constant (20--40%) resulting in an inverse relationship between ELM amplitude and frequency. Measurements on DIII-D and ASDEX-Upgrade indicate that 50--80% of the ELM energy is deposited on the target. There is currently no evidence for a large fraction of the ELM energy being dissipated through radiation. Profiles of the ELM heat flux are typically 1--2 times the width of steady heat flux between ELMs, with the ELM amplitude usually larger on the inboard target. The ELM deposition time varies from about 0.1 ms in JET to as high as 1.0 ms in ASDEX-Upgrade and DIII-D. The ELM deposition time for ITER will depend upon the level of conductive versus convective transport determined by the ratio of energy to particles released by the ELM. Preliminary analysis suggests that large Type 1 ELMs for low recycling H-mode may exceed the ablation parameter by a factor of 5. Promising regimes with smaller ELMS have been found at other edge operational regimes, including high density with gas puffing, use of rf heating and operation with Type 3 ELMs.

  15. The impact of ELMS on the ITER divertor

    International Nuclear Information System (INIS)

    Edge-Localized-Modes (ELMs) are expected to present a significant transient flux of energy and particles to the ITER divertor. The threshold for ablation of the graphite target will be reached if the ELM transient exceeds Q/t1/2 ∼ 45 MJ-m-2-s-1/2 where Q is the ELM deposition energy density and t is the ELM deposition time. The ablation parameter in ITER can be determined by scaling four factors from present experiments: the ELM energy loss from the core plasma, the fraction of ELM energy deposited on the divertor target, the area of the ELM profile onto the target, and finally the time for the ELM deposition. Review of the ELM energy loss of Type 1 ELM data suggests an ITER ELM energy loss of 2--6% of the stored energy or 25--80 MJ. The fraction of heating power crossing the separatrix due to ELMs is nearly constant (20--40%) resulting in an inverse relationship between ELM amplitude and frequency. Measurements on DIII-D and ASDEX-Upgrade indicate that 50--80% of the ELM energy is deposited on the target. There is currently no evidence for a large fraction of the ELM energy being dissipated through radiation. Profiles of the ELM heat flux are typically 1--2 times the width of steady heat flux between ELMs, with the ELM amplitude usually larger on the inboard target. The ELM deposition time varies from about 0.1 ms in JET to as high as 1.0 ms in ASDEX-Upgrade and DIII-D. The ELM deposition time for ITER will depend upon the level of conductive versus convective transport determined by the ratio of energy to particles released by the ELM. Preliminary analysis suggests that large Type 1 ELMs for low recycling H-mode may exceed the ablation parameter by a factor of 5. Promising regimes with smaller ELMS have been found at other edge operational regimes, including high density with gas puffing, use of rf heating and operation with Type 3 ELMs

  16. 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)

  17. Development of a Method for Local Electron Temperature and Density Measurements in the Divertor of the JET Tokamak

    Science.gov (United States)

    Jupen, C.; Meigs, A.; Bhatia, A. K.; Brezinsek, S.; OMullane, M.

    2004-01-01

    Plasma volume recombination in the divertor, a process in which charged particles recombine to neutral atoms, contributes to plasma detachment and hence cooling at the divertor target region. Detachment has been observed at JET and other tokamaks and is known to occur at low electron temperatures (T(sub e)10(exp 20)/m(exp 3)). The ability to measure such low temperatures is therefore of interest for modelling the divertor. In present work we report development of a new spectroscopic technique for investigation of local electron density (n(sub e)) and temperature (T,) in the outer divertor at JET.

  18. 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

  19. 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

  20. 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.

  1. 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.)

  2. Design of impurity influx monitor (divertor) for ITER

    International Nuclear Information System (INIS)

    Because of the changes of interfaces between ITER and the Impurity Influx Monitor (divertor) accompanied with the change of the ITER design to a reduced size machine, changes in the design of the monitor are required. In this design work, optics compatible with new interfaces, a calibration system and an alignment system of the optical axis and the focus were designed and investigated. The design of the optical systems was simplified to save the cost. To simplify the optics, the design of the collection optics was changed from an off-axis aspherical mirror system, which is a previous design, to a simple Cassegrain telescope system composed of simple spherical mirrors and lenses. In addition, a micro lens array is inserted just in front of the fiber bundle to increase the light detected. The ray-trace analysis shows that the spatial resolution of ITER requirement (50 mm) will be achieved by these optical systems designed here. The in-situ sensitivity calibration will be realized by applying the light on a micro retro-reflector array installed in front of the plasma facing first mirror from the outside the vacuum chamber through the same optics for plasma measurement and measuring the intensity of the reflected light from the array by using the same optics. In addition, optical design of an adjustment system and a focusing system, and a conceptual design of the shutter system of the monitor were carried out. Moreover, the detector system for the monitor was investigated and designed. (author)

  3. Small steady-state tokamak (TST) for divertor testing

    International Nuclear Information System (INIS)

    The TST is a small steady-state tokamak designed for testing diverters under conditions similar to those anticipated in future large tokamaks. An initial design has R0/a = 2.5, R0 = 0.75 m, a = 0.3 m, and Bt0 = 2.2 T with full inductive capability. With heating and current drive power of 4.5 MW, the heat flux at the plasma edge Q perpendicular can be as high as 0.3 MW/m2. Plasma currents Ip above 500 kA can be maintained by 1 MW of lower hybrid power (2.45 GHz) for average densities ne up to 3 x 1019 m-3. Additional power via ICRF (2 MW) and neutral beams (1.5 MW) maintain current for ne up to 5 x 1019 m-3. Fully demountable, actively cooled, steady-state toroidal field coils permit ample access for the auxiliary systems and diverter cassettes. The toroidal field magnets require a steady-state supply of less than 40 MW. The size and cost of the TST can be reduced by eliminating the solenoid, reducing Bt0 to 1.4 T, and lowering R0/a to 1.7. This option permits low-R0/a experimentation while maintaining the capability for testing divertors but requires successful noninductive current initiation and maintenance in the low-R0/a regime

  4. Technologies for the ITER divertor vertical target plasma facing components

    International Nuclear Information System (INIS)

    The ITER divertor vertical target (VT) has to sustain heat fluxes up to 20 MW/m2. The concept developed for this Plasma Facing Component (PFC) working at steady state is based on Carbon Fibre Composite (CFC) armour for the lower straight part and tungsten (W) for the curved upper part. The main challenges of such components are to be able to remove the high deposited heat fluxes and to join mechanically and thermally armour to the metallic heat sink, despite of the mismatch of the thermal expansions. Two solutions based on the use of CuCrZr hardened copper alloy and active metal casting (AMC) process were investigated during the ITER EDA phase: the first one called 'flat tile geometry' was mainly developed for Tore Supra pumped limiter, the second one called 'monoblock geometry' was developed by the EU Participating Team for the ITER project. This paper presents a review of these two solutions and analyses their assets and drawbacks: pressure drop, critical heat flux, surface temperature and expected behaviour during operation, risks during the manufacture, controls of the armour defects during the manufacture and at the reception and possibility to repair defected tiles. (author)

  5. 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. __________________________________________________

  6. 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.

  7. 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

  8. 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)

  9. 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.

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

    International Nuclear Information System (INIS)

    Production of a full-size divertor cassette for the International Thermonuclear Experimental Reactor (ITER) involves eight major components. All of the components are mounted on the cassette body. Inner divertor channel components for both the vertical target and the gas box design are being provided by the Japan Home Team. Outer divertor channel components for the vertical target design are being provided by the European and United States Home Teams. Gas box liners are being provided by the Russian Home Team. The full-size components manufactured by the four parties will be shipped to the US Home Team for assembly into a full-size divertor cassette. The techniques for assembly and maintenance for the cassette will be demonstrated during this process. The assembled cassette will be tested for proper flow distribution and proof of the filling and draining procedures. The testing will include vacuum leak, cyclic heating to 150 deg. C, and verification of dimensional accuracy of the assembled components. The development of the divertor for ITER depends on successful R and D efforts on materials, joining and plasma-materials interactions. Results of the development program are presented. The scale-up of the process developed in the basic R and D tasks is accomplished by producing medium and full-scale mock-ups and testing them at high heat flux. The design of these mock-ups is discussed. (author). 18 refs, 3 figs

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

    International Nuclear Information System (INIS)

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

  12. 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.

  13. 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

  14. Impact of Resonant Magnetic Perturbation Fields on NSTX-U Advanced Divertor Topologies

    Science.gov (United States)

    Waters, Ian; Frerichs, Heinke; Schmitz, Oliver; Ahn, Joon-Wook; Canal, Gustavo; Evans, Todd; Soukhanovskii, Vlad

    2015-11-01

    Explorations are under way to optimize the magnetic topology in the plasma edge of NSTX-U with the goal of improving neutral and impurity fueling and exhaust. The use of magnetic perturbation fields is being considered to spread heat and particle fluxes in the divertor, adjust plasma refueling, control impurity transport, and improve coupling to the exhaust systems. Also, advanced divertor configurations are being considered to improve peak heat loads on divertors. An assessment is made of the topologies of a number of representative NSTX-U advanced divertor configurations: lower single null, exact snowflake, and snowflake minus. Wall to wall magnetic connection lengths for each configuration are assessed in both their perturbed and axisymmetric configurations with perturbation coil currents of 1kA and 3kA. The magnetic perturbations yield complex strike patterns on divertor elements that are expected to be resolvable experimentally. The EMC3-EIRENE fluid plasma and kinetic neutral transport code will be used to study neutral and impurity transport and exhaust in these topologies. This work was funded in part by the Department of Energy under grant DE-SC0012315 and by startup funds of the Department of Engineering Physics at the University of Wisconsin-Madison.

  15. Conceptual design of divertor and first wall for DEMO-FNS

    Science.gov (United States)

    Sergeev, V. Yu.; Kuteev, B. V.; Bykov, A. S.; Gervash, A. A.; Glazunov, D. A.; Goncharov, P. R.; Dnestrovskij, A. Yu.; Khayrutdinov, R. R.; Klishchenko, A. V.; Lukash, V. E.; Mazul, I. V.; Molchanov, P. A.; Petrov, V. S.; Rozhansky, V. A.; Shpanskiy, Yu. S.; Sivak, A. B.; Skokov, V. G.; Spitsyn, A. V.

    2015-11-01

    Key issues of design of the divertor and the first wall of DEMO-FNS are presented. A double null closed magnetic configuration was chosen with long external legs and V-shaped corners. The divertor employs a cassette design similar to that of ITER. Water-cooled first wall of the tokamak is made of Be tiles and CuCrZr-stainless steel shells. Lithium injection and circulation technologies are foreseen for protection of plasma facing components. Simulations of thermal loads onto the first wall and divertor plates suggest a possibility to distribute heat loads making them less than 10 MW m-2. Evaluations of sputtering and evaporation of plasma-facing materials suggest that lithium may protect the first wall. To prevent Be erosion at the outer divertor plates either the full detached divertor operation or arrangement of the renewal lithium flow on targets should be implemented. Test bed experiments on the Tsefey-M facility with the first wall mockup coated by Ве tiles and cooled by water are presented. The temperature of the surface of tiles reached 280-300 °С at 5 MW m-2 and 600-650 °С at 10.5 MW m-2. The mockup successfully withstood 1000 cycles with the lower thermal loading and 100 cycles with higher thermal loading.

  16. 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

  17. Helium exhaust and forced flow effects with both-leg pumping in W-shaped divertor of JT-60U

    International Nuclear Information System (INIS)

    The W-shaped divertor of JT-60U was modified from inner-leg pumping to both-leg pumping. After the modification, the pumping rate was improved from 3% with inner-leg pumping to 5% with both-leg pumping in a divertor-closure configuration, which means both separatrixes close to the divertor slots. Efficient helium exhaust was realized in the divertor-closure configuration with both-leg pumping. A global particle confinement time of τ*He=0.4s and τ*He/τE=3 was achieved in attached ELMy H-mode plasmas. The helium exhaust efficiency with both-leg pumping was extended by 45% as compared with inner-leg pumping. By using central helium fueling with He-beam injection, the helium removal from the core plasma inside the internal transport barrier (ITB) in reversed shear plasmas in the divertor-closure configuration was investigated for the first time. The helium density profiles inside the ITB were peaked as compared with those in ELMy H-mode plasmas. In the case of low recycling divertor, it was difficult to achieve good helium exhaust capability in reversed shear plasmas with ITB. However, the helium exhaust efficiency was improved with high recycling divertor. Carbon impurity reduction was observed by the forced flow with gas puff and effective divertor pumping. (author)

  18. Optimization of tungsten castellated structures for the ITER divertor

    International Nuclear Information System (INIS)

    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/m2. 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

  19. 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.

  20. Concept for spectrally resolved ITER divertor thermography with fibres

    International Nuclear Information System (INIS)

    Infrared thermography on tokamak target plates under plasma impact performed at a single wavelength may be misleading because the temperature at the surface of a target is not homogeneous. Since the existing ITER divertor thermography diagnostic proposal did not include the possibility to measure at multiple wavelengths at one place, a study was performed to remedy this with a diagnostic proposal based on a fibre-optics approach. We have found an inverse matrix method to deduce the distribution of the target temperature from the spectral radiance distribution. The method seems to be robust against calibration errors and may allow to discriminate thermal radiation against the Bremsstrahlung from the plasma. Fibres are a natural choice for spectroscopic diagnostics. They minimise movements problems and they offer good possibilities for laser methods for calibration and active measurements as presumed necessary for an environment containing deposited layers and low emissivity, high reflection materials as tungsten and beryllium. Due to the high environmental temperature of 150 Celsius degrees the choice of fibres is limited. An optical study was performed to conceive an all mirror optical front-end design suitable to a fibre solution. The optical resolution of the design is about 3 mm on the targets which fits ITER requirements. About 500 fibres are necessary to exploit this fully. Looking only at the centre of the tiles (20 mm pitch) reduces the number of fibres to 100. The mirrors (and their box) and the fibres should be cooled. A detection system similar to the existing Tore Supra multi-fibre sapphire prism spectrometer coupled to a focal plane array InSb infrared camera is a viable detection solution for such a system. The logical next step is to perform radiation tests of true infrared fibres. (A.C.)

  1. Small steady-state tokamak (TST) for divertor testing

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Y.M.; Colchin, R.J.; Swain, D.W.; Nelson, B.E.; Monday, J.F. (Oak Ridge National Lab., TN (United States)); Blevins, J.; Delisle, M.; Stringer, J. (Canadian Fusion Fuels Technology Project, Mississauga, ON (Canada)); Bonoli, P.; Luckhardt, S. (Massachusetts Inst. of Tech., Cambridge, MA (United States)); Pauletti, R. (Sao Paulo Univ., SP (Brazil))

    1992-01-01

    The TST is a small steady-state tokamak designed for testing diverters under conditions similar to those anticipated in future large tokamaks. An initial design has R{sub 0}/a = 2.5, R{sub 0} = 0.75 m, a = 0.3 m, and Bt{sub 0} = 2.2 T with full inductive capability. With heating and current drive power of 4.5 MW, the heat flux at the plasma edge Q{perpendicular} can be as high as 0.3 MW/m{sup 2}. Plasma currents I{sub p} above 500 kA can be maintained by 1 MW of lower hybrid power (2.45 GHz) for average densities n{sub e} up to 3 {times} 10{sup 19} m{sup {minus}3}. Additional power via ICRF (2 MW) and neutral beams (1.5 MW) maintain current for n{sub e} up to 5 {times} 10{sup 19} m{sup {minus}3}. Fully demountable, actively cooled, steady-state toroidal field coils permit ample access for the auxiliary systems and diverter cassettes. The toroidal field magnets require a steady-state supply of less than 40 MW. The size and cost of the TST can be reduced by eliminating the solenoid, reducing Bt{sub 0} to 1.4 T, and lowering R{sub 0}/a to 1.7. This option permits low-R{sub 0}/a experimentation while maintaining the capability for testing divertors but requires successful noninductive current initiation and maintenance in the low-R{sub 0}/a regime.

  2. 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)

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

    International Nuclear Information System (INIS)

    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 cm3 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

  4. Evaluation of target-plate heat flux for a possible snowflake divertor in CFETR using SOLPS

    Energy Technology Data Exchange (ETDEWEB)

    Mao, S.F., E-mail: sfmao@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, 96 Jinzhai Road, Hefei, Anhui 230026 (China); Guo, Y.; Peng, X.B.; Luo, Z.P. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Xiao, B.J.; Song, Y.T. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, 96 Jinzhai Road, Hefei, Anhui 230026 (China); Yao, D.M.; Zhu, S.Z. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Ye, M.Y., E-mail: yemy@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, 96 Jinzhai Road, Hefei, Anhui 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

    2015-08-15

    China Fusion Engineering Test Reactor (CFETR) is proposed as a good complement to ITER for demonstration of fusion energy. CFETR is based on both physics and some technologies of ITER. The main goals of CFETR are fusion power P{sub f} = 50–200 MW, duty cycle time ⩾0.3–0.5, and a tritium breeding ratio ∼1.2. To explore a more effective way to manage heat exhaust in a future fusion reactor, which will have higher heating power (auxiliary heating power plus 20% of the fusion power) than ITER, a snowflake divertor (SFD) is an optional choice considered for CFETR. In this paper, the preliminary design of a SFD for CFETR is presented, including the divertor magnetic configuration and geometry. A numerical simulation is then performed to evaluate the heat flux onto the divertor targets by using SOLPS.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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)

  11. Fast pedestal, SOL and divertor measurements from DIII-D to validate BOUT++ nonlinear ELM simulations

    Energy Technology Data Exchange (ETDEWEB)

    Fenstermacher, M.E., E-mail: fenstermacher@fusion.gat.com [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Xu, X.Q.; Joseph, I.; Lanctot, M.J.; Lasnier, C.J.; Meyer, W.H. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Tobias, B. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Zeng, L. [University of California-Los Angeles, Los Angeles, CA 90095-7099 (United States); Leonard, A.W.; Osborne, T.H. [General Atomics, San Diego, CA 92186-5608 (United States)

    2013-07-15

    This paper documents first work toward validation of BOUT++ nonlinear edge localized mode (ELM) simulations in X-point geometry, at experimental pedestal collisionality, against multiple diagnostic measurements of a well-characterized ELM event in DIII-D. The key to the BOUT++ simulations is the use of a hyper-resistivity model that effectively spreads the very thin current sheets that form in low collisionality nonlinear simulations, and allows for ELM driven magnetic reconnection at finite current density. Experimental ELM characterization includes multiple fast line-integrated diagnostic measurements revealing in–out divertor asymmetric response to ELMs, IRTV imaging at the divertor targets, visible emission in the divertor volume to test the extension of BOUT++ to X-point geometry, and forward modeling of new electron cyclotron emission imaging to test predictions of ELM filaments in the edge pedestal. Initial comparisons suggest optimized BOUT boundary conditions and model parameters, and show similarities between initial BOUT++ results and several measurements.

  12. 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.

  13. Estimates of EAST Operation Window with LHCD by Using a Core-SOL-Divertor Model

    International Nuclear Information System (INIS)

    An experimental advanced superconducting tokamak (EAST) operation window with the lower hybrid current drive (LHCD) in H-mode is estimated by using a core-SOL-divertor (C-S-D) model validated by the present EAST divertor experiments. The operation window consists of four limits including two usual limits, one of which is the maximum allowable heat load onto the divertor plate, and two additional limits associated with the LHCD. The predictive EAST operation window is not qualified to fulfill its mission for high input power. To extend the operation window, gas puffing and impurity seeding are presented as two effective methods. In addition, the effect of the LHCD current on the operation window is also discussed. Our numerical analysis results provide a reference for the safe operation of EAST experiments with LHCD in future. (magnetically confined plasma)

  14. 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.

  15. Initial Study Comparing the Radiating Divertor Behavior in Single-Null and Double-Null Plasmas in DIII-D

    International Nuclear Information System (INIS)

    'Puff and pump' radiating divertor scenarios [1,2] were applied to upper SN and DN H-mode plasmas. Under similar operating conditions, argon (Ar) accumulated in the main plasma of single-null (SN) plasmas more rapidly and reached a higher steady-state concentration when the B x (del)B ion drift direction was toward the divertor than when the B x (del)B ion drift direction was out of the divertor. The initial rate that Ar accumulated inside double-null (DN) plasmas was more than twice that of comparably-prepared SNs with the same B x (del)B direction. One way to reduce power loading at the divertor targets is to 'seed' the divertor plasma with impurities that radiatively reduce the conducted power. Studies have shown that the concentration of impurities in the divertor are increased by raising the flow of deuterium ions (D+) into the divertor by a combination of upstream deuterium gas puffing and active particle exhaust at the divertor targets, i.e., puff-and-pump. An enhanced D+ particle flow toward the divertor targets exerts a frictional drag on impurities, and inhibits their escape from the divertor. A puff-and-pump approach using Ar as the impurity was successfully applied in recent DIII-D experiments to SN plasmas [3] while maintaining good H-mode performance. Studies on DIII-D and other tokamaks have shown that both the direction of the toroidal magnetic field BT and the degree of magnetic balance between divertors [i.e., the degree to which the plasma shape is considered SN or DN] are important factors in determining recycling and particle pumping [4,5]. It is unclear whether the favorable results of Ref. [3] can be extended to cases with different magnetic balance and/or BT direction. We show in this paper that reversing the direction of BT or altering the divertor magnetic balance does have an impact on how plasmas behave under puff-and-pump conditions. Our study takes advantage of DIII-D's capabilities to actively pump SN and DN shapes with high

  16. 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)

  17. 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.

  18. 3D effects of edge magnetic field configuration on divertor/SOL transport and optimization possibilities for a future reactor

    International Nuclear Information System (INIS)

    Recent progress on the experimental identification and physics interpretation of 3D effects of magnetic field geometry/topology on divertor transport is overviewed. In this paper, 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 process has strong impacts on the divertor functions, such as 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, are discussed, suggesting demanding issues to be addressed for divertor optimization in future reactors. (author)

  19. The installation of the JET pumped divertor systems inside the vacuum vessel

    International Nuclear Information System (INIS)

    The JET Pumped Divertor, designed to study impurity control will be installed during 1992/1993. The related in-vessel components are essentially: four magnetic coils, forty-eight beryllium divertor plate assemblies, one toroidal LHe cryopump, eight ICRH antennae, and poloidal limiters. With the exception of the magnetic coils, which will be would and brazed inside the vessel, all other components will be pre-assembled and prepared before installation. The paper describes the organization and the installation principles to undertake and to accomplish all the activities

  20. Divertor ‘death-ray’ explained: An artifact of a Langmuir probe operating at negative bias in a high-recycling divertor

    International Nuclear Information System (INIS)

    The divertor ‘death-ray’, enhanced plasma pressure near the outer strike-point relative to ‘upstream’ values, was thought to correspond to axisymmetric increased divertor heat flux. Recent measurements on Alcator C-Mod show that the ‘death-ray’ is localized to biased Langmuir probes. Heat fluxes deduced from plasma-sheath theory and surface thermocouples agree in sheath-limited and moderate-recycling regimes. They diverge in high-recycling and detached regimes; surface thermocouples measure reduced heat flux while a ‘death-ray’ appears on Langmuir probes. The ‘death-ray’ is caused by the probe’s negative bias affecting the local flux tube. With the bias, electron heat flux to the probe surface is reduced. Thus, the local electron temperature is raised, enhancing neutral ionization and increasing the ion flux to the probe. The plasma fluid code UEDGE is used to simulate and reproduce many of the features of this integrated biased probe/divertor system

  1. 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.

  2. 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.

  3. 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.

  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. 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

  7. 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.)

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

    Czech Academy of Sciences Publication Activity Database

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

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

  9. 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. The...

  10. The WEST project mechanical analysis of the divertor structure according to the nuclear construction code

    Energy Technology Data Exchange (ETDEWEB)

    Larroque, S., E-mail: sebastien.larroque@cea.fr [CEA Cadarache, IRFM, F-13108 Saint-Paul-lez-Durance (France); Portafaix, C. [ITER Organization, 13108 Saint-Paul-lez-Durance (France); Saille, A.; Doceul, L.; Bucalossi, J.; Samaille, F.; Freslon, S. de [CEA Cadarache, IRFM, F-13108 Saint-Paul-lez-Durance (France)

    2014-10-15

    Highlights: • Divertor structure is mainly loaded by electromagnetical forces. • A simplified FEM analysis give the stresses in the structure. • RCCM criteria are required for the sizing. • Refined finite element models are used for local overstresses. - Abstract: The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST project, launched in support to the ITER tungsten divertor strategy. The installation of coils inside the vacuum vessel led to the design of a divertor supporting platform able to meet the project requirements and the associated electromagnetic loads. This paper illustrates the design, the method and the results of the thermomechanical elastic stress analyses performed in 2012. The validation of the integrity of the structure is based on the compliance with RCCMR design criteria (even though these Design and Construction rules for Mechanical Components of nuclear installations are not required for such experimental fusion device). Several 3D analyses are performed with the ANSYS code. The major one is a global analysis of half structure which determinates the stresses in the main part of the components. It gives an idea of the areas which needs local analyses. It also provides the interface loads for junction studies or simplified local model.

  11. Finite element modelling of transport and drift effects in tokamak divertor and SOL

    International Nuclear Information System (INIS)

    A finite element code is used to simulate transport of a single-species plasma in the edge and divertor of a tokamak. The physical model is based on Braginskii's fluid equations for the conservation of particles, parallel momentum, ion and electron energy. In modelling recycling, transport of neutral density and energy is treated in the diffusion approximation. The electrostatic potential is obtained from the generalized Ohm's law. It is used to compute the electric field and the associated E x B drift. In a first approximation, transport is assumed to be ambipolar. The system of equations is discretized on an unstructured triangular mesh, thus permitting good spatial resolution near the X-point and an accurate description of divertor plates of arbitrary shape. Special care must be taken to prevent numerical corruption of the highly anisotropic thermal diffusion. Comparisons will be made between simulations and experimental results from TdeV. This will focus, in particular, on density and temperature profiles at the divertor plates, and on the plasma parallel velocity in the SOL. The asymmetry in the power deposited to the inner and outer divertors and the effect of magnetic field reversal will be considered. Comparisons with B2-Eirene simulation results will also be presented

  12. 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

  13. Safety implications of an integrated boiling curve model for water-cooled divertor channels

    International Nuclear Information System (INIS)

    The international fusion community is actively researching advanced heat transfer methods for removal of high thermal loads from next-generation divertor assemblies. Such advanced techniques may indeed optimize the operational and economical performance of future divertor designs. However, with its extensive operational database, water-cooling remains as one of the optimum choices for near-term divertor designs. Critical heat flux (CHF) is the maximum heat flux that water, at a given set of inlet conditions, can remove via fully developed nucleate boiling. Accordingly, an accurate CHF calculation is of the utmost importance for maintaining adequate safety margins in divertor operation. This paper uses the integrated boiling curve model developed at Sandia National Laboratories to examine the safety implications of calculating the CHF. In particular, this paper focuses on the influence of the finite element peaking factor (FEPF) that converts the heat flux predicted by CHF correlations into a plasma heat flux that can be measured. The analyses illustrate that the FEPF is proportional to the plasma heat flux and thus accurate calculation of the CHF requires the use of the appropriate FEPF for the given water conditions and plasma heat flux. It is shown that using a geometric peaking factor is inadequate since the true peak factor is dependent upon the plasma heat flux. The conclusion is that a finite element analysis incorporating an integrated boiling curve is required for accurate calculation of the CHF

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  16. 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 swellin...

  17. 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

  18. Evaluation of performance for the EAST upgraded divertor targets during type I ELMy H-mode

    Science.gov (United States)

    Qian, X. Y.; Peng, X. B.; Wang, L.; Song, Y. T.; Ye, M. Y.; Zhang, J. W.; Li, W. X.; Zhu, C. C.

    2016-02-01

    The long-pulse high-confinement (H-mode) plasma regime is considered to be a preferable scenario in future fusion devices, and in the period of normal operation during H-mode, edge-localised modes (ELMs) are one of the most serious threats to the performance and capability of divertor targets. The EAST recently achieved a variety of H-mode regimes with ELMs. For the purpose of studying the performance of the EAST upgraded divertor during type I ELMs, a series of simulations were performed by using three-dimensional (3D) finite element code. To make a visible outcome of the direct ELM impact on the divertor targets, a preliminary evaluation system with three indices to exhibit the influence has been developed. The indices that comprise temperature evolution, thermal penetration depth and crack initiation life, which could reveal the process of micro-crack formation, are calculated in both low and high-power scenarios for type I ELMs. The initial results indicate that the transient heat load has a significant influence in a very short thickness layer along the direction perpendicular to the plasma-facing surface throughout its duration. The conclusion could offer a pertinent guide to the next-step high-power long-pulse operation in EAST and would also be helpful for scientifically studying the damage and fatigue mechanism of the divertor in ITER and future fusion power reactors.

  19. 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

  20. The Design and Development of Divertor Remote Handling Equipment for ITER

    International Nuclear Information System (INIS)

    A key ITER maintenance activity is the complete exchange of the divertor system at scheduled intervals, typically after every 3-4 years of plasma operations. In view of this, ITER divertor maintenance is classified as an RH Class 1 activity and as such, detailed design of the associated RH equipment and verification of its operation before ITER construction by way of prototypes and mock-ups, is considered an essential activity. Throughout the course of the ITER design activities one of the major focuses of the EU contribution has been the study and development of remote handling equipment (RHE) necessary for divertor exchange. This suite of RHE will include a number of heavy in-vessel robotic transporters (known as '' ssette movers ''), ex-vessel transfer casks and several general purpose dextrous manipulators used to deploy and operate task-specific RH tooling. The current major step in the divertor RH development programme for ITER involves the construction of a full scale physical test facility in which to demonstrate and refine RHE designs through the operation of prototypes closely replicating those proposed for ITER. This facility, designated the '' Divertor Test Platform 2 (DTP2) '', will be constructed in Tampere, Finland and operated by the Finnish Fusion Association, TEKES. Four separate procurement contracts are currently being executed within European Industry for the supply of the major DTP2 sub-systems namely, a mock-up of the ITER divertor region, a mock-up cassette, a prototype cassette mover and the mover control hardware. The control system software development is being carried out in parallel by staff of the DTP2 host organisation. These DTP2 sub-systems will be brought together in Tampere during the Autumn 2006 / Spring 2007 and the system is expected to be ready for RH trials on the second cassette handling process in the summer of that year. Measures to extend the facility in subsequent years to allow more extensive trials in a 30 degree

  1. Design, Fabrication, and Installation of the Lower Divertor for DIII-D

    International Nuclear Information System (INIS)

    The geometry of the lower divertor of the DIII-D tokamak was modified to provide improved density control of the tokamak plasma during operation in a high-triangularity double null plasma. This divertor replaces the low triangularity Advanced Divertor in use since 1990. The design, analysis and fabrication were completed in 2005 and the installation was completed in March 2006. Plasma operations are planned for June 2006. The primary component of the lower divertor is a toroidally continuous flat cooling plate. Three rows of graphite tiles are mechanically attached to the plate to shield it from plasma impingement. The plate is water cooled for heat removal between shots and is heated to 350oC with hot air and inductive current during vessel baking. The divertor ring is supported 100 mm from the vacuum vessel floor by two rows of 24 supports that must react the vertical loads due to halo currents. The space below the ring forms a pumping plenum between the floor strike point and the lower cryopump. The divertor plate was fabricated by the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) in four 90 degree sectors from Type 316 stainless. Each sector consists of two plate halves with three machined coolant channels connected in parallel. Two plate halves are joined together by spot welds and perimeter TIG welds. During installation, the vacuum-tight 90 degree panel sectors were aligned and welded together inside the vessel forming a toroidally continuous ring. The water cooling/air bakeout lines connecting the 4 sectors into two 180 degree cooling circuits were then welded in place. The vacuum boundary for the cooling/air bakeout lines uses a reverse conflat design with the tubes welded into a modified, outward facing conflat flange. This design provides for copper gasket seal replacement without disturbing any welds. Plasma facing tile designs have been modified from previous designs to eliminate holes in high heat flux areas. Upgraded floor tiles

  2. Ergodic divertor experiments on the route to steady-state operation of Tore Supra

    International Nuclear Information System (INIS)

    Ergodic Divertor operation on Tore Supra is characterised by good performance in terms of divertor physics. Control of particle recirculation and impurity screening are related to the symmetry both poloidally and toroidally of the shell of open field lines and to its radial extent, Δr∼0.16m. Feedback control of the divertor plasma temperature has led to controlled radiative divertor experiments. In particular, good performance is obtained when the plasma is controlled to be a temperature comparable to the energy involved in the atomic processes, (15 to 20 eV). For standard discharges with 5 MW total power and ICRH heating, the low parallel energy flux ∼10 MW m-2 is reduced to ∼ 3MWm-2 with nitrogen injection. This is achieved at a modest cost in core dilution, Δ Zeff∼0.3. Despite the large volume of open field lines (∼36%) the Ergodic Divertor does not reduce the possible current in the discharge since stable discharges are achieved with qsep∼2. It is shown that the reorganisation of the current profile in conjunction with a transport barrier in the electron temperature on the separatrix stabilises the (2,1) tearing mode. Confinement follows the standard L-mode confinement. In a few cases at high density and with no gas injection (wall fuelled discharges), 'RI-like' modes are reported with modest increase in confinement (∼40%). Despite the lack of core fuelling on Tore Supra, high densities during ICRH pulses can be achieved with Greenwald fractions fG∼1. Compatibility with both ICRH and LH is demonstrated. In particular long pulse operation with flat top in excess of 20 s are achieved with LHCD. (author)

  3. 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)

  4. 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)

  5. 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)

  6. 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

  7. 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

  8. Self-consistent numerical analyses for scrape-off plasmas and neutral particles in a fer divertor chamber

    International Nuclear Information System (INIS)

    The present report deals with a numerical analysis of characteristics of poloidal divertors in the Fusion Experimental Reactor (abbreviated FER) now under the design study in JAERI. Diverted scrape-off plasmas are formulated and analyzed based on a fluid model including the interaction of the plasma with neutral particles through ionization and charge exchange reactions. The neutral particle behavior is calculated using Monte Carlo methods. The possibility of high density operation of the FER divertor is examined numerically and the pumping requirement for the helium ash exhaust is discussed. It is also shown that the same numerical model gives the results qualitatively consistent with the DIII divertor experiments. (orig.)

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

    International Nuclear Information System (INIS)

    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

  10. The WEST project: Testing ITER divertor high heat flux component technology in a steady state tokamak environment

    Energy Technology Data Exchange (ETDEWEB)

    Bucalossi, J., E-mail: jerome.bucalossi@cea.fr; Missirlian, M.; Moreau, P.; Samaille, F.; Tsitrone, E.; Houtte, D. van; Batal, T.; Bourdelle, C.; Chantant, M.; Corre, Y.; Courtois, X.; Delpech, L.; Doceul, L.; Douai, D.; Dougnac, H.; Faïsse, F.; Fenzi, C.; Ferlay, F.; Firdaouss, M.; Gargiulo, L.; and others

    2014-10-15

    The WEST project recently launched at Cadarache consists in transforming Tore Supra in an X-point divertor configuration while extending its long pulse capability, in order to test the ITER divertor technology. The implementation of a full tungsten actively cooled divertor with plasma facing unit representative of ITER divertor targets will allow addressing risks both in terms of industrial-scale manufacturing and operation of such components. Relevant plasma scenarios are foreseen for extensive testing under high heat load in the 10–20 MW/m{sup 2} range and ITER-like fluences (1000 s pulses). Plasma facing unit monitoring and development of protection strategies will be key elements of the WEST program. WEST is scheduled to enter into operation in 2016, and will provide a key facility to prepare and be prepared for ITER.

  11. 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.

  12. Nuclear modules of ITER tokamak systems code

    International Nuclear Information System (INIS)

    Nuclear modules were developed to model various reactor components in the ITER systems code. These modules include first wall, tritium breeding blanket (or shield), bulk shield, reactor vault, impurity control, and tritium system. The function of these modules is to define the performance parameters for each component as a function of the reactor operating conditions. Several design options and cost algorithms are included for each component. The first wall, blanket and shield modules calculate the beryllium zone thickness, the disruptions results, the nuclear responses in different components including the toroidal field coils. Tungsten shield/water coolant/steel structure and steel shield/water coolant are the shield options for the inboard and outboard sections of the reactor. Lithium nitrate dissolved in the water coolant with a variable beryllium zone thickness in the outboard section of the reactor provides the tritium breeding capability. The reactor vault module defines the thickness of the reactor wall and the roof based on the dose equivalent during operation including skyshine contribution. The impurity control module provides the design parameters for the divertor including plate design, heat load, erosion rate, tritium permeation through the plate material to the coolant, plasma contamination by sputtered impurities, and plate lifetime. Several materials: Be, C, V, Mo, and W can be used for the divertor plate to cover a range of plasma edge temperatures. The tritium module calculates tritium and deuterium flow rates for the reactor plant. The tritium inventory in the fuelers, neutral beams, vacuum pumps, impurity control, first wall, and blanket is calculated. Tritium requirements are provided for different operating conditions. The nuclear models are summarized in this paper including the different design options and key analyses of each module

  13. 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

  14. 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

  15. 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

  16. 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

  17. Progress on radiative transfer modelling in optically thick divertor plasmas

    International Nuclear Information System (INIS)

    The physical model used in the photon transport module of the Monte-Carlo code EIRENE (http://www.eirene.de) is presented. A critical assessment of the spectral line broadening mechanisms (which give the shape of the photon-atom reaction rates) and of their relevance in a transport simulation is carried out (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  18. 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.

  19. Design and analysis of a low edge temperature divertor for INTOR

    International Nuclear Information System (INIS)

    A low plasma edge temperature regime has been proposed for the INTOR divertor in order to minimize erosion on the collector plate. We have examined the design and lifetime issues of a divertor plate operating in such a regime. These issues include the choice of materials, erosion due to sputtering and disruptions, thermal response, and tritium permeation. We have concluded that a tungsten coated, copper substrate plate offers the possibility of low erosion coupled with good thermal properties. However, there are many unresolved issues with this and other plate configurations. These issues include the transfer of first wall sputtered material to the plate, tritium permeation into the coolant, materials problems such as radiation swelling, embrittlement, and joining

  20. Simulation of hydrogen retention and re-emission from tungsten exposed to divertor plasmas

    International Nuclear Information System (INIS)

    In order to study the diffusion, surface recombination and trapping of hydrogen isotopes in tungsten, the processes are incorporated into a Monte Carlo code, EDDY. After estimating the relevant parameters by simulating an existing thermal desorption spectroscopy (TDS) experiment with the code, the deuterium retention characteristics of a tungsten divertor plate are investigated for an ITER plasma configuration. At the inner divertor plate and the early stage of discharge, the number of mobile D atoms reaches a steady-state value due to the balance between implanted and re-emitted atoms. During discharge, most of the implanted D atoms are retained in traps, and after saturating available traps, inward diffusion and subsequent trapping increase the inventory. The saturation distributes deeply into W at the strike point of the plate, where the surface temperature is higher than that at distance away from it. Most of the trapped atoms remain in the bulk even after discharge, whereas mobile atoms are strongly reduced.

  1. 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

  2. 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.

  3. 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.

  4. A 250 GHz microwave interferometer for divertor experiments on DIII-D

    International Nuclear Information System (INIS)

    A new 250 GHz, two-frequency microwave interferometer system has been developed to diagnose divertor plasmas on DIII-D. This diagnostic will measure the line-averaged density across both the inner and outer, lower divertor legs. With a cut-off density of over 7 x 1014 cm-3, temporal measurements of ELMs, MARFs and plasma detachment are expected. The outer leg system will use a double pass method while the inner leg system will be single pass. Two special 3D carbon composite tiles are used, one to protect the microwave antennas mounted directly under the strike point and the other as the outer leg reflecting surface. Performance, design constraints, and the thermalmechanical design of the 3D carbon composite tiles are discussed

  5. 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

  6. 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

  7. 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

  8. 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)

  9. Activation of TZM and stainless steel divertor materials in the NET fusion machine

    International Nuclear Information System (INIS)

    This paper presents the results of the activation and decay heat calculations for the divertor plate materials of the Next European Torus (NET). The basic option assessed enables molybdenum alloy TZM and AISI 316L as material for divertor cooling channels. Burn time, effective irradiation time history, and fluence dependence on activation, decay heat, and contact dose is assessed. Impact of the material impurity level on the radioactive inventory is also investigated. The ANITA code is used, with updated cross sections and decay data libraries based on EFF-2 and EAF-3 evaluation files. The flux-weighted spectrum provided by XSDRNPM or ANISN 1-D codes has been used. The real NET geometry was modelled with the 3-D MCNP Monte Carlo neutron transport code. ((orig.))

  10. 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.

  11. The development of in-situ calibration method for divertor IR thermography in ITER

    International Nuclear Information System (INIS)

    For the development of the calibration method of the emissivity in IR light on the divertor plate in ITER divertor IR thermography system, the laboratory experiments have been performed by using IR instruments. The calibration of the IR camera was performed by the plane black body in the temperature of 100–600 degC. The radiances of the tungsten heated by 280 degC were measured by the IR camera without filter (2.5–5.1 μm) and with filter (2.95 μm, 4.67 μm). The preliminary data of the scattered light of the laser of 3.34 μm that injected into the tungsten were acquired

  12. 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.

  13. 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.

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

    International Nuclear Information System (INIS)

    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

  15. Divertor target profiles and recycling studies in TCV single null lower standard discharges

    International Nuclear Information System (INIS)

    A 'standard', single null lower diverted discharge has been developed to enable continuous monitoring of the first wall conditions and to characterise the effectiveness and influence of wall conditioning in the TCV tokamak. Measurements over a period encompassing nearly 2000 ohmic discharges of varying configuration and input power show the global confinement time and main plasma impurity concentrations to be good general indicators of the first wall condition, whilst divertor target profiles demonstrate strikingly the short term beneficial effects of He glow. Good agreement, consistent with a reduction in recycling at the plates is found between the predictions of the fluid code UEDGE and the observed outer divertor profiles of Te and ne before and after He glow. (author) 5 figs., 7 refs

  16. 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.

  17. 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.

  18. 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.

  19. Thermal fatigue resistance of W-Cu divertor plates for fusion reactors

    International Nuclear Information System (INIS)

    The results of thermal cycling tests of W-Cu pseudo-alloy as a candidate material for fusion reactor divertor plates showed that the material resisted without any damages to radiation and thermal shock effects of cyclic electron beam of 6, 8 and 10 MW/m2 power density when a good thermal sink was provided. Practically ideal thermal contact between thermally loaded sample and cooled substrate was shown can be obtained using various spelters vacuum brazing. 2 refs.; 4 figs

  20. A tangentially viewing VUV TV system for the DIII-D divertor

    Energy Technology Data Exchange (ETDEWEB)

    Nilson, D.G.; Ellis, R.; Fenstermacher, M.E. [Lawrence Livermore National Lab., CA (United States); Brewis, G. [Allied Optical (United States); Jalufka, N. [Hampton Univ., VA (United States). Center for Fusion Research and Training

    1998-07-01

    A video camera system capable of imaging VUV emission in the 120--160 nm wavelength range, from the entire divertor region in the DIII-D tokamak, was designed. The new system has a tangential view of the divertor similar to an existing tangential camera system which has produced two dimensional maps of visible line emission (400--800 nm) from deuterium and carbon in the divertor region. However, the overwhelming fraction of the power radiated by these elements is emitted by resonance transitions in the ultraviolet, namely the C IV line at 155.0 nm and Ly-{alpha} line at 121.6 nm. To image the ultraviolet light with an angular view including the inner wall and outer bias ring in DIII-D, a 6-element optical system (f/8.9) was designed using a combination of reflective and refractive optics. This system will provide a spatial resolution of 1.2 cm in the object plane. An intermediate UV image formed in a secondary vacuum is converted to the visible by means of a phosphor plate and detected with a conventional CID camera (30 ms framing rate). A single MgF{sub 2} lens serves as the vacuum interface between the primary and secondary vacuums; a second lens must be inserted in the secondary vacuum to correct the focus at 155 nm. Using the same tomographic inversion method employed for the visible TV, they reconstruct the poloidal distribution of the UV divertor light. The grain size of the phosphor plate and the optical system aberrations limit the best focus spot size to 60 {micro}m at the CID plane. The optical system is designed to withstand 350 C vessel bakeout, 2 T magnetic fields, and disruption-induced accelerations of the vessel.

  1. 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)

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

    International Nuclear Information System (INIS)

    Re-evaluation of tritium permeation through vertical target of divertor under the ITER operation condition was carried out using tritium properties in the candidate materials such as the diffusion coefficient and the trapping factors in tungsten for armor, and the surface recombination coefficient on copper for the heat sink obtained by authours' recent investigation (authors' data), which simulated the plasma-facing conditions of ITER. Evaluation with the data set of previous evaluation was also carried out for comparison (previous data). The permeation analysis was carried out individually by classifying into the armor region (Carbon Fiber Composites and tungsten) and the slit region without armor (3% of armor surface area) assuming the incident flux and temperature for each region. As the results of the permeation analysis, estimated permeation amount with the authors' data was one order less than that with the previous data at the end of lifetime of the divertor due to authors' small diffusion coefficient of tritium in tungsten. It also indicated the possibility that permeation through the slit region of the armor tiles could dominate total permeation through the vertical target, since tritium permeation amount through tungsten armor with the authors' data was estimated to be reduced drastically smaller than that with the previous evaluation data. The result of a little tritium permeation amount through the vertical target with the authors' data ensured the conservatism of the current evaluation of tritium concentration in the primary cooling water in ITER divertor, as it indicated the possibility of direct drainage of the divertor primary cooling water. (author)

  3. Development of tungsten and tungsten alloys for DEMO divertor applications via MIM technology

    Energy Technology Data Exchange (ETDEWEB)

    Blagoeva, D.T., E-mail: blagoeva@nrg.eu [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Opschoor, J. [Energy Research Center of the Netherlands (ECN), Petten (Netherlands); Laan, J.G. van der [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Sârbu, C. [National Institute for Materials Physics (NIMP), Măgurele-Bucharest (Romania); Pintsuk, G. [Forschungszentrum Jülich GmbH, Jülich (Germany); Jong, M.; Bakker, T.; Ten Pierick, P.; Nolles, H. [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands)

    2013-11-15

    This paper is an overview of the very first results obtained on pure tungsten (W) and oxide dispersed strengthened (ODS) W alloys produced by the Metal Injection Molding (MIM) technique for fusion applications. An extensive mechanical and physical characterization was performed, together with microstructural material investigation. The reported work was accomplished within the framework of the European Fusion Development Agreement work program. The main objective was to develop suitable tungsten grades for structural and armor divertor applications in the future DEMO fusion reactor.

  4. 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.

  5. 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.

  6. 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.

  7. 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

  8. 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)

  9. 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.

  10. 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

  11. 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.

  12. The role of plasma response in divertor footprint modification by 3D fields in NSTX

    Science.gov (United States)

    Ahn, Joonwook; Kim, Kimin; Canal, Gustavo; Gan, Kaifu; Gray, Travis; McLean, Adam; Park, Jong-Kyu; Scotti, Filippo

    2015-11-01

    In NSTX, the divertor footprints of both heat and particle fluxes are found to be significantly modified by externally applied 3D magnetic perturbations. Striations on the divertor surface, indicating separatrix splitting and formation of magnetic lobes, are observed for both n = 1 and n = 3 perturbation fields. These striations can lead to localized heating of the divertor plates and to the re-attachment of detached plasmas, both of which have to be avoided in ITER for successful heat flux management. In this work, the role of plasma response on the formation of separatrix splitting has been investigated in the ideal framework by comparing measured heat and particle flux footprints with field line tracing calculations with and without contributions from the plasma response calculated by the ideal code IPEC. Simulations show that, n = 3 fields are slightly shielded by the plasma, with the measured helical pattern of striations in good agreement with the results from the vacuum approximation. The n = 1 fields are, however, significantly amplified by the plasma response, which provides a better agreement with the measurements. Resistive plasma response calculations by M3D-C1 are also in progress and the results will be compared with those from the ideal code IPEC. This work was supported by DoE Contracts: DE-AC05-00OR22725, DE-AC52-07NA27344 and DE-AC02-09CH11466.

  13. 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.

  14. Critical issues identified by the ASDEX Upgrade edge and divertor modelling

    International Nuclear Information System (INIS)

    A detailed comparison between the ASDEX Upgrade (AUG) experimental data and results of the SOLPS 2D edge code simulations has recently been performed. High quality upstream profiles of electron density and ion and electron temperatures in the scrape-off layer (SOL) of AUG have been collected for two shots with different upstream collisionalities: a low density ELMy H-mode shot (low collisionality) and a medium density Ohmic shot (higher collisionality). A generally broad agreement, within a factor 2, considering basic parameters characterising the divertor, has been reached between simulations and experiment. In both Ohmic and H-mode shots, however, the tendency of SOLPS solutions to underestimate the divertor electron temperature and overestimate its density has been reliably established. Two main possible causes of the discrepancies have been considered: some deficiencies in the neutral modelling (e.g. missing atomic and molecular reactions in EIRENE, the Monte-Carlo neutral part of SOLPS), and the presence of a significant population of supra-thermal ions and electrons in the SOL and divertor plasma. The results of dedicated SOLPS runs where the sensitivity of the code solution to various assumptions of the neutral model and parallel heat transport of ions and electrons are described. A comparison between simulated and experimentally measured Mach numbers of the parallel ion flow in the SOL is presented, and conditions necessary for obtaining fast flows in the code are analysed. (author)

  15. 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

  16. 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.)

  17. Analysis of a multi-machine database on divertor heat fluxes

    Energy Technology Data Exchange (ETDEWEB)

    Makowski, M. A.; Lasnier, C. J. [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States); Elder, D.; Stangeby, P. C. [University of Toronto Institute of Aerospace Studies, Toronto M3H 5T6 (Canada); Gray, T. K.; Maingi, R. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); LaBombard, B.; Terry, J. L. [Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Leonard, A. W.; Osborne, T. H. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Watkins, J. [Sandia National Laboratory, Albuquerque, New Mexico 87185 (United States)

    2012-05-15

    A coordinated effort to measure divertor heat flux characteristics in fully attached, similarly shaped H-mode plasmas on C-Mod, DIII-D, and NSTX was carried out in 2010 in order to construct a predictive scaling relation applicable to next step devices including ITER, FNSF, and DEMO. Few published scaling laws are available and those that have been published were obtained under widely varying conditions and divertor geometries, leading to conflicting predictions for this critically important quantity. This study was designed to overcome these deficiencies. Analysis of the combined data set reveals that the primary dependence of the parallel heat flux width is robustly inverse with I{sub p}, which all three tokamaks independently demonstrate. An improved Thomson scattering system on DIII-D has yielded very accurate scrape off layer (SOL) profile measurements from which tests of parallel transport models have been made. It is found that a flux-limited model agrees best with the data at all collisionalities, while a Spitzer resistivity model agrees at higher collisionality where it is more valid. The SOL profile measurements and divertor heat flux scaling are consistent with a heuristic drift based model as well as a critical gradient model.

  18. Analysis of a multi-machine database on divertor heat flux

    Energy Technology Data Exchange (ETDEWEB)

    Makowski, M. A. [Lawrence Livermore National Laboratory (LLNL); Elder, J. D. [University of Toronto, Toronto, ON, Canada; Gray, Travis K [ORNL; LaBombard, Brian [Massachusetts Institute of Technology (MIT); Lasnier, C. J. [Lawrence Livermore National Laboratory (LLNL); Leonard, A. W. [General Atomics; Maingi, Rajesh [ORNL; Osborne, T. H. [General Atomics; Stangeby, P. C. [University of Toronto Institute for Aerospace Studies; Terry, J. L. [MIT Plasma Science & Fusion Center, Cambridge, MA 02139 USA; Watkins, J. G. [Sandia National Laboratories (SNL)

    2012-01-01

    A coordinated effort to measure divertor heat flux characteristics in fully attached, similarly shaped H-mode plasmas on C-Mod, DIII-D, and NSTX was carried out in 2010 in order to construct a predictive scaling relation applicable to next step devices including ITER, FNSF, and DEMO. Few published scaling laws are available and those that have been published were obtained under widely varying conditions and divertor geometries, leading to conflicting predictions for this critically important quantity. This study was designed to overcome these deficiencies. Analysis of the combined data set reveals that the primary dependence of the parallel heat flux width is robustly inverse with I-p, which all three tokamaks independently demonstrate. An improved Thomson scattering system on DIII-D has yielded very accurate scrape off layer (SOL) profile measurements from which tests of parallel transport models have been made. It is found that a flux-limited model agrees best with the data at all collisionalities, while a Spitzer resistivity model agrees at higher collisionality where it is more valid. The SOL profile measurements and divertor heat flux scaling are consistent with a heuristic drift based model as well as a critical gradient model.

  19. 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

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

    International Nuclear Information System (INIS)

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

  1. 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.

  2. 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

  3. Extreme Ultraviolet Spectra of Few-Times Ionized Tungsten for Divertor Plasma Diagnostics

    Directory of Open Access Journals (Sweden)

    Joel Clementson

    2015-09-01

    Full Text Available The extreme ultraviolet (EUV emission from few-times ionized tungsten atoms has been experimentally studied at the Livermore electron beam ion trap facility. The ions were produced and confined during low-energy operations of the EBIT-I electron beam ion trap. By varying the electron-beam energy from around 30–300 eV, tungsten ions in charge states expected to be abundant in tokamak divertor plasmas were excited, and the resulting EUV emission was studied using a survey spectrometer covering 120–320 Å. It is found that the emission strongly depends on the excitation energy; below 150 eV, it is relatively simple, consisting of strong isolated lines from a few charge states, whereas at higher energies, it becomes very complex. For divertor plasmas with tungsten impurity ions, this emission should prove useful for diagnostics of tungsten flux rates and charge balance, as well as for radiative cooling of the divertor volume. Several lines in the 194–223 Å interval belonging to the spectra of five- and seven-times ionized tungsten (Tm-like W VI and Ho-like W VIII were also measured using a high-resolution spectrometer.

  4. ITER operation window determined from mutually consistent core-SOL-divertor simulations: definition and application

    International Nuclear Information System (INIS)

    An operating window for ITER is defined based on mutually consistent core-SOL-divertor modelling, in which the core turbulent transport is based on the Weiland formulation as incorporated into the multi-mode model. The window consists of five limits, one of which is the edge-based density limit based on divertor detachment. The predicted operating space is ample for ITER to fulfil its mission, reaching a maximum Q ∼ 60 at Palpha ∼ 150 MW (65% of the edge-based density limit, 1.1 times the Greenwald limit), and a maximum Palpha of 220 MW at Q ∼ 15 (90% of the edge-based density limit, 1.45 times the Greenwald limit). This operating window takes into account physics constraints and the technical constraints imposed by the divertor system, i.e. peak power load and attainable pumping speed, but does not include further constraints arising from other technological aspects of the ITER design, such as first wall cooling or shielding, which may further limit operation at high fusion power. The operating window is still compatible with the ITER mission if the magnetic field were reduced by 5% or if the underlying core transport were GLF-like rather than Weiland-like. A moderate reduction in helium exhaust or in pumping speed could be accommodated. Other changes in the operating window resulting from different technical or physical hypotheses are also evaluated

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

    International Nuclear Information System (INIS)

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

  6. 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

  7. Quantification of chemical erosion in the divertor of the DIII-D tokamak

    Science.gov (United States)

    McLean, Adam Gordon

    The International Thermonuclear Experimental Reactor (ITER) is currently designed to use graphite targets in the divertor for power handling and impurity control. Understanding and quantifying chemical sputtering is therefore key to the success of fusion as a clean energy source. The principal goal of this thesis is to design and carry out experiments, then analyze and interpret the results in order to elucidate the role of chemical sputtering in carbon sources in the DIII-D tokamak. A self-contained gas puff system has been designed, constructed, and employed for in-situ study of chemical erosion. The porous plug injector (PPI) releases methane through a porous graphite surface into the divertor plasma at a precisely calibrated rate, minimizing perturbation to local plasma while replicating the immediate environment of methane molecules released from a solid graphite surface more accurately than done previously. For the first time in a tokamak environment, the methane flow rate used in a puffing experiment was the same order of magnitude as that expected from laboratory experiments for intrinsic chemical sputtering. Effective photon efficiencies for CH4 injection are reported; results are found to have significant dependencies on surface conditions and the divertor operating regime. The contribution of sputtering processes to sources of C0 and C+ are assessed through measurement of background and incremental spectroscopic emissions of both physically and chemically-released sputtering products and by CI, 910 nm line profile fitting. Comparison of background and incremental emissions of chemically-released products demonstrate a dramatic drop in production of CH in cold and detached conditions. Finally, the chemical erosion yield is calculated in both attached and cold-divertor conditions and found to be much closer to that measured ex-situ in ion beam experiments than previously determined in DII-D. These observations represent a positive result for ITER which

  8. 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)

  9. 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.

  10. Acceleration and redeposition of a dust particle in SOL/divertor plasma of HL-2A Tokamak

    International Nuclear Information System (INIS)

    Acceleration of a iron dust particle is studied in the SOL (Scrape-Off Layer)/divertor plasma of the HL-2A tokamak in Southwestern Institute of Physics, China, with a single-null configuration. In this study the simplest model of the dust dynamics is applied: spherical shape of a dust, ion drag force due to Coulomb scattering and drag force due to the plasma ion absorption as dominant forces, and spontaneous charging of a dust particle to the equilibrium charge. In the outer region near the plasma-facing wall the parallel plasma flow along the magnetic field pushes the dust particle to the divertor plates. It is clarified that the dust particle with a radius of 1mm from the top of the SOL/divertor region is accelerated up to around 100 m/s in a few hundreds milliseconds to the divertor plates. In our model even the small dust particle with 1nm size can overcome the sheath potential and redeposite on the divertor plate (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  11. Results and consequences of high heat flux testing as quality assessment of the Wendelstein 7-X divertor

    International Nuclear Information System (INIS)

    Highlights: • One third of the series production of the actively cooled divertor target elements for W7-X is completed. • A statistical analysis of the CFC bonding quality was developed on the basis of high heat flux tests of 100 cycles at 10 MW/m2. • The application of this method on the 14,000 CFC tiles allows a reliable quality assessment with reasonable test effort. • The results confirm the high thermal performance of the delivered divertor elements. -- Abstract: The series manufacturing of the first 282 Wendelstein 7-X divertor elements was concluded in 2011. The divertor is designed to remove a steady-state heat load of 10 MW/m2. 940 target elements of five different types made of CuCrZr heat sinks and covered with 16,000 CFC NB31 flat-tiles have to be produced. Additional to quality assessment during the manufacturing process, a final assessment of the delivered elements with operational heat load is indispensable to ensure a constant high thermal performance of the installed divertor. Based on the results of the pre-series testing a statistical quality assessment method has been developed for the series production. The application of this method to the series elements ensures their thermal performance with reasonable high heat flux test effort

  12. 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

  13. 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

  14. Detailed electromagnetic analysis for optimization of a tungsten divertor plate for JET

    International Nuclear Information System (INIS)

    The ITER-like wall project at JET involves the replacement of the divertor tiles by either tungsten-coated carbon fibre composite (CFC) or solid tungsten. The background is a full replacement of CFC in order to avoid tritium retention due to co-deposition of carbon. In a R-and-D phase (T.Hirai et al., R-and-D on full tungsten divertor and beryllium wall for JET ITER-like Wall Project.), both tungsten coating and solid tungsten are investigated. Tungsten has a high electrical conductivity, exceeding that of graphite or CFC by two orders of magnitude. This drawback has to be compensated by a proper design (Ph. Mertens et al., Conceptual Design for a Bulk Tungsten Divertor Tile in JET (both citations: this conference)). This report shows how detailed electromagnetic consideration has influenced the design of the solid tungsten divertor for JET. Patterns and sum values were calculated for: (1) eddy currents induced by variation of two orthogonal magnetic fields; (2) toroidal eddy current induced by variation of the poloidal magnetic flux, (3) eddy-current related loads in three orthogonal magnetic fields; (4) Halo current pattern for two cases; (5) Halo-current related loads in three orthogonal magnetic fields; (6) the worst loads combinations; (7) stresses in fixtures. Analytical and numerical methods were combined and cross-checked. The load-bearing septum replacement plate (LB-SRP) which is currently used in the JET divertor consists of two large CFC tiles attached to two superimposed Inconel frames, namely wedge and adapter. The present design is quite loaded by eddy-currents and does not allow for simple replacement of the CFC with solid tungsten. A tree-like shape, which excludes large contours of eddy currents, is proposed. In realization of the tree-like shape, the wedge has a narrow middle part, elongated in radial direction, and eight wings, elongated in toroidal direction. Eight feet form the Halo current path. Each wing carries one tungsten lamellae stack

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

    International Nuclear Information System (INIS)

    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, - ∼ 1021 m-3, - plasma) typical for large tokamak devices. This is especially critical for the modelling of the divertor plasma of ITER: an international project of an experimental tokamak fusion reactor to be built in Cadarache, France by 2016. This present work is devoted to a major upgrade of the B2-EIRENE package, which is routinely used for ITER modelling, essentially with a significantly revised version of EIRENE: the Monte-Carlo neutral transport code. The main part of the thesis address three major groups of the new physical effects which have been added to the model in frame of this work: the neutral-neutral collisions, the up-to date hydrogen molecular reaction kinetics and the line radiation transport. The impact of the each stage of the upgrade on the self-consistent (between plasma, the neutral gas and the radiation field) solution for the reference ITER case is analysed. The strongest effect is found to be due to the revised molecular collision kinetics, in particular due to hitherto neglected elastic collisions of hydrogen molecules with ions. The newly added non-linear effects (neutral-neutral collisions, radiation opacity) are

  16. 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.

  17. COMPARISON OF ELM PULSE PROPAGATION IN THE DIII-D SOL AND DIVERTORS WITH AN ION CONVECTION MODEL

    International Nuclear Information System (INIS)

    OAK-B135 Results from dedicated ELM experiments, performed in DIII-D with fast diagnostics to measure the evolution of Type-I ELM effects in the SOL and divertor, are compared with a simple ion convection model and with initial time-dependent UEDGE simulations. Delays between ELM effects observed in the inner versus the outer divertor regions in the experiments scale, as a function of density, with the difference in ion convection time along field lines from the outer midplane to the divertor targets. The ELM perturbation was modeled as an instantaneous radially uniform increase of diffusion coefficients from the top of the pedestal to the outer SOL. The perturbation was confined to a low field side poloidal zone ± 40o from the outer midplane. The delays in the simulations are similar to those observed in the experiments

  18. Assessment of the effect of parallel temperature gradients in the JET SOL on Te measured by divertor target Langmuir probes

    International Nuclear Information System (INIS)

    Higher than expected electron temperatures (Te) are often measured by divertor Langmuir probes (LP) in high recycling and detached regimes in JET and other tokamaks. As a possible mechanism to explain this discrepancy, we investigate the effect of penetration of fast, almost collisionless electrons connecting the hot upstream scrape-off layer (SOL) region to the divertor targets in JET. We simulate the electron velocity distribution function (EVDF) near the divertor targets using a simple 1D kinetic model using parallel SOL profiles from EDGE2D-EIRENE simulations. The resulting EVDF is used to construct synthetic LP current–voltage (IV) characteristics and evaluation of Te is performed in the same way as for experimental data. Results indicate that the process does not explain the anomalously high Te values estimated from the target probe measurements if the EDGE2D-EIRENE simulated parallel profiles are a good representation of reality

  19. 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.

  20. 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.; Groth, M.; Brezinsek, S.; Naulin, Volker

    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 gas...... fuelled ramp-up scenario modelling of the full transient from attached divertor into the high-recycling regime, following a target flux roll over into divertor detachment, ultimately ending in a density limit is presented. A strong dependence on divertor geometry is observed which can mask features of the...... new transport model: a smoothly decaying target recycling flux roll over, an asymmetric drop of temperature and pressure along the field lines as well as macroscopic power dependent plasma oscillations near the density limit which had been previously observed also experimentally. The latter effect is...