WorldWideScience

Sample records for be-cu divertor modules

  1. Manufacturing and testing of a Be/OFHC-Cu divertor module

    International Nuclear Information System (INIS)

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

    1996-01-01

    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 x 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. (orig.)

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

    International Nuclear Information System (INIS)

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

    1992-01-01

    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 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)

  3. 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.

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

    International Nuclear Information System (INIS)

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

    1994-08-01

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

  5. Handling and maintenance procedure for replaceable parts, (blanket modules and divertors) of DEMO-DN

    International Nuclear Information System (INIS)

    Gaffka, R.C.; Hopkins, M.; Lavender, K.E.; McCarthy, C.E.A.

    1987-01-01

    The task of examining the reactor relevance of the NET-DN design must include a handling and maintenance procedure for parts which require replacement since high machine availability is an important requirement for a commercial reactor. This paper considers one certain requirement for the handling and maintenance of a DEMO-DN reactor and lists the constraints imposed by following the NET-DN design. A support and handling system for the replaceable parts (blanket modules and divertors) is described and an estimate is given for the maintenance down-time. (author)

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

    International Nuclear Information System (INIS)

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

    1992-01-01

    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 2π 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 ( ∼ 2π ). (Author) 2 refs

  7. 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.

  8. Experimental and numerical investigation of the thermal performance of gas-cooled divertor modules

    Science.gov (United States)

    Crosatti, Lorenzo

    Divertors are in-vessel, plasma-facing, components in magnetic-confinement fusion reactors. Their main function is to remove the fusion reaction ash (alpha-particles), unburned fuel, and eroded particles from the reactor, which adversely affect the quality of the plasma. A significant fraction (˜15 %) of the total fusion thermal power is removed by the divertor coolant and must, therefore, be recovered at elevated temperature in order to enhance the overall thermal efficiency. Helium is the leading coolant because of its high thermal conductivity, material compatibility, and suitability as a working fluid for power conversion systems using a closed high temperature Brayton cycle. Peak surface heat fluxes on the order of 10 MW/m2 are anticipated with surface temperatures in the region of 1,200 °C to 1,500 °C. Recently, several helium-cooled divertor designs have been proposed, including a modular T-tube design and a modular "finger" configuration with jet impingement cooling from perforated end caps. Design calculations performed using the FLUENTRTM CFD software package have shown that these designs can accommodate a peak heat load of 10 MW/m2. Extremely high heat transfer coefficients (˜50,000 W/(m2•K)) were predicted by these calculations. Since these values of heat transfer coefficient are considered to be "outside of the experience base" for gas-cooled systems, an experimental investigation has been undertaken to validate the results of the numerical simulations. Attention has been focused on the thermal performance of the T-tube and the "finger" divertor designs. Experimental and numerical investigations have been performed to support both divertor geometries. Excellent agreement has been obtained between the experimental data and model predictions, thereby confirming the predicted performance of the leading helium-cooled divertor designs for near- and long-term magnetic fusion reactor designs. The results of this investigation provide confidence in the

  9. Magnetic divertors

    International Nuclear Information System (INIS)

    Keilhacker, M.

    1978-01-01

    The different needs for divertors in large magnetic confinement experiments and prospective fusion reactors are summarized, special emphasis being placed on the problem of impurities. After alternative concepts for reducing the impurity level are touched on, the basic principle and the different types of divertors are described. The various processes in the scrape-off and divertor regions are discussed in greater detail. The dependence of the effectiveness of the divertor on these processes is illustrated from the examples of an ASDEX/PDX-size and a reactor-size tokamak. Various features determining the design of a divertor are dealt with. Among the physical requirements are the stability of the plasma column and divertor throat and the problems relating to the start-up phase. On the engineering side, there are requirements on the pumping speed and energy deposition, and for a reactor, the need for superconducting coils, neutron shields and remote disassembly

  10. 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.

  11. 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)

  12. 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.

  13. Complex investigation of several silver-less brazed Be/CuCrZr joints

    International Nuclear Information System (INIS)

    Komarov, A.; Gervash, A.; Komarov, V.; Mazul, I.; Litounovski, N.; Fedotov, V.; Sevrukov, O.; Ganenko, A.

    1998-01-01

    One of the main problems for ITER divertor target technology is to provide a reliable joint between Be as armour material and copper alloy as heat-sink structure. Such joints should satisfy the different requirements. In particular, these joints should successfully withstand cyclic heat fluxes and should have good properties under neutron irradiation. To study such complex of problems several investigation stages were planned in Russia. This paper presents the results of complex investigation of several silver-less brazed Be/CuCrZr joint candidates. (author)

  14. Advanced divertor concepts

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  15. Actively convected liquid metal divertor

    International Nuclear Information System (INIS)

    Shimada, Michiya; Hirooka, Yoshi

    2014-01-01

    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. (letter)

  16. 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.

  17. Conceptual design of CFETR divertor remote handling compatible structure

    Energy Technology Data Exchange (ETDEWEB)

    Dai, Huaichu, E-mail: yaodm@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei (China); Yao, Damao; Cao, Lei; Zhou, Zibo; Li, Lei [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2016-11-15

    Highlights: • Conceptual design for the CFETR divertor have been proposed, especially the divertor remote handling compatible structure. • The degrees of freedom of the divertor are analyzed in order to validate the design the divertor supports structure. • Besides the ITER-like scheme, a new scheme for the divertor remote handling compatible supports is proposed, that is the rack and pinion mechanism. • The installation/removel process is verified through simulation in Delmia in order to check design quality for remote handling requirements. - Abstract: Divertor is one of key components of tokamak fusion reactor. The CFETR is China Fusion Engineering Test Reactor. Its divertor will expose to tritium environment and neutron radiation. Materials of the divertor will be radioactived, and cannot be handled by personnel directly. To develop structure which compatible with robots handle for installation, maintenance and removing is required. This paper introduces a conceptual design of CFETR divertor module which compatible with remote handling end-effectors. The divertor module is confined by inner and outer support. The inner support is only confined divertor module radial, toroidal and vertical moving freedom degrees, but not confined rotating freedom degrees. The outer support is the structure that can confine rotating freedom degrees and should also be compatible with remote handling end-effectors.

  18. Conceptual design of CFETR divertor remote handling compatible structure

    International Nuclear Information System (INIS)

    Dai, Huaichu; Yao, Damao; Cao, Lei; Zhou, Zibo; Li, Lei

    2016-01-01

    Highlights: • Conceptual design for the CFETR divertor have been proposed, especially the divertor remote handling compatible structure. • The degrees of freedom of the divertor are analyzed in order to validate the design the divertor supports structure. • Besides the ITER-like scheme, a new scheme for the divertor remote handling compatible supports is proposed, that is the rack and pinion mechanism. • The installation/removel process is verified through simulation in Delmia in order to check design quality for remote handling requirements. - Abstract: Divertor is one of key components of tokamak fusion reactor. The CFETR is China Fusion Engineering Test Reactor. Its divertor will expose to tritium environment and neutron radiation. Materials of the divertor will be radioactived, and cannot be handled by personnel directly. To develop structure which compatible with robots handle for installation, maintenance and removing is required. This paper introduces a conceptual design of CFETR divertor module which compatible with remote handling end-effectors. The divertor module is confined by inner and outer support. The inner support is only confined divertor module radial, toroidal and vertical moving freedom degrees, but not confined rotating freedom degrees. The outer support is the structure that can confine rotating freedom degrees and should also be compatible with remote handling end-effectors.

  19. LHD helical divertor

    International Nuclear Information System (INIS)

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

    1993-07-01

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

  20. Comparative thermal cyclic testing and strength investigation of different Be/Cu joints

    International Nuclear Information System (INIS)

    Gervash, A.; Giniyatulin, R.; Komarov, V.; Mazul, I.; Litunovsky, N.; Ganenko, A.; Vainerman, A.; Fedotov, V.; Davydov, D.; Zalavutdinov, R.

    1998-01-01

    One of the main problems for ITER divertor target technology is to provide a reliable joint between Be as armour material and copper alloy heat-sink structure. Such joints are to satisfy numerous requirements. In particular these joints should successfully withstand cyclic hear fluxes and should have good properties after neutron irradiation. To study such a complex problem, several investigation stages were planed in Russia. This paper presents the results of comparative thermal cyclic testing of different Be/Cu candidates. Summarising the thermal cyclic test results and analysing the metallography of those joints it was found that the life-time of all tested joints is limited by rather thick brittle intermetallic layers in the bonding zone caused by relatively long brazing time using heating and cooling down in traditional ohmic furnace. This paper thus presents attempts of using a unique brazing technique with fast e-beam heating. Metallographic investigation as well as X-ray spectrometric analysis of joints produced using the new technique were done. The recent results of testing of Be/Cu joints produced by fast e-beam brazing are discussed and some ideas for the nearest future investigations are presented. (orig.)

  1. Plasma detachment in divertor tokamaks

    Science.gov (United States)

    Leonard, A. W.

    2018-04-01

    Observations of divertor plasma detachment in tokamaks are reviewed. Plasma detachment is characterized in terms of transport and dissipation of power, momentum and particle flux along the open field lines from the midplane to the divertor. Asymmetries in detachment onset and other characteristics between the inboard and outboard divertor plasmas is found to be primarily driven by plasma E× B drifts. The effect of divertor plate geometry and magnetic configuration on divertor detachment is summarized. Control of divertor detachment has progressed with a development of a number of diagnostics to characterize the detached state in real-time. Finally the compatibility of detached divertor operation with high performance core plasmas is examined.

  2. Divertor coil device

    International Nuclear Information System (INIS)

    Hanai, Satoru.

    1990-01-01

    The present invention concerns a divertor coil device used in a tokamak type thermonuclear device and the object thereof is to reduce thermal loads in the heat receiving portion. An auxiliary power source is disposed, in addition to a main power source, for supplying main electric current for changing electric current ratio between each of the divertor coils. Then, the null point for forming plasmas is made controllable. As a result, a power source for a part of coils connected to the auxiliary power source of the divertor coils can be changed by controlling the voltage of the auxiliary power source. Accordingly, the electric current distribution in the divertor coils is changed and the position for the null point high thermal load region can be moved laterally. The area of the heat receiving portion can be increased by moving the high thermal load region, thereby decreasing the thermal load density. (I.S.)

  3. 'EU divertor celebration day'

    International Nuclear Information System (INIS)

    Merola, M.

    2002-01-01

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

  4. Plans of LHD divertor experiment

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  5. Innovative divertor concepts for LHD

    International Nuclear Information System (INIS)

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

    1994-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-01-01

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

  7. The MAST improved divertor

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  8. 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)

  9. Boundary plasma control with the ergodic divertor

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  10. Boundary plasma control with the ergodic divertor

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  11. Numerical studies on divertor experiments

    International Nuclear Information System (INIS)

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

    1988-04-01

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

  12. Innovations in the LHD divertor program

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  13. High temperature divertor plasma operation

    International Nuclear Information System (INIS)

    Ohyabu, Nobuyoshi.

    1991-02-01

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

  14. The divertor remote maintenance project

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  15. The divertor remote maintenance project

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  16. Advanced divertor experiments on DIII-D

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  17. Assessment of the W7-X high heat flux divertor with thermo-mechanical analysis

    International Nuclear Information System (INIS)

    Qian, Xinyuan; Peng, Xuebing; Fellinger, Joris; Boscary, Jean; Bykov, Victor; Wang, Zhongwei; Ye, Minyou; Song, Yuntao

    2016-01-01

    Highlights: • Thermo-mechanical analysis of HHF divertor module, TM2H. • Temperature of all parts is acceptable for long pulse operation. • Stress in different parts is mainly caused by different load. • Radial displacement need to be improved based on FE calculations. - Abstract: The Wendelstein 7-X is an experimental device designed with a stellarator magnetic confinement for stationary plasma operation (up to 30 min). At the first stage, it is scheduled to start with an inertially cooled test divertor unit and a shorter plasma pulse operation up to 10 s. After the completion of this stage, a water-cooled high heat flux (HHF) divertor will be installed for the steady-state operation phase. The divertor consists of individual target modules, which are sets of target elements armored with CFC tiles supported by a stainless steel structure and fed in parallel with manifolds. Detailed thermo-mechanical analysis of the target modules using the finite element method has been performed to validate and/or improve the elected design of the HHF divertor under operation. Different operating conditions have been studied and the effect of the variation of the convective heat flux pattern with localized heating loads as high as 10 MW/m 2 onto the target elements has been computed. The analysis of the thermal response, stress distribution and deformation allowed a better understanding of the behavior of the divertor modules under operation and confirmed the suitability of the design.

  18. Assessment of the W7-X high heat flux divertor with thermo-mechanical analysis

    Energy Technology Data Exchange (ETDEWEB)

    Qian, Xinyuan [School of Nuclear Science and Technology, University of Science and Technology of China, Jinzhai Road 96, 230026 Hefei,Anhui (China); Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei, Anhui (China); Peng, Xuebing, E-mail: pengxb@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei, Anhui (China); Fellinger, Joris [Max Planck Institute for Plasma Physics, Wendelsteinstr. 1, 17491 Greifswald (Germany); Boscary, Jean [Max Planck Institute for Plasma Physics, Boltzmannstr. 2, 85748 Garching (Germany); Bykov, Victor [Max Planck Institute for Plasma Physics, Wendelsteinstr. 1, 17491 Greifswald (Germany); Wang, Zhongwei [Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei, Anhui (China); Ye, Minyou; Song, Yuntao [School of Nuclear Science and Technology, University of Science and Technology of China, Jinzhai Road 96, 230026 Hefei,Anhui (China); Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei, Anhui (China)

    2016-11-01

    Highlights: • Thermo-mechanical analysis of HHF divertor module, TM2H. • Temperature of all parts is acceptable for long pulse operation. • Stress in different parts is mainly caused by different load. • Radial displacement need to be improved based on FE calculations. - Abstract: The Wendelstein 7-X is an experimental device designed with a stellarator magnetic confinement for stationary plasma operation (up to 30 min). At the first stage, it is scheduled to start with an inertially cooled test divertor unit and a shorter plasma pulse operation up to 10 s. After the completion of this stage, a water-cooled high heat flux (HHF) divertor will be installed for the steady-state operation phase. The divertor consists of individual target modules, which are sets of target elements armored with CFC tiles supported by a stainless steel structure and fed in parallel with manifolds. Detailed thermo-mechanical analysis of the target modules using the finite element method has been performed to validate and/or improve the elected design of the HHF divertor under operation. Different operating conditions have been studied and the effect of the variation of the convective heat flux pattern with localized heating loads as high as 10 MW/m{sup 2} onto the target elements has been computed. The analysis of the thermal response, stress distribution and deformation allowed a better understanding of the behavior of the divertor modules under operation and confirmed the suitability of the design.

  19. Divertor retention for recycling impurities

    International Nuclear Information System (INIS)

    Krieger, K.; Roth, J.; Fussmann, G.

    1992-01-01

    As an important issue for fusion devices with divertor configurations the retention capability for both recycling and non-recycling impurities receives increasing interest. In the case of recycling, gaseous, impurities the retention capability is usually investigated by means of short impurity gas puffs into the plasma vessel and the analysis of the time dependence of the observed line radiation. The detailed understanding of the impurity transport processes related to the retention capability of a certain divertor structure will require modelling of the experimental results with 2D or 3D transport code simulations. However, for the comparison of the global behavior of different configurations a much simpler description of the divertor retention in terms of global time constants may be sufficient. We will give a summary of experimental results from ASDEX for the dependence of the retention capability on parameters like divertor plasma density and temperature and the distance along field lines between main plasma and divertor. In addition we will compare some of these results with similar experiments on DIIID. (author) 8 refs., 2 figs., 2 tabs

  20. The ITER divertor cassette project meeting

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  1. 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.

  2. Diagnostic measurements of the pumped divertor plasma

    International Nuclear Information System (INIS)

    Gondhalekar, A.; Bartlett, D.; Costley, A.

    1989-01-01

    The scope of plasma diagnostic capability needed for the pumped divertor is determined by the measurement objectives, which are: (i) to demonstrate feasibility of impurity control using a high flow rate divertor, (ii) to validate a model of the divertor action, and (iii) to optimize pumped divertor performance. Installation of diagnostics, with spatial resolution along the separatrix in the divertor region between the x-point and the target plates, is proposed. Difficult access, small plasma size, large dynamic range, and interpretational issues determine the choices of diagnostic methods which have been made. (author)

  3. Engineering structure and thermal-technical analysis of fusion experimental breeder FEB divertor

    International Nuclear Information System (INIS)

    Feng Kaiming; Huang Jinhua; Zhu Yukun; Deng Peizhi; Zhou Xiaobing; Wang Min; Huo Tiejun

    1999-10-01

    On the basis of the physical study of FEB divertor, the engineering structure and thermal-technical analysis of FEB divertor are presented. In order to improve the impurity control and to increase ion-neutral interactions in the divertor, the configuration of the divertor is optimized to be the close type in the engineering design activity compared with the open type in the early conceptual activity. The operation mode of the divertor is designed to be partial detached plasma mode under conditions of combination gas-puffing with impurity injection. The position of gas-puffing is optimized at the torus mid-plane with NEWT1D code from the view point of impurity retention and radiation in the scrape-off layer/divertor region. The divertor structure is consisted of 48 rounded cassette modules. The thermal-technical calculations are carried out with COSMOS/M-HSTAR code for target plates. The result showed that the He-cooled target with 4 MPa coolant pressure and radial flowing is feasible

  4. Evaluation of helium cooling for fusion divertors

    International Nuclear Information System (INIS)

    Baxi, C.B.

    1993-09-01

    The divertors of future fusion reactors will have a power throughput of several hundred MW. The peak heat flux on the diverter surface is estimated to be 5 to 15 MW/m 2 at an average heat flux of 2 MW/m 2 . The divertors have a requirement of both minimum temperature (100 degrees C) and maximum temperature. The minimum temperature is dictated by the requirement to reduce the absorption of plasma, and the maximum temperature is determined by the thermo-mechanical properties of the plasma facing materials. Coolants that have been considered for fusion reactors are water, liquid metals and helium. Helium cooling has been shown to be very attractive from safety and other considerations. Helium is chemically and neutronically inert and is suitable for power conversion. The challenges associated with helium cooling are: (1) Manifold sizes; (2) Pumping power; and (3) Leak prevention. In this paper the first two of the above design issues are addressed. A variety of heat transfer enhancement techniques are considered to demonstrate that the manifold sizes and the pumping power can be reduced to acceptable levels. A helium-cooled diverter module was designed and fabricated by GA for steady-state heat flux of 10 MW/m 2 . This module was recently tested at Sandia National Laboratories. At an inlet pressure of 4 MPa, the module was tested at a steady-state heat flux of 10 MW/m 2 . The pumping power required was less than 1% of the power removed. These results verified the design prediction

  5. VUV Spectroscopy in DIII-D Divertor

    International Nuclear Information System (INIS)

    Alkesh Punjabi; Nelson Jalufka

    2004-01-01

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

  6. Dipole Map For Divertor Tokamaks

    International Nuclear Information System (INIS)

    Ali, Halima; Punjabi, Alkesh; Boozer, Allen

    2003-01-01

    Heat flux impinging on the collector plates of divertor tokamaks can be prodigious. Therefore, the problem of spreading the heat flux on plates is a crucial issue for divertor tokamaks such as ITER. Here we use method of maps /1,2/ to investigate this problem. Magnetic field lines in non-axisymmetric divertor tokamaks are a one and a half degree of freedom Hamiltonian system /1-3/. We represent the unperturbed magnetic topology by the Symmetric Simple Map (SSM) /4/ given by yn+1 = yn + 2kxn - 2k2yn (1 - yn), xn+1 = xn - kyn (1 - yn) - 2k2yn+1 (1 - yn+1). The effects of a current carrying coil placed externally across from X-point is represented by Dipole Map (DP) /4,5/ given by x n+1 = x n + 2δs 3 x n+1 (y n - y s + s/[x n+1 2 + (y n - y s + s) 2 ] 2 ), y n+1 = y n + δs 3 x n+1 ((y n - y s + s) 2 - x n+1 2 /[x n+1 2 + (y n - y s + s) 2 ] 2 ) δ is amplitude of high MN magnetic perturbation, s is the distance of coil from last good surface across from X point, and is the y coordinate of last good surface where it crosses the axis joining X point and O point across from X point. We fix k=0.3 and s = (1/2)|y s |. We calculate the increase in width of stochastic layer and area of footprint of field lines on divertor plate as δ is increased. We also calculate how connection length, toroidal and poloidal circuits and their fractal structures, the number, location and density of hot spots change with δ. Finally, we make conclusions about how the heat flux can be possibly controlled and reduced by applying external magnetic perturbation in divertor tokamaks

  7. Progress in ergodic divertor operation on Tore Supra

    International Nuclear Information System (INIS)

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

    1999-09-01

    Upgrade of the Tore ergodic divertor has led to significant progress in ergodic divertor physics. The disruptive limit governed by the stochastization of the outer magnetic surfaces is found to occur for a value of the Chirikov parameter reaching 2 on the magnetic surface q = 2 + 3 / 12. This experimentally observed robustness allows one to operate at very low safety factor on the separatrix (q ∼ 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)

  8. Plasma facing components integration studies for the WEST divertor

    Energy Technology Data Exchange (ETDEWEB)

    Ferlay, Fabien, E-mail: fabien.ferlay@cea.fr; Missirlian, Marc; Guilhem, Dominique; Firdaouss, Mehdi; Richou, Marianne; Doceul, Louis; Faisse, Frédéric; Languille, Pascal; Larroque, Sébastien; Martinez, André; Proust, Maxime; Louison, Céphise; Jeanne, Florian; Saille, Alain; Samaille, Frank; Verger, Jean-Marc; Bucalossi, Jérôme

    2015-10-15

    Highlights: • The divertor PFU integration has been studied regarding existing environment. • Magnetic, electric, thermal, hydraulic, mechanical loads and assembly are considered. - Abstract: In the context of the Tokamak Tore-Supra evolution, the CEA aims at transforming it into a test bench for ITER actively cooled tungsten (ACW) plasma facing components (PFC). This project named WEST (Tungsten Environment in Steady state Tokamak) is especially focused on the divertor target. The modification of the machine, by adding two axisymmetric divertors will make feasible an H-mode with an X-point close to the lower divertor. This environment will allow exposing the divertor ACW components up to 20 MW/m{sup 2} heat flux during long pulse. These specifications are well suited to test the ITER-like ACW target elements, respecting the ITER design. One challenge in such machine evolution is to integrate components in an existing vacuum vessel in order to obtain the best achievable performance. This paper deals with the design integration of ITER ACW target elements into the WEST environment considering magnetic, electric, thermal and mechanical loads. The feasibility of installation and maintenance has to be strongly considered as these PFC could be replaced several times. The ports size allows entering a 30° sector of pre-installed tungsten targets which will be plugged as quickly and easily as possible. The main feature of steady state operation is the active cooling, which leads to have many embedded cooling channels and bulky pipes on the PFC module including many connections and sealings between vacuum and water channels. The 30° sector design is now finalized regarding the ITER ACW elements specifications. No major modifications are expected.

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  11. 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.

  12. Development of divertor remote maintenance system

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-04-01

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

  13. Development of divertor remote maintenance system

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  14. A Lithium Vapor Box Divertor Similarity Experiment

    Science.gov (United States)

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

    2017-10-01

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

  15. Controlling marginally detached divertor plasmas

    Science.gov (United States)

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

    2017-06-01

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

  16. The ITER divertor cassette project

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  17. The ITER divertor cassette project

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    Science.gov (United States)

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

    2012-10-01

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

  19. Optimization of a bundle divertor for FED

    International Nuclear Information System (INIS)

    Hively, L.M.; Rothe, K.E.; Minkoff, M.

    1982-01-01

    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

  20. Comparison between stellarator and tokamak divertor transport

    International Nuclear Information System (INIS)

    Feng, Y.; Lunt, T.; Kobayashi, M.; Reiter, D.

    2010-11-01

    The paper compares the essential divertor transport features of the poloidal divertor, which is well-developed for tokamaks, and the non-axisymmetric divertors currently investigated on helical devices. It aims at surveying the fundamental similarities and differences in divertor concept and geometry, and their consequences for how the divertor functions. In particular, the importance of various transport terms governing axisymmetric and helical scrape-off-layers (SOLs) is examined, with special attention being paid to energy, momentum and impurity transport. Tokamak and stellarator SOLs are compared by identifying key geometric parameters through which the governing physics can be illustrated by simple models and estimates. More quantitative assessments rely nevertheless on the modeling using EMC3-EIRENE code. Most of the theoretical results are discussed in conjunction with experimental observations. (author)

  1. Reactor application of an improved bundle divertor

    International Nuclear Information System (INIS)

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

    1978-11-01

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

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

    International Nuclear Information System (INIS)

    1994-03-01

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

  3. First divertor operation on the HL-2A tokamak

    International Nuclear Information System (INIS)

    Yang Qingwei; Ding Xuantong; Yan Longwen; Xuan Weimin; Liu Dequan; Chen Liaoyuan; Song Xianming; Yuan Baoshan; Zhang Jinhua; Cao Zeng; Li Xiaodong; Mao Weicheng; Zhou Caipin; Wang Enyao; Yan Jiancheng; Liu Yong

    2004-01-01

    HL-2A device is the first divertor tokamak in China. One of its main subjects is to study the features of the divertor plasma. In the last campaign, the first divertor configuration has been achieved and sustained on the HL-2A tokamak. Here authors give a brief description about the HL-2A tokamak, diagnostics arrangements, and the equilibrium analysis results on divertor configuration. The main results of divertor experiments are also presented. (author)

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

    International Nuclear Information System (INIS)

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

    1995-10-01

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

  5. Optical design study for divertor observation at the stellarator W7-X

    NARCIS (Netherlands)

    König, R.; Hildebrandt, D.; Hübner, T.; Klinkhamer, J.F.F.; Moddemeijer, K.; Vliegenthart, W.A.

    2006-01-01

    The stellarator W7-X will be capable of running in a quasicontinuous operating mode with 10 MW of electron cyclotron heating (ECRH) heating for 30 min, the duration only being limited by the capacity of the available cooling reservoir. The integrated ten discrete water cooled divertor modules need

  6. 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

  7. Recent developments towards ITER 2001 divertor maintenance

    International Nuclear Information System (INIS)

    Palmer, J.; Siuko, M.; Agostini, P.; Gottfried, R.; Irving, M.; Martin, E.; Tesini, A.; Uffelen, M. Van

    2005-01-01

    One of the key maintenance operations for the ITER tokamak is the remote replacement of its divertor system. The components making up this system are expected to be activated to a level of several hundred gyrations per hour and contaminated with hazardous and/or activated dust (beryllium, carbon, tungsten) and tritium. A suite of specialized remote handling (RH) equipment is, therefore, necessary to facilitate divertor exchange. This paper describes the ITER divertor maintenance approach together with recent European efforts towards the design and development of the associated remote handling equipment and procedures

  8. Manufacture and installation of JET MKII divertor support structure

    International Nuclear Information System (INIS)

    Celentano, G.; Altmann, H.; Macklin, B.; Miele, P.; Pick, M.A.; Tait, J.; Moletta, L.; Romagnolo, A.; Shaw, R.

    1995-01-01

    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

  9. Thermomechanical behavior of actively cooled, brazed divertor components under cyclic high heat flux loads

    International Nuclear Information System (INIS)

    You, J.H.; Bolt, H.; Duwe, R.; Linke, J.; Nickel, H.

    1997-01-01

    Actively cooled divertor mock-ups consisting of various low-Z armor tiles brazed to refractory metal heat sinks were tested in the electron beam test facility at Juelich. Screening and thermal cycling tests were perfomed on the mock-ups to estimate the overall thermal performance under cyclic high heat flux (HHF) loadings. By detecting the temperature of the armor surface and the braze layer, it was possible to assess the heat removal capability and the accumulation of interfacial damage. Microstructures were investigated to elucidate the degradation of the joints. Finite element analyses are carried out for the simulated HHF test conditions. Temperature fields and thermal stresses are calculated for a typical divertor module. The nature of thermomechanical behavior of the divertor mock-ups under cyclic HHF loadings is discussed. (orig.)

  10. Thermomechanical behavior of actively cooled, brazed divertor components under cyclic high heat flux loads

    Science.gov (United States)

    You, J. H.; Bolt, H.; Duwe, R.; Linke, J.; Nickel, H.

    1997-12-01

    Actively cooled divertor mock-ups consisting of various low- Z armor tiles brazed to refractory metal heat sinks were tested in the electron beam test facility at Jülich. Screening and thermal cycling tests were perfomed on the mock-ups to estimate the overall thermal performance under cyclic high heat flux (HHF) loadings. By detecting the temperature of the armor surface and the braze layer, it was possible to assess the heat removal capability and the accumulation of interfacial damage. Microstructures were investigated to elucidate the degradation of the joints. Finite element analyses are carried out for the simulated HHF test conditions. Temperature fields and thermal stresses are calculated for a typical divertor module. The nature of thermomechanical behavior of the divertor mock-ups under cyclic HHF loadings is discussed.

  11. Stability, divertors and innovative concepts

    International Nuclear Information System (INIS)

    Mirnov, S.

    2003-01-01

    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)

  12. ARIES-III divertor engineering design

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Schultz, K.R.; Cheng, E.T.; Grotz, S.; Hasan, M.A.; Najmabadi, F.; Sharafat, S.; Herring, J.S.; Valenti, M.; Steiner, D.

    1992-01-01

    This paper reports the engineering design of the ARIES-III double- null divertor. The divertor coolant tubes are made from W-3Re alloy and cooled by subcooled flow boiling of organic coolant. A coating of 4 mm thick tungsten is plasma sprayed onto the divertor surface. This W layer can withstand the thermal deposition of a few disruptions. At a maximum surface heat flux of 5.4 MW/m 2 , a conventional divertor design can be used. The divertor surface is contoured to have a constant heat flux of 5.4 MW/m 2 . The net erosion of the W-surface was found to be negligible at about 0.1 mm/year. After 3 years of operation, the W-3Re alloy ARIES-III divertor can be disposed of as Class A waste. In order to control the prompt dose release at site boundary to less than 200 Rem, isotopic tailoring of the W-alloy will be needed

  13. Thermal stress analysis of FIRE divertor

    International Nuclear Information System (INIS)

    Baxi, C.B.; Reis, E.E.; Ulrickson, M.A.; Heizenroeder, P.; Driemeyer, D.

    2003-01-01

    The fusion engineering research experiment (FIRE) device is designed for high power density and advanced physics operating modes. Due to the short distance of the divertor from the X-point, the connection lengths are short and the scrape off layer thickness is small. A relatively high peak heat flux of 25 MW/m 2 is expected on the divertor. The FIRE divertor engineering design is based on the design approaches developed for international thermonuclear experimental reactor (ITER). The geometry of the FIRE divertor consists of water cooled copper fingers and a tungsten brush armor as plasma facing material. The divertor assembly consists of modular units for remote handling. A 316 stainless steel back plate is used for support and manifolding. The backing plate is joined to the copper fingers by pins. The coolant channel diameter is 8 mm at a pitch of 14 mm. The total power flow to the outer divertor is 35 MW. Water at an inlet temperature of 30 deg.C, 1.5 MPa and a flow velocity of 10 m/s is used with two channels in series. A margin of ∼1.6 is obtained on the critical heat flux. A three dimensional thermal stress finite element (FE) analysis of this geometry was performed. Thermal hydraulic correlations derived for ITER were used to perform the thermal analysis. Design changes were implemented to reduce the stresses and temperatures to acceptable levels

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

    International Nuclear Information System (INIS)

    O'NEIL, RC; STAMBAUGH, RD

    2002-01-01

    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

  15. Snowflake divertor configuration studies for NSTX-Upgrade

    International Nuclear Information System (INIS)

    Soukhanovskii, V.A.

    2011-01-01

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

  16. Integrated core-edge-divertor modeling studies

    International Nuclear Information System (INIS)

    Stacey, W.M.

    2001-01-01

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

  17. Geometrical properties of a 'snowflake' divertor

    International Nuclear Information System (INIS)

    Ryutov, D. D.

    2007-01-01

    Using a simple set of poloidal field coils, one can reach the situation in which the null of the poloidal magnetic field in the divertor region is of second order, not of first order as in the usual X-point divertor. Then, the separatrix in the vicinity of the null point splits the poloidal plane not into four sectors, but into six sectors, making the whole structure look like a snowflake (hence the name). This arrangement allows one to spread the heat load over a much broader area than in the case of a standard divertor. A disadvantage of this configuration is that it is topologically unstable, and, with the current in the plasma varying with time, it would switch either to the standard X-point mode, or to the mode with two X-points close to each other. To avoid this problem, it is suggested to have a current in the divertor coils that is roughly 5% higher than in an ''optimum'' regime (the one in which a snowflake separatrix is formed). In this mode, the configuration becomes stable and can be controlled by varying the current in the divertor coils in concert with the plasma current; on the other hand, a strong flaring of the scrape-off layer still remains in force. Geometrical properties of this configuration are analyzed. Potential advantages and disadvantages of this scheme are discussed

  18. Nondestructive Evaluation of a Be/Cu Diffusion Bond using a Shear Horizontal Wave

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hyun Kyu; Cheong, Yong Moo; Lee, Dong Won; Hong, Bong Keun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    The International Thermo-nuclear Experimental Reactor (ITER) blanket first wall includes Beryllium(Be) amour tiles joined to a CuCrZr heat sink with stainless steel cooling tubes. This first wall's panels are one of the critical components in the ITER which is exposed with a surface heat flux of 0.5 MW/m2. As a qualification program, ultrasonic test (UT) of a Be/CuCrZr diffusion bond has to be applied according to the proper procedure. Ultrasonic test can detect the presence of unbonded regions and is based on an amplitude change and a phase inversion in a signal reflected from a bond interface. The purpose of this study is to investigate the feasibility of EMAT (Electro-Magnetic Acoustic Transducer) technology for an in-situ inspection of a Be/Copper alloy joining interface under a high temperature and high radiation environment.

  19. Nondestructive Evaluation of a Be/Cu Diffusion Bond using a Shear Horizontal Wave

    International Nuclear Information System (INIS)

    Jung, Hyun Kyu; Cheong, Yong Moo; Lee, Dong Won; Hong, Bong Keun

    2009-01-01

    The International Thermo-nuclear Experimental Reactor (ITER) blanket first wall includes Beryllium(Be) amour tiles joined to a CuCrZr heat sink with stainless steel cooling tubes. This first wall's panels are one of the critical components in the ITER which is exposed with a surface heat flux of 0.5 MW/m2. As a qualification program, ultrasonic test (UT) of a Be/CuCrZr diffusion bond has to be applied according to the proper procedure. Ultrasonic test can detect the presence of unbonded regions and is based on an amplitude change and a phase inversion in a signal reflected from a bond interface. The purpose of this study is to investigate the feasibility of EMAT (Electro-Magnetic Acoustic Transducer) technology for an in-situ inspection of a Be/Copper alloy joining interface under a high temperature and high radiation environment

  20. Thermal effects of runaway electrons in an armoured divertor

    International Nuclear Information System (INIS)

    Stad, R.C.L. van der.

    1993-12-01

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

  1. Experimental studies of the snowflake divertor in TCV

    NARCIS (Netherlands)

    Labit, B.; Canal, G. P.; Christen, N.; Duval, B. P.; Lipschultz, B.; Lunt, T.; Nespoli, F.; Reimerdes, H.; Sheikh, U.; Theiler, C.; Tsui, C. K.; Verhaegh, K.; Vijvers, W. A. J.

    2017-01-01

    To address the risk that, in a fusion reactor, the conventional single-null divertor (SND) configuration may not be able to handle the power exhaust, alternative divertor configurations, such as the Snowflake divertor (SFD), are investigated in TCV. The expected benefits of the SFD-minus in terms of

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

    Science.gov (United States)

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

    2018-05-01

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

  3. Divertor and first wall design for TIBER-II

    International Nuclear Information System (INIS)

    Gallix, R.; Bourque, R.; Baxi, C.; Creedon, L.; Schultz, K.; Vance, D.

    1987-01-01

    The conceptual design of the double-null divertor plates and the first wall armor for the TIBER-II fusion engineering test reactor is presented. The divertor plates, which receive a steady-state heat flux of up to 4.3 MW/m 2 , are actively cooled. They consist of small carbon or tungsten tiles brazed on water-cooled copper tubes which are fed by a dedicated cooling system. The first wall armor protects the water-cooled shield and blanket modules which form the walls of the plasma chamber. The armor receives an average, steady-state heat flux of 0.23 MW/m 2 and up to 2.4 MJ/m 2 during plasma disruptions. It consists of radiation-cooled, carbon-carbon composite tiles. The tiles cover the entire inboard wall and form 16 discrete poloidal limiters attached to the bare stainless steel outboard wall. Due to potentially severe plasma erosion, the components are designed for remote replacement

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  5. Towards the procurement of the ITER divertor

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  6. Safety characteristics of the monolithic CFC divertor

    International Nuclear Information System (INIS)

    Zucchetti, M.; Merola, M.; Matera, R.

    1994-01-01

    The main distinguishing feature of the monolithic CFC divertor is the use of a single material, a carbon fibre reinforced carbon, for the protective armour, the heat sink and the cooling channels. This removes joint interface problems which are one of the most important concerns related to the reference solutions of the ITER CDA divertor. An activation analysis of the different coolant options for this concept is presented. It turns out that neither short-term nor long-term activation are a concern for any coolants investigated. Therefore the proposed concept proves to be attractive from a safety stand-point also. ((orig.))

  7. Safety characteristics of the monolithic CFC divertor

    Science.gov (United States)

    Zucchetti, M.; Merola, M.; Matera, R.

    1994-09-01

    The main distinguishing feature of the monolithic CFC divertor is the use of a single material, a carbon fibre reinforced carbon, for the protective armour, the heat sink and the cooling channels. This removes joint interface problems which are one of the most important concerns related to the reference solutions of the ITER CDA divertor. An activation analysis of the different coolant options for this concept is presented. It turns out that neither short-term nor long-term activation are a concern for any coolants investigated. Therefore the proposed concept proves to be attractive from a safety stand-point also.

  8. Local island divertor experiments on LHD

    International Nuclear Information System (INIS)

    Morisaki, T.; Masuzaki, S.; Komori, A.; Ohyabu, N.; Kobayashi, M.; Feng, Y.; Sardei, F.; Narihara, K.; Tanaka, K.; Ida, K.; Peterson, B.J.; Yoshinuma, M.; Ashikawa, N.; Emoto, M.; Funaba, H.; Goto, M.; Ikeda, K.; Inagaki, S.; Kaneko, O.; Kawahata, K.; Kubo, S.; Miyazawa, J.; Morita, S.; Nagaoka, K.; Nagayama, Y.; Nakanishi, H.; Ohkubo, K.; Oka, Y.; Osakabe, M.; Shimozuma, T.; Shoji, M.; Takeiri, Y.; Sakakibara, S.; Sakamoto, R.; Sato, K.; Toi, K.; Tsumori, K.; Watababe, K.Y.; Yamada, H.; Yamada, I.; Yoshimura, Y.; Motojima, O.

    2005-01-01

    A local island divertor (LID) experiment has begun on LHD, with the aims of controlling edge recycling and improving the plasma confinement. The fundamental divertor functions of the LID have been demonstrated in the recent experiments. From the particle flux profile measurements on the LID head it was found that the particles diffusing out from the core region are well guided along the island separatrix to the LID head. Owing to the closed configuration around the LID head, evidence of the high efficient pumping was observed, together with a strong capacity to screen impurities. The first results of edge modeling using the EMC3-EIRENE code are also presented

  9. Local island divertor experiments on LHD

    Energy Technology Data Exchange (ETDEWEB)

    Morisaki, T. [National Institute for Fusion Science, 322-6 Orosi, Toki, Gifu 509-5292 (Japan)]. E-mail: morisaki@nifs.ac.jp; Masuzaki, S.; Komori, A; Ohyabu, N.; Kobayashi, M.; Narihara, K.; Tanaka, K.; Ida, K.; Peterson, B.J.; Yoshinuma, M.; Ashikawa, N.; Emoto, M.; Funaba, H.; Goto, M.; Ikeda, K.; Inagaki, S.; Kaneko, O.; Kawahata, K.; Kubo, S.; Miyazawa, J.; Morita, S.; Nagaoka, K.; Nagayama, Y.; Nakanishi, H.; Ohkubo, K.; Oka, Y.; Osakabe, M.; Shimozuma, T.; Shoji, M.; Takeiri, Y.; Sakakibara, S.; Sakamoto, R.; Sato, K.; Toi, K.; Tsumori, K.; Watababe, K.Y.; Yamada, H.; Yamada, I.; Yoshimura, Y.; Motojima, O. [National Institute for Fusion Science, 322-6 Orosi, Toki, Gifu 509-5292 (Japan); Feng, Y.; Sardei, F. [Max-Planck-Institute fuer Plasmaphysik, Euratom Association Teilinstitut Greifswald, Wendelsteinstrasse 1, D-17491 Greifswald (Germany)

    2005-03-01

    A local island divertor (LID) experiment has begun on LHD, with the aims of controlling edge recycling and improving the plasma confinement. The fundamental divertor functions of the LID have been demonstrated in the recent experiments. From the particle flux profile measurements on the LID head it was found that the particles diffusing out from the core region are well guided along the island separatrix to the LID head. Owing to the closed configuration around the LID head, evidence of the high efficient pumping was observed, together with a strong capacity to screen impurities. The first results of edge modeling using the EMC3-EIRENE code are also 0015present.

  10. Conceptual design of divertor cassette handling by remote handling system for JT-60SA

    International Nuclear Information System (INIS)

    Hayashi, Takao; Sakurai, Shinji; Masaki, Kei; Tamai, Hiroshi; Yoshida, Kiyoshi; Matsukawa, Makoto

    2007-01-01

    The JT-60SA aims to contribute and supplement ITER toward DEMO reactor based on tokamak concept. One of the features of JT-60SA is its high power long pulse heating, causing the large annual neutron fluence. Because the expected dose rate at the vacuum vessel (VV) may exceed 1 mSv/hr after 10 years operation and three month cooling, the human access inside the VV is prohibited. Therefore a remote handling (RH) system is necessary for the maintenance and repair of in-vessel components. This paper described the RH system of JT-60SA, especially the expansion of the RH rail and exchange of the divertor modules. The RH rail is divided into nine and three-point mounting. The nine sections can cover 225 degrees in toroidal direction. A divertor module, which is 10 degrees wide in toroidal direction and weighs 500kg itself due to the limitations of port width and handling weight, can be exchanged by heavy weight manipulator (HWM). The HWM brings the divertor module to the front of the other RH port, which is used for supporting the rail and/or carrying in and out equipments. Then another RH device receives and brings out the module by a pallet installed from outside the VV. (author)

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

    Science.gov (United States)

    Norajitra, P.; Antusch, S.; Giniyatulin, R.; Mazul, I.; Ritz, G.; Ritzhaupt-Kleissl, H.-J.; Spatafora, L.

    2011-10-01

    A divertor concept for DEMO has been investigated at Karlsruhe Institute of Technology (KIT) which has to withstand a heat flux of 10 MW/m 2. 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 °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&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.

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-12-15

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

  14. Conical surface textures formed by ion bombarding 2% Be Cu alloy

    International Nuclear Information System (INIS)

    Panitz, J.K.G.

    1990-01-01

    A homogeneous, micrometer-sized conical surface texture forms on 2% Be-Cu alloy which is bombarded with an argon beam produced by a Kaufman ion source. The dimensions of the features that form strongly depend on: (1) argon energy (from 250 to 1500 eV), (2) fluence (10 19 to 10 20 ions/cm 2 ), and (3) flux (0.1 to 1 mA/cm 2 ). The texture morphology depends less strongly on the background ambient (Mo vs graphite), earlier alloy heat treatments and the temperature during bombardment (100 degree C and 450 degree C). As the texture matures with increasing fluence, the number of large features increases at the expense of the number of small features. The observed relationship between texture formation and ion flux suggests that the evolution of these features is not adequately described by theories predicting that the mature conical sidewall angle is related to the angle of the maximum sputtering yield. These textured surfaces can be coated with other metals for a variety of possible applications including: (1) pulsed power Li+ beam anodes, (2) cold cathode field emission devices, (3) optical absorbers and (4) catalysis supports. 18 refs., 5 figs

  15. Mechanical design and manufacture of magnetic ergodic divertor for the TORE SUPRA tokamak

    International Nuclear Information System (INIS)

    Lipa, M.; Aymar, R.; Deschamps, P.; Hertout, P.; Portafaix, C.; Samain, A.

    1989-01-01

    A configuration of six equally spaced ergodic divertors has been chosen to control the plasma impurities in the TORE SUPRA tokamak since the control of these impurities is essential to the long pulse duration envisioned for the machine. Each of the six indentical modules is composed of (8) conductor bars arranged in a poloidal direction forming a resonant helical winding. The proximity of the conductors to the plasma requires that each copper assembly be water cooled, enclosed in a stainless steel casing and protected by pure graphite tiles attaches to the inner surface of the casing. Particles which drift between the coil bars are neutralized on actively water cooled neutralizer plates and then pumped out by titanium getter pumps which are located on each toroidal end of a divertor modul. (author). 5 refs.; 7 figs.; 1 tab

  16. Divertor design for the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    Hill, D.N.; Braams, B.

    1994-05-01

    In this paper we discuss the present divertor design for the planned TPX tokamak, which will explore the physics and technology of steady-state (1000s pulses) heat and particle removal in high confinement (2--4x L-mode), high beta (β N ≥ 3) divertor plasmas sustained by non-inductive current drive. The TPX device will operate in the double-null divertor configuration, with actively cooled graphite targets forming a deep (0.5 m) slot at the outer strike point. The peak heat flux on, the highly tilted (74 degrees from normal) re-entrant (to recycle ions back toward the separatrix) will be in the range of 4--6 MW/m 2 with 18 MW of neutral beams and RF heating power. The combination of active pumping and gas puffing (deuterium plus impurities), along with higher heating power (45 MW maximum) will allow testing of radiative divertor concepts at ITER-like power densities

  17. Liquid metal cooled divertor for ARIES

    International Nuclear Information System (INIS)

    Muraviev, E.

    1994-01-01

    The paper represents an overview of the design study of a divertor system with liquid metal coolant (gallium) related to ARIES project. The work has been conducted by a group of specialists from Institute of Nuclear Fusion of Russian Scientific Center Kurchatov Institute within the scope of subcontract No. E212601 with General Atomics, San Diego, CA, USA. The key features of the proposed divertor design concept based on the specific LM coolant properties are as follows: (1) the requirement of the vacuum tightness of the divertor cooling tract is dismissed; (2) the pressurized coolant ducts can be separated from the plasma facing structure (PFS) elements which are subject to the thermal loads, and with this feature PFS can be replaced independently, without disturbing the cooling system; this is achieved with using free LM jets sprayed on the back side of the PFS elements, free LM film cooling and free LM draining under the action of gravity force. The divertor design has been developed formally as particularly applicable to ARIES-II reactor, the major reason for this being the choice of a vanadium-based alloy as the structural material compatible with gallium. Though there are some good prospects that carbon based materials including SiC-composite might be compatible with gallium as well. Then this concept could be used also in ARIES-IV and this possibility should be kept in mind for future

  18. The ITER Divertor Cassette Project meeting

    International Nuclear Information System (INIS)

    Akiba, M.; Tivey, R.

    2000-01-01

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

  19. Divertor erosion in DIII-D

    International Nuclear Information System (INIS)

    Whyte, D.G.; Bastasz, R.; Wampler, W.R.; Brooks, J.N.; West, W.P.; Wong, C.P.C.

    1998-05-01

    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/m 2 ) 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)

    Wong, C.P.; West, W.P.; Whyte, D.G.; Bastasz, R.J.; Brooks, J.; Wampler, W.R.

    1997-11-01

    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 3 He(d,p) 4 He nuclear reaction. For carbon, these measurements showed peak deuterium areal density of about 8 x 10 18 D/cm 2 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. Operating conditions of the BPX divertor

    International Nuclear Information System (INIS)

    Hill, D.N.; Milovich, J.; Rognlien, T.; Braams, B.J.; Brooks, J.N.; Campbell, R.; Haines, J.; Knoll, D.; Prinja, A.; Stotler, D.P.; Ulrickson, M.

    1991-01-01

    In this paper we discuss the expected operating conditions at the divertor of the BPX tokamak (Burning Plasma Experiment), the next- step US tokamak proposed for the study of self-heated plasmas at Q ≅ 5 to ignition. In this double-null device (κ ≅ 2), the predicted first-wall loading is high because of is compact size (R = 2.6m, α = 0.8m, I p = 10.6 MA, and B T ) and its high projected fusion power output (100--500 MW with up to 20 MW of ICRH). Present designs call for inertially cooled carbon-based target plate material and X-point sweeping to handle the divertor heat flux during the 3--5 s flat-top at full power. The X-point is maintained about 15--20 cm off the target plates (a distance of ∼5m along field lines), which represents a reasonable compromise between lowering the divertor electron temperature (T e,d ) by increasing the connection length, and lowering the peak divertor heat flux (q d ) by increasing the magnetic flux expansion (which is about 15--20 in this case). It is planned for the BPX device to operate with H-mode confinement; ELMs are expected because of the relatively high power flow through the edge plasma (P sep ≅ 0.6 MW/m 2 for P fus = 500 MW). The ELMs will help reduce the impurity concentration in the core plasma (Z eff ≅ 1.7) and keep the density down, but should not add significantly to the divertor heat flux since their measured contribution to the global power balance drops with increasing input power

  2. Divertor erosion in DIII-D

    International Nuclear Information System (INIS)

    Whyte, D.G.; Bastasz, R.; Wampler, W.R.; Brooks, J.N.; West, W.P.; Wong, C.P.C.; Buzhinskij, O.I.; Opimach, I.V.

    1998-08-01

    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 (OSP) of two divertor plasma conditions: attached (T e > 40 eV) ELMing plasmas, and detached (T e 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. In the attached cases, physical sputtering (with enhancement from self-sputtering and oblique incidence) is dominant, and the effective sputtering yield, Y, is greater than 10%. In ELM-free discharges, the total OSP net erosion rate is equal to the rate of carbon accumulation in the core plasma. For the detached divertor cases, the cold incident plasma eliminates physical sputtering. Attempts to measure chemically eroded hydrocarbon molecules spectroscopically indicate an upper limit of Y ≤ 0.1% for the chemical sputtering yield. Net erosion is suppressed at the outer strike-point, which becomes a region of net redeposition (∼ 4 cm/exposure-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/m 2 ) have very high net erosion rates at the OSP of an attached plasma (∼ 10 microm/s > 1,000x erosion rate of aligned surfaces). Leading edge erosion, and subsequent carbon redeposition, caused by tile gaps can account for half of the deuterium codeposition in the DIII-D divertor

  3. Divertor erosion in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Whyte, D.G. [Univ. of California, San Diego, CA (United States); Bastasz, R.; Wampler, W.R. [Sandia National Labs., Albuquerque, NM (United States); Brooks, J.N. [Argonne National Lab., IL (United States); West, W.P.; Wong, C.P.C.; Buzhinskij, O.I. [General Atomics, San Diego, CA (United States); Opimach, I.V. [TRINITI Lab. (United States)

    1998-08-01

    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 (OSP) of two divertor plasma conditions: attached (T{sub e} > 40 eV) ELMing plasmas, and detached (T{sub e} < 2 eV) ELMing plasmas. For the attached cases, the erosion rates exceed 10 cm/exposure-year, even with incident heat flux < 1 MW/m{sup 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. In the attached cases, physical sputtering (with enhancement from self-sputtering and oblique incidence) is dominant, and the effective sputtering yield, Y, is greater than 10%. In ELM-free discharges, the total OSP net erosion rate is equal to the rate of carbon accumulation in the core plasma. For the detached divertor cases, the cold incident plasma eliminates physical sputtering. Attempts to measure chemically eroded hydrocarbon molecules spectroscopically indicate an upper limit of Y {le} 0.1% for the chemical sputtering yield. Net erosion is suppressed at the outer strike-point, which becomes a region of net redeposition ({approximately} 4 cm/exposure-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 ({approximately} 50 MW/m{sup 2}) have very high net erosion rates at the OSP of an attached plasma ({approximately} 10 {micro}m/s > 1,000x erosion rate of aligned surfaces). Leading edge erosion, and subsequent carbon redeposition, caused by tile gaps can account for half of the deuterium codeposition in the DIII-D divertor.

  4. Divertors for Helical Devices: Concepts, Plans, Results, and Problems

    International Nuclear Information System (INIS)

    Koenig, R.; Grigull, P.; McCormick, K.

    2004-01-01

    With Large Helical Device (LHD) and Wendelstein 7-X (W7-X), the development of helical devices is now taking a large step forward on the path to a steady-state fusion reactor. Important issues that need to be settled in these machines are particle flux and heat control and the impact of divertors on plasma performance in future continuously burning fusion plasmas. The divertor concepts that will initially be explored in these large machines were prepared in smaller-scale devices like Heliotron E, Compact Helical System (CHS), and Wendelstein 7-AS (W7-AS). While advanced divertor scenarios relevant for W7-X were already studied in W7-AS, other smaller-scale experiments like Heliotron-J, CHS, and National Compact Stellarator Experiment will be used for the further development of divertor concepts. The two divertor configurations that are being investigated are the helical and the island divertor, as well as the local island divertor, which was successfully demonstrated on CHS and just went into operation on LHD. At present, on its route to a fully closed helical divertor, LHD operates in an open helical divertor configuration. W7-X will be equipped right from the start with an actively cooled discrete island divertor that will allow quasi-continuous operation. The divertor design is very similar to the one explored on W7-AS. For sufficiently large island sizes and not too long field line connection lengths, this divertor gives access to a partially detached quasi-steady-state operating scenario in a newly found high-density H-mode operating regime, which benefits from high energy and low impurity confinement times, with edge radiation levels of up to 90% and sufficient neutral compression in the subdivertor region (>10) for active pumping. The basic physics of the different divertor concepts and associated implementation problems, like asymmetries due to drifts, accessibility of essential operating scenarios, toroidal asymmetries due to symmetry breaking error fields

  5. Septum assessment of the JET gas box divertor

    Energy Technology Data Exchange (ETDEWEB)

    Rapp, J; Huber, A [Institut fuer Plasmaphysik, Forschungszentrum Juelich GmbH, EURATOM Association, TEC, Juelich (Germany); Fundamenski, W; Matthews, G F; Morgan, P; Stamp, M F [EURATOM-UKAEA/Fusion Association, Culham Science Centre, Abingdon, OXON (United Kingdom); Ingesson, L C [FOM Instituut voor Plasma Fysica Rijnhuizen, EURATOM Association, TEC, Nieuwegein (Netherlands); Jachmich, S [LPP-ERM/KMS, EURATOM-Belgian State Association, TEC, Brussels (Belgium)

    2008-09-15

    The influence of the physical isolation of inner and outer divertor volumes by a septum plate of the Mk-II gas box divertor, thus increasing divertor closure and neutral compression, on the plasma and divertor performance has been studied at the Joint European Torus (JET). The septum plate was installed in 1999, together with the original Mk-II gas box divertor, and was then replaced by a simple protection plate in 2001. This removal reduced the closure of the divertor by opening a line of sight path for neutrals to travel between the inner to the outer divertor volumes. Comparison of identical discharges with and without the septum thus provides direct evidence of the effect of divertor closure on plasma behaviour. With this aim, following septum removal, several dedicated L-mode and H-mode discharges have been performed, in each case repeating earlier discharges when the septum was still in place. In each case, the fuelling location was varied between the inner/outer divertor and the main chamber, and differences in detachment in the inner and outer divertors were studied. Under L-mode conditions, differences in detachment dynamics were indeed observed between closed (with septum) and open (without septum) divertor configurations, although the differences were only significant in the medium density range. In contrast, the ultimate density limit was not affected, being determined in each case by the formation of a wall multifacedted asymmetric radiation from the edge (MARFE), rather than an X-point MARFE. Under H-mode conditions, the differences were more subtle. Although the ion fluxes to the targets were unaffected, the target electron temperatures were found to be lower in the closed divertor configuration. In this case, the fuelling efficiency was the largest when the gas injected from the inner divertor, with implications on global energy confinement and ELM frequency. Otherwise, no difference in the confinement of the discharges with and without septum was

  6. Divertors for helical devices: Concepts, plans, results and problems

    International Nuclear Information System (INIS)

    Koenig, R.; Grigull, P.; McCormick, K.

    2003-01-01

    With LHD and W7-X stellarator development is now taking a large leap forward on the path to a steady-state fusion reactor. Important issues that need to be settled in these machines are particle flux and heat control, and the impact of divertors on plasma performance in future continuously burning fusion plasmas. The divertor concepts that will initially be explored in these large stellarators were carefully prepared in smaller scale devices like Heliotron E, CHS and W7-AS. While advanced divertor scenarios relevant for W7-X were already studied in W7-AS, other smaller scale experiments like Heliotron-J, CHS and NCSX will be used for the further development of divertor concepts. The two divertor configurations that are presently being investigated, are the helical and the island divertor, as well as the local island divertor (LID), which was successfully demonstrated on CHS and just went into operation on LHD. Presently, on its route to a fully closed helical divertor, LHD operates in an open helical divertor configuration. W7-X will be equipped right from the start with an actively cooled discrete island divertor which will allow quasi continuous operation. The divertor design is very similar to the one explored on W7-AS. For sufficiently large island sizes and not too long field line connection lengths, this divertor gives access to a partially detached quasi steady-state operating scenario in a newly found high density H-mode operating regime, which benefits from high energy and extremely low impurity confinement times, with edge radiation levels of up to 90 % and sufficient neutral compression in the subdivertor region (> 10) for active pumping. The basic physics of the different divertor concepts and associated implementation problems, like asymmetries due to drifts, accessibility of essential operating scenarios and toroidal asymmetries due to symmetry breaking error fields, etc. will be discussed. (orig.)

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

    Directory of Open Access Journals (Sweden)

    J. Uljanovs

    2017-08-01

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

  8. 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.

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

    International Nuclear Information System (INIS)

    Ikegami, Hideo.

    1995-01-01

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

  10. Development of a radiative divertor for DIII-D

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  11. Role of molecular effects in divertor plasma recombination

    Directory of Open Access Journals (Sweden)

    A.S. Kukushkin

    2017-08-01

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

  12. Comparative divertor-transport study for helical devices

    International Nuclear Information System (INIS)

    Feng, Y.; Sardei, F.; Kobayashi, M.

    2008-10-01

    Using the island divertors (ID) of W7-AS and W7-X and the helical divertor (HD) of LHD as examples, the paper presents a comparative divertor transport study for three typical helical devices of different machine-size following two distinct divertor concepts, aiming at identifying common physics issues/effects for mutual validation and combined studies. Based on EMC3/EIRENE simulations supported by experimental results, the paper first reviews and compares the essential transport features of the W7-AS ID and the LHD HD in order to build a base and framework for a predictive study of W7-X. Revealed is the fundamental role of the low-order magnetic islands in both divertor concepts. Preliminary EMC3/EIRENE simulation results for W7-X are presented and discussed with respect to W7-AS and LHD in order to show how the individual field and divertor topologies affect the divertor transport and performance. For instance, a high recycling regime which is absent from W7-AS and LHD is expected for W7-X. Topics addressed are restricted to the basic function elements of a divertor such as particle flux enhancement and impurity retention. In particular, the divertor function on reducing the influx of intrinsic impurities is examined for all the three devices under different divertor plasma conditions. Special attention is paid to examining the island screening potential of intrinsic impurities which has been predicted for all the three devices under high divertor collisionality conditions. The results are discussed in conjunction with the experimental observations for high density divertor plasmas in W7-AS and LHD. (author)

  13. Experimental studies of the snowflake divertor in TCV

    Directory of Open Access Journals (Sweden)

    B. Labit

    2017-08-01

    Full Text Available To address the risk that, in a fusion reactor, the conventional single-null divertor (SND configuration may not be able to handle the power exhaust, alternative divertor configurations, such as the Snowflake divertor (SFD, are investigated in TCV. The expected benefits of the SFD-minus in terms of power load and peak heat flux are discussed and compared to experimental measurements. In addition, key results obtained during the last years are summarized.

  14. NSTX plasma response to lithium coated divertor

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  15. He-cooled divertor development for DEMO

    International Nuclear Information System (INIS)

    Norajitra, P.; Giniyatulin, R.; Ihli, T.; Janeschitz, G.; Krauss, W.; Kruessmann, R.; Kuznetsov, V.; Mazul, I.; Widak, V.; Ovchinnikov, I.; Ruprecht, R.; Zeep, B.

    2007-01-01

    Goal of the He-cooled divertor development for future fusion power plants is to resist a high heat flux of at least 10 MW/m 2 . The development includes the fields of design, analyses, and experiments. A helium-cooled modular jet concept (HEMJ) has been defined as reference solution, which is based on jet impingement cooling. In cooperation with the Efremov Institute, work was aimed at construction and high heat flux tests of prototypical tungsten mockups to demonstrate their manufacturability and their performances. A helium loop was built for this purpose to simulate the realistic thermo-hydraulics conditions close to those of DEMO (10 MPa He, 600 deg. C). The first high heat flux test results confirm the feasibility and the performance of the divertor design

  16. 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.

  17. Divertor and scoop limiter experiments on PDX

    International Nuclear Information System (INIS)

    McGuire, K.; Beirsdorfer, P.; Bell, M.

    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 tausub(E)/Isub(p)) to average H-mode plasma. Two new instabilities are observed near the plasma edge in PDX during H-mode operation. The first, a quasi-coherent fluctuation, occurred in bursts at well-defined frequencies (Δω/ω<=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 ERPs are characterized by sharp spikes in the divertor plasma density, Hsub(α) emission, and on the X-ray signals they appear as sawtooth-like relaxations at the plasma edge with an inversion radius near the separatrix. Attempts to obtain high βsub(T) in the H-mode discharges were hampered by a deterioration in the H-mode confinement and major disruptions which limited the achievable βsub(T). A study of the stability of both the limiter L-mode and divertor H-mode discharge 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. (author)

  18. Divertor and scoop limiter experiments on PDX

    International Nuclear Information System (INIS)

    McGuire, K.; Beiersdorfer, P.; Bell, M.

    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 (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

  19. Two-point model for divertor transport

    International Nuclear Information System (INIS)

    Galambos, J.D.; Peng, Y.K.M.

    1984-04-01

    Plasma transport along divertor field lines was investigated using a two-point model. This treatment requires considerably less effort to find solutions to the transport equations than previously used one-dimensional (1-D) models and is useful for studying general trends. It also can be a valuable tool for benchmarking more sophisticated models. The model was used to investigate the possibility of operating in the so-called high density, low temperature regime

  20. 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

  1. 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.

  2. Low energy neutral particle fluxes in the JET divertor

    International Nuclear Information System (INIS)

    Reichle, R.; Horton, L.D.; Ingesson, L.C.; Jaeckel, H.J.; McCormick, G.K.; Loarte, A.; Simonini, R.; Stamp, M.F.

    1997-01-01

    First measurements are presented of the total power loss through neutral particles and their average energy in the JET divertor. The method used distinguishes between the heat flux and the electromagnetic radiation on bolometers. This is done by comparing measurements from inside the divertor either with opposite lines of sight or with a tomographic reconstruction of the radiation. The typical value of the total power loss in the divertor through neutrals is about 1 MW. The average energy of the neutral particles at the inner divertor leg is 1.5-3 eV when detachment is in progress, which agrees with EDGE2D/NIMBUS modelling. (orig.)

  3. Island divertor studies on W7-AS

    International Nuclear Information System (INIS)

    Sardei, F.; Feng, Y.; Grigull, P.; Herre, G.; Hildebrandt, D.; Hofmann, J.V.; Kisslinger, J.; Brakel, R.; Das, J.; Geiger, J.; Heinrich, O.; Kuehner, G.; Niedermeyer, H.; Reiter, D.; Richter-Gloetzl, M.; Runov, A.; Schneider, R.; Stroth, U.; Verbeek, H.; Wagner, F.; Wolf, R.

    1997-01-01

    Basic topological features of the island divertor concept for low shear stellarators are discussed with emphasis on the differences to tokamak divertors. Extensive measurements of the edge structures by two-dimensional plasma spectroscopy and by target calorimetry are in excellent agreement with predicted vacuum and equilibrium configurations, which are available up to central β values of ∝1%. For this β value the calculated field-line pitch inside the islands is twice that of the corresponding vacuum case. Video observations of the strike points indicate stability of the island structures for central β values up to ∝3.7%. The interpretation of the complex island divertor physics of W7-AS has become possible by the development of the three-dimensional plasma transport code EMC3 (Edge Monte Carlo 3D), which has been coupled self-consistently to the EIRENE neutral gas code. Analysis of high density NBI discharges gives strong indications of stable high recycling conditions for n e ≥10 20 m -3 . The observations are reproduced by the EMC3/EIRENE code and supported by calculations with the B2/EIRENE code adapted to W7-AS. Improvement of recycling, pumping and target load distribution is expected from the new optimized target plates and baffles to be installed in W7-AS. (orig.)

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  5. The DIII-D Radiative Divertor Project: Status and plans

    International Nuclear Information System (INIS)

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

    1996-10-01

    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 second and final phase, in 1998. The phased approach enables the continuation of the divertor characterization research in the lower divertor while providing pumping for density control in high triangularity, single- or double-null advanced tokamak discharges. 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. By puffing neutral gas into the divertor region, a reduction in the heat flux on the target plates will be be demonstrated without a large rise in core density. This reduction in heat flux is accomplished by dispersing the power with radiation in the divertor region. Experiments and modeling have formed the basis for the new design. The capability of the DIII-D cryogenic system is being upgraded as part of this project. The increased capability of the cryogenic system will allow delivery of liquid helium and nitrogen to three new cryopumps. Physics studies on the effects of slot width and length can be accomplished easily with the design of the Radiative Divertor. The slot width can be varied by installing graphite tiles of different geometry. The change in slot length, the distance from the X-point to the target plate, requires relocating the structure vertically and can be completed in about 6-8 weeks. Radiative Divertor diagnostics are being designed to provide comprehensive measurements for diagnosing the divertor. Required diagnostic modifications will be minimal for Phase 1, but extensive for Phase 2 installation. These Phase 2 diagnostics will be required to fully diagnose the high triangularity discharges in the divertor slots

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

    International Nuclear Information System (INIS)

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

    1996-09-01

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

  7. ELM-induced transient tungsten melting in the JET divertor

    Science.gov (United States)

    Coenen, J. W.; Arnoux, G.; Bazylev, B.; Matthews, G. F.; Autricque, A.; Balboa, I.; Clever, M.; Dejarnac, R.; Coffey, I.; Corre, Y.; Devaux, S.; Frassinetti, L.; Gauthier, E.; Horacek, J.; Jachmich, S.; Komm, M.; Knaup, M.; Krieger, K.; Marsen, S.; Meigs, A.; Mertens, Ph.; Pitts, R. A.; Puetterich, T.; Rack, M.; Stamp, M.; Sergienko, G.; Tamain, P.; Thompson, V.; Contributors, JET-EFDA

    2015-02-01

    The original goals of the JET ITER-like wall included the study of the impact of an all W divertor on plasma operation (Coenen et al 2013 Nucl. Fusion 53 073043) and fuel retention (Brezinsek et al 2013 Nucl. Fusion 53 083023). ITER has recently decided to install a full-tungsten (W) divertor from the start of operations. One of the key inputs required in support of this decision was the study of the possibility of W melting and melt splashing during transients. Damage of this type can lead to modifications of surface topology which could lead to higher disruption frequency or compromise subsequent plasma operation. Although every effort will be made to avoid leading edges, ITER plasma stored energies are sufficient that transients can drive shallow melting on the top surfaces of components. JET is able to produce ELMs large enough to allow access to transient melting in a regime of relevance to ITER. Transient W melt experiments were performed in JET using a dedicated divertor module and a sequence of IP = 3.0 MA/BT = 2.9 T H-mode pulses with an input power of PIN = 23 MW, a stored energy of ˜6 MJ and regular type I ELMs at ΔWELM = 0.3 MJ and fELM ˜ 30 Hz. By moving the outer strike point onto a dedicated leading edge in the W divertor the base temperature was raised within ˜1 s to a level allowing transient, ELM-driven melting during the subsequent 0.5 s. Such ELMs (δW ˜ 300 kJ per ELM) are comparable to mitigated ELMs expected in ITER (Pitts et al 2011 J. Nucl. Mater. 415 (Suppl.) S957-64). Although significant material losses in terms of ejections into the plasma were not observed, there is indirect evidence that some small droplets (˜80 µm) were released. Almost 1 mm (˜6 mm3) of W was moved by ˜150 ELMs within 7 subsequent discharges. The impact on the main plasma parameters was minor and no disruptions occurred. The W-melt gradually moved along the leading edge towards the high-field side, driven by j × B forces. The evaporation rate determined

  8. Advantages and Challenges of Radiative Liquid Lithium Divertor

    Science.gov (United States)

    Ono, Masayuki

    2017-10-01

    Steady-state fusion power plant designs present major divertor technology challenges, including high divertor heat flux both in steady-state and during transients. In addition to these concerns, there are the unresolved technology issues of long term dust accumulation and associated tritium inventory and safety issues. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid Li divertor (RLLD) concept and its variant, the active liquid Li divertor concept (ARLLD), taking advantage of the enhanced Li radiation in relatively poorly confined divertor plasmas. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest Li-loop could provide a possible solution for the outstanding fusion reactor technology issues such as divertor heat flux mitigation and real time dust removal, while potentially improving the reactor plasma performance. Laboratory tests are also planned to investigate the Li-T recover efficiency and other relevant research topics of the RLLD. This work supported by DoE Contract No. DE-AC02-09CH11466.

  9. Engineering design of the Aries-IV gaseous divertor

    International Nuclear Information System (INIS)

    Hasan, M.Z.; Najmabadi, F.; Sharafat, S.

    1994-01-01

    ARIES-IV is a conceptual, D-T burning, steady-state tokamak fusion reactor producing 1000 MWe net. It operates in the second plasma stability regime. The structural material is SiC composite and the primary coolant is helium at 10MPa base pressure. ARIES-IV uses double-null divertors for particle control. Total thermal power recovered from the divertors is 425MW, which is 16% of the total reactor thermal power. Among the desirable goals of divertor design were to avoid the use of tungsten and to use the same structural material and primary coolant as in the blanket design. In order to reduce peak heat flux, the innovative gaseous divertor has been used in ARIES-IV. A gaseous divertor reduces peak heat flux by increasing the surface area and by distributing particle and radiation energy more uniformly. Another benefit of gaseous divertor is the reduction of plasma temperature in the divertor chamber, so that material erosion due to sputtering, can be diminished. This makes the use of low-Z material possible in a gaseous divertor

  10. Divertor heat flux control and plasma-material interaction

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  11. Alternative divertor target concepts for next step fusion devices

    Science.gov (United States)

    Mazul, I. V.

    2016-12-01

    The operational conditions of a divertor target in the next steps of fusion devices are more severe in comparison with ITER. The current divertor designs and technologies have a limited application concerning these conditions, and so new design concepts/technologies are required. The main reasons which practically prevent the use of the traditional motionless solid divertor target are analyzed. We describe several alternative divertor target concepts in this paper. The comparative analysis of these concepts (including the advantages and the drawbacks) is made and the prospects for their practical implementation are prioritized. The concept of the swept divertor target with a liquid metal interlayer between the moving armour and motionless heat-sink is presented in more detail. The critical issues of this design are listed and outlined, and the possible experiments are presented.

  12. Experience gained with the 3D machining of the W7-X HHF divertor target elements

    Energy Technology Data Exchange (ETDEWEB)

    Junghanns, P. [Max Planck Institute for Plasma Physics, Greifswald (Germany); Boscary, J., E-mail: jean.boscary@ipp.mpg.de [Max Planck Institute for Plasma Physics, Garching (Germany); Peacock, A. [Max Planck Institute for Plasma Physics, Garching (Germany)

    2015-10-15

    Highlights: • The Wendelstein 7-X surface of the actively cooled divertor is built up of 890 individually 3D machined target elements. • To date 300 target elements have been 3D machined with an accuracy of ±0.015 mm. • Copper discovered on the surface of few elements is no risk to operation. - Abstract: The high heat flux (HHF) divertor of W7-X consists of 100 target modules assembled from 890 actively water-cooled target elements protected with CFC tiles. The divertor surface will be built up of individually 3D machined target elements with 89 individual element types. To date 300 of the 890 target elements have been 3D machined with a very good accuracy. To achieve this successful result, a prototyping phase has been conducted to qualify the manufacturing route and to define the acceptance criteria with measures taken to minimize the risk of unacceptable damage during the manufacturing. After the 3D-machining, during the incoming inspection, copper infiltration from the interface between the CFC tiles and the CuCrZr heat sink to the plasma facing surface was detected in a small number of elements.

  13. Experience gained with the 3D machining of the W7-X HHF divertor target elements

    International Nuclear Information System (INIS)

    Junghanns, P.; Boscary, J.; Peacock, A.

    2015-01-01

    Highlights: • The Wendelstein 7-X surface of the actively cooled divertor is built up of 890 individually 3D machined target elements. • To date 300 target elements have been 3D machined with an accuracy of ±0.015 mm. • Copper discovered on the surface of few elements is no risk to operation. - Abstract: The high heat flux (HHF) divertor of W7-X consists of 100 target modules assembled from 890 actively water-cooled target elements protected with CFC tiles. The divertor surface will be built up of individually 3D machined target elements with 89 individual element types. To date 300 of the 890 target elements have been 3D machined with a very good accuracy. To achieve this successful result, a prototyping phase has been conducted to qualify the manufacturing route and to define the acceptance criteria with measures taken to minimize the risk of unacceptable damage during the manufacturing. After the 3D-machining, during the incoming inspection, copper infiltration from the interface between the CFC tiles and the CuCrZr heat sink to the plasma facing surface was detected in a small number of elements.

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

    Science.gov (United States)

    Frerichs, Heinke; Schmitz, Oliver; Covele, Brent; Guo, Houyang; Hill, David; Feng, Yuhe

    2017-10-01

    In the Small Angle Slot (SAS) divertor in DIII-D, the combination of misaligned slot structure and non-axisymmetric perturbations to the magnetic field causes the strike point to vary radially along the divertor slot and even leave it at some toroidal locations. This effect essentially introduces an opening in the divertor slot from where recycling neutrals can easily escape, and thereby degrade performance of the slot divertor. This effect has been approximated by a finite gap in the divertor baffle. Simulations with EMC3-EIRENE show that a toroidally localized loss of divertor closure can result in non-axisymmetric divertor densities and temperatures. This introduces a density window of 10-15% on top of the nominal threshold separatrix density during which a non-axisymmetric onset of local detachment occurs, initially leaving the gap and up to 60 deg beyond that still attached. Conversely, the impact of such toroidally localized divertor perturbations on the toroidal symmetry of midplane separatrix conditions is small. This work has been funded by the U.S. Department of Energy under Early Career Award Grant DE-SC0013911, and Grant DE-FC02-04ER54698.

  15. Design analysis of the ITER divertor

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  16. On the refuelling of large divertor experiments

    International Nuclear Information System (INIS)

    Staebler, A.; Haas, G.; Ott, W.; Speth, E.

    1976-01-01

    The use of fast hydrogen atoms, molecules and clusters for refuelling large divertor-experiments like ASDEX is investigated. Three criteria for the choice among the various methods are discussed. It is shown that clusters suffer from lack of penetration. Molecules, created by fragmentation of clusters, offer the advantage of plasma-like energy combined with appreciable penetration. Large penetration and high ionization efficiency can only be achieved at energies for above the plasma temperature with H 0 -atoms of several tens of keV

  17. Reversal of plasma flow in tokamak divertors

    International Nuclear Information System (INIS)

    Maddison, G.P.; Reiter, D.; Stangeby, P.C.; Prinja, A.K.

    1993-01-01

    In a magnetic divertor, retention of impurity ions is expected to be dependent on an expulsive thermal force directed up the gradient of ion temperature being opposed by frictional entrainment in a plasma flow towards the target. Preferred conditions of high recycling, however, can induce a reversal of usual plasma flow, with consequent reinforcement of thermal forces potentially leading to damaging contamination of the core. Backflow in diverted plasmas was first anticipated theoretically by Nedospasov and Tokar', subsequently observed experimentally in DITE and JET, and has been seen in a number of numerical studies. We report briefly on a systematic investigation of steady-state divertor flow reversal for ITER-relevant conditions, by detailed numerical modelling. The BRAAMS 'B2' edge plasma transport code is used, with both analytic approximations and EIRENE Monte Carlo simulation of neutral particle recycling. The flexibility of numerical models regarding physics admitted is exploited to expose the key role of redistribution of recycling sources across magnetic surfaces in flow reversal. Concomitant amplification of cross-field ion diffusion in the SOL is also examined. (author) 10 refs., 4 figs

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

    International Nuclear Information System (INIS)

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

    2012-11-01

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

  19. Operation method for thermonuclear device and divertor for it

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  20. Physical study of experimental fusion breeder FEB divertor

    International Nuclear Information System (INIS)

    Zhu Yukun; Zhou Xiaobing; Huang Jinhua; Feng Kaiming; Deng Peizhi; Huo Tiejun

    1999-10-01

    The physical study of FEB divertor is presented. In order to improve the impurity control and increase ion-neutral interactions in the divertor, the configuration of the divertor is optimized to be the close type in the engineering design activity compared with the open type in the early conceptual activity. The operation mode of the divertor is designed to be partial detached plasma mode under conditions of combination gas-puffing with impurity injection. The position of gas-puffing is optimized to be at the torus mid-plane with NEWT1D code from the viewpoint of impurity retention and radiation in the scrape-off layer/divertor region. Boron is chosen as the injected impurity. The effect of boron impurity injection is evaluated from the reduced heat load on the divertor target. The plasma pressure drop along the scrape-off layer/divertor region is estimated with the two-point transport model and impurity radiation model in the dynamic gas target concept. The simulation results show that the plasma pressure drop factor f p is not only related to the radiation fraction f rad but also related greatly to the stagnation point density n s

  1. Development of divertor simulation research in the GAMMA 10/PDX tandem mirror

    International Nuclear Information System (INIS)

    Nakashima, Y.; Sakamoto, M.; Yoshikawa, M.; Oki, K.; Takeda, H.; Ichimura, K.; Hosoi, K.; Hirata, M.; Ichimura, M.; Ikezoe, R.; Imai, T.; Kariya, T.; Katanuma, I.; Kohagura, J.; Minami, R.; Numakura, T.; Wang, X.; Iwamoto, M.; Hosoda, Y.; Asakura, Nobuyuki; Fukumoto, Masakatsu; Kubo, Hirotaka; Hatayama, A.; Hirooka, Y.; Masuzaki, S.; Sagara, A.; Shoji, M.; Kado, S.; Matsuura, H.; Nagata, S.; Shikama, T.; Nishino, N.; Ohno, N.; Tonegawa, A.; Ueda, Y.

    2014-10-01

    This paper describes the recent development of divertor simulation research towards the characterization and control of the detached plasma. In the end-mirror of large tandem mirror device GAMMA 10/PDX, additional ICRF heating experiments in the anchor-cells significantly increases the density in both the anchor and the central cells, which attained the highest particle flux up to 1.7×10 23 particles/s·m 2 at the end-mirror exit. Massive gas injection (H 2 and noble gases) to enhance the radiation cooling in divertor simulation experimental module (D-module) was performed and we have succeeded for the first time in achieving detachment of high temperature plasma equivalent to the SOL plasma of tokamaks by using linear device. A remarkable reduction of the electron temperature (from few tens eV to < 3 eV) on the target plate was successfully achieved associated with the strong reduction of particle and heat fluxes. Two-dimensional image of Hα emission in D-module observed with high-speed camera showed the bright emission in upstream region and strong reduction near the target plate. These results indicate radiation cooling and formation of detached plasma due to gas injection. It is also found that Xe gas is much effective on achieving detached plasma than Ar gas. Simultaneous injection of noble gas and hydrogen gas showed the most effective results on detached plasma generation, which indicates the effect of molecular activated recombination (MAR) processes. The above results will contribute to establishment of detached plasma control and clarification of radiation cooling mechanism towards the development of future divertor systems. (author)

  2. Resonant island divertor experiments on text

    International Nuclear Information System (INIS)

    deGrassie, J.S.; Evans, T.E.; Jackson, G.L.

    1988-09-01

    The first experimental tests of the resonant island divertor (RID) concept have been carried out on the Texas Experimental Tokamak (TEXT). Modular perturbation coils produce static resonant magnetic fields at the tokamak boundary. The resulting magnetic islands are used to guide heat and particle fluxes around a small scoop limiter head. An enhancement in the limiter collection efficiency over the nonisland operation, as evidenced by enhanced neutral density within the limiter head, of up to a factor of 4 is obtained. This enhancement is larger than one would expect given the measured magnitude of the cross-field particle transport in TEXT. It is proposed that electrostatic perturbations occur which enhance the ion convection rate around the islands. Preliminary experiments utilizing electron cyclotron heating (ECH) in conjunction with RID operation have also have been performed. 6 refs., 3 figs

  3. Development of liquid lithium divertor for fusion reactor

    International Nuclear Information System (INIS)

    Evtihkin, V. A.; Lyublinskij, I. E.; Vertkov, A.V.; Chumanov, A.V.; Shpolyanskij, V.N.

    2000-01-01

    Development of divertor is one of the most acute problems of the tokamak fusion reactor. The use of such materials as tungsten, beryllium, graphite and CFC's enabled to solve the problem to a certain extent fulfilling the need of the ITER project. The problem still rests unsolved for the DEMO-type reactors. Lithium if used as a material for high heat flux components may provide a successful solution of the problem. A concept of Li divertor based on the use of capillary-pore structures (CPS) is proposed and is being validated by a complex of experimental research and engineering developments. An optional concept of Li divertor for power removal at 400 MW in steady-state (DEMO-S project) is presented. The complex of experimental research is under way to prove the serviceability of the Li CPS in different conditions that would be realized in divertor

  4. Divertor IR thermography on Alcator C-Moda)

    Science.gov (United States)

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

    2010-10-01

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

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

    International Nuclear Information System (INIS)

    Sugie, Tatsuo; Ogawa, Hiroaki; Ebisawa, Katsuyuki; Ando, Toshiro; Kasai, Satoshi; 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 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)

  6. Design of divertor impurity monitoring system for ITER

    International Nuclear Information System (INIS)

    Sugie, Tatsuo; Ogawa, Hiroaki; Ando, Toshiro; Kasai, Satoshi; Katsunuma, Jun; Maruo, Mitsumasa.

    1996-12-01

    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)

  7. Thermomechanical simulation of WEST actively cooled upper divertor

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  8. Thermomechanical simulation of WEST actively cooled upper divertor

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-15

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

  9. 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.

  10. The cascading pebble divertor for the spherical tokamak power plant

    International Nuclear Information System (INIS)

    Voss, G.M.; Bond, A.; Davis, S.; Harte, M.; Watson, R.

    2006-01-01

    The design of a power plant based on the spherical tokamak (ST) is being developed in order to explore its potential advantages. The plasma is operated in a double null configuration, forming both an upper and lower divertor. In order to accommodate the high erosion rates and heat fluxes developed in the divertors, a system based on a cascading flow of silicon carbide pebbles is being developed. The pebbles flow into the upper divertor where they fall as a toroidal curtain, which intercepts the divertor particle flux. The pebbles then flow under gravity through ducts to the lower divertor where they form a similar curtain. The bulk temperature of the pebbles rises to about 1150 deg. C although the outer surface is transiently heated to about 1800 deg. C. The pebbles pass out of the vacuum chamber into holding tanks and then into a fluidised bed heat exchanger. Here the pebbles are cooled down to about 340 deg. C and dust and damaged pebbles are removed. The pebbles are transferred to an upper tank by a pneumatic conveyor where the remaining gas is removed and the pebbles flow into the upper divertor again

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

    International Nuclear Information System (INIS)

    Allen, S.L.; Hill, D.N.; Carlstrom, T.N.; Nilson, D.G.

    1996-10-01

    We have obtained the first measurements of n e and T e 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 D 2 and impurities. Mapping of the data to either the (L pol , φ) or (R, Z) planes with the EFIT equilibrium is used to analyze the 2D profiles. We find that in ELMing H-mode: n e , T e , and P e are relatively constant along field lines from the X-point to the divertor plate, especially near the separatrix field line. With D 2 puffing, the DTS profiles indicate that T e 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-T e divertor, with an increased electron density in the range of 2 to 4 x 10 20 m -3 . Detailed comparisons of the spatial profiles of n e , T e , and electron pressure P e , are presented for several operating modes. In addition, these data are compared with initial calculations from the UEDGE fluid code

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

    Energy Technology Data Exchange (ETDEWEB)

    Ryutov, D. D., E-mail: ryutov1@llnl.gov; Cohen, R. H.; Rognlien, T. D.; Soukhanovskii, V. A.; Umansky, M. V. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States)

    2014-05-15

    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.

  13. Efficiency of water coolant for DEMO divertor

    Energy Technology Data Exchange (ETDEWEB)

    Fetzer, Renate, E-mail: renate.fetzer@kit.edu; Igitkhanov, Yuri; Bazylev, Boris

    2015-10-15

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  14. Efficiency of water coolant for DEMO divertor

    International Nuclear Information System (INIS)

    Fetzer, Renate; Igitkhanov, Yuri; Bazylev, Boris

    2015-01-01

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  15. Divertor conceptual designs for a fusion power plant

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  16. Analysis of sweeping heat loads on divertor plate materials

    International Nuclear Information System (INIS)

    Hassanein, A.

    1991-01-01

    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/m 2 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

  17. Toroidal asymmetries in divertor impurity influxes in NSTX

    Directory of Open Access Journals (Sweden)

    F. Scotti

    2017-08-01

    Full Text Available Toroidal asymmetries in divertor carbon and lithium influxes were observed in NSTX, due to toroidal differences in surface composition, tile leading edges, externally-applied three-dimensional (3D fields and toroidally-localized edge plasma modifications due to radio frequency heating. Understanding toroidal asymmetries in impurity influxes is critical for the evaluation of total impurity sources, often inferred from measurements with a limited toroidal coverage. The toroidally-asymmetric lithium deposition induced asymmetries in divertor lithium influxes. Enhanced impurity influxes at the leading edge of divertor tiles were the main cause of carbon toroidal asymmetries and were enhanced during edge localized modes. Externally-applied 3D fields led to strike point splitting and helical lobes observed in divertor impurity emission, but marginal changes to the toroidally-averaged impurity influxes. Power coupled to the scrape-off layer SOL plasma during radio frequency (RF heating of H-mode discharges enhanced impurity influxes along the non-axisymmetric divertor footprint of flux tubes connecting to plasma in front of the RF antenna.

  18. Optimal thermal-hydraulic performance for helium-cooled divertors

    International Nuclear Information System (INIS)

    Izenson, M.G.; Martin, J.L.

    1996-01-01

    Normal flow heat exchanger (NFHX) technology offers the potential for cooling divertor panels with reduced pressure drops (<0.5% Δp/p), reduced pumping power (<0.75% pumping/thermal power), and smaller duct sizes than conventional helium heat exchangers. Furthermore, the NFHX can easily be fabricated in the large sizes required for divertors in large tokamaks. Recent experimental and computational results from a program to develop NFHX technology for divertor coolings using porous metal heat transfer media are described. We have tested the thermal and flow characteristics of porous metals and identified the optimal heat transfer material for the divertor heat exchanger. Methods have been developed to create highly conductive thermal bonds between the porous material and a solid substrate. Computational fluid dynamics calculations of flow and heat transfer in the porous metal layer have shown the capability of high thermal effectiveness. An 18-kW NFHX, designed to meet specifications for the international Thermonuclear Experimental Reactor divertor, has been fabricated and tested for thermal and flow performance. Preliminary results confirm design and fabrication methods. 11 refs., 12 figs., 1 tab

  19. Upgraded divertor Thomson scattering system on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Glass, F., E-mail: glassf@fusion.gat.com; Carlstrom, T. N.; Du, D.; Taussig, D. A.; Boivin, R. L. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); McLean, A. G. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94550 (United States)

    2016-11-15

    A design to extend the unique divertor Thomson scattering system on DIII-D to allow measurements of electron temperature and density in high triangularity plasmas is presented. Access to this region is selectable on a shot-by-shot basis by redirecting the laser beam of the existing divertor Thomson system inboard — beneath the lower floor using a moveable, high-damage threshold, in-vacuum mirror — and then redirecting again vertically. The currently measured divertor region remains available with this mirror retracted. Scattered light is collected from viewchords near the divertor floor using in-vacuum, high temperature optical elements and relayed through the port window, before being coupled into optical fiber bundles. At higher elevations from the floor, measurements are made by dynamically re-focusing the existing divertor system collection optics. Nd:YAG laser timing, analysis of the scattered light spectrum via polychromators, data acquisition, and calibration are all handled by existing systems or methods of the current multi-pulse Thomson scattering system. Existing filtered polychromators with 7 spectral channels are employed to provide maximum measurement breadth (T{sub e} in the range of 0.5 eV–2 keV, n{sub e} in the range of 5 × 10{sup 18}–1 × 10{sup 21} m{sup 3}) for both low T{sub e} in detachment and high T{sub e} measurement up beyond the separatrix.

  20. Developing physics basis for the snowflake divertor in the DIII-D tokamak

    Science.gov (United States)

    Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; Lasnier, C. J.; Makowski, M. A.; McLean, A. G.; Meyer, W. H.; Ryutov, D. D.; Kolemen, E.; Groebner, R. J.; Hyatt, A. W.; Leonard, A. W.; Osborne, T. H.; Petrie, T. W.; Watkins, J.

    2018-03-01

    Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (see standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power P_NBI ≤slant 4 -5 MW and a range of plasma currents I_p=0.8-1.2 MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10-15% of total outer divertor power. Measurements of electron pressure and poloidal beta βp support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4-4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and the standard

  1. 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

  2. Stability of the plasma in a bundle divertor

    International Nuclear Information System (INIS)

    Yang, T.F.; Callen, J.D.

    1979-02-01

    Due to the pressure and magnetic field gradients and curvature of the magnetic field lines in a bundle divertor of a tokamak device, the plasma may be unstable to local interchange modes. Turbulent transport could be quite large and lead to a thick scrape-off layer which is as large as the radius of curvature of the diverted flux bundle. Such turbulence would be beneficial for lowering the energy and particle fluxes on the collector in a bundle divertor. The effect of a bundle divertor on the β limit resulting from the ballooning modes of instability in the central plasma is also estimated. The critical β is reduced by less than one percent

  3. 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.

  4. The simple map for a single-null divertor tokamak

    International Nuclear Information System (INIS)

    Punjabi, A.; Verma, A.; Boozer, A.

    1996-01-01

    We present the simple map for a single-null divertor tokamak. The simple map is an area-preserving map based on the idea that magnetic field lines are a single-degree-of-freedom time-dependent Hamiltonian system, and that the basic features of such systems near the X-point are generic. We obtain the properties of this map and the resulting footprints of field lines on the divertor plate. These include the width of the stochastic layer, the edge safety factor, the area of the footprint and the amount of magnetic flux diverted. We give the safety factor profile, the average and median values of strike angles, lengths and the Liapunov exponents. We describe how the effects of magnetic perturbations can be included in the simple map. We show how the map can be applied to the problem of the determination of heat flux on the divertor plate in tokamaks. (Author)

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

    Clever, Meike

    2010-01-01

    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

  7. Radiative divertor plasmas with convection in DIII-D

    International Nuclear Information System (INIS)

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

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    Jaenicke, R.; Anton, M.; Baldzuhn, J.

    2001-01-01

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

  9. Divertor plate concept with carbon based armour for NET

    International Nuclear Information System (INIS)

    Moons, F.; Howard, R.; Kneringer, G.; Stickler, R.

    1989-01-01

    A series of tests has been performed on simulated divertor elements for NET at the JET neutral beam injector test bed. The test section consisted of a water cooled main structure, the surface of which was protected with a carbon based armour in the form of tiles. The scope of these was to study the thermal behaviour of mechanically attached tiles with the use of an intermediate soft carbon layer to improve the thermal contact under divertor relevant conditions. (author). 4 refs.; 4 figs.; 1 tab

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  12. Engineering design of a toroidal divertor for the EBT-S fusion device. Final report, Phase II. EBT-S divertor project

    International Nuclear Information System (INIS)

    Mai, L.P.; Malick, F.S.

    1981-01-01

    The mechanical, structural, thermal, electrical, and vacuum design of a magnetic toroidal divertor system for the Elmo Bumpy Torus (EBT-S) is presented. The EBT-S is a toroidal magnetic fusion device located at the ORNL that operates under steady state conditions. The engineering of the divertor was performed during the second of three phases of a program aimed at the selection, design, fabrication, and installation of a magnetic divertor for EBT-S. The magnetic analysis of the toroidal divertor was performed during Phase I of the program and has been reported in a separate document. In addition to the details of the divertor design, the modest modifications that are required to the EBT-S device and facility to accommodate the divertor system are presented

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

    Science.gov (United States)

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

    2018-04-01

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

  14. 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.

  15. Stochastic broadening of the scrapeoff layer of a single-null divertor tokamak

    International Nuclear Information System (INIS)

    Punjabi, A.; Verma, A.; Boozer, A.

    1992-01-01

    Magnetic perturbations cause the region near the separatrix of a magnetic divertor to become stochastic. The last magnetic surface to provide magnetic confinement passes inside the X-point a distance that is proportional to the square root of the applied perturbation. Particles that diffuse across the last confining surface can follow open magnetic lines to the divertor plates. The strike points of these field lines on the divertor plates lie in helical discrete stripes. The properties of these stripes is important for determining if one can control the heat loads on divertor plates as well as assessing the effects of natural perturbations, such as MHD activity, on divertor designs

  16. 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.

  17. Enhancing the DEMO divertor target by interlayer engineering

    International Nuclear Information System (INIS)

    Barrett, T.R.; McIntosh, S.C.; Fursdon, M.; Hancock, D.; Timmis, W.; Coleman, M.; Rieth, M.; Reiser, J.

    2015-01-01

    Highlights: • The European ‘near-term’ DEMO forsees a water-cooled divertor. • Divertor targets typically use an interlayer between the armour and structure. • Engineering the properties of the interlayer can yield large gains in performance. • A response surface based design search and optimisation method is used. • A new design passes linear-elastic code rules up to applied heat flux of 18 MW/m 2 . - Abstract: A robust water-cooled divertor target plate solution for DEMO has to date remained elusive. Common to all contemporary concepts is an interlayer at the boundary between the tungsten armour and the cooling structure. In this paper we show by design optimisation that an effectively designed interlayer can produce dramatic gains in power handling. By engineering the interlayer as part of the design study, it is found that divertor performance is enhanced by either a low conductivity ‘Thermal Break’ interlayer or an ‘Ultra-Compliant’ interlayer. For a 10 MW/m 2 surface heat flux we find that a thermal conductivity of 15 W/mK and elastic modulus of 1 GPa are effective. A design is proposed which passes linear-elastic code rules up to an applied heat flux of 18 MW/m 2 .

  18. Enhancing the DEMO divertor target by interlayer engineering

    Energy Technology Data Exchange (ETDEWEB)

    Barrett, T.R., E-mail: tom.barrett@ccfe.ac.uk [CCFE, Culham Science Centre, Oxfordshire OX14 3DB (United Kingdom); McIntosh, S.C.; Fursdon, M.; Hancock, D.; Timmis, W.; Coleman, M. [CCFE, Culham Science Centre, Oxfordshire OX14 3DB (United Kingdom); Rieth, M.; Reiser, J. [Karlsruhe Institute for Technology, IMF-I, D-7602 Karlsruhe (Germany)

    2015-10-15

    Highlights: • The European ‘near-term’ DEMO forsees a water-cooled divertor. • Divertor targets typically use an interlayer between the armour and structure. • Engineering the properties of the interlayer can yield large gains in performance. • A response surface based design search and optimisation method is used. • A new design passes linear-elastic code rules up to applied heat flux of 18 MW/m{sup 2}. - Abstract: A robust water-cooled divertor target plate solution for DEMO has to date remained elusive. Common to all contemporary concepts is an interlayer at the boundary between the tungsten armour and the cooling structure. In this paper we show by design optimisation that an effectively designed interlayer can produce dramatic gains in power handling. By engineering the interlayer as part of the design study, it is found that divertor performance is enhanced by either a low conductivity ‘Thermal Break’ interlayer or an ‘Ultra-Compliant’ interlayer. For a 10 MW/m{sup 2} surface heat flux we find that a thermal conductivity of 15 W/mK and elastic modulus of 1 GPa are effective. A design is proposed which passes linear-elastic code rules up to an applied heat flux of 18 MW/m{sup 2}.

  19. Visible spectroscopy in the DIII-D divertor

    International Nuclear Information System (INIS)

    Brooks, N.H.; Fehling, D.; Hillis, D.L.; Klepper, C.C.; Naumenko, N.; Tugarinov, S.; Whyte, D.G.

    1996-06-01

    Spectroscopy measurements in the DIII-D divertor have been carried out with a survey spectrometer which provides simultaneous registration of the visible spectrum over the region 400--900 nm with a resolution of 0.2 nm. Broad spectral coverage is achieved through use of a fiberoptic transformer assembly to map the curved focal plane of a fast (f/3) Rowland spectrograph into a rastered format on the rectangular sensor area of a two-dimensional CCD camera. Vertical grouping of pixels during CCD readout integrates the signal intensity over the height of each spectral segment in the rastered image, minimizing readout time. For the full visible spectrum, readout time is 50 ms. Faster response time (< 10 ms) may be obtained by selecting for readout just a small number of the twenty spectral segments in the image on the CCD. Simultaneous recording of low charge states of carbon, oxygen and injected impurities has yielded information about gas recycling and impurity behavior at the divertor strike points. Transport of lithium to the divertor region during lithium pellet injection has been studied, as well as cumulative deposition of lithium on the divertor targets from pellet injection over many successive discharges

  20. Visible spectroscopy in the DIII endash D divertor

    International Nuclear Information System (INIS)

    Brooks, N.H.; Fehling, D.; Hillis, D.L.; Klepper, C.C.; Naumenko, N.N.; Tugarinov, S.N.; Whyte, D.G.

    1997-01-01

    Spectroscopy measurements in the DIII endash D divertor have been carried out with a survey spectrometer that provides simultaneous registration of the visible spectrum over the region 400 endash 900 nm with a resolution of 0.25 nm. Broad spectral coverage is achieved through the use of a fiberoptic transformer assembly to map the curved focal plane of a fast (f/3) Rowland spectrograph [Tugarinov et al., Rev. Sci. Instrum. 66, 603 (1995)] into a rastered format on the rectangular sensor area of a two-dimensional charge-coupled device (CCD) camera. Vertical grouping of pixels during CCD readout integrates the signal intensity over the height of each spectral segment in the rastered image, minimizing readout time. For the full visible spectrum, readout time is 50 ms. Faster response time (<10 ms) may be obtained by selecting for readout just a small number of the 20 spectral segments in the image on the CCD. Simultaneous recording of low charge states of carbon, oxygen, and injected impurities has yielded information about gas recycling and impurity behavior at the divertor strike points. Transport of lithium to the divertor region during lithium pellet injection has been studied, as well as cumulative deposition of lithium on the divertor targets from pellet injection over many successive discharges. copyright 1997 American Institute of Physics

  1. 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.

  2. Edge and divertor physics with reversed toroidal field in JET

    Czech Academy of Sciences Publication Activity Database

    Pitts, R. A.; Andrew, P.; Bonnin, X.; Chankin, A.V.; Corre, Y.; Corrigan, G.; Coster, D.; Ďuran, Ivan; Eich, T.; Erents, S. K.; Fundameski, W.; Huber, A.; Jachmich, S.; Kirnev, G.; Lehnen, M.; Lomas, P. J.; Loarte, A.; Matthews, G. F.; Rapp, J.; Silva, C.; Stamp, M.F.; Strachan, J.D.; Tsitrone, E.

    337-339, č. 16 (2005), s. 146-153 ISSN 0022-3115. [Plasma Surface Interactions /16./. Portland, 24.5.2005-28.5.2005] Institutional research plan: CEZ:AV0Z20430508 Keywords : SOL * Particle drifts * JET * Plasma flow * Divertor Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.414, year: 2005

  3. 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.

  4. Modeling results for a linear simulator of a divertor

    International Nuclear Information System (INIS)

    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.

    1993-01-01

    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/m 2 along the magnetic fieldlines and > 10 MW/m 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

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

    International Nuclear Information System (INIS)

    Smid, I.; Akiba, Masato; Araki, Masanori; Suzuki, Satoshi; Satoh, Kazuyoshi

    1993-07-01

    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/m 2 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. Modular He-cooled divertor for power plant application

    International Nuclear Information System (INIS)

    Diegele, Eberhard; Kruessmann, R.; Malang, S.; Norajitra, P.; Rizzi, G.

    2003-01-01

    Gas cooled divertor concepts are regarded as a suitable option for fusion power plants because of an increased thermal efficiency for power conversion systems and the use of a coolant compatible with all blanket systems. A modular helium cooled divertor concept is proposed with an improved heat transfer. The concept employs small tiles made of tungsten and brazed to a finger-like structure made of Mo-alloy (TZM). Design goal was a heat flux of at least 15 MW/m 2 and a minimum temperature of the structure of 600 deg.C. The divertor has to survive a number of cycles (100-1000) between operating temperature and room temperature even for the steady state operation assumed. Thermo-hydraulic design requirements for the concepts include to keep the pumping power below 10% of the thermal power to the divertor plates, and simultaneously achieving a heat transfer coefficient in excess of 60 kW/m 2 K. Inelastic stress analysis indicates that design allowable stress limits on primary and secondary (thermal) stresses as required by the ITER structural design criteria are met even under conservative assumptions. Finally, critical issues for future development are addressed

  7. Manufacturing and joining technologies for helium cooled divertors

    International Nuclear Information System (INIS)

    Aktaa, J.; Basuki, W.W.; Weber, T.; Norajitra, P.; Krauss, W.; Konys, J.

    2014-01-01

    Highlights: • The manufacturing and joining technologies developed at KIT for helium cooled divertors are reviewed and critically discussed. • Various technologies have been pursued and further developed aiming divertor components with very high quality and sufficient reliability. • Very promising routes have been found for which however still R and D works are necessary. • Technologies developed are also useful for other divertor and even blanket concepts, particularly those with tungsten armor. - Abstract: In the helium cooled (HC) divertor, developed at KIT for a fusion power plant, tungsten has been selected as armor as well as structural material due to its crucial properties: high melting point, very low sputtering yield, good thermal conductivity, high temperature strength, low thermal expansion and low activation. Thereby the armor tungsten is attached to the structural tungsten by thermally conductive joint. Due to the brittleness of tungsten at low temperatures its use as structural material is limited to the high temperature part of the component and a structural joint to the reduced activation ferritic martensitic steel EUROFER97 is foreseen. Hence, to realize the selected hybrid material concept reliable tungsten–steel and tungsten–tungsten joints have been developed and will be reported in this paper. In addition, the modular design of the HC divertor requires tungsten armor tiles and tungsten structural thimbles to be manufactured in high numbers with very high quality. Due to the high strength and low temperature brittleness of tungsten special manufacturing techniques need to be developed for the production of parts with no cavities inside and/or surface flaws. The main achievement in developing the respective manufacturing technologies will be presented and discussed. To achieve the objectives mentioned above various manufacturing and joining technologies are pursued. Their later applicability depends on the level of development

  8. Optimization and limitations of known DEMO divertor concepts

    International Nuclear Information System (INIS)

    Reiser, Jens; Rieth, Michael

    2012-01-01

    Highlights: ► Limitations of the materials. ► Improved H 2 O cooled divertor. ► Improved He cooled divertor. - Abstract: In this work we will introduce and discuss improvements for two types of DEMO divertors based on known designs: (i) gas cooled designs and (ii) liquid coolant concepts. In a first step, the advantages and disadvantages of gas cooling as well as the necessity of a jet impingement to increase the heat transfer coefficients will be discussed. Further discussion deals with the pros and cons of liquid coolant concepts, like for example, liquid metal or water cooling. Thereafter, we will present two rather contrary DEMO divertor concepts which are based on today's knowledge on refractory materials science, fabrication and joining technology. The first improved concept uses water flowing through steel pipes, typically made of Eurofer steel. It is well known that using Eurofer at low temperatures is critical due to its severe embrittlement under neutron irradiation. Here we make a proposal how it could be possible to use the Eurofer steel anyway: the solution could consist in a limited operation period followed by an annealing cycle at 550 °C for a few hours during any maintenance shut down phases. The second design is based on the known helium cooling concept using jet impingement. Drawbacks of the actual He-cooled divertor design are small scale parts as well as the necessary high helium inlet temperature of about 600–800 °C which leads to the question: How can we deal with such high helium temperatures? This paper shows a solution for large scale components as well as a new thermal management for the helium outlet gas that we call ‘cooling of the coolant’. Both concepts are discussed in terms of materials selection due to material limits and joining technology with a special focus on the material issue using already existing and available materials.

  9. Hydrogen molecules in the divertor of ASDEX Upgrade

    International Nuclear Information System (INIS)

    Fantz, U.; Reiter, D.; Heger, B.; Coster, D.

    2001-01-01

    In order to reduce the power load onto the target plates detached divertor conditions are often preferred. These are characterized by volume recombination, i.e. three-body and radiative recombination. Due to low T e (few eV) hydrogen molecules can penetrate into the plasma and may play a role in divertor dynamics. In particular, it was suggested, that molecules may assist the volume recombination process. The role of molecules in the divertor is examined here by a combination of experimental results with plasma edge simulations (B2-EIRENE) and a collisional-radiative model for hydrogen molecules. Spectroscopic diagnostics of the Fulcher transition carried out at the divertor of ASDEX Upgrade yield estimates of molecular hydrogen fluxes and the vibrational population in the ground state in detached and attached hydrogen plasmas. Good agreement with B2-EIRENE is achieved only if vibrational levels are treated as distinct (metastable) particles in the model and if the collisional-radiative model is applied to the electronically excited levels. On this basis the contribution of molecules to plasma recombination was determined to be in the order of a few 10%. The dominant molecular process is the dissociation process via H 2 + . As a consequence initially detached divertor plasmas can even re-attach if vibrationally resolved molecules are properly included in plasma edge models. A set of B2-EIRENE calculations carried out for ASDEX Upgrade is discussed. In particular the threshold upstream density for detachment was found to be up to a factor 1.5 higher than that originally expected due to these molecular effects. The transferability of the results to deuterium will be discussed

  10. 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.

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  12. Design and characterization of a prototype divertor viewing infrared video bolometer for NSTX-U

    Science.gov (United States)

    van Eden, G. G.; Reinke, M. L.; Peterson, B. J.; Gray, T. K.; Delgado-Aparicio, L. F.; Jaworski, M. A.; Lore, J.; Mukai, K.; Sano, R.; Pandya, S. N.; Morgan, T. W.

    2016-11-01

    The InfraRed Video Bolometer (IRVB) is a powerful tool to measure radiated power in magnetically confined plasmas due to its ability to obtain 2D images of plasma emission using a technique that is compatible with the fusion nuclear environment. A prototype IRVB has been developed and installed on NSTX-U to view the lower divertor. The IRVB is a pinhole camera which images radiation from the plasma onto a 2.5 μm thick, 9 × 7 cm2 Pt foil and monitors the resulting spatio-temporal temperature evolution using an IR camera. The power flux incident on the foil is calculated by solving the 2D+time heat diffusion equation, using the foil's calibrated thermal properties. An optimized, high frame rate IRVB, is quantitatively compared to results from a resistive bolometer on the bench using a modulated 405 nm laser beam with variable power density and square wave modulation from 0.2 Hz to 250 Hz. The design of the NSTX-U system and benchtop characterization are presented where signal-to-noise ratios are assessed using three different IR cameras: FLIR A655sc, FLIR A6751sc, and SBF-161. The sensitivity of the IRVB equipped with the SBF-161 camera is found to be high enough to measure radiation features in the NSTX-U lower divertor as estimated using SOLPS modeling. The optimized IRVB has a frame rate up to 50 Hz, high enough to distinguish radiation during edge-localized-modes (ELMs) from that between ELMs.

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

    International Nuclear Information System (INIS)

    Whyte, D.G.; West, W.P.; Wong, C.P.C.

    2001-01-01

    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/m 2 /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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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).

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  16. Study of Lithium Vapor Flow In a Detached Divertor using DSMC code

    Science.gov (United States)

    Emdee, Eric; Schwartz, Jacob; Goldston, Robert; Jaworski, Michael

    2017-10-01

    A detached divertor is predicted to be necessary to handle the heat fluxes of a demonstration fusion power plant. The lithium vapor box divertor has poloidal baffles to form distinct chambers and contains dense lithium vapor to cause detachment. These chambers would be differentially pumped via condensation, resulting in flow at Knudsen numbers 0.01-0.5 and densities 1019-1023m-3. This divertor geometry is predicted to handle the estimated heat flux while also localizing the vapor in the divertor. We provide a simulation of the divertor's lithium vapor flow using the SPARTA Direct Simulation Monte Carlo (DSMC) code. Lithium mass flow, vapor pressures, and temperatures within each chamber are given. Preliminary simulations of a vapor box divertor similarity experiment are within 30% of an ideal-gas choked nozzle flow calculation. This work supported by DOE Contract No. DE-AC02-09CH11466.

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

    International Nuclear Information System (INIS)

    Mahdavi, M.A.; Baker, D.R.; Allen, S.L.

    1994-05-01

    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 D 2 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

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

    International Nuclear Information System (INIS)

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

    1999-11-01

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

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

    DEFF Research Database (Denmark)

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

    2016-01-01

    carbon divertor. For both sets of divertors, the power threshold was found to decrease with gradual accumulation of the lithium wall coating, suggesting the important role played by the low Z impurities and/or the edge neutral density on the L-H power threshold. When operating in the upper single null......The power threshold for low (L) to high (H) confinement mode transition achieved by radio-frequency heating and molybdenum first wall with lithium coating has been experimentally investigated on the EAST tokamak for two sets of divertor geometries and materials: tungsten/carbon divertor and full...... configuration, with the ion grad-B drift direction away from the primary X-point, a lower normalized power threshold is observed in EAST with the tungsten/carbon divertor, compared to the carbon divertor after intensive lithium wall coating. A newly installed cryopump increasing the pumping efficiency also...

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

    International Nuclear Information System (INIS)

    Maggi, C.F.; Horton, L.D.; Summers, H.P.

    2000-01-01

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

  1. Studies of impurity deposition/implantation in JET divertor tiles using SIMS and ion beam techniques

    International Nuclear Information System (INIS)

    Likonen, J.; Lehto, S.; Coad, J.P.; Renvall, T.; Sajavaara, T.; Ahlgren, T.; Hole, D.E.; Matthews, G.F.; Keinonen, J.

    2003-01-01

    At the end of C4 campaign at JET, a 1% SiH 4 /99% D 2 mixture and pure 13 CH 4 were injected into the torus from the outer divertor wall and from the top of the vessel, respectively, in order to study material transport and scrape-off layer (SOL) flows. A set of MkIIGB tiles was removed during the 2001 shutdown for surface analysis. The tiles were analysed with secondary ion mass spectrometry (SIMS) and time-of-flight elastic recoil detection analysis (TOF-ERDA). 13 C was detected in the inner divertor wall tiles implying material transport from the top of the vessel. Silicon was detected mainly at the outer divertor wall tiles and very small amounts were found in the inner divertor wall tiles. Si amounts in the inner divertor wall tiles were so low that rigorous conclusions about material transport from divertor outboard to inboard cannot be made

  2. Engineering, installation, testing, and initial operation of the DIII-D Advanced Divertor

    International Nuclear Information System (INIS)

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

    1990-09-01

    The Advanced Divertor (AD) for General Atomics tokamak, DIII-D, was installed in the summer of 1990. The AD has enabled two classes of physics experiments to be run: divertor biasing and divertor baffling. Both are new experiments for DIII-D. The AD has two principal components: (1) a continuous ring electrode; and (2) a toroidally symmetric baffle. The tokamak can be run in bias baffle or standard DIII-D divertor modes by accurate positioning of the outer divertor strike point through the use of the DIII-D control system. The paper covers design, analysis, fabrication, installation, instrumentation, testing, initial operation, and future plans for the Advanced Divertor from an engineering viewpoint. 2 refs., 5 figs

  3. 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.

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  5. Alignment systems for pumped divertor installation at JET

    International Nuclear Information System (INIS)

    Macklin, B.; Celentano, G.; Israel, G.; Tait, J.; Lente, E. van; Cordier, J.J.

    1994-01-01

    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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    International Nuclear Information System (INIS)

    Pestchanyi, S.

    2009-01-01

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

  8. Thermal and structural design study of divertor collector plates

    International Nuclear Information System (INIS)

    Kitamura, Kazunori; Iida, Hiromasa; Sako, Kiyoshi

    1982-01-01

    Thermal and structural design study of divertor collector plates for a Swimming Pool Type Tokamak Reactor (SPTR) is carried out. Co-axial tube type divertor plate is employed for the reduction of electromagnetic force caused by plasma disruption. Maximum heat flux on cooling surface is sufficiently below burn-out heat flux. High thermal stress appers at the brazing region between copper cooling tube and tungsten armor. Some measures are required to decrease the thermal stress for extending the life time of the plate. These will be decreasing the heat flux on the plate by the reduction of beam angle to the plate or promoting the boiling in the tube by the reduction of coolant pressure. The life time of the plate by erosion due to ion sputtering is estimated to be about 4 years. (author)

  9. SOLPS simulations of X-divertor in NSTX-U

    Science.gov (United States)

    Chen, Zhongping; Kotschenreuther, Mike; Mahajan, Swadesh

    2017-10-01

    The X-divertor (XD) geometry in NSTX-U has demonstrated, in SOLPS simulations, a better performance than the standard divertor (SD) regarding detachment: achieving detachment with a lower upstream density and stabilizing the detachment front near the target. The benefits of such a localized front is that the power exhaust requirement can be satisfied without the radiation front encroaching on the core plasma. It is also found by our simulations that at similar states of detachment the XD outperforms the SD by reducing the heat fluxes to the target and maintaining higher upstream temperatures. These advantages are attributed to the unique geometric characteristics of XD - poloidal flaring near the target. The detailed physical mechanisms behind the better XD performance that is found in the simulations will be examined. Work supported by US DOE under DE-FG02-04ER54742 and SC 0012956.

  10. Plasma Parameters in the COMPASS Divertor During Ohmic Plasmas

    Czech Academy of Sciences Publication Activity Database

    Dimitrova, Miglena; Dejarnac, Renaud; Popov, Tsv.K.; Ivanova, P.; Kovačič, J.; Stöckel, Jan; Havlíček, Josef; Janky, Filip; Pánek, Radomír

    2014-01-01

    Roč. 54, č. 3 (2014), s. 255-260 ISSN 0863-1042. [International Workshop on Electric Probes in Magnetized Plasmas/10./. Madrid, 09.07.2013-12.07.2013] R&D Projects: GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : Langmuir divertor probes * EEDF * first-derivative technique Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.838, year: 2014 http://onlinelibrary.wiley.com/doi/10.1002/ctpp.201410073/abstract

  11. Surface heat loads on the ITER divertor vertical targets

    Czech Academy of Sciences Publication Activity Database

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

    2017-01-01

    Roč. 57, č. 4 (2017), č. článku 046025. ISSN 0029-5515 Institutional support: RVO:61389021 Keywords : ITER * divertor * ELM heat load * inter-ELM heat load * tungsten Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 3.307, year: 2016 http://iopscience.iop.org/article/10.1088/1741-4326/aa5e2a

  12. DiMES divertor erosion experiments on DIII-D

    International Nuclear Information System (INIS)

    Whyte, D.G.; Brooks, J.N.; Wong, C.P.C.; West, W.P.; Bastasz, R.; Wampler, W.R.; Rubinstein, J.

    1996-01-01

    The DiMES (Divertor Material Evaluation Studies) mechanism allows insertion of material samples to the lower divertor floor of the DIII-D tokamak. The main purpose of these studies is to measure erosion rates and redeposition mechanisms under tokamak divertor plasma conditions in order to obtain a physical understanding of the erosion/redeposition processes and to determine its implications for fusion power plant plasma facing components. Thin metal films of Be, W, V, and Mo, were deposited on a Si depth-marked graphite sample and exposed to the steady-state outer strike point on DIII-D. A variety of surface analysis techniques are used to determine the erosion/redeposition of the metals and the carbon after 5--15 seconds of exposure. These short exposure times ensure controlled exposure conditions and the extensive array of DIII-D divertor diagnostics provide a well characterized plasma for modeling efforts. Erosion rates and redeposition lengths are found to decrease with the atomic number of the metallic species, as expected. Under these conditions, the peak net erosion rate for carbon is ∼ 4 nm/s, with the erosion following the ion flux profile. Comparisons of the measured carbon erosion with REDEP code calculations show good agreement for both the absolute net erosion rate and its spatial variation. Measured erosion rates of the metals are smaller than predicted for sputtering from a bare metal surface, apparently due to effects of carbon deposition on the metal surface. Visible spectroscopic measurements of singly ionized Be have determined that the erosion process reaches steady-state during the exposure

  13. Thermal and structural analysis of the TPX divertor

    International Nuclear Information System (INIS)

    Reis, E.E.; Baxi, C.B.; Chin, E.; Redler, K.M.

    1995-01-01

    The high heat flux on the surfaces of the TPX divertor will require a design in which a carbon-carbon (C-C) tile material is brazed to water cooled copper tubes. Thermal and structural analyses were performed to assist in the design selection of a divertor tile concept and C-C material. The relevancy of finite element analysis (FEA) for evaluating tile design was examined by conducting a literature survey to compare FEA stress results to subsequent brazing and thermal test results. The thermal responses for five tile concepts and four C-C materials were analyzed for a steady-state heat flux of 7.5 MW/m 2 . Elastic-plastic stress analyses were performed to calculate the residual stresses due to brazing C-C tiles to soft copper heat sinks for the various tile designs. Monoblock and archblock divertor tile concepts were analyzed for residual stresses in which elevated temperature creep effects were included with the elastic-plastic behavior of the copper heat sink for an assumed braze cooldown cycle. As a result of these 2D studies, the archblock concept with a 3D fine weave C-C was initially found to be a preferred design for the divertor. A 3D elastic-plastic analysis for brazing of the arch block tile was performed to investigate the singularity effects at the C-C to copper interface in the direction of the tube axis. This analysis showed that the large residual stresses at the tube and tile edge intersection would produce cracks in the C-C and possible delamination along the braze interface. These results, coupled with the difficulties experienced in brazing archblocks for the Tore Supra Limiter, required that other tile designs be considered

  14. Remote bolting tools for the JET divertor exchange

    International Nuclear Information System (INIS)

    Mills, S.F.; Loving, A.B.

    1998-01-01

    With over 2500 bolting operations to perform remotely, the selection of appropriate tooling was fundamental to the efficiency of the recent JET Divertor Exchange. During mock-up trials using the Mascot force reflecting manipulator it became apparent that many traditional manual techniques were inappropriate. Hence, it was necessary to design special bolting tools to satisfy the remote handling needs. This paper discusses the evolution of this tooling. (authors)

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

    Czech Academy of Sciences Publication Activity Database

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

    2017-01-01

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

  16. Comparative studies of inner and outer divertor discharges and a fueling study in QUEST

    Energy Technology Data Exchange (ETDEWEB)

    Mitarai, O., E-mail: omitarai@ktmail.tokai-u.jp [Kumamoto Liberal Arts Education Center, Tokai University, 9-1-1 Toroku, Higashi-ku, Kumamoto 862-8652 (Japan); Nakamura, K.; Hasegawa, M.; Onchi, T.; Idei, H.; Fujisawa, A.; Hanada, K.; Zushi, H.; Higashijima, A.; Nakashima, H.; Kawasaki, S. [Research Institute for Applied Mechanics, Kyushu University, 6-1 Kasugakoen, Kasuga 816-8580 Japan (Japan); Matsuoka, K. [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292 (Japan); Koike, S.; Takahashi, T. [Division of Electronics and Informatics, Faculty of Science and Technology, Gunma University, 1-5-1 Tenjin-cho, Kiryu, Gunma 376-8515 (Japan); Tsutsui, H. [Research Laboratory for Nuclear Reactors, Tokyo Inst. Tech, 2-12-1 Ookayama, Tokyo 152-8550 (Japan)

    2016-11-01

    Highlights: • Central solenoid has a small flux in QUEST. • Large plasma current is obtained when the position is shifted to the inboard side. • Two types of divertor operation are compared. • Novel merging fueling methods are proposed. • Coaxial helicity injection (CHI) fueling was examined in QUEST divertor configuration. - Abstract: As QUEST has a small central solenoid (CS), a larger Ohmic discharge current has been obtained when the plasma shifts to the inboard side. This tendency restricts a divertor operation to the smaller plasma current regime. As the inner divertor coil has a smaller mutual inductance, it would be expected that its utilization seems to be better for easier plasma current ramp-up for a divertor operation. In this work, we made comparative studies on the plasma current ramp-up for two divertor coils. It is found that while the inner divertor coil with smaller mutual inductance needs a larger coil current, the outer divertor coil with larger mutual inductance needs a smaller coil current for divertor operation. Thus we have found that the plasma current ramp-up characteristics are almost similar for both configurations. We also propose a new fueling method for spherical tokamak (ST) using the coaxial helicity injection (CHI). The main plasma current would be generated at first, and then the CHI plasma current is created between bottom two electrode plates and merged into the main plasma current for fueling.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    Reiter, D.; May, C.; Baelmans, M.; Boerner, P.

    1997-01-01

    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.)

  19. General properties of the magnetic field in a snowflake divertor

    Science.gov (United States)

    Ryutov, D. D.; Makowski, M. A.; Umanski, M. V.

    2010-11-01

    The power-law series for the poloidal magnetic flux function, up to the third order terms, are presented for the case where two nulls of the poloidal magnetic field are separated by a small distance, as in a snowflake divertor. Distinct from the earlier results, no assumptions about the field symmetry are made. Conditions for the realization of an exact snowflake are expressed in terms of the coefficients of the power series. It is shown that, by a proper choice of the coordinate frame in the poloidal plane, one can obtain efficient similarity solutions for the separatrices and flux surfaces in the divertor region: the whole variety of flux surface shapes can be characterized by a single dimensionless parameter. Transition from a snowflake-minus to snowflake-plus configuration in the case of no particular symmetry is described. The effect of the finite toroidal current density in the divertor region is assessed. A possibility of creating a near-snowflake configuration in the ITER-scale facilities is discussed. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

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

    International Nuclear Information System (INIS)

    Haines, J.R.; Kitamura, Kazunori; Kobayashi, Takeshi; Iida, Hiromasa

    1986-07-01

    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/m 2 for the first wall up to 5.0 MW/m 2 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)

  1. Automated magnetic divertor design for optimal power exhaust

    International Nuclear Information System (INIS)

    Blommaert, Maarten

    2017-01-01

    The so-called divertor is the standard particle and power exhaust system of nuclear fusion tokamaks. In essence, the magnetic configuration hereby 'diverts' the plasma to a specific divertor structure. The design of this divertor is still a key issue to be resolved to evolve from experimental fusion tokamaks to commercial power plants. The focus of this dissertation is on one particular design requirement: avoiding excessive heat loads on the divertor structure. The divertor design process is assisted by plasma edge transport codes that simulate the plasma and neutral particle transport in the edge of the reactor. These codes are computationally extremely demanding, not in the least due to the complex collisional processes between plasma and neutrals that lead to strong radiation sinks and macroscopic heat convection near the vessel walls. One way of improving the heat exhaust is by modifying the magnetic confinement that governs the plasma flow. In this dissertation, automated design of the magnetic configuration is pursued using adjoint based optimization methods. A simple and fast perturbation model is used to compute the magnetic field in the vacuum vessel. A stable optimal design method of the nested type is then elaborated that strictly accounts for several nonlinear design constraints and code limitations. Using appropriate cost function definitions, the heat is spread more uniformly over the high-heat load plasma-facing components in a practical design example. Furthermore, practical in-parts adjoint sensitivity calculations are presented that provide a way to an efficient optimization procedure. Results are elaborated for a fictituous JET (Joint European Torus) case. The heat load is strongly reduced by exploiting an expansion of the magnetic flux towards the solid divertor structure. Subsequently, shortcomings of the perturbation model for magnetic field calculations are discussed in comparison to a free boundary equilibrium (FBE) simulation

  2. Automated magnetic divertor design for optimal power exhaust

    Energy Technology Data Exchange (ETDEWEB)

    Blommaert, Maarten

    2017-07-01

    The so-called divertor is the standard particle and power exhaust system of nuclear fusion tokamaks. In essence, the magnetic configuration hereby 'diverts' the plasma to a specific divertor structure. The design of this divertor is still a key issue to be resolved to evolve from experimental fusion tokamaks to commercial power plants. The focus of this dissertation is on one particular design requirement: avoiding excessive heat loads on the divertor structure. The divertor design process is assisted by plasma edge transport codes that simulate the plasma and neutral particle transport in the edge of the reactor. These codes are computationally extremely demanding, not in the least due to the complex collisional processes between plasma and neutrals that lead to strong radiation sinks and macroscopic heat convection near the vessel walls. One way of improving the heat exhaust is by modifying the magnetic confinement that governs the plasma flow. In this dissertation, automated design of the magnetic configuration is pursued using adjoint based optimization methods. A simple and fast perturbation model is used to compute the magnetic field in the vacuum vessel. A stable optimal design method of the nested type is then elaborated that strictly accounts for several nonlinear design constraints and code limitations. Using appropriate cost function definitions, the heat is spread more uniformly over the high-heat load plasma-facing components in a practical design example. Furthermore, practical in-parts adjoint sensitivity calculations are presented that provide a way to an efficient optimization procedure. Results are elaborated for a fictituous JET (Joint European Torus) case. The heat load is strongly reduced by exploiting an expansion of the magnetic flux towards the solid divertor structure. Subsequently, shortcomings of the perturbation model for magnetic field calculations are discussed in comparison to a free boundary equilibrium (FBE) simulation

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

    International Nuclear Information System (INIS)

    Allen, S.L.; Brooks, N.H.; Bastasz, R.

    1998-11-01

    The authors 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. They 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 D 2 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 D 2 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 n e /n gw ∼ 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

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

    NARCIS (Netherlands)

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

    2018-01-01

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

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

    International Nuclear Information System (INIS)

    Seon, C.; An, Y.; Lee, H.; Pak, S.; Cheon, M.S.; Choi, J.; Kim, H.; Hong, J.; Song, I.; Jang, J.; Lee, H.; Jeon, T.; Park, J.; Choe, W.; Kim, B.; Biel, W.; Bernascolle, P.; Barnsley, R.; O'Mullane, M.

    2017-01-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    International Nuclear Information System (INIS)

    Cohen, S.A.

    1991-12-01

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

  9. Proposal of an alternative upper divertor in ASDEX Upgrade supported by EMC3-EIRENE simulations

    Directory of Open Access Journals (Sweden)

    T. Lunt

    2017-08-01

    Full Text Available We discuss the benefits of installing a pair of in-vessel coils with currents |Ifx| ≲ 50 kAt in the upper divertor of ASDEX Upgrade (AUG to study a series of ‘alternative’ divertor configurations, like the Snowflake (SF and the X-divertor (XD, that are currently considered as alternative solutions for the power exhaust problem. The possibility of operating the standard lower single-null (SN and double-null (DN would be preserved. Potential effects to reduce the peak parallel- and/or perpendicular heat flux are predicted from a simple geometrical-diffusive model as well as by numerical EMC3-EIRENE simulations for pure deuterium attached conditions with spatially constant diffusion coefficients. Beyond that a series of other potential transport- and radiation related heat flux mitigation effects are identified and could be studied experimentally with the modified upper divertor in the high-power divertor Tokamak AUG.

  10. Recent advances towards a lithium vapor box divertor

    Directory of Open Access Journals (Sweden)

    R.J. Goldston

    2017-08-01

    Full Text Available Fusion power plants are likely to require near complete detachment of the divertor plasma from the divertor target plates, in order to have both acceptable heat flux at the target to avoid prompt damage and also acceptable plasma temperature at the target surface, to minimize long-term erosion. However hydrogenic and impurity puffing experiments show that detached operation leads easily to x-point MARFEs, impure plasmas, degradation in confinement, and lower helium pressure at the exhaust. The concept of the Lithium Vapor Box Divertor is to use local evaporation and strong differential pumping through condensation to localize low-Z gas-phase material that absorbs the plasma heat flux and so achieve detachment while avoiding these difficulties. The vapor localization has been confirmed using preliminary Navier–Stokes calculations. We use ADAS calculations of εcool, the plasma energy lost per injected lithium atom, to estimate the lithium vapor pressure, and so temperature, required for detachment, taking into account power balance. We also develop a simple model of detachment to evaluate the required upstream density, based on further taking into account dynamic pressure balance. A remarkable general result is found, not just for lithium-vapor-induced detachment, that the upstream density divided by the Greenwald-limit density scales as nup/nGW ∝ (P5/8/B3/8 Tdet1/2/(εcool+γTdet, with no explicit size scaling. Tdet is the temperature just before strong pressure loss, assumed to be ∼ ½ of the ionization potential of the dominant recycling species, and γ is the sheath heat transmission factor.

  11. Surface erosion issues and analysis for dissipative divertors

    International Nuclear Information System (INIS)

    Brooks, J.N.; Ruzic, D.N.; Hayden, D.B.; Turkot, R.B. Jr.

    1994-05-01

    Erosion/redeposition is examined for the sidewall of a dissipative divertor using coupled impurity transport, charge exchange, and sputtering codes, applied to a plasma solution for the ITER design. A key issue for this regime is possible runaway self-sputtering, due to the effect of a low boundary density and nearly parallel field geometry on redeposition parameters. Net erosion rates, assuming finite self-sputtering, vary with wall location, boundary conditions, and plasma solution, and are roughly of the following order: 200--2000 angstrom/s for beryllium, 10--100 angstrom/s for vanadium, and 0.3--3 angstrom/s for tungsten

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  13. ITER divertor, design issues and research and development

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    International Nuclear Information System (INIS)

    Brooks, J.N.

    1996-01-01

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

  16. Surface heat flux feedback controlled impurity seeding experiments with Alcator C-Mod’s high-Z vertical target plate divertor: performance, limitations and implications for fusion power reactors

    Science.gov (United States)

    Brunner, D.; Wolfe, S. M.; LaBombard, B.; Kuang, A. Q.; Lipschultz, B.; Reinke, M. L.; Hubbard, A.; Hughes, J.; Mumgaard, R. T.; Terry, J. L.; Umansky, M. V.; The Alcator C-Mod Team

    2017-08-01

    The Alcator C-Mod team has recently developed a feedback system to measure and control surface heat flux in real-time. The system uses real-time measurements of surface heat flux from surface thermocouples and a pulse-width modulated piezo valve to inject low-Z impurities (typically N2) into the private flux region. It has been used in C-Mod to mitigate peak surface heat fluxes  >40 MW m-2 down to    1. While the system works quite well under relatively steady conditions, use of it during transients has revealed important limitations on feedback control of impurity seeding in conventional vertical target plate divertors. In some cases, the system is unable to avoid plasma reattachment to the divertor plate or the formation of a confinement-damaging x-point MARFE. This is due to the small operational window for mitigated heat flux in the parameters of incident plasma heat flux, plasma density, and impurity density as well as the relatively slow response of the impurity gas injection system compared to plasma transients. Given the severe consequences for failure of such a system to operate reliably in a reactor, there is substantial risk that the conventional vertical target plate divertor will not provide an adequately controllable system in reactor-class devices. These considerations motivate the need to develop passively stable, highly compliant divertor configurations and experimental facilities that can test such possible solutions.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  19. Scoping studies for small steady-state tokamaks for divertor testing

    International Nuclear Information System (INIS)

    Galambos, J.D.; Peng, Y.K.M.; Nelson, B.E.; Hirshman, S.P.; Fogarty, P.J.

    1991-01-01

    A prime uncertainty in next-generation devices is the divertor performance. For the International Thermonuclear Experimental Reactor (ITER), the divertor limit often plays a more critical role in the operational scenario definition than do beta limit and energy confinement constraints. Hence, it is desirable to test the divertors in an environment as close as possible to that expected in next-step burning plasma experiments. Initial global scoping studies are done for small, steady-state, copper coil, beam-driven tokamaks that are dedicated to divertor testing. The usual ITER global physics models (beta limit, energy confinement, and analytic divertor heat load calculation) are incorporated, and for performance criteria we require that the divertor heat load and plasma collisionality in the edge region be similar to those expected in ITER. The smallest, lowest-cost devices satisfying these constraints tend to have major radius below 1 m, plasma current of 0.5 to 1 MA, and low aspect ratio and costs of a few tens of millions of dollars. Injection powers of about 4 to 5 MW are needed to sustain the plasma current, maintain plasma power balance, and provide the required divertor heat load. 6 refs., 5 tabs

  20. Analytical and experimental evaluation of simulated sweeping heat load on the divertor plate for ITER

    International Nuclear Information System (INIS)

    Araki, M.; Akiba, M.; Sugihara, M.; Suzuki, S.; Nishio, S.; Yokoyama, K.

    1993-01-01

    The magnitude of the heat flux on the surface of the divertor plate of ITER is one of the most limiting constraints on its lifetime. A technique for sweeping the separatrix across the divertor surface will be applied to reduce the mean surface heat fluxes and erosion damages due to intense fluxes over 15 MW/m 2 . As a first step for the evaluation of the sweeping effects, the thermal response of the divertor plate has been analyzed under the ITER relevant heat flux conditions that a peak heat flux on the divertor plate and full width at half maximum are expected to be 30 MW/m 2 and 3 cm, respectively. The analytical results show that the application of sweeping is very effective for reducing the surface temperature of the divertor plate. To realize these benefits for ITER, the divertor separatrix must be swept with a frequency of higher than 3.0 Hz over a distance of ±10 cm. Based on the analytical results, thermal response experiments with a divertor mock-up are carried out using the JAERI Electron Beam Irradiation Stand (JEBIS). The conditions for this experiment were a peak heat flux of 30 MW/m 2 with a sweeping frequency of 1.0 Hz over a distance of ±10 cm for a 30 s long cycle. Experimental results show that the divertor mock-up has successfully endured for more than 1000 major thermal cycles without an increase of the surface temperature. Therefore, it has been experimentally demonstrated that application of the sweeping technique is very effective for improvement of the power handling capability of the divertor plate. Experimental results showed a good agreement with analytical results. (orig.)

  1. 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

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

    International Nuclear Information System (INIS)

    Yoshizawa, S.; Adachi, J.; Morishita, H.; Kakudate, S.; Taguchi, H.; Tada, E.

    1995-01-01

    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.)

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  4. Control of the configuration in JT-60 lower X-point divertor plasma

    International Nuclear Information System (INIS)

    Yoshino, Ryuji

    1988-11-01

    This paper presents experimental results of the configuration-control of lower X-point divertor plasmas in JT-60. Vertical positional instability is well stabilized by the combination control of horizontal magnetic field coil current and divertor coil current, where the latter one flows just below a vacuum vessel and reinforces the control of vertical plasma position. Therefore plasma elongation of ∼1.40 with n index of -1.80 has been obtained, Where -1.80 is almost same level with -n S SD , that is the passive index obtained from the combination of horizontal magnetic coil and divertor magnetic coil. (author)

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  6. Stable sheath formation in expanding magnetic field to divertor plate

    International Nuclear Information System (INIS)

    Tomita, Y.; Takayama, A.; Takamaru, H.; Sato, T.

    2002-01-01

    The stable sheath formation and the effects of charge exchange collisions of ions with cold neutrals for the stable presheath formation in an expanding magnetic field towards a divertor plate is studied by a one-dimensional analysis. The requirement for flow velocity of ions at a plasma-sheath boundary is more restricted than that of the uniform magnetic field, which should be greater than the ion sound speed. The difference, however, between both cases is an order of the Debye length to plasma radius, which is negligibly small. The requirement for ion flow velocity inside a quasi-neutral plasma is investigated by taking into account the effects of the charge exchange collisions. Without neutrals in the quasi-neutral plasma, the ion flow velocity at an injection point should be much greater than the ion sound speed. The unisotropic velocity distribution of injected ions with coupling the expanding magnetic field and the charge exchange collisions might mitigate this requirement. (orig.)

  7. Power transport to the poloidal divertor experiment scoop limiter

    International Nuclear Information System (INIS)

    Kugel, H.W.; Budny, R.; Fonck, R.

    1987-01-01

    Power transport to the Poloidal Divertor Experiment graphite scoop limiter was measured during both ohmic- and neutral-beam-heated discharges by observing its front face temperatures using an infrared camera. Measurements were made as a function of a plasma density, current, position, fueling mode, and heating power for both co- and counter-neutral beam injection. The measured thermal load on the scoop limiter was 25 to 50%. of the total plasma heating power. The measured peak front face midplane temperature was 1500 0 C, corresponding to a peak surface power density of 3 kW/cm/sup 2/. This power density implies an effective parallel power flow of 54 kW/cm/sup 2/ in agreement with the radial power distribution extrapolated from television Thomson scattering and calorimetry measurements

  8. Stability and heating of a poloidal divertor tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Biddle, A. P.; Dexter, R. N.; Holly, D. T.; Lipschultz, B.; Osborne, T. H.; Prager, S. C.; Shepard, D.A., Sprott, J.C.; Witherspoon, F. D.

    1980-06-01

    Five experimental studies - two stability and three heating investigations - have been carried out on Tokapole II, a Tokamak with a four node poloidal divertor. First, discharges have been attained with safety factor q as low as 0.6 over most of the column without degradation of confinement, and correlation of helical instability onset with current profile shape is being studied. Second, the axisymmetric instability has been investigated in detail for various noncircular cross-sectional shapes, and results have been compared with a numerical stability code adapted to the Tokapole machine. Third, application of high power fast wave ion cyclotron resonance heating doubles the ion temperature and permits observation of heating as a function of harmonic number and spatial location of the resonance. Fourth, low power shear Alfven wave propagation is underway to test the applicability of this heating method to tokamaks. Fifth, preionization by electron cyclotron heating has been employed to reduce the startup loop voltage by approx. 60%.

  9. Ergodic divertor impact on Tore Supra plasma edge

    International Nuclear Information System (INIS)

    Grosman, A.; Ghendrih, P.; Agostini, E.; Bruneau, J.L.; Michelis, C. De; Fall, T.; Gil, C.; Guilhem, D.; Hess, W.; Hutter, T.

    1990-01-01

    Present ergodic divertor experiments in TORE SUPRA have been devoted to benchmarking the operational regimes of the apparatus. Two major effects are reported; on 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

  10. Structural evaluation of a DTHR bundle divertor particle collector

    International Nuclear Information System (INIS)

    Prevenslik, T.V.

    1980-09-01

    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

  11. Behaviour of the W7-AS island divertor

    International Nuclear Information System (INIS)

    Grigull, P.; McCormick, K.; Feng, Y.

    2003-01-01

    The W7-AS island divertor enables quasi steady state operation with NBI at very high density (HDH-mode), including regimes with stable partial detachment. Stable partial detachment is restricted to configurations with relatively large boundary magnetic islands and characterized by strong radiation from the x-point regions (radiated power fractions up to 90%), and moderate core recycling. With small islands the behaviour is more limiter-like. Above a critical density the discharges switch abruptly from attached, stable HDH regimes to unstable complete detachment with strongly reduced stored energy, drastically increased core recycling and radiation (often by MARFE formation), and subsequent collapse. The neutral dynamics as well as the energy and particle deposition on the targets in attached discharges show strong top/bottom asymmetries which invert with B-field reversal thus indicating plasma drift effects. (orig.)

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

    International Nuclear Information System (INIS)

    Skinner, C.H.; Federici, G.

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Krat, S.; Coad, J.P.; Gasparyan, Yu.; Hakola, A.; Likonen, J.; Mayer, M.; Pisarev, A.; Widdowson, A.

    2013-01-01

    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 3

  14. A snowflake divertor: a possible solution to the power exhaust problem for tokamaks

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-11-21

    This paper summarizes recent progress in the theory of a snowflake divertor, a possible path to reduce both steady-state and intermittent heat loads on the divertor plates to an acceptable level. The most important feature of a SF divertor is the presence of a large zone of a very weak poloidal magnetic field around the poloidal field (PF) null. Qualitative explanation of a variety of new features characteristic of a SF divertor is provided based on simple scaling relations. The main part of the paper is focused on the concept of spreading of the heat flux by curvature-driven convection near the PF null. References to experimental results from the NSTX and TCV tokamaks are provided.

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

    International Nuclear Information System (INIS)

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

    2006-10-01

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    DEFF Research Database (Denmark)

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

    2012-01-01

    were made between the H-mode plasmas with lower hybrid current drive (LHCD) and those with combined ion cyclotron resonance heating (ICRH). The particle and heat flux profiles between and during ELMs were obtained from Langmuir triple-probe arrays embedded in the divertor target plates. And isolated...... ELMs were chosen for analysis in order to reduce the uncertainty resulting from the influence of fast electrons on Langmuir triple-probe evaluation during ELMs. The power deposition obtained from Langmuir triple probes was consistent with that from the divertor infra-red camera during an ELM...... significantly broadening the SOL width and plasma-wetted area on the divertor target in both LHCD and LHCD + ICRH H-modes, thus posing a great challenge for the next-step high-power, long-pulse operation in EAST. Increasing the divertor-wetted area was also observed to reduce the peak heat flux and particle...

  18. Experimental testing and theoretical analysis of samples of a divertor plate proposed for NET

    International Nuclear Information System (INIS)

    Brossa, F.; Federici, G.; Renda, V.; Papa, L.

    1986-01-01

    This paper presents the JRC-Ispra effort to support the design of a divertor concept for future reactors. The reference frame used in this work, i.e. divertor geometry and wall loading, is that of the NET (Next European Torus) reactor, which constitutes the European collaboration in the fusion reactor technology Program. Because of its main function of plasma impurity control, the divertor is submitted to high thermal fluxes, severe sputtering rates and electromagnetic forces. The present proposal for the divertor plate is the following: 1) W-5Re for the armour; 2) Cu for the heat sink. This choice is due to the low sputtering rate and favourable high temperature mechanical properties of the W-5Re, and the high thermal conductivity of copper

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

    International Nuclear Information System (INIS)

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

    1997-06-01

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

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

    Directory of Open Access Journals (Sweden)

    R.A. Pitts

    2017-08-01

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

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

    Science.gov (United States)

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

    2017-12-01

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

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

    International Nuclear Information System (INIS)

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

    1995-10-01

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

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

    International Nuclear Information System (INIS)

    Willem Janssens, Ir.; Crutzen, Y.; Farfaletti-Casali, F.; Matera, R.

    1991-01-01

    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

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

    International Nuclear Information System (INIS)

    Kumagai, Akira

    1999-03-01

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

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  6. Magnetic field models and their application in optimal magnetic divertor design

    Energy Technology Data Exchange (ETDEWEB)

    Blommaert, M.; Reiter, D. [Institute of Energy and Climate Research (IEK-4), FZ Juelich GmbH, Juelich (Germany); Baelmans, M. [KU Leuven, Department of Mechanical Engineering, Leuven (Belgium); Heumann, H. [TEAM CASTOR, INRIA Sophia Antipolis (France); Marandet, Y.; Bufferand, H. [Aix-Marseille Universite, CNRS, PIIM, Marseille (France); Gauger, N.R. [TU Kaiserslautern, Chair for Scientific Computing, Kaiserslautern (Germany)

    2016-08-15

    In recent automated design studies, optimal design methods were introduced to successfully reduce the often excessive heat loads that threaten the divertor target surface. To this end, divertor coils were controlled to improve the magnetic configuration. The divertor performance was then evaluated using a plasma edge transport code and a ''vacuum approach'' for magnetic field perturbations. Recent integration of a free boundary equilibrium (FBE) solver allows to assess the validity of the vacuum approach. It is found that the absence of plasma response currents significantly limits the accuracy of the vacuum approach. Therefore, the optimal magnetic divertor design procedure is extended to incorporate full FBE solutions. The novel procedure is applied to obtain first results for the new WEST (Tungsten Environment in Steady-state Tokamak) divertor currently under construction in the Tore Supra tokamak at CEA (Commissariat a l'Energie Atomique, France). The sensitivities and the related divertor optimization paths are strongly affected by the extension of the magnetic model. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  7. Divertor plasma flow near the lower x-point in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Tsalas, M; Herrmann, A; Kallenbach, A; Mueller, H W; Neuhauser, J; Rohde, V; Tsois, N; Wischmeier, M

    2007-01-01

    A reciprocating probe in the lower divertor of ASDEX Upgrade, capable of accessing the low-field (LFS) and high-field side (HFS) scrape-off layers (SOLs) as well as the private flux region, was equipped with a Mach probe and used to measure flows in the vicinity of the lower x-point. We report on our measurements from ohmic and low-power H-mode discharges with ion B x ∇B drift towards the bottom x-point, and discuss their relevance to the current SOL/divertor flow understanding. In ohmic discharges, we present the evolution of divertor SOL and private flux flow profiles for increasing central (n e ). We show that the private flux flow is mainly directed from the HFS to the LFS at low densities. At medium-high densities the flow profile becomes more symmetric, and at very high densities the flow direction reverses on the LFS separatrix, having a LFS to HFS direction inside the private flux. We discuss the possible mechanisms that could affect divertor flows and produce such behaviour and conclude that pressure asymmetry between the two divertor legs combined with an E x B drift towards the inner divertor is a likely driving mechanism. At the HFS SOL, very large Mach numbers (typically exceeding M = 1) were observed in most cases. In low-power H-mode discharges inter-ELM flows were observed to be very similar to ohmic ones

  8. Thermal strain measurement of EAST W/Cu divertor structure using electric resistance strain gauges

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Xingli [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, 230031 (China); Science Island Branch of Graduate School, University of Science & Technology of China, Hefei, 230031 (China); Wang, Wanjing, E-mail: wjwang@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, 230031 (China); Wang, Jichao [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, 230031 (China); Wei, Ran; Sun, Zhaoxuan [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, 230031 (China); Science Island Branch of Graduate School, University of Science & Technology of China, Hefei, 230031 (China); Li, Qiang; Xie, Chunyi [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, 230031 (China); Chen, Hong-En; Wang, Kaiqiang; Wu, Lei; Chen, Zhenmao [State Key Lab for Strength and Vibration of Mechanical Structures, Xi’an Jiaotong University (China); Luo, Guang-Nan [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, 230031 (China); Science Island Branch of Graduate School, University of Science & Technology of China, Hefei, 230031 (China); Hefei Center for Physical Science and Technology, Hefei, 230022 (China); Hefei Science Center of Chinese Academy of Sciences, Hefei, 230027 (China)

    2016-12-15

    Highlights: • To understand the service behavior of W/Cu divertor, an electrical resistance strain gauge system had been introduced in a thermal strain measurement experiment. • The measurement system successfully finished the experiment and obtained valued thermal strain data. • Two thermomechanical analyses had also been carried out and compared with the measurement results. • Experiment results corresponded well to simulations and threw a light upon the failure of W/Cu divertor in the previous baking tests. - Abstract: W/Cu divertor has complex structure and faces extreme work environment in EAST Tokamak device. To measure its thermal strain shall be a valued way to understand its service behavior and then optimize its design and manufacturing process. This work presents a preliminary study on measuring thermal strain of EAST W/Cu divertor structure using electric resistance strain gauges. Eight gauges had been used in the experiment and the heating temperature had been set to 230 °C with respect to the work temperature. To realize the measuring experiment, an appropriate fixing method of gauges in divertor narrow spaces had been taken and tested, which could not only withstand high temperature but also had no damage to the divertor sample. The measurement results were that three gauges showed positive strain while other three showed negative strain after having been compensated, which corresponded to tensile stress and compressed stress respectively. Two thermomechanical simulations had also been carried out and used for comparing with the experiment.

  9. Experimental investigation of density regimes in the helical divertor at TEXTOR

    International Nuclear Information System (INIS)

    Clever, M.; Brezinsek, S.; Frerichs, H.; Lehnen, M.; Pospieszczyk, A.; Reiter, D.; Samm, U.; Schmitz, O.; Schweer, B.

    2012-01-01

    Using the capabilities of the dynamic ergodic divertor in the limiter tokamak TEXTOR, we have experimentally investigated the hydrogen recycling in a complex, three-dimensional, helical divertor structure similar to divertor structures found in stellarators. The observations were then compared with results from modelling with the three-dimensional transport code EMC3-EIRENE. The measurements showed that the recycling flux at the divertor target increases linearly with increasing plasma density, a high recycling regime is not observed. At highest plasma densities before the density limit disruption, the formation of a poloidally structured and helically inclined radiating belt, a helical divertor MARFE, is observed. The radial penetration depth of the neutral hydrogen particles (λ n ≈ 3 cm) estimated from spectroscopic measurements was found to be often larger than the varying radial extent of the scrape-off layer of the helical divertor (few mm up to 6 cm) which points to convective heat transport reducing parallel temperature gradients and inhibiting flux amplification. The detailed comparison of the experimental observations and the modelling results showed agreement in this high density behaviour confirming the absence of a high recycling regime. Also agreement in the absolute values of the calculated and measured target particle fluxes was observed. Simulations using different cross-field transport coefficients showed, that this agreement is only found above a certain level of cross-field transport (D ⊥ = 1 m 2 s −1 ). (paper)

  10. Stable sheath formation in expanding magnetic field to divertor plate

    International Nuclear Information System (INIS)

    Tomita, Y.; Takayama, A.; Takamaru, H.; Sato, T.

    2001-01-01

    The stable sheath formation in expanding magnetic field to a divertor plate was studied theoretically by one-dimensional analysis. In fusion devices the magnetic field is expanding in the direction of the plate, i.e. the magnitude of magnetic field is decreasing to the plate. In this configuration ions are accelerated to the plate due to the gradient of the magnetic field strength, so called a mirror force. The bombardment of accelerated ions to the plate may cause several severe problems to fusion plasmas, for example, release of large amount of impurities from the diverter plate. Limited research efforts have been carried out describing magnetic field effects on various potential formation and particle and heat fluxes to the divertor plate. The plasma-wall interaction in an oblique to the plate but uniform magnetic field has been studied by means of 1D-PIC numerical simulation. This analysis shows the formation of a quasi-neutral magnetic pre-sheath preceding the electrostatic Debye sheath, which scales to the ion gyroradius at the sound speed and to the incidence angle of the magnetic field. Sato clarifies this magnetic pre-sheath is attributed to the ion polarisation drift by the two dimensional kinetic analysis. None of effects, however, of non-uniformity of the magnetic field has been taken into account on the stable electrostatic potential and sheath formation. In this paper, we consider a collisionless sheath model between an infinite metal plate and a quasi-neutral plasma in the expanding magnetic field to the plate. One dimensional kinetic analysis leads that a condition for flow velocity of ions at a plasma-sheath boundary is more restricted than that of the uniform magnetic field, which should be larger than the ion sound speed. The difference, however, between both cases is an order of the Debye length to a plasma radius, which is negligible small. The requirement for the ion flow velocity inside the plasma is obtained from the condition of the quasi

  11. Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices

    Science.gov (United States)

    Guo, H. Y.; Hill, D. N.; Leonard, A. W.; Allen, S. L.; Stangeby, P. C.; Thomas, D.; Unterberg, E. A.; Abrams, T.; Boedo, J.; Briesemeister, A. R.; Buchenauer, D.; Bykov, I.; Canik, J. M.; Chrobak, C.; Covele, B.; Ding, R.; Doerner, R.; Donovan, D.; Du, H.; Elder, D.; Eldon, D.; Lasa, A.; Groth, M.; Guterl, J.; Jarvinen, A.; Hinson, E.; Kolemen, E.; Lasnier, C. J.; Lore, J.; Makowski, M. A.; McLean, A.; Meyer, B.; Moser, A. L.; Nygren, R.; Owen, L.; Petrie, T. W.; Porter, G. D.; Rognlien, T. D.; Rudakov, D.; Sang, C. F.; Samuell, C.; Si, H.; Schmitz, O.; Sontag, A.; Soukhanovskii, V.; Wampler, W.; Wang, H.; Watkins, J. G.

    2016-12-01

    A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). This paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.

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

    International Nuclear Information System (INIS)

    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 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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  14. Diagnostics for Evaluating Performance of NSTX Liquid Lihium Divertor

    Science.gov (United States)

    Kaita, R.; Kugel, H.; Kallman, J.; Leblanc, B.; Paul, S.; Roquemore, A. L.; Skinner, C.; Soukhanovskii, V.; Maingi, R.; Ahn, J.-W.; Wilgen, J.; Allain, J.-P.; Taylor, C.

    2009-11-01

    A Liquid Lithium Divertor (LLD) is being installed on NSTX to investigate particle control and power handling with liquid lithium as plasma-facing component (PFC). The LLD is expected to provide a low-recycling plasma-facing component (PFC). To study the effects of such a PFC on plasma performance, a variety of edge measurements are required. Since its surface is highly reflective at visible wavelengths, a Lyman-alpha detector array will be used to monitor the recycling. To understand changes in edge transport, electron temperature and density measurements will be made with Langmuir probes mounted in PFC's near the LLD, and the edge sightlines of a multipoint Thomson scattering system. A frequency-scanning reflectometer will also provide scrapeoff layer electron density profiles. The LLD response to heat loads will be examined with infrared cameras and thermocouples. Diagnostics are also needed to measure the erosion and codeposition of lithium. They include quartz deposition monitors and a retractable probe for exposing samples to the plasma.

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  16. Effects of ELMs on ITER divertor armour materials

    Science.gov (United States)

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

    2007-06-01

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

  17. Technologies for ITER divertor vertical target plasma facing components

    International Nuclear Information System (INIS)

    Schlosser, J.; Escourbiac, F.; Merola, M.; Fouquet, S.; Bayetti, P.; Cordier, J.J.; Grosman, A.; Missirlian, M.; Tivey, R.; Roedig, M.

    2005-01-01

    The ITER divertor vertical target has to sustain heat fluxes up to 20 MW m -2 . The concept developed for this plasma facing component working at steady state is based on carbon fibre composite armour for the lower straight part and tungsten for the curved upper part. The main challenges involved in the use of such components include the removal of the high heat fluxes deposited and mechanically and thermally joining the armour to the metallic heat sink, despite the mismatch in the thermal expansions. Two solutions based on the use of a CuCrZr hardened copper alloy and an active metal casting (AMC (registered) ) process were investigated during the ITER EDA phase: the first one called 'flat tile geometry' was mainly developed for the Tore Supra pumped limiter, the second one called 'monoblock geometry' was developed by the EU Participating Team for the ITER project. This paper presents a review of these two solutions and analyses their assets and drawbacks: pressure drop, critical heat flux, surface temperature and expected behaviour during operation, risks during the manufacture, control of the armour defects during the manufacture and at the reception, and the possibility of repairing defective tiles

  18. DITE Mk2 bundle divertor power supplies and control system

    International Nuclear Information System (INIS)

    Bayes, D.V.; Bell, D.; Gray, J.W.; Browning, J.L.; Poole, E.G.

    1981-01-01

    The power supplies and control system currently being constructed for the DITE Tokamak MK 2 Bundle Divertor Field System meet the following specifications: maximum current 20 kA; rise time 25 ms; flat top time 200 ms; load parameters 2.0 mH, 21 mΩ. This is accomplished by a 2.5 kV, 800 kJ electrolytic 'start' bank of dimensions 2.9 m x 1.8 m x 4.4 m high, and a 1080 V, 2.8 MJ electrolytic 'sustain' bank of dimensions 3.4 m x 1.8 m x 5.7 m high. The start bank is discharged into the load to establish the current and the 8 'sustain' bank sections are fired sequentially to maintain the 'flat top'. The power supply and load assembly is controlled by a microprocessor based Programmable Logic System, which is a logical progression from the modular hard-wired, DITE system using solid state elements. (author)

  19. Advances in optical thermometry for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-01-15

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

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

    International Nuclear Information System (INIS)

    Lorenzetto, P.; Bednarek, M.; Gavila, P.; Riccardi, B.; Saibene, G.; Escourbiac, F.; Hirai, T.; Merola, M.; Pitts, R.; Suzuki, S.; Mazul, I.

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

  2. Design of the Wendelstein 7-X inertially cooled Test Divertor Unit Scraper Element

    Energy Technology Data Exchange (ETDEWEB)

    Lumsdaine, Arnold, E-mail: lumsdainea@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Boscary, Jean [Max Planck Institute for Plasma Physics, Garching (Germany); Fellinger, Joris [Max Planck Institute for Plasma Physics, Greifswald (Germany); Harris, Jeff [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Hölbe, Hauke; König, Ralf [Max Planck Institute for Plasma Physics, Greifswald (Germany); Lore, Jeremy; McGinnis, Dean [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Neilson, Hutch; Titus, Peter [Princeton Plasma Physics Lab, Princeton, NJ (United States); Tretter, Jörg [Max Planck Institute for Plasma Physics, Garching (Germany)

    2015-10-15

    Highlights: • The justification for the installation of the Test Divertor Unit Scraper Element is given. • Specially designed operational scenarios for the component are presented. • Plans for the design of the component are detailed. - Abstract: The Wendelstein 7-X stellarator is scheduled to begin operation in 2015, and to achieve full power steady-state operation in 2019. Computational simulations have indicated that for certain plasma configurations in the steady-state operation, the ends of the divertor targets may receive heat fluxes beyond their qualified technological limit. To address this issue, a high heat-flux “scraper element” (HHF-SE) has been designed that can protect the sensitive divertor target region. The surface profile of the HHF-SE has been carefully designed to meet challenging engineering requirements and severe spatial limitations through an iterative process involving physics simulations, engineering analysis, and computer aided design rendering. The desire to examine how the scraper element interacts with the plasma, both in terms of how it protects the divertor, and how it affects the neutral pumping efficiency, has led to the consideration of installing an inertially cooled version during the short pulse operation phase. This Test Divertor Unit Scraper Element (TDU-SE) would replicate the surface profile of the HHF-SE. The design and instrumentation of this component must be completed carefully in order to satisfy the requirements of the machine operation, as well as to support the possible installation of the HHF-SE for steady-state operation.

  3. 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.

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  6. Effect of the magnetic topology of a tokamak divertor on the power exhaust properties

    Science.gov (United States)

    Pericoli Ridolfini, V.; Ambrosino, R.; Calabrò, G.; Crisanti, F.; Lombroni, R.; Mastrostefano, S.; Rubino, G.; Zagórski, R.

    2017-08-01

    The peculiarities of various advanced divertor magnetic configurations that could be adopted for a tokamak reactor are investigated with the 2D edge code TECXY applied to the different divertor options of the projected tokamak DTT (Divertor Test Tokamak). The analysis highlights very interesting features for those configurations that realize a wide region with significantly depressed poloidal field in between the main X point and the target. Here, the energy cross-field diffusion can become so fast to extend up to ≈10 times the width of the power flow channel, in terms of the poloidal flux coordinates. This can spread the power over a long length and then drop the peak heat load below the technologically safe value, even with no help from impurities. Furthermore, the strongly enlarged effective divertor volume can favour the dissipative processes and lead to plasma detachment from the associated target. The driving mechanism appears to rest on the strongly increased connection lengths. This reduces the parallel thermal gradient and then slows down the power streaming, hence forcing the flow channel to widen in order to convey the same amount of power. However, the other target can be significantly penalized by an unbalance in the power sharing between the two divertor plates. Similarly, modifying the topology of this region also could overcome this problem.

  7. 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.

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  9. 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

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

    Science.gov (United States)

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

    2017-12-01

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

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

    International Nuclear Information System (INIS)

    Ou Jing; Gan Chunyun; Ye Lei

    2014-01-01

    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)

  12. Method development for detecting divertor failures during steady state operation of Wendelstein 7X

    Energy Technology Data Exchange (ETDEWEB)

    Rodatos, Alexander; Jakubowski, Marcin; Sunn Pedersen, Thomas [Max Planck Institute for Plasma Physics, Wendelsteinstr. 1, Greifswald (Germany); Greuner, Henri [Max Planck Institute for Plasma Physics, Boltzmannstr. 2, Garching (Germany)

    2015-05-01

    Wendelstein 7-X (W7-X) is stellarator fusion experiment, which will start operation in 2015. One of its goals is the demonstration of the stellarator concepts steady state capability while operating with fusion relevant plasma parameters. For particle and heat exhaust from the plasma a set of 10 island divertor units is installed in the machine. During the steady state operation they are exposed to a heat flux of up to 10MW/m{sup 2} for up to 30 min. Transient, even higher heat fluxes are possible. To guarantee the save operation of W7-X a continues surveillance of the divertor is mandatory, which is realized by a set of 10 infrared cameras observing the divertor surface. These data needs to be evaluated during the experiment identifying defects, surface layers and actual hot spots caused by overheating.

  13. Definition of acceptance criteria for the ITER divertor plasma-facing components through systematic experimental analysis

    Energy Technology Data Exchange (ETDEWEB)

    Escourbiac, F; Richou, M; Guigon, R; Durocher, A; Schlosser, J; Grosman, A [CEA/IRFM, F-13108, Saint-Paul-lez-Durance (France); Constans, S [AREVA-NP, Le Creusot (France); Merola, M [ITER Organization, Cadarache (France); Riccardi, B [Fusion For Energy, Barcelona (Spain)], E-mail: frederic.escourbiac@cea.fr

    2009-12-15

    Experience has shown that a critical part of the high-heat flux (HHF) plasma-facing component (PFC) is the armour to heat sink bond. An experimental study was performed in order to define acceptance criteria with regards to thermal hydraulics and fatigue performance of the International Thermonuclear Experimental Reactor (ITER) divertor PFCs. This study, which includes the manufacturing of samples with calibrated artificial defects relevant to the divertor design, is reported in this paper. In particular, it was concluded that defects detectable with non-destructive examination (NDE) techniques appeared to be acceptable during HHF experiments relevant to heat fluxes expected in the ITER divertor. On the basis of these results, a set of acceptance criteria was proposed and applied to the European vertical target medium-size qualification prototype: 98% of the inspected carbon fibre composite (CFC) monoblocks and 100% of tungsten (W) monoblock and flat tiles elements (i.e. 80% of the full units) were declared acceptable.

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

    Bak, J.G.; Kim, H.S.; Bae, M.K.; Juhn, J.W.; Seo, D.C.; Bang, E.N.; Shim, S.B.; Chung, K.S.; Lee, H.J.; Hong, S.H.

    2015-01-01

    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

  16. Tungsten divertor erosion in all metal devices: Lessons from the ITER like wall of JET

    Energy Technology Data Exchange (ETDEWEB)

    Rooij, G.J. van, E-mail: g.j.vanrooij@differ.nl [Dutch Institute For Fundamental Energy Research, Association EURATOM-FOM (Netherlands); Coenen, J.W. [Institute of Energy and Climate Research, Forschungszentrum Jülich, Association EURATOM-FZJ, Jülich (Germany); Aho-Mantila, L. [VTT, P.O. Box 1000, FI-02044 VTT (Finland); Brezinsek, S.; Clever, M. [Institute of Energy and Climate Research, Forschungszentrum Jülich, Association EURATOM-FZJ, Jülich (Germany); Dux, R. [Max-Planck-Institut für Plasmaphysik, Association EURATOM-IPP (Germany); Groth, M. [Aalto University, Association EURATOM-Tekes, Espoo (Finland); Krieger, K.; Marsen, S. [Max-Planck-Institut für Plasmaphysik, Association EURATOM-IPP (Germany); Matthews, G.F.; Meigs, A. [Culham Centre for Fusion Energy, EURATOM-CCFE Association, Abingdon (United Kingdom); Neu, R.; Potzel, S.; Pütterich, T. [Max-Planck-Institut für Plasmaphysik, Association EURATOM-IPP (Germany); Rapp, J. [Oak Ridge National Laboratory, Oak Ridge (United States); Stamp, M.F. [Culham Centre for Fusion Energy, EURATOM-CCFE Association, Abingdon (United Kingdom)

    2013-07-15

    Tungsten erosion in the outer divertor of the JET ITER like wall was quantified by spectroscopy. Effective sputtering yields of typically 10{sup −4} were measured in L-mode at ∼30 eV attached divertor conditions and beryllium was identified as the main cause of sputtering. The signature of prompt redeposition was observed in the analysis of WI 400.9 nm and WII 364 nm line ratios and indicative of >50% redeposition fractions. Inter- and intra-ELM sputtering were compared for an example of 10 Hz ELMs with 13 MW NBI heating, in which intra-ELM sputtering was found to dominate by a factor of 5. Nitrogen seeding initially increased the tungsten sputtering threefold due to higher extrinsic impurity levels and effectively reduced the tungsten sputtering when the divertor plasma temperature was decreased from the initial 25 eV down to 15 eV.

  17. D III-D divertor target heat flux measurements during Ohmic and neutral beam heating

    International Nuclear Information System (INIS)

    Hill, D.N.; Petrie, T.; Mahdavi, M.A.; Lao, L.; Howl, W.

    1988-01-01

    Time resolved power deposition profiles on the D III-D divertor target plates have been measured for Ohmic and neutral beam injection heated plasmas using fast response infrared thermography (τ ≤ 150 μs). Giant Edge Localized Modes have been observed which punctuate quiescent periods of good H-mode confinement and deposit more than 5% of the stored energy of the core plasma on the divertor armour tiles on millisecond time-scales. The heat pulse associated with these events arrives approximately 0.5 ms earlier on the outer leg of the divertor relative to the inner leg. The measured power deposition profiles are displaced relative to the separatrix intercepts on the target plates, and the peak heat fluxes are a function of core plasma density. (author). Letter-to-the-editor. 11 refs, 7 figs

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  19. Comparison between a pumped-limiter and a divertor for the next step machines

    International Nuclear Information System (INIS)

    Harrison, F.F.A.

    1985-01-01

    The paper presents a simple description of the physics issues which influence the conceptual design of a pumped-limiter and single-null poloidal divertor in a next step, long burn tokamak of NET/INTOR scale. Predicted performance of the limiter and divertor are compared in regard to localised recycling, sputtering of the plasma collection surfaces, penetration of sputtered impurities into the fusion plasma, surface power loading and exhaust of helium ash. It is concluded that the performance of the divertor is superior and that it can be predicted with a reasonable degree of confidence. The viability of the limiter remains in doubt but the concept cannot be rejected at the present time

  20. Numerical modeling and validation of helium jet impingement cooling of high heat flux divertor components

    International Nuclear Information System (INIS)

    Koncar, Bostjan; Simonovski, Igor; Norajitra, Prachai

    2009-01-01

    Numerical analyses of jet impingement cooling presented in this paper were performed as a part of helium-cooled divertor studies for post-ITER generation of fusion reactors. The cooling ability of divertor cooled by multiple helium jets was analysed. Thermal-hydraulic characteristics and temperature distributions in the solid structures were predicted for the reference geometry of one cooling finger. To assess numerical errors, different meshes (hexagonal, tetra, tetra-prism) and discretisation schemes were used. The temperatures in the solid structures decrease with finer mesh and higher order discretisation and converge towards finite values. Numerical simulations were validated against high heat flux experiments, performed at Efremov Institute, St. Petersburg. The predicted design parameters show reasonable agreement with measured data. The calculated maximum thimble temperature was below the tile-thimble brazing temperature, indicating good heat removal capability of reference divertor design. (author)

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

    International Nuclear Information System (INIS)

    D'Agata, E.; Tivey, R.

    2005-01-01

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

  2. Boundary layer physics considerations for the TEXTOR dynamic ergodic divertor experiment

    Energy Technology Data Exchange (ETDEWEB)

    Evans, T.E. [General Atomics, San Diego, CA (United States)

    1997-09-01

    The proposed arrangement of the TEXTOR dynamic ergodic divertor coils is quite different from that of previous stochastic boundary layer experiments. These coils are located on the high field side and have continuous helical windings. The basic resonant magnetic field line structure is not substantially affected by the location of the perturbation coils but the non-resonant near field effects, such as those related to helical divertor effects and the interaction of the near field with the stochastic layer, are different. Differences with the TORE SUPRA ergodic divertor coils are discussed. Based on these differences we expect that the plasma positioning will play an important role in: (i) the onset of MARFE`s; (ii) the impurity screening efficiency; (iii) the onset of disruptions; and (iii) the transport properties of the boundary layer plasma. (orig.) 17 refs.

  3. C transport studies in L-mode divertor plasmas on DIII-D.

    Energy Technology Data Exchange (ETDEWEB)

    Brooks, N. (General Atomics, San Diego, CA); Nagy A. (Princeton Plasma Physics Laboratory, Princeton NJ); McLean, A. G. (University of Toronto Institute for Aerospace Studies, Toronto, Canada); Whyte, D. G. (University of Wisconsin, Madison, WI); Rudakov, D. L. (University of California, San Diego, La Jolla, CA); Bozek, A. (General Atomics, San Diego, CA); Allen, S. L. (Lawrence Livermore National Laboratory, Livermore, CA); West, W.P. (General Atomics, San Diego, CA); Stangeby, P. C. (University of Toronto Institute for Aerospace Studies, Toronto, Canada); Wampler, William R.; Matthews, G. F. (Euratom/UKAEA Fusion Association, Culham Science Center, Abingdon, UK); Phillips, V. (FZJ Julich GmbH/Euratom Institut fur Plasmaphysik, Julich, Germany); Ellis, R. (Lawrence Livermore National Laboratory, Livermore, CA)

    2004-06-01

    {sup 13}CH{sub 4} was injected with a toroidally-symmetric gas system into 22 identical lower-single-null L-mode discharges on DIII-D. The injection level was adjusted so that it did not significantly perturb the core or divertor plasmas, with a duration of {approx}3 s on each shot, for a total of {approx}300 T L of injected particles. The plasma shape remained very constant; the divertor strike points were controlled to {approx}1 cm at the divertor plate. At the beginning of the subsequent machine vent, 29 carbon tiles were removed for nuclear reaction analysis of {sup 13}C content to determine regions of carbon deposition. It was found that only the tiles inboard of the inner strike point had appreciable {sup 13}C above background. Visible spectroscopy measurements of the carbon injection and comparisons with modeling are consistent with carbon transport by means of scrape-off layer flow.

  4. {sup 13}C transport studies in L-mode divertor plasmas on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Allen, S.L. [Lawrence Livermore National Laboratory, P.O. Box 808, 7000 East Avenue, L-637, Livermore, CA 94550 (United States)]. E-mail: allens@fusion.gat.com; Wampler, W.R. [Sandia National Laboratories, Albuquerque, NM 87185-1129 (United States); McLean, A.G. [University of Toronto Institute for Aerospace Studies, Toronto, Canada MH3 5T6 (Canada); Whyte, D.G. [University of Wisconsin, Madison, WI 53706 (United States); West, W.P. [General Atomics, San Diego, CA 92186-5608 (United States); Stangeby, P.C. [University of Toronto Institute for Aerospace Studies, Toronto, Canada MH3 5T6 (Canada); Brooks, N.H. [General Atomics, San Diego, CA 92186-5608 (United States); Rudakov, D.L. [University of California, San Diego, La Jolla, CA 92093-0417 (United States); Phillips, V. [FZJ Juelich GmbH/EURATOM Institut fuer Plasmaphysik, TEC, D-52425 Juelich (Germany); Rubel, M. [Alfven Laboratory, Royal Institute of Technology, Association EURATOM-VR, Stockholm (Sweden); Matthews, G.F. [EURATOM/UKAEA Fusion Association, Culham Science Center, 0X14 3DB Abingdon (United Kingdom); Nagy, A. [Princeton Plasma Physics Laboratory, Princeton NJ 08543-0451 (United States); Ellis, R. [Lawrence Livermore National Laboratory, P.O. Box 808, 7000 East Avenue, L-637, Livermore, CA 94550 (United States); Bozek, A.S. [General Atomics, San Diego, CA 92186-5608 (United States)

    2005-03-01

    {sup 13}CH{sub 4} was injected with a toroidally-symmetric gas system into 22 identical lower-single-null L-mode discharges on DIII-D. The injection level was adjusted so that it did not significantly perturb the core or divertor plasmas, with a duration of {approx}3 s on each shot, for a total of {approx}300 T L of injected particles. The plasma shape remained very constant; the divertor strike points were controlled to {approx}1 cm at the divertor plate. At the beginning of the subsequent machine vent, 29 carbon tiles were removed for nuclear reaction analysis of {sup 13}C content to determine regions of carbon deposition. It was found that only the tiles inboard of the inner strike point had appreciable {sup 13}C above background. Visible spectroscopy measurements of the carbon injection and comparisons with modeling are consistent with carbon transport by means of scrape-off layer flow.

  5. Definition of acceptance criteria for the ITER divertor plasma-facing components through systematic experimental analysis

    Science.gov (United States)

    Escourbiac, F.; Richou, M.; Guigon, R.; Constans, S.; Durocher, A.; Merola, M.; Schlosser, J.; Riccardi, B.; Grosman, A.

    2009-12-01

    Experience has shown that a critical part of the high-heat flux (HHF) plasma-facing component (PFC) is the armour to heat sink bond. An experimental study was performed in order to define acceptance criteria with regards to thermal hydraulics and fatigue performance of the International Thermonuclear Experimental Reactor (ITER) divertor PFCs. This study, which includes the manufacturing of samples with calibrated artificial defects relevant to the divertor design, is reported in this paper. In particular, it was concluded that defects detectable with non-destructive examination (NDE) techniques appeared to be acceptable during HHF experiments relevant to heat fluxes expected in the ITER divertor. On the basis of these results, a set of acceptance criteria was proposed and applied to the European vertical target medium-size qualification prototype: 98% of the inspected carbon fibre composite (CFC) monoblocks and 100% of tungsten (W) monoblock and flat tiles elements (i.e. 80% of the full units) were declared acceptable.

  6. Definition of acceptance criteria for the ITER divertor plasma-facing components through systematic experimental analysis

    International Nuclear Information System (INIS)

    Escourbiac, F; Richou, M; Guigon, R; Durocher, A; Schlosser, J; Grosman, A; Constans, S; Merola, M; Riccardi, B

    2009-01-01

    Experience has shown that a critical part of the high-heat flux (HHF) plasma-facing component (PFC) is the armour to heat sink bond. An experimental study was performed in order to define acceptance criteria with regards to thermal hydraulics and fatigue performance of the International Thermonuclear Experimental Reactor (ITER) divertor PFCs. This study, which includes the manufacturing of samples with calibrated artificial defects relevant to the divertor design, is reported in this paper. In particular, it was concluded that defects detectable with non-destructive examination (NDE) techniques appeared to be acceptable during HHF experiments relevant to heat fluxes expected in the ITER divertor. On the basis of these results, a set of acceptance criteria was proposed and applied to the European vertical target medium-size qualification prototype: 98% of the inspected carbon fibre composite (CFC) monoblocks and 100% of tungsten (W) monoblock and flat tiles elements (i.e. 80% of the full units) were declared acceptable.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    International Nuclear Information System (INIS)

    Nagasaki, Kazunobu; Itoh, Kimitaka

    1990-02-01

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

  9. The design and development of divertor remote handling equipment for ITER

    International Nuclear Information System (INIS)

    Palmer, J.; Irving, M.; Jaervenpaeae, J.; Maekinen, H.; Saarinen, H.; Siuko, M.; Timperi, A.; Verho, S.

    2007-01-01

    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. ITER divertor replacement is classified as a remote handling (RH) Class 1 activity and as such, detailed design of the associated equipment and verification of its operation before ITER construction by way of prototypes and mock-ups, is considered an essential activity. With this in mind, a major step in the EU RH development programme for ITER involves the construction of a full-scale physical test facility in which to verify and refine divertor RH equipment designs through the operation of prototypes closely replicating those proposed for ITER. This paper reports on the design of one such prototype, namely the cassette multifunctional mover (CMM), and outlines the current build status and planning for the new RH mock-up facility in which to test this device

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

    International Nuclear Information System (INIS)

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

    1997-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    You, J.H., E-mail: you@ipp.mpg.de [Max Planck Institute for Plasma Physics, Boltzmann Str. 2, 85748 Garching (Germany); Mazzone, G.; Visca, E. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Bachmann, Ch. [EUROfusion PMU, c/o IPP, Boltzmann Str. 2, 85748 Garching (Germany); Autissier, E. [CEA, IRFM, F-13108 Saint Paul Lez Durance (France); Barrett, T. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Cocilovo, V.; Crescenzi, F. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Domalapally, P.K. [Research Cnter Rez, Hlavní 130, 250 68 Husinec–Řež (Czech Republic); Dongiovanni, D. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Entler, S. [Institute of Plasma Physics CAS, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Federici, G. [EUROfusion PMU, c/o IPP, Boltzmann Str. 2, 85748 Garching (Germany); Frosi, P. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Fursdon, M. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Greuner, H. [Max Planck Institute for Plasma Physics, Boltzmann Str. 2, 85748 Garching (Germany); Hancock, D. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Marzullo, D. [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); McIntosh, S. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Müller, A.V. [Max Planck Institute for Plasma Physics, Boltzmann Str. 2, 85748 Garching (Germany); Porfiri, M.T. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); and others

    2016-11-01

    Highlights: • A brief overview is given on the overall R&D activities of the work package Divertor which is a project of the EUROfusion Consortium. • The rationale of the hydraulic, thermal and structural design scheme is described. • The first results obtained for the preliminary DEMO divertor cassette model are presented. - Abstract: In the European fusion roadmap, reliable power handling has been defined as one of the most critical challenges for realizing a commercially viable fusion power. In this context, the divertor is the key in-vessel component, as it is responsible for power exhaust and impurity removal for which divertor target is subjected to very high heat flux loads. To this end, an integrated R&D project was launched in the EUROfusion Consortium in order to deliver a holistic conceptual design solution together with the core technologies for the entire divertor system of a DEMO reactor. The work package ‘Divertor’ consists of two project areas: ‘Cassette design and integration’ and ‘Target development’. The essential mission of the project is to develop and verify advanced design concepts and the required technologies for a divertor system being capable of meeting the physical and system requirements defined for the next-generation European DEMO reactor. In this contribution, a brief overview is presented of the works from the first project year (2014). Focus is put on the loads specification, design boundary conditions, materials requirements, design approaches, and R&D strategy. Initial ideas and first estimates are presented.

  12. Changes in divertor conditions in response to changing core density with RMPs

    Science.gov (United States)

    Briesemeister, A. R.; Ahn, J.-W.; Canik, J. M.; Fenstermacher, M. E.; Frerichs, H.; Lasnier, C. J.; Lore, J. D.; Leonard, A. W.; Makowski, M. A.; McLean, A. G.; Meyer, W. H.; Schmitz, O.; Shafer, M. W.; Unterberg, E. A.; Wang, H. Q.; Watkins, J. G.

    2017-07-01

    The effects of changes in core density on divertor electron temperature, density and heat flux when resonant magnetic perturbations (RMPs) are applied are presented, notably a reduction in RMP induced secondary radial peaks in the electron temperature profile at the target plate is observed when the core density is increased, which is consistent with modeling. RMPs is used here to indicate non-axisymmetric magnetic field perturbations, created using in-vessel control coils, which have at least one but typically many resonances with the rotational transform of the plasma (Evans et al 2006 Phys. Plasmas 13 056121). RMPs are found to alter inter-ELM heat flux to the divertor by modifying the core plasma density. It is shown that applying RMPs reduces the core density and increases the inter-ELM heat flux to both the inner and outer targets. Using gas puffing to return the core density to the pre-RMP levels more than eliminates the increase in inter-ELM heat flux, but a broadening of the heat flux to the outer target remains. These measurements were made at a single toroidal location, but the peak in the heat flux profile was found near the outer strike point where simulations indicate little toroidal variation should exist and tangentially viewing diagnostics showed no evidence of strong asymmetries. In experiments where divertor Thomson scattering measurements were available it is shown that local secondary peaks in the divertor electron temperature profile near the target plate are reduced as the core density is increased, while peaks in the divertor electron density profile near the target are increased. These trends observed in the divertor electron temperature and density are qualitatively reproduced by scanning the upstream density in EMC3-Eirene modeling. Measurements are presented showing that higher densities are needed to induce detachment of the outer strike point in a case where an increase in electron temperature, likely due to a change in MHD activity

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  15. JET ICRH antenna for pumped-divertor geometry

    International Nuclear Information System (INIS)

    Bhatnagar, V.P.; Jacquinot, J.; Kaye, A.

    1989-01-01

    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)

  16. Development of divertor tungsten coatings for the JET ITER-like wall

    Energy Technology Data Exchange (ETDEWEB)

    Matthews, G.F., E-mail: guy.matthews@jet.u [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon (United Kingdom); Coad, P. [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon (United Kingdom); Greuner, H. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Hill, M. [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon (United Kingdom); Hirai, T. [Forschungszentrum Juelich, Euratom Association, Juelich (Germany); Likonen, J. [Association Euratom-TEKES, VTT (Finland); Maier, H.; Mayer, M.; Neu, R. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Philipps, V. [Forschungszentrum Juelich, Euratom Association, Juelich (Germany); Pitts, R. [EPFL SP CRPP Batiment PPB, Station 13 CH-1015 Lausanne (Switzerland); Riccardo, V. [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon (United Kingdom)

    2009-06-15

    The main objectives of the JET ITER-like Wall Project are to provide a beryllium main wall and tungsten divertor with at least a 4 year lifetime to allow full evaluation of the materials and related plasma scenarios for ITER. Tungsten coatings will be used over most of the divertor area and this paper describes the latest developments in the coating technology and an analysis of the implications for the coating lifetime and machine operation. Both steady state and transient heat loads are assessed.

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  18. Energy system for the generation of divertor magnetic fields in the PDX fusion research device

    International Nuclear Information System (INIS)

    Turitzin, N.M.

    1976-05-01

    One of the major problems encountered in the development of Tokamak type fusion reactors is the presence of impurities in the plasma. The PDX device is designed to study the operation of poloidal magnetic field divertors and consequent magnetic limiters for controlling and reducing the amount of impurities. A system of coils placed at specific locations produces a required field configuration for the poloidal divertor. This paper describes the system of energy supplies required and the interrelations of field coil currents during plasma current initiation, growth and steady state

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

  20. Impurity transport in the divertor of the Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    McCracken, G.M.; Bombarda, F.; Graf, M.; Goetz, J.A.; Jablonski, D.; Kurz, C.; Labombard, B.; Lipschultz, B.; Rice, J.; Welch, B.

    1995-01-01

    Argon has been injected into the private flux zone of the Alcator C-Mod divertor and its transport into the confined plasma studied. The injected gas pulse lasts for about 200 ms, but the impurity concentration increases to an equilibrium level and remains there for similar 200 ms. It is clear that the argon is acting as a recycling impurity. Nevertheless it has been observed that less than 0.5% of the injected atoms reach the confined plasma and that the percentage decreases with increasing plasma density. The good impurity screening is discussed in terms of the ionisation mean free path in the divertor. ((orig.))

  1. Status of R and D of the plasma facing components for the ITER divertor

    International Nuclear Information System (INIS)

    Mazul, I.V.; Akiba, M.; Arkhipov, I.

    2001-01-01

    The paper reports the progress made by the ITER Home Teams in the development of robust carbon and tungsten armoured plasma facing components for the ITER divertor. The activities on the development and study of armour materials, joining technologies, non-destructive evaluation techniques, high heat flux testing of manufactured components and neutron irradiation resistance studies are presented. The results of these activities confirm the feasibility of the main divertor components. Examples of the fruitful collaboration between Parties and future R and D needs are also described. (author)

  2. Nuclear modules of ITER tokamak systems code

    International Nuclear Information System (INIS)

    Gohar, Y.; Baker, C.; Brooks, J.

    1987-10-01

    Nuclear modules were developed to model various reactor components in the ITER systems code. 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. 39 refs., 3 tabs

  3. Thermal fatigue tests with actively cooled divertor mock-ups for ITER

    International Nuclear Information System (INIS)

    Roedig, M.; Duwe, R.; Linke, J.; Schuster, A.; Wiechers, B.; Ibbott, C.; Jacobson, D.; Le Marois, G.; Lind, A.; Lorenzetto, P.; Vieider, G.; Peacock, A.; Ploechl, L.; Severi, Y.; Visca, E.

    1998-01-01

    Mock-ups for high heat flux components with beryllium and CFC armour materials have been tested by means of the electron beam facility JUDITH. The experiments concerned screening tests to evaluate heat removal efficiency and thermal fatigue tests. CFC monoblocks attached to DS-Cu (Glidcop Al25) and CuCrZr tubes by active metal casting and Ti brazing showed the best thermal fatigue behaviour. They survived more than 1000 cycles at heat loads up to 25 MW m -2 without any indication of failure. Operational limits are given only by the surface temperature on the CFC tiles. Most of the beryllium mock-ups were of the flat tile type. Joining techniques were brazing, hot isostatic pressing (HIP) and diffusion bonding. HIPed and diffusion bonded Be/Cu modules have not yet reached the standards for application in high heat flux components. The limit of this production method is reached for heat loads of approximately 5 MW m -2 . Brazing with and without silver seems to be a more robust solution. A flat tile mock-up with CuMnSnCe braze was loaded at 5.4 MW m -2 for 1000 cycles without damage The first test with a beryllium monoblock joined to a CuCrZr tube by means of Incusil brazing shows promising results; it survived 1000 cycles at 4.5 MW m -2 without failure. (orig.)

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

    International Nuclear Information System (INIS)

    Takizuka, T.; Tokunaga, S.; Hoshino, K.; Shimizu, K.; Asakura, N.

    2015-01-01

    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

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

    International Nuclear Information System (INIS)

    Heinola, K.

    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 reaction. The amount of tritium retained in the wall materials should be minimized and its recycling back to the plasma must be unrestrained, otherwise it cannot be used for fueling the plasma. A very expensive and thus economically not viable solution is to replace the first walls quite often. A better solution is to heat the walls to temperatures where tritium is released. Unfortunately, the exact mechanisms of hydrogen release in tungsten are not known. In this thesis both experimental and computational methods have been used for studying the release and retention of hydrogen in tungsten. The experimental work consists of hydrogen implantations into pure polycrystalline tungsten, the determination of the hydrogen concentrations using ion beam analyses (IBA) and monitoring the out-diffused hydrogen gas with thermodesorption spectrometry (TDS) as the tungsten samples are heated at elevated temperatures. Combining IBA methods with TDS, the retained amount of hydrogen is obtained as well as the temperatures needed for the hydrogen release. With computational methods the hydrogen-defect interactions and implantation-induced irradiation damage can be examined at the atomic level. The method of multiscale modelling combines the results obtained from computational methodologies applicable at different length and time scales. Electron density functional theory calculations were used for determining the energetics of the elementary processes of hydrogen in tungsten, such as diffusivity and

  6. A helical hydrogen-MARFE-like phenomenon in the divertor of the Wendelstein 7-AS stellarator

    Science.gov (United States)

    Wenzel, U.; König, R.; Pedersen, T. Sunn; the W7-AS Team

    2015-01-01

    In the island divertor of the W7-AS stellarator a high-density zone (HDZ) near the divertor plates was discovered some years ago (Ramasubramanian et al 2004 Nucl. Fusion 44 992-8) with electron densities up to 7 × 1020 m-3. We shed further light on this phenomenon by determining the poloidal and radial location of this zone and discussing potential implications of these findings. The HDZ is in the vicinity of, but clearly separated from the nearest X-point line. The carbon emission is clearly spatially separated, residing near or at the X-point lines. The HDZ shows many similarities with the hydrogen or wall MARFE in Textor-94 (Samm et al 1999 J. Nucl. Mater. 266-269 666). The structure is associated with a strongly increased neutral pressure, thus enabling efficient pumping. This offers the possibility for a very efficient exhaust regime in a stellarator with island divertor such as W7-X, simultaneously with significantly reduced convective heat loads onto the divertor itself. The spatial separation of the HDZ and the carbon radiation region may imply that such a state can be reached even in a non-carbon machine, and might therefore be DEMO-relevant.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bruhn, M.L.

    1988-04-01

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

  8. Numerical Simulation on Subcooled Boiling Heat Transfer Characteristics of Water-Cooled W/Cu Divertors

    Science.gov (United States)

    Han, Le; Chang, Haiping; Zhang, Jingyang; Xu, Tiejun

    2015-04-01

    In order to realize safe and stable operation of a water-cooled W/Cu divertor under high heating condition, the exact knowledge of its subcooled boiling heat transfer characteristics under different design parameters is crucial. In this paper, subcooled boiling heat transfer in a water-cooled W/Cu divertor was numerically investigated based on computational fluid dynamic (CFD). The boiling heat transfer was simulated based on the Euler homogeneous phase model, and local differences of liquid physical properties were considered under one-sided high heating conditions. The calculated wall temperature was in good agreement with experimental results, with the maximum error of 5% only. On this basis, the void fraction distribution, flow field and heat transfer coefficient (HTC) distribution were obtained. The effects of heat flux, inlet velocity and inlet temperature on temperature distribution and pressure drop of a water-cooled W/Cu divertor were also investigated. These results provide a valuable reference for the thermal-hydraulic design of a water-cooled W/Cu divertor. supported by the National Magnetic Confinement Fusion Science Program of China (No. 2010GB104005), Funding of Jiangsu Innovation Program for Graduate Education (CXLX12_0170), the Fundamental Research Funds for the Central Universities of China

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  10. Parametric analyses of DEMO Divertor using two dimensional transient thermal hydraulic modelling

    Science.gov (United States)

    Domalapally, Phani; Di Caro, Marco

    2017-11-01

    Among the options considered for cooling of the Plasma facing components of the DEMO reactor, water cooling is a conservative option because of its high heat removal capability. In this work a two-dimensional transient thermal hydraulic code is developed to support the design of the divertor for the projected DEMO reactor with water as a coolant. The mathematical model accounts for transient 2D heat conduction in the divertor section. Temperature-dependent properties are used for more accurate analysis. Correlations for single phase flow forced convection, partially developed subcooled nucleate boiling, fully developed subcooled nucleate boiling and film boiling are used to calculate the heat transfer coefficients on the channel side considering the swirl flow, wherein different correlations found in the literature are compared against each other. Correlation for the Critical Heat Flux is used to estimate its limit for a given flow conditions. This paper then investigates the results of the parametric analysis performed, whereby flow velocity, diameter of the coolant channel, thickness of the coolant pipe, thickness of the armor material, inlet temperature and operating pressure affect the behavior of the divertor under steady or transient heat fluxes. This code will help in understanding the basic parameterś effect on the behavior of the divertor, to achieve a better design from a thermal hydraulic point of view.

  11. Erosion/redeposition analysis of the ITER [International Tokamak Engineering Reactor] divertor

    International Nuclear Information System (INIS)

    Brooks, J.N.

    1989-07-01

    Sputtering erosion of the proposed ITER divertor has been analyzed using the REDEP computer code. A carbon coated plate at medium and low plasma edge temperatures, as well as beryllium and tungsten plates, have been examined. Peak net erosion rates for C and Be are very high (∼20--80 cm/burn·yr) though an order of magnitude less than the gross rates. Tritium buildup rates in codeposited carbon surface layers may also be high (∼50--250 kg/burn·yr). Plasma contamination, however, from divertor sputtering is low (approx lt.5%). Operation with low Z divertor plates, at high duty factors, therefore appears unacceptable due to erosion, but may work for low duty factor (∼2%) ''physics phase'' operation. Sweeping of the poloidal field lines at the divertor can reduce erosion, by typical factors of ∼2--8. A tungsten coated plate works well, from the erosion standpoint, for plasma plate temperatures of ∼40 eV or less. 18 refs., 11 figs., 3 tabs

  12. Suppression of erosion in the DIII-D divertor with detached plasmas

    International Nuclear Information System (INIS)

    Wampler, William R.; Bastasz, Robert J.; Whyte, D.G.; Wong, C.P.C.; West, W.P.

    2000-01-01

    The ability to withstand disruptions makes carbon-based materials attractive for use as plasma-facing components in divertors. However, such materials suffer high erosion rates during attached plasma operation which, in high power long pulse machines, would give short component lifetimes and high tritium inventories. The authors present results from recent experiments in DIII-D, in which the Divertor Materials Evaluation System (DiMES) was used to examine erosion and deposition during short exposures to well defined plasma conditions. These studies show that during operation with detached plasmas, produced by gas injection, net erosion is suppressed everywhere in the divertor. Net deposition of carbon with deuterium was observed at the inner and outer strikepoints and in the private-flux region between strikepoints. For these low temperature plasmas (T e < 2eV), physical sputtering is eliminated. These results show that with detached plasmas, the location of carbon net erosion and the carbon impurity source, probably lies outside the divertor. Physical or chemical sputtering by charge-exchange neutrals or ions in the main plasma chamber is a probable source of carbon under these plasma conditions

  13. European development of He-cooled divertors for fusion power plants

    International Nuclear Information System (INIS)

    Norajitra, P.; Giniyatulin, R.; Kuznetsov, V.; Mazul, I.; Ovchinnikov, I.; Ihli, T.; Janeschitz, G.; Krauss, W.; Kruessmann, R.; Karditsas, P.; Maisonnier, D.; Sardain, P.; Nardi, C.; Papastergiou, S.; Pizzuto, A.

    2005-01-01

    Helium-cooled divertor concepts are considered suitable for use in fusion power plants for safety reasons, as they enable the use of a coolant compatible with any blanket concept, since water would not be acceptable e.g. in connection with ceramic breeder blankets using large amounts of beryllium. Moreover, they allow for a high coolant exit temperature for increasing the efficiency of the power conversion system. Within the framework of the European power plant conceptual study (PPCS), different helium-cooled divertor concepts based on different heat transfer mechanisms are being investigated at ENEA Frascati, Italy, and Forschungszentrum Karlsruhe, Germany. They are based on a modular design which helps reduce thermal stresses. The design goal is to withstand a high heat flux of about 10-15 MW/m 2 , a value which is considered relevant to future fusion power plants to be built after ITER. The development and optimisation of the divertor concepts require an iterative design approach with analyses, studies of materials and fabrication technologies, and the execution of experiments. These issues and the state of the art of divertor development shall be the subject of this report. (author)

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

    International Nuclear Information System (INIS)

    Simard, M.; Marchand, R.; Boucher, C.; Gunn, J.P.

    1996-01-01

    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

  15. Preliminary analysis of the efficiency of non-standard divertor configurations in DEMO

    Directory of Open Access Journals (Sweden)

    F. Subba

    2017-08-01

    Full Text Available The standard Single Null (SN divertor is currently expected to be installed in DEMO. However, a number of alternative configurations are being evaluated in parallel as backup solutions, in case the standard divertor does not extrapolate successfully from ITER to a fusion power plant. We used the SOLPS code to produce a preliminary analysis of two such configurations, the X-Divertor (XD and the Super X-Divertor (SX, and compare them to the SN solution. Considering the nominal power flowing into the SOL (PSOL = 150 MW, we estimated the amplitude of the acceptable DEMO operational space. The acceptability criterion was chosen as plasma temperature at the target lower than 5eV, providing low sputtering and at least partial detachment, while the operational space was defined in terms of the electron density at the outboard mid-plane separatrix and of the seeded impurity (Ar only in the present study concentration. It was found that both the XD and the SXD extend the DEMO operational space, although the advantages detected so far are not dramatic. The most promising configuration seems to be the XD, which can produce acceptable target temperatures at moderate outboard mid-plane electron density (nomp=4.5×1019 m−3 and Zeff= 1.3.

  16. 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/

  17. A tangentially viewing visible TV system for the DIII-D divertor

    International Nuclear Information System (INIS)

    Fenstermacher, M.E.; Meyer, W.H.; Wood, R.D.; Nilson, D.G.; Ellis, R.; Brooks, N.H.

    1997-01-01

    A video camera system has been installed on the DIII-D tokamak for two-dimensional spatial studies of line emission in the lower divertor region. The system views the divertor tangentially at approximately the height of the X point through an outer port. At the tangency plane, the entire divertor from the inner wall to outside the DIII-D bias ring is viewed with spatial resolution of ∼1 cm. The image contains information from ∼90 deg of toroidal angle. In a recent upgrade, remotely controllable filter changers were added which have produced images from nominally identical discharges using different spectral lines. Software was developed to calculate the response function matrix of the optical system using distributed computing techniques and assuming toroidal symmetry. Standard sparse matrix algorithms are then used to invert the three-dimensional images onto a poloidal plane. Spatial resolution of the inverted images is 2 cm; higher resolution simply increases the size of the response function matrix. Initial results from a series of experiments with multiple identical discharges show that the emission from CII and CIII, which appears along the inner scrape-off layer above and below the X point during ELMing H mode, moves outward and becomes localized near the X point in radiative divertor operation induced by deuterium injection. copyright 1997 American Institute of Physics

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    International Nuclear Information System (INIS)

    Edmonds, P.H.; Medley, S.S.

    1996-05-01

    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

  1. Nuclear analysis of structural damage and nuclear heating on enhanced K-DEMO divertor model

    Science.gov (United States)

    Park, J.; Im, K.; Kwon, S.; Kim, J.; Kim, D.; Woo, M.; Shin, C.

    2017-12-01

    This paper addresses nuclear analysis on the Korean fusion demonstration reactor (K-DEMO) divertor to estimate the overall trend of nuclear heating values and displacement damages. The K-DEMO divertor model was created and converted by the CAD (Pro-Engineer™) and Monte Carlo automatic modeling programs as a 22.5° sector of the tokamak. The Monte Carlo neutron photon transport and ADVANTG codes were used in this calculation with the FENDL-2.1 nuclear data library. The calculation results indicate that the highest values appeared on the upper outboard target (OT) area, which means the OT is exposed to the highest radiation conditions among the three plasma-facing parts (inboard, central and outboard) in the divertor. Especially, much lower nuclear heating values and displacement damages are indicated on the lower part of the OT area than others. These are important results contributing to thermal-hydraulic and thermo-mechanical analyses on the divertor and also it is expected that the copper alloy materials may be partially used as a heat sink only at the lower part of the OT instead of the reduced activation ferritic-martensitic steel due to copper alloy’s high thermal conductivity.

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

    International Nuclear Information System (INIS)

    1989-09-01

    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.)

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

    Science.gov (United States)

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

    2018-02-01

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

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

    International Nuclear Information System (INIS)

    Anderson, P.M.; Murphy, C.J.; Reis, E.E.; Hu, Q.; Song, Y.; Yao, D.

    2006-01-01

    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 350 o C 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

  5. Manufacturing monitoring and mock-ups validation of the WEST divertor structure and coils

    Energy Technology Data Exchange (ETDEWEB)

    Doceul, Louis, E-mail: louis.doceul@cea.fr [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Bucalossi, Jérôme; Decool, Patrick; Dougnac, Hubert; Ferlay, Fabien; Gargiulo, Laurent; Keller, Delphine; Larroque, Sébastien; Lipa, Manfred; Martino, Patrick; Pilia, Arnaud; Poli, Serge [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Portafaix, Christophe [ITER Organization, Route de Vinon-sur-Verdon, 13115 Saint-Paul-lez-Durance (France); Saille, Alain [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Salami, Michael [AVANTIS Engineering Groupe, ZI de l’Aiguille, 46100 Figeac (France); Samaille, Frank; Soler, Bernard; Thouvenin, Didier; Verger, Jean-Marc [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Voyard, Olivier [CNIM, ZI de Brégaillon, 83500 La Seyne-sur-Mer (France); and others

    2015-10-15

    Highlights: • The mechanical design and integration of the divertor structure have been performed. • The design of the casing and the winding-pack has been optimized. • The coil assembly process has been assessed. • The realization of a coil mock-up scale one is scheduled. - Abstract: In order to fully validate “ITER-like” actively water cooled tungsten plasma facing units, the implementation of an axisymmetric divertor structure in the Tokamak Tore-Supra has been studied. With this major upgrade, the so-called WEST (Tungsten Environment in Steady state Tokamak), Tore-Supra will be able to address the issues of long plasma discharges using a tungsten divertor based on monoblock targets. The divertor structure and coils assembly are made up of two stainless steel casings containing a copper winding pack cooled by a pressurized hot water circuit (up to 180 °C, 4 MPa) in which a total divertor current of up to 16 × 13 kA is circulating in steady state. The conductor is electrically insulated and wedged inside the casing in order to be mechanically protected. The divertor which is designed to perform steady state plasma operation (up to 1000 s), must sustain harsh environmental conditions in terms of ultra light vacuum conditions, electromagnetical loads and electrical insulation (5 kV ground voltage) under high temperature (180 °C). Therefore, a feasibility study of such a complex structure has been performed. It implied activities on a scale one dummy coil, such as installation, assembly issues and representative tests (electric, thermal and hydraulic). The manufacturing of the divertor structure, which is a large assembly of 4-m diameter representing a total weight of around 20 tonnes, started in the second half of 2013 and is expected to be delivered by the end of 2014. The paper will illustrate the technical developments and tests performed during 2013 and beginning of 2014 in order to fully validate the design concept before the industrial phase

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

    International Nuclear Information System (INIS)

    Bucalossi, J.; Traverse, J.M.; Corre, Y.; Courtois, X.; Firdaouss, M.; Grosman, A.; Missirlian, M.; Nardon, E.; Salasca, S.; Tsitrone, E.; Van Houtte, D.; Aumeunier, M.H.

    2015-01-01

    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/m 2 ) 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

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

    International Nuclear Information System (INIS)

    GROTH, M.; PORTER, G.D.; PETRIE, T.W.; FENSTERMACHER, M.E.; BROOKS, N.H.

    2003-01-01

    OAK-B135 Measurements of the D α 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. 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)

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  11. Thermal performance and 3D analysis of advanced mechanically joined divertor plate for LHD under steady state high heat flux

    International Nuclear Information System (INIS)

    Kubota, Y.; Noda, N.; Sagara, A.; Sakamoto, R.; Yamazaki, K.; Satow, T.; Motojima, O.

    2001-01-01

    Thermal performances of an advanced mechanically joined module (MJM) under steady state high heat fluxes of 2.0-4.25 MW/m 2 and thermo-mechanical analysis are described. The advanced MJM was designed to apply for a helical divertor plate of a large helical device (LHD) at the next experimental phase II. The advanced MJM has a unified armor/heat sink made of carbon/carbon composite different from the normal MJM with a separated copper heat sink. To evaluate the thermal performance of the advanced MJM, short pulse high heat flux test up to 5.4 MW/m 2 and steady state high heat flux tests up to 4.25 MW/m 2 have been carried out using a test facility ACT. Moreover, the thermal fatigue test of the advanced MJM up to 150 cycles under steady state high heat flux of 2.5 MW/m 2 has been performed. After theses tests, no apparent damage and no cracking on the armor tile were observed although there was a little increase in the armor/heat sink temperature during the thermal fatigue test. To evaluate the thermal stress in the armor/heat sink of the MJM during steady state high heat flux test and to optimize the structure, thermo-mechanical analyses are done using a 3D CAD

  12. Toroidally symmetric/asymmetric effect on the divertor flux due to neon/nitrogen seeding in LHD

    Directory of Open Access Journals (Sweden)

    H. Tanaka

    2017-08-01

    Full Text Available Toroidal distributions of divertor particle flux during neon (Ne and nitrogen (N2 seeded discharges were investigated in the Large Helical Device (LHD. By using 14 toroidally distributed divertor probe arrays, which were positioned at radially inner side where the divertor flux concentrates in the inward-shifted magnetic axis configuration, it is found that Ne puffing leads to toroidally quasi-uniform reduction of divertor particle fluxes; whereas toroidally localized reductions were observed with N2 puffing. The toroidally asymmetric reduction pattern with N2 puffing is strongly related to the magnetic field structure around the N2 puffing port. Assuming that nitrogen particles do not recycle, EMC3-EIRENE simulation shows similar reduction pattern with the experiment around the N2 puffing port.

  13. Development of a non destructive evaluation system using infrared images for divertor on nuclear fusion experiment reactor

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji; Enoeda, Mikio; Akiba, Masato

    2008-01-01

    An infrared thermography NDE facility which is utilized in the acceptance test of ITER divertor components has been developed in JAEA. This NDE facility can inspect the integrity of the bonding interface of the divertor components based on its surface temperature response by means of switching of hot (95 deg C)/cold (5 deg C) water. The advantages of this facility are 1) to have active coolant purging system which enables rapid temperature change and 2) to inspect the surface and the both side walls of three components at a time. We have conduct test operation for the divertor mockups and have found sufficient performance to implement the required acceptance test of the ITER divertor components. (author)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  16. Development of integrated SOL/Divertor code and simulation study of the JT-60U/JT-60SA tokamaks

    International Nuclear Information System (INIS)

    Kawashima, H.; Shimizu, K.; Takizuka, T.

    2007-01-01

    To predict the particle and heat controllability in the divertor of tokamak reactors such as ITER and to optimize the divertor design, comprehensive simulations by integrated modelling with taking in various physical processes are indispensable. For the design study of ITER divertor, the modelling codes such as B2, UEDGE and EDGE2D have been developed, and their results have contributed to the evolution of the divertor concept. In Japan Atomic Energy Agency (JAEA), SOL/divertor codes have also been developed for the interpretation and the prediction on behaviours of plasmas, neutrals and impurities in the SOL/divertor regions. The code development is originally carried out since physics models can be verified quickly and flexibly under the circumstance of close collaboration with JT-60 team. Figure 1 shows our code system, which consists of the 2 dimensional fluid code SOLDOR, the neutral Monte Carlo (MC) code NEUT2D, and the impurity MC code IMPMC. The particle simulation code PARASOL has also been developed in order to establish the physics modelling used in fluid simulations. Integration of SOLDOR, NEUT2D and IMPMC, called the '' SONIC '' code, is being carried out to simulate self-consistently the SOL/divertor plasmas in present tokamaks and in future devices. Combination of the SOLDOR and NEUT2D was completed, which has the features such as 1) high-resolution oscillation-free scheme in solving fluid equations, 2) neutral transport calculation under the fine meshes, 3) success in reduction of MC noise, 4) optimization on the massive parallel computer, etc. The simulation reproduces the X-point MARFE in the JT-60U experiment. It is found that the chemically sputtered carbon at the dome causes the radiation peaking near the X-point. The performance of divertor pumping in JT-60U is evaluated from the particle balances. We also present the divertor designing of JT-60SA, which is the modification program of JT-60U to establish high beta steady-state operation. To

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

    International Nuclear Information System (INIS)

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

    1992-09-01

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

  18. Experimental study of heating scheme effect on the inner divertor power footprint widths in EAST lower single null discharges

    Science.gov (United States)

    Deng, G. Z.; Xu, J. C.; Liu, X.; Liu, X. J.; Liu, J. B.; Zhang, H.; Liu, S. C.; Chen, L.; Yan, N.; Feng, W.; Liu, H.; Xia, T. Y.; Zhang, B.; Shao, L. M.; Ming, T. F.; Xu, G. S.; Guo, H. Y.; Xu, X. Q.; Gao, X.; Wang, L.

    2018-04-01

    A comprehensive work of the effects of plasma current and heating schemes on divertor power footprint widths is carried out in the experimental advanced superconducting tokamak (EAST). The divertor power footprint widths, i.e., the scrape-off layer heat flux decay length λ q and the heat spreading S, are crucial physical and engineering parameters for fusion reactors. Strong inverse scaling of λ q and S with plasma current have been demonstrated for both neutral beam (NB) and lower hybrid wave (LHW) heated L-mode and H-mode plasmas at the inner divertor target. For plasmas heated by the combination of the two kinds of auxiliary heating schemes (NB and LHW), the divertor power widths tend to be larger in plasmas with higher ratio of LHW power. Comparison between experimental heat flux profiles at outer mid-plane (OMP) and divertor target for NB heated and LHW heated L-mode plasmas reveals that the magnetic topology changes induced by LHW may be the main reason to the wider divertor power widths in LHW heated discharges. The effect of heating schemes on divertor peak heat flux has also been investigated, and it is found that LHW heated discharges tend to have a lower divertor peak heat flux compared with NB heated discharges under similar input power. All these findings seem to suggest that plasmas with LHW auxiliary heating scheme are better heat exhaust scenarios for fusion reactors and should be the priorities for the design of next-step fusion reactors like China Fusion Engineering Test Reactor.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-04-01

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

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  1. Improved structural strength and lifetime of monoblock divertor targets by using doped tungsten alloys under cyclic high heat flux loading

    Science.gov (United States)

    Nogami, S.; Guan, W. H.; Hattori, T.; James, K.; Hasegawa, A.

    2017-12-01

    The divertor is one of the most important components of a fusion reactor, which performs the function of the removal of waste material from fusion plasma. Because the divertor is subjected to cyclic high heat flux loading up to about 20 MW m-2 induced by the plasma, the plasma facing material of the divertor should exhibit good thermo-mechanical properties. In this work, the possibility of improving the structural strength and the lifetime of fusion reactor divertors by using K-doped W and K-doped W-3%Re as plasma facing material instead of ordinary pure W was evaluated by thermo-mechanical finite element analysis (FEA). These materials have been developed for divertor applications in Japan and show higher recrystallization temperature and strength than pure W. The results of the present study indicated that K-doped W and K-doped W-3%Re render lower applied strain to the divertor and longer fatigue life of the plasma facing material. The evaluation results regarding the macro-crack formation life based on the FEA analyses indicated the possibility of an extension of the fatigue life by using K-doped W and K-doped W-3%Re.

  2. Determination of volumetric plasma parameters from spectroscopic N II and N III line ratio measurements in the ASDEX Upgrade divertor

    Science.gov (United States)

    Henderson, S. S.; Bernert, M.; Brezinsek, S.; Carr, M.; Cavedon, M.; Dux, R.; Lipschultz, B.; O’Mullane, M. G.; Reimold, F.; Reinke, M. L.; The ASDEX Upgrade Team; The MST1 Team

    2018-01-01

    The diagnosis of tokamak divertor plasmas is limited in the ability to understand the behaviour and role of impurities, central to the overall understanding of how the divertor plasma can be utilised to control the power exhaust. New methods have been developed to extract the N concentration as well as plasma characteristics; the use of three visible N II lines has been shown to provide a unique solution of the background plasma density and temperature. Those techniques are applied to data from two sightlines sampling horizontally across the outer divertor plasma. The plasma densities obtained from the N II line ratios during a scan of the divertor temperature in a partially detached H-mode plasma suggest that, as the temperature drops, the plasma density decreases further up the divertor leg while closer to the strike point the plasma density increases. The former is consistent with the emission zone moving from the private flux region into the scrape-off-layer plasma, and therefore sampling two different density regimes, while the latter is consistent with electron pressure conservation along a field line. With an approximate model of the length of the emission region, the N II divertor concentration is calculated in this discharge to be  ≈5-25% . The single N III line ratio measurement available within the same spectral range is dependent on temperature and density and therefore cannot provide a unique solution of both.

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

    International Nuclear Information System (INIS)

    Strauss, H.R.

    1995-08-01

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

  4. Algorithm development for safeguarding the Wendelstein 7-X divertor during steady state operation

    Energy Technology Data Exchange (ETDEWEB)

    Rodatos, A.; Jakubowski, M. [Max-Planck-Institut fuer Plasmaphysik, Wendelsteinstrasse 1, D-17491 Greifswald (Germany); Greuner, H.; Sunn Pedersen, T. [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2, D-85748 Garching (Germany); Wurden, G.A. [Los Alamos National Laboratory, Los Alamos, NM 87544 (United States)

    2014-07-01

    The divertor of Wendelstein 7X is designed to withstand steady state heat fluxes of 10 MW/m{sup 2} and 15 MW/m{sup 2} transiently. However higher local heat fluxes are possible. 10 thermographic infrared (IR) observation systems will be installed to monitor the divertor and its center goal is the detection of overheated areas in real time. Besides an increased plasma heat flux, there are at least two potential causes of an elevated diverter surface temperature. First, redeposited eroded material forming surface layers with a poor thermal connection to the underlying water-cooled tiles. Second, delaminated CFC tiles will exhibit an elevated surface temperature relative to properly bonded tiles. Using the measured characteristic time scales for the thermal response, gained from experiments at GLADIS, we have concluded that it is possible to distinguish between healthy, delaminated, surface-coated and delaminated surface-coated tiles.

  5. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-15

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

  7. Erosion and deposition of metals and carbon in the DIII-D divertor

    International Nuclear Information System (INIS)

    Wampler, W.R.; Bastasz, R.; Buchenauer, D.

    1995-01-01

    Net erosion rates at the outer strike point of the DIII-D divertor plasma were measured for several materials during quiescent H-mode operation with deuterium plasmas. Materials examined include graphite, beryllium, tungsten, vanadium and molybdenum. For graphite, net erosion rates up to 4 nm/sec were found. Erosion rates for the metals were much smaller than for carbon. Ion fluxes from Langmuir probe measurements were used to predict gross erosion by sputtering. Measured net erosion was much smaller than predicted gross erosion. Transport of metal atoms by the plasma across the divertor surface was also examined. Light atoms were transported farther than heavy atoms as predicted by impurity transport models

  8. Static and dynamic behaviors of plasma detachment in divertor simulator NAGDIS-II

    International Nuclear Information System (INIS)

    Takamura, S.; Ohno, N.; Uesugi, Y.; Nishijima, D.; Motoyama, M.; Hattori, N.; Arakawa, H.; Krasheninnikov, S.I.; Pigarov, A.; Wenzel, U.

    2001-01-01

    We have performed comprehensive investigation on the static and dynamic behaviors in detached recombining plasmas in the linear divertor plasma simulator, NAGDIS-II. For the stationary plasma detachment, the transition from electron-ion recombination (EIR) to molecular activated recombination (MAR) has been observed by injecting hydrogen gas to high density He plasmas. The particle loss rate due to MAR is found to be comparable to that of EIR. We have also performed experiments on injection of a plasma heat pulse produced by rf heating to the detached recombining He plasma to demonstrate the dynamic behavior of the volumetric plasma recombination. Negative spikes in Balmer series line emissions were observed similar to the so called negative ELM observed in tokamak divertors, which were analyzed with collisional-radiative model in detail. Rapid increase of the ion flux to the target plate was observed associated with the re-ionization of the highly excited atoms generated by EIR. (author)

  9. Electron Beam Design and Calibration for the Solid/Liquid Lithium Divertor Experiment

    Science.gov (United States)

    Jaworski, Michael; Flauta, R.; Gray, T. K.; Kim, J.; Lau, C. Y.; Lee, M. B.; Neumann, M. J.; Surla, V.; Ruzic, D. N.

    2008-11-01

    An electron beam has been developed as part of the Solid/Liquid Lithium Divertor Experiment (SLiDE) at the University of Illinois at Urbana-Champaign. The purpose of the SLiDE apparatus is to examine the motion of liquid lithium under fusion relevant heat loads and magnetic fields. To mimic the heat fluxes present in the divertor of a fusion machine, a linear sheet beam is utilized which can operate over a range of applied magnetic fields and power levels. With steady state operation up to 15kW input power, the beam can produce peak heat fluxes of 10 MW/m^2 and heat flux gradients comparable to those found in fusion experiments. The design of the electron beam was developed using commercial beam transport codes and the final design is diagnosed with a two-lead Faraday cup. Beam performance and characteristics are presented.

  10. Tungsten nano-tendril growth in the Alcator C-Mod divertor

    International Nuclear Information System (INIS)

    Wright, G.M.; Brunner, D.; Labombard, B.; Lipschultz, B.; Terry, J.L.; Whyte, D.G.; Baldwin, M.J.; Doerner, R.P.

    2012-01-01

    Growth of tungsten nano-tendrils (‘fuzz’) has been observed for the first time in the divertor region of a high-power density tokamak experiment. After 14 consecutive helium L-mode discharges in Alcator C-Mod, the tip of a tungsten Langmuir probe at the outer strike point was fully covered with a layer of nano-tendrils. The thickness of the individual nano-tendrils (50–100 nm) and the depth of the layer (600 ± 150 nm) are consistent with observations from experiments on linear plasma devices. The observation of tungsten fuzz in a tokamak may have important implications for material erosion, dust formation, divertor lifetime and tokamak operations in next-step devices. (letter)

  11. Thermal strain measurement of EAST tungsten divertor component with bare fiber Bragg grating sensors

    Science.gov (United States)

    Wang, Xingli; Wang, Wanjing; Wang, Jichao; Wei, Ran; Sun, Zhaoxuan; Li, Qiang; Xie, Chunyi; Luo, Guang-Nan

    2017-12-01

    Fiber Bragg Gratings (FBGs) have been widely used in the sensor field to monitor temperature and strain. However, the weak mechanical property of optical fibers and insufficient heat-resistant property of general optic-fiber sensors have prevented it from being widely used, such as in some extreme engineering situations. In this work, a bare FBG sensor system had been introduced to measure thermal strain of an Experimental Advanced Superconducting Tokamak tungsten divertor component under baking condition. This strain measurement system had withstood as high temperature as 210 °C and finished the measurement experiment successfully. Meaningful measurement results had been obtained and analyzed, which showed the applicability of such a bare fiber grating sensor system and as well contributed to studying on tungsten divertor's thermal strain conditions.

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

    Science.gov (United States)

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

    2007-06-01

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

  13. He-cooled divertor for DEMO: Experimental verification of the conceptual modular design

    International Nuclear Information System (INIS)

    Norajitra, P.; Gervash, A.; Giniyatulin, R.; Ihli, T.; Krauss, W.; Kruessmann, R.; Kuznetsov, V.; Makhankov, A.; Mazul, I.; Ovchinnikov, I.

    2006-01-01

    A modular He-cooled divertor concept is being developed at the Forschungszentrum Karlsruhe. The design goal is to withstand a high heat flux of 10 MW/m 2 at least. The work programme of 2004 focused on experiments to verify the design and thermohydraulics layout. In cooperation with the Efremov Institute, experimental investigations were performed for the joining of tungsten parts and/or tungsten parts with steel and the fabrication of divertor components from tungsten. Moreover, gas puffing experiments were carried out with a stationary approach to measuring pressure loss and heat transfer for the purpose of screening the design options and verifying the computational fluid dynamics (CFD) calculations. The status and results of the technological and helium experiments shall be outlined in this report

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

    International Nuclear Information System (INIS)

    Hiwatari, R.; Hatayama, A.; Zhu, S.; Takizuka, T.; Tomita, Y.

    2005-01-01

    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)

  15. Design feasibility study of a divertor component reinforced with fibrous metal matrix composite laminate

    Energy Technology Data Exchange (ETDEWEB)

    You, J.-H. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Boltzmannstr. 2, D-85748 Garching (Germany)]. E-mail: j.h.you@ipp.mpg.de

    2005-01-01

    Fibrous metal matrix composites possess advanced mechanical properties compared to conventional alloys. It is expected that the application of these composites to a divertor component will enhance the structural reliability. A possible design concept would be a system consisting of tungsten armour, copper composite interlayer and copper heat sink where the composite interlayer is locally inserted into the highly stressed domain near the bond interface. For assessment of the design feasibility of the composite divertor concept, a non-linear multi-scale finite element analysis was performed. To this end, a micro-mechanics algorithm was implemented into a finite element code. A reactor-relevant heat flux load was assumed. Focus was placed on the evolution of stress state, plastic deformation and ductile damage on both macro- and microscopic scales. The structural response of the component and the micro-scale stress evolution of the composite laminate were investigated.

  16. Design feasibility study of a divertor component reinforced with fibrous metal matrix composite laminate

    International Nuclear Information System (INIS)

    You, J.-H.

    2005-01-01

    Fibrous metal matrix composites possess advanced mechanical properties compared to conventional alloys. It is expected that the application of these composites to a divertor component will enhance the structural reliability. A possible design concept would be a system consisting of tungsten armour, copper composite interlayer and copper heat sink where the composite interlayer is locally inserted into the highly stressed domain near the bond interface. For assessment of the design feasibility of the composite divertor concept, a non-linear multi-scale finite element analysis was performed. To this end, a micro-mechanics algorithm was implemented into a finite element code. A reactor-relevant heat flux load was assumed. Focus was placed on the evolution of stress state, plastic deformation and ductile damage on both macro- and microscopic scales. The structural response of the component and the micro-scale stress evolution of the composite laminate were investigated

  17. Design manufacturing and thermo-mechanical testing of a relevant size mono block divertor prototype

    International Nuclear Information System (INIS)

    Cardella, A.; Vieider, G.; Di Pietro, E.; Orsini, A.; Febvre, M.; Guerreschi, U.; Reheis, N.; Bruno, L.

    1994-01-01

    Following a technological development of joining techniques between carbon fibre composite tiles and metallic tubes, and the manufacturing and testing of small size actively cooled mock-ups, a relevant size divertor prototype has been designed, manufactured and tested. The prototype consisted of a series of metallic tubes surrounded by CFC tiles, cooling collectors and a supporting system representative of a divertor dump plate for high power reactors. The tubes have been preliminary tested at the CEA 200 kW electron beam facility with uniform fluxes up to 5 MW/m 2 to select the best five tubes, which together with a sixth non tested tube have been then assembled to form the prototype. This has been tested at the JET high power neutral beam injector test facility. After screening tests the prototype has been subjected to thermal cycling at more than 15 MW/m 2 . (author) 12 refs.; 4 figs

  18. 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...... strengthened copper (Cu-Al2O3) is considered to be the best candidate for high heat flux structural applications, followed by CuNiBe and CuCrZr.......This paper presents a critical analysis of the main factors of radiation damage limiting the possibility to use copper alloys in the ITER divertor and first wall structure. In copper alloys the most significant types of radiation damage in the proposed temperature-dose operation range are swelling......, creep, and low-temperature radiation embrittlement. Low-temperature radiation embrittlement at T-irr alloys, as their uniform elongation at T-test - T-irr - 100 degrees C drops...

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    Czech Academy of Sciences Publication Activity Database

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

    2017-01-01

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

  1. HRP facility for fabrication of ITER vertical target divertor full scale plasma facing units

    International Nuclear Information System (INIS)

    Visca, Eliseo; Roccella, S.; Candura, D.; Palermo, M.; Rossi, P.; Pizzuto, A.; Sanguinetti, G.P.; Mancini, A.; Verdini, L.; Cacciotti, E.; Cerri, V.; Mugnaini, G.; Reale, A.; Giacomi, G.

    2015-01-01

    Highlights: • R&D activities for the manufacturing of ITER divertor high heat flux plasma-facing components (HHFC). • ENEA and Ansaldo have jointly manufactured several actively cooled monoblock mock-ups and prototypical components. • ENEA and ANSALDO NUCLEARE jointly participate to the European program for the qualification of the manufacturing technology for the ITER divertor IVT. • Successful manufacturing by HRP (Hot Radial Pressing) of first full-scale full-W armored IVT qualification prototype. - Abstract: ENEA and Ansaldo Nucleare S.p.A. (ANN) have being deeply involved in the European development activities for the manufacturing of the ITER Divertor Inner Vertical Target (IVT) plasma-facing components. During normal operation the heat flux deposited on the bottom segment of divertor is 5–10 MW/m 2 but the capability to remove up to 20 MW/m 2 during transient events of 10 s must also be demonstrated. In order to fulfill ITER requirements, ENEA has set up and widely tested a manufacturing process, named Hot Radial Pressing (HRP). The last challenge is now to fabricate full-scale prototypes of the IVT, aimed to be qualified for the next step, i.e. the series production. On the basis of the experience of manufacturing hundreds of small mock-ups, ENEA designed and installed a new suitable HRP facility. The objective of getting a final shaped plasma facing unit (PFU) that satisfies these requirements is an ambitious target because tolerances set by ITER/F4E are very tight. The setting-up of the equipment started with the fabrication of full scale and representative ‘dummies’ in which stainless steel instead of CFC or W was used for monoblocks. The results confirmed that dimensions were compliant with the required tolerances. The paper reports a brief description of the innovative HRP equipment and the dimensional check results after HRP of the first full-scale full-W PFU.

  2. HRP facility for fabrication of ITER vertical target divertor full scale plasma facing units

    Energy Technology Data Exchange (ETDEWEB)

    Visca, Eliseo, E-mail: eliseo.visca@enea.it [Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, IT-00044 Frascati (Roma) (Italy); Roccella, S. [Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, IT-00044 Frascati (Roma) (Italy); Candura, D.; Palermo, M. [Ansaldo Nucleare S.p.A., Corso Perrone 25, IT-16152 Genova (Italy); Rossi, P.; Pizzuto, A. [Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, IT-00044 Frascati (Roma) (Italy); Sanguinetti, G.P. [Ansaldo Nucleare S.p.A., Corso Perrone 25, IT-16152 Genova (Italy); Mancini, A.; Verdini, L.; Cacciotti, E.; Cerri, V.; Mugnaini, G.; Reale, A.; Giacomi, G. [Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, IT-00044 Frascati (Roma) (Italy)

    2015-10-15

    Highlights: • R&D activities for the manufacturing of ITER divertor high heat flux plasma-facing components (HHFC). • ENEA and Ansaldo have jointly manufactured several actively cooled monoblock mock-ups and prototypical components. • ENEA and ANSALDO NUCLEARE jointly participate to the European program for the qualification of the manufacturing technology for the ITER divertor IVT. • Successful manufacturing by HRP (Hot Radial Pressing) of first full-scale full-W armored IVT qualification prototype. - Abstract: ENEA and Ansaldo Nucleare S.p.A. (ANN) have being deeply involved in the European development activities for the manufacturing of the ITER Divertor Inner Vertical Target (IVT) plasma-facing components. During normal operation the heat flux deposited on the bottom segment of divertor is 5–10 MW/m{sup 2} but the capability to remove up to 20 MW/m{sup 2} during transient events of 10 s must also be demonstrated. In order to fulfill ITER requirements, ENEA has set up and widely tested a manufacturing process, named Hot Radial Pressing (HRP). The last challenge is now to fabricate full-scale prototypes of the IVT, aimed to be qualified for the next step, i.e. the series production. On the basis of the experience of manufacturing hundreds of small mock-ups, ENEA designed and installed a new suitable HRP facility. The objective of getting a final shaped plasma facing unit (PFU) that satisfies these requirements is an ambitious target because tolerances set by ITER/F4E are very tight. The setting-up of the equipment started with the fabrication of full scale and representative ‘dummies’ in which stainless steel instead of CFC or W was used for monoblocks. The results confirmed that dimensions were compliant with the required tolerances. The paper reports a brief description of the innovative HRP equipment and the dimensional check results after HRP of the first full-scale full-W PFU.

  3. Mock-up test results of monoblock-type CFC divertor armor for JT-60SA

    Energy Technology Data Exchange (ETDEWEB)

    Higashijima, S. [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)], E-mail: higashijima.satoru@jaea.go.jp; Sakurai, S.; Suzuki, S.; Yokoyama, K.; Kashiwa, Y.; Masaki, K.; Shibama, Y.K.; Takechi, M.; Shibanuma, K.; Sakasai, A.; Matsukawa, M.; Kikuchi, M. [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)

    2009-06-15

    The JT-60 Super Advanced (JT-60SA) tokamak project starts under both the Japanese domestic program and the international program 'Broader Approach'. The maximum heat flux to JT-60SA divertor is estimated to {approx}15 MW/m{sup 2} for 100 s. Japan Atomic Energy Agency (JAEA) has developed a divertor armor facing high heat flux in the engineering R and D for ITER, and it is concluded that monoblock-type CFC divertor armor is promising for JT-60SA. The JT-60SA armor consists of CFC monoblocks, a cooling CuCrZr screw-tube, and a thin oxygen-free high conductivity copper (OFHC-Cu) buffer layer between the CFC monoblock and the screw-tube. CFC/OFHC-Cu and OFHC-Cu/CuCrZr joints are essential for the armor, and these interfaces are brazed. Needed improvements from ITER engineering R and D are good CFC/OFHC-Cu and OFHC-Cu/CuCrZr interfaces and suppression of CFC cracking. For these purposes, metalization inside CFC monoblock is applied, and we confirmed again that the mock-up has heat removal capability in excess of ITER requirement. For optimization of the fabrication method and understanding of the production yield, the mock-ups corresponding to quantity produced in one furnace at the same time is also produced, and the half of the mock-ups could remove 15 MW/m{sup 2} as required. This paper summarizes the recent progress of design and mock-up test results for JT-60SA divertor armor.

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

    International Nuclear Information System (INIS)

    Nilson, D.G.; Ellis, R.; Fenstermacher, M.E.; Brewis, G.; Jalufka, N.

    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-α 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 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 microm 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

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

    International Nuclear Information System (INIS)

    Azeroual, A.

    2000-01-01

    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 D α light emission and the neutral pressure in the divertor plenum. Last, one question how such a system can be extrapolated to next step machines, for which one must account for technical constraints linked to the presence of the shield protecting the coils from the high neutron flux. (author)

  6. Divertor-localized fluctuations in NSTX-U L-mode discharges

    Science.gov (United States)

    Scotti, Filippo; Soukhanovskii, V. A.; Zweben, S.; Myra, J.; Baver, D.; Sabbagh, S. A.

    2017-10-01

    The 3-D structure of divertor turbulence is characterized in NSTX-U by means of fast camera imaging. Edge and divertor turbulence can be important in determining the heat flux width in fusion devices. Field-aligned filaments are found on the divertor legs via imaging of C III and D- α emission in NBI-heated diverted L-mode discharges, similar to observations in Alcator C-Mod and MAST. These flute-like fluctuations of up to 10-20% in RMS/mean are radially localized around the separatrix and limited to the region below the X-point. Poloidal and parallel correlation lengths are a few cm (10-50ρi) and several meters, respectively. For the outer leg filaments, poloidal correlation lengths decrease along the leg away from the strike point and typical effective toroidal mode numbers are in the range of 10-20. Opposite toroidal rotation is observed for inner (co-current rotation) and outer leg (counter-current rotation) filaments with apparent poloidal propagation of 1 km/s. The poloidal motion of outer leg filaments is opposite to the one typically observed for NSTX upstream blobs in the scrape-off layer. The shape, dynamics and absence of correlation with upstream turbulence suggest that these fluctuations are generated and localized in the divertor region. Supported by US DOE DE-AC52-07NA27344, DE-AC02-09CH11466, DE-FG02- 02ER54678, DE-FG02-99ER54524.

  7. The magnitude and direction of forces on impurities near a divertor plate. An exploratory investigation

    International Nuclear Information System (INIS)

    Stangeby, P.C.; Toronto Univ., ON

    1990-01-01

    Most divertor impurity modelling gives the result that negligible leakage occurs of impurities produced at the plates, to the confined plasma and that negligible impurity radiation occurs, even in the divertor plasma. This is not, however, found experimentally. A Monte Carlo impurity code has been employed in exploratory studies aimed at identifying possible leakage pathways to the main plasma. It was found that for a substantial range of divertor plasma densities and temperatures, the friction force dominated all other forces on the impurities, including the temperature gradient forces which are generally directed away from the plates. Thus, if it was assumed that the deuterium plasma flow was everywhere directed toward the plates, negligible leakage to the confined plasma occurred. The possibility of deuterium flow reversal was also considered where it was assumed that over some radial fraction of the SOL, the deuterium plasma flow was directed away from the plates, starting at a distance from the plates equal to the deuterium average ionization distance. The spatial distribution of impurity (carbon) physically sputtered neutrals was modelled and it was found that a fraction of the impurity neutrals were ionized in the deuterium flow reversal zone. When these impurity particles were then tracked through further ionization, and their parallel and cross-field transport was followed, significant impurity radiation resulted and a substantial impurity density was found in the confined plasma with Z eff values of 3 or more. It was thus concluded that flow reversal of the deuterium plasma is a promising possible explanation of impurity radiation and of the leakage of divertor plate impurities to the confined plasma

  8. Energy confinement of beam-heated divertor and limiter discharges in Doublet III

    International Nuclear Information System (INIS)

    Nagami, M.; Kasai, M.; Kitsuezaki, A.

    1984-01-01

    Observation of the intensity of the recycling particle flux at the main plasma edge for various limiter and divertor discharges indicates that the gross energy confinement of beam-heated discharges is closely related to the intensity of the edge particle flux. In limiter discharges, the global particle confinement time and the energy confinement time tausub(E) show many similarities: 1) linear Isub(p) dependence at Isub(p) 13 cm -3 , tausub(E)*approx.=70 ms). (author)

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

    Czech Academy of Sciences Publication Activity Database

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

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

  10. Steady-state tokamak reactor with non-divertor impurity control: STARFIRE

    International Nuclear Information System (INIS)

    Baker, C.C.

    1980-01-01

    STARFIRE is a conceptual design study of a commercial tokamak fusion electric power plant. Particular emphasis has been placed on simplifying the reactor concept by developing design concepts to produce a steady-state tokamak with non-divertor impurity control and helium ash removal. The concepts of plasma current drive using lower hybrid rf waves and a limiter/vacuum system for reactor applications are described

  11. Simulation of neutral gas flow in a tokamak divertor using the Direct Simulation Monte Carlo method

    International Nuclear Information System (INIS)

    Gleason-González, Cristian; Varoutis, Stylianos; Hauer, Volker; Day, Christian

    2014-01-01

    Highlights: • Subdivertor gas flows calculations in tokamaks by coupling the B2-EIRENE and DSMC method. • The results include pressure, temperature, bulk velocity and particle fluxes in the subdivertor. • Gas recirculation effect towards the plasma chamber through the vertical targets is found. • Comparison between DSMC and the ITERVAC code reveals a very good agreement. - Abstract: This paper presents a new innovative scientific and engineering approach for describing sub-divertor gas flows of fusion devices by coupling the B2-EIRENE (SOLPS) code and the Direct Simulation Monte Carlo (DSMC) method. The present study exemplifies this with a computational investigation of neutral gas flow in the ITER's sub-divertor region. The numerical results include the flow fields and contours of the overall quantities of practical interest such as the pressure, the temperature and the bulk velocity assuming helium as model gas. Moreover, the study unravels the gas recirculation effect located behind the vertical targets, viz. neutral particles flowing towards the plasma chamber. Comparison between calculations performed by the DSMC method and the ITERVAC code reveals a very good agreement along the main sub-divertor ducts

  12. Deuterium retention in the divertor tiles of JET ITER-Like wall

    Directory of Open Access Journals (Sweden)

    A. Lahtinen

    2017-08-01

    Full Text Available Divertor tiles removed after the second JET ITER-Like Wall campaign 2013–2014 (ILW-2 were studied using Secondary Ion Mass Spectrometry (SIMS. Measurements show that the thickest beryllium (Be dominated deposition layers are located at the upper part of the inner divertor and are up to ∼40µm thick at the lower part of Tile 0 exposed in 2011–2014. The highest deuterium (D amounts (>8 · 1018at./cm2, in contrast, were found on the upper part of Tile 1 (2013–2014, where the Be deposits are ∼10µm thick. D was mainly retained in the near-surface layer of the Be deposits but also deeper in tungsten (W and molybdenum (Mo layers of the marker coated tiles, especially at W–Mo layer interfaces. D retention for the ILW-2 divertor tiles is higher than for the first campaign 2011–2012 (ILW-1 and probable reasons for the difference are that SIMS measurements for the ILW-2 samples were done deeper than for the ILW-1 samples, some of the tiles were exposed during both ILW-1 and ILW-2 and therefore had a longer exposure time, and the differences between ILW-1 and ILW-2 campaigns e.g. in strike point distributions and injected powers.

  13. A new divertor plates design concept for the double null NET configuration

    International Nuclear Information System (INIS)

    Farfaletti-Casali, F.; Renda, V.; Federici, G.; Papa, L.

    1986-01-01

    A new divertor plate design concept for the Double Null NET configuration (NET-DN) is presented. This concept applies to the plasma configuration of NET and takes advantage by the maintenance scheme of the internal components adopted in NET. According to this maintenance approach, which uses the top loading of the internal segments, 48 inboard removable segments, 3 for each of the 16 reactor sectors, act as simple protective panels, gathering together in only one piece the plates of both the upper and lower divertor regions and the intermediate portion of the inboard first wall. They are cooled by water flowing inside a set of hairpin-shaped, stainless steel tubes, arranged in poloidal direction inside a copper heat sink, and fed by supply lines at the top of the reactor. The surface facing the plasma is covered by a tungsten alloy layer. In such a way, the maintenance of the two divertor regions and of the inboard first wall can be easily achieved by removing the inboard panels from the top of the reactor. The layout of the cooling system and preliminary thermohydraulics and thermomechanical calculations, carried out for assessing the feasibility of the proposed system for the NET reference configuration, are reported in this paper. (author)

  14. Conceptual design of divertor cassette handling by remote handling system of JT-60SA

    International Nuclear Information System (INIS)

    Hayashi, Takao; Sakurai, Shinji; Masaki, Kei; Tamai, Hiroshi; Yoshida, Kiyoshi; Matsukawa, Makoto

    2008-01-01

    The JT-60SA aims to contribute and supplement ITER toward demonstration fusion reactor based on tokamak concept. One of the features of JT-60SA is its high power long pulse heating, causing the large annual neutron fluence. Because the expected dose rate at the vacuum vessel (VV) may exceed 1 mSv/hr after 10 years operation and three month cooling, the human access inside the VV is restricted. Therefore a remote handling (RH) system is necessary for the maintenance and repair of in-vessel components. This paper described the RH system of JT-60SA, especially the expansion of the RH rail and exchange of the divertor cassettes. The RH rail is divided into nine and three-point mounting. The nine sections can cover 225 degrees in toroidal direction. A divertor cassette, which is 10 degrees wide in toroidal direction and weighs 500 kg itself due to the limitations of port width and handling weight, can be exchanged by heavy weight manipulator (HWM). The HWM brings the divertor cassette to the front of the other RH port, which is used for supporting the rail and/or carrying in and out equipments. Then another RH device receives and brings out the cassette by a pallet installed from outside the VV. (author)

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

    International Nuclear Information System (INIS)

    Kaye, S.M.; Bell, M.; Bol, K.

    1983-11-01

    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

  16. High heat flux Langmuir probe array for the DIII-D divertor platesa)

    Science.gov (United States)

    Watkins, J. G.; Taussig, D.; Boivin, R. L.; Mahdavi, M. A.; Nygren, R. E.

    2008-10-01

    Two modular arrays of Langmuir probes designed to handle a heat flux of up to 25 MW/m2 for 10 s exposures have been installed in the lower divertor target plates of the DIII-D tokamak. The 20 pyrolytic graphite probe tips have more than three times higher thermal conductivity and 16 times larger mass than the original DIII-D isotropic graphite probes. The probe tips have a fixed 12.5° surface angle to distribute the heat flux more uniformly than the previous 6 mm diameter domed collectors and a symmetric "rooftop" design to allow operation with reversed toroidal magnetic field. A large spring-loaded contact area improves heat conduction from each probe tip through a ceramic insulator into a cooled graphite divertor floor tile. The probe tips, brazed to molybdenum foil to ensure good electrical contact, are mounted in a ceramic tray for electrical isolation and reliable cable connections. The new probes are located 1.5 cm radially apart in a staggered arrangement near the entrance to the lower divertor pumping baffle and are linearly spaced 3 cm apart on the shelf above the in-vessel cryopump. Typical target plate profiles of Jsat, Te, and Vf with 4 mm spatial resolution are shown.

  17. High heat flux Langmuir probe array for the DIII-D divertor plates.

    Science.gov (United States)

    Watkins, J G; Taussig, D; Boivin, R L; Mahdavi, M A; Nygren, R E

    2008-10-01

    Two modular arrays of Langmuir probes designed to handle a heat flux of up to 25 MW/m(2) for 10 s exposures have been installed in the lower divertor target plates of the DIII-D tokamak. The 20 pyrolytic graphite probe tips have more than three times higher thermal conductivity and 16 times larger mass than the original DIII-D isotropic graphite probes. The probe tips have a fixed 12.5 degree surface angle to distribute the heat flux more uniformly than the previous 6 mm diameter domed collectors and a symmetric "rooftop" design to allow operation with reversed toroidal magnetic field. A large spring-loaded contact area improves heat conduction from each probe tip through a ceramic insulator into a cooled graphite divertor floor tile. The probe tips, brazed to molybdenum foil to ensure good electrical contact, are mounted in a ceramic tray for electrical isolation and reliable cable connections. The new probes are located 1.5 cm radially apart in a staggered arrangement near the entrance to the lower divertor pumping baffle and are linearly spaced 3 cm apart on the shelf above the in-vessel cryopump. Typical target plate profiles of J(sat), T(e), and V(f) with 4 mm spatial resolution are shown.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  20. Effects of neutral gas collisions on the power transmission factor at the divertor sheath

    International Nuclear Information System (INIS)

    Futch, A.H.; Matthews, G.F.; Buchenauer, D.; Hill, D.N.; Jong, R.A.; Porter, G.D.

    1992-01-01

    We show that charge-exchange and other ion-neutral collision can reduce the power transmission factor of the plasma sheath, thereby lowering the ion impact energy and target plate sputtering. The power transmission factor relates the heat flux reaching the divertor target to the plasma density and temperature just in front of the surface: δ=Q surf /J ew k T e . Experimental data from the DIII-D tokamak suggests that δ could be as low as 2-3 near the region of peak divertor particle flux, instead of the 7-8 expected from usual sheath theory. Several effects combine to allow ion-neutral interactions to be important in the divertor plasma sheath. The shallow angle of incidence of the magnetic field (1-3deg in DIII-D) leads to the spatial extension of the sheath from approximately ρ i ∝1 mm normal to the plate to several centimeters along the field lines. Ionization reduces the sheath potential, and collisions reduce the ion impact energy. (orig.)

  1. Fluctuations measured by flush mounted versus proud divertor Langmuir probes - why are they different?

    Science.gov (United States)

    Garcia, O. E.; Kuang, A. Q.; Brunner, D.; Labombard, B.; Kube, R.

    2017-10-01

    A flush-mounted, toroidally-elongated, and field-aligned divertor `rail' Langmuir probe array was installed in Alcator C-Mod in 2015. This geometry is heat flux tolerant and effectively mitigates sheath expansion effects down to incident field line angles of 0.5 degree. Further complications have arisen that cannot be explained by sheath-expansion. In particular, the `rail' probe geometry measures significantly higher plasma fluctuation levels in the common flux region compared to traditional proud probes, whereas they are similar in the private flux zone. In some instances, the amplitudes of ion saturation current fluctuations normalized to the mean are a factor of 2 higher; probability distribution functions correspondingly record large amplitude events that are not seen by the proud probes. This discrepancy also appears to depend on divertor plasma regime. For example, fluctuations become similar near the strikepoint when the electron temperature is low. To ensure that these discrepancies were not due to perturbations caused by the voltage bias or currents collected by the probes, the two Langmuir probe systems were left to `float' and the fluctuation statistics analyzed. Yet, even in this non-perturbative situation, there exist clear differences in the fluctuation characteristics. The raises two questions: how does the probe geometry affect plasma fluctuations measurements and what are the true plasma fluctuations experienced by the divertor surface? Supported by USDoE awards DE-FC02-99ER54512.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  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. A new divertor plates design concept for the double null net configuration

    International Nuclear Information System (INIS)

    Farfaletti-Casali, F.; Iop, O.; Renda, V.; Federici, G.; Papa, L.

    1987-01-01

    A new divertor plate design concept for the Double Null NET configuration (NET-DN) is presented in this paper. This concept applies to the plasma configuration of NET and takes advantage by the maintenance scheme of the internal components adopted in NET. According to this maintenance approach, which uses the top loading of the internal segments, 48 inboard removable segments, 3 for each of the 16 reactor sectors, act as simple protective panels, gathering together in only one piece the plates of both the upper and lower divertor regions and the intermediate portion of the inboard first wall. They are cooled by water flowing inside a set of hairpin-shaped, stainless steel tubes, arranged in poloidal direction inside a copper heat sink, and fed by supply lines at the top of the reactor. The surface facing the plasma is covered by a tungsten alloy layer. In such a way, the maintenance of the two divertor regions and of the inboard first wall can be easily achieved by removing the inboard panels from the top of the reactor. The layout of the cooling system and preliminary thermohydraulics and thermomechanical calculations, carried out for assessing the feasibility of the proposed system for the NET reference configuration, are reported in this paper

  5. Secondary radiation effects during transient events on ITER divertor and nearby hidden components

    Science.gov (United States)

    Hassanein, Ahmed; Sizyuk, Valeryi

    2016-10-01

    The secondary plasma developed as a result of disruptions and ELMs on divertor plate is composed mainly from high-Z divertor materials. This mini-plasma will greatly increase the radiation flux to nearby and hidden components. Our simulations showed significant increase in radiation fluxes and heat loads in the high-Z (i.e., tungsten) generated secondary plasma. These radiations could seriously damage hidden nearby components such as umbrella and dome structure. We have implemented enhanced models in our comprehensive integrated HEIGHTS package for 3D simulation of detailed photon and particle transport in the evolved secondary plasma during instabilities. HEIGHTS can simulate full 3D realistic ITER geometry, including the reflector plate, etc. to assess performance of all components resulting from plasma instabilities. HEIGHTS predicted, for the first time, details of heat loads and temperatures evolution of all nearby components due to transient events. Secondary radiation evolution and expansion cause serious damage to hidden internal components that were not directly exposed to DT plasma. Current ITER divertor design may require changes to mitigate such effects and protect interior components that is hard and costly to repair in case of transient disruptive events. Work is partially supported by Purdue University.

  6. Physics Design Requirements for the National Spherical Torus Experiment Liquid Lithium Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Kugel, W.; Bell, M.; Berzak,L.; Brooks, A.; Ellis, R.; Gerhardt, S.; Harjes, H.; Kaita, R.; Kallman, J.; Maingi, R.; Majeski, R.; Mansfield, D.; Menard, J.; Nygren,R. E.; Soukhanovskii, V.; Stotler, D.; Wakeland, P.; Zakharov L. E.

    2008-09-26

    Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on PFC's 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 installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15-25% ne decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, ne/nGW~1), to enable ne scan capability (x2) in the H-mode, to test the ability to operate at significantly lower density for future ST-CTF reactor designs (e.g., ne/nGW = 0.25), and eventually to investigate high heat-flux power handling (10 MW/m2) with longpulse discharges (>1.5s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

  7. The impact of the biasing radial electric field on the SOL in a divertor tokamak

    International Nuclear Information System (INIS)

    Rozhansky, V.; Tendler, M.

    1993-01-01

    Strong radial electric field can be induced within the SOL in a divertor tokamak by applying a voltage to divertor plates with respect to the first wall. This biasing scheme results in the strong radial electric field which is much larger than the natural electric field, usually of the order T e /e. Experiments employing this biasing scheme were carried out on the tokamak TdeV. Many interesting effects such as - modifications of the density profile and radial transport of impurities as a function of the polarity and the magnitude of the biasing voltage, the generation of the flux surface average toroidal rotation proportional to the applied voltage, redistribution of the plasma outflow onto divertor plates and so on - were demonstrated to result from the biasing. Furthermore, in contrast to studies carried out employing a different biasing scheme which primarily results in a poloidal electric field, the strong radial electric field impacts more significantly within SOL than the poloidal electric field. Here, we aim to show that the main effects observed experimentally follow from the analysis, provided continuity and momentum balances are employed invoking anomalous viscosity and inertia. (author) 4 refs

  8. Design and construction of a lithium vapor box divertor similarity experiment

    Science.gov (United States)

    Schwartz, J. A.; Cohen, R. A.; Emdee, E. D.; Jaworski, M. A.; Goldston, R. J.

    2017-10-01

    Future fusion devices will require handling extreme heat fluxes. The lithium vapor box divertor is a concept to manage this heat flux. The divertor plasma impinges on a dense cloud of lithium vapor, leading to volumetric cooling, radiation, and recombination. The vapor is localized by baffles and condensation on the divertor slot walls upstream of the target, limiting the lithium reaching the main chamber. A series of test stand experiments will study vapor confinement and plasma plugging in a simplified baffled-pipe geometry. A first experiment without plasma will validate a DSMC model for evaporation, flow, and condensation of lithium vapor. Three stainless steel cylindrical cans will be heated to 550C, 600C, and 650C respectively inside a vacuum chamber. Lithium flow will be measured by weighing the cans before and after heating and by calorimetry of the latent heat of the vapor. Progress on the experiment will be presented. This work supported by DOE Contract No. DE-AC02-09CH11466.

  9. Time-dependent modeling of dust injection in semi-detached ITER divertor plasma

    Science.gov (United States)

    Smirnov, Roman; Krasheninnikov, Sergei

    2017-10-01

    At present, it is generally understood that dust related issues will play important role in operation of the next step fusion devices, i.e. ITER, and in the development of future fusion reactors. Recent progress in research on dust in magnetic fusion devises has outlined several topics of particular concern: a) degradation of fusion plasma performance; b) impairment of in-vessel diagnostic instruments; and c) safety issues related to dust reactivity and tritium retention. In addition, observed dust events in fusion edge plasmas are highly irregular and require consideration of temporal evolution of both the dust and the fusion plasma. In order to address the dust-related fusion performance issues, we have coupled the dust transport code DUSTT and the edge plasma transport code UEDGE in time-dependent manner, allowing modeling of transient dust-induced phenomena in fusion edge plasmas. Using the coupled codes we simulate burst-like injection of tungsten dust into ITER divertor plasma in semi-detached regime, which is considered as preferable ITER divertor operational mode based on the plasma and heat load control restrictions. Analysis of transport of the dust and the dust-produced impurities, and of dynamics of the ITER divertor and edge plasma in response to the dust injection will be presented. This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, under Award Number DE-FG02-06ER54852.

  10. Migration of Artificially Introduced Micron Size Carbon Dust in the DIII-D Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Rudakov, D; West, W; Wong, C; Brooks, N; Evans, T; Fenstermacher, M; Groth, M; Krasheninnikov, S; Lasnier, C; McLean, A; Pigarov, A Y; Solomon, W; Antar, G; Boedo, J; Doerner, R; Hollmann, E; Hyatt, A; Maingi, R; Moyer, R; Nagy, A; Nishino, N; Roquemore, L; Stangeby, P; Watkins, J

    2006-05-15

    Migration of pre-characterized carbon dust in a tokamak environment was studied by introducing about 30 milligrams of dust flakes 5-10 {micro}m in diameter in the lower divertor of DIII-D using the DiMES sample holder. The dust was exposed to high power ELMing Hmode discharges in lower-single-null magnetic configuration with the strike points swept across the divertor floor. When the outer strike point (OSP) passed over the dust holder exposing it to high particle and heat fluxes, part of the dust was injected into the plasma. In about 0.1 sec following the OSP pass over the dust, 1-2% of the total dust carbon content (2-4 x 10{sup 19} carbon atoms, equivalent to a few million dust particles) penetrated the core plasma, raising the core carbon density by a factor of 2-3. When the OSP was inboard of the dust holder, the dust injection continued at a lower rate. Individual dust particles were observed moving at velocities of 10-100 m/s, predominantly in the toroidal direction for deuteron flow to the outer divertor target, consistent with the ion drag force. The observed behavior of the dust is in qualitative agreement with modeling by the 3D DustT code.

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

    International Nuclear Information System (INIS)

    Kitazawa, Sin-iti; Ogawa, Hiroaki

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kitazawa, Sin-iti, E-mail: kitazawa.siniti@qst.go.jp; Ogawa, Hiroaki

    2016-11-15

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

  13. Use of isotopically-enriched carbon probes for erosion/deposition measurements in the ASDEX divertor

    International Nuclear Information System (INIS)

    Roberto, J.B.; Roth, J.; Taglauer, E.; Holland, O.W.

    1984-01-01

    An isotopic marker technique has been developed and used to measure near-surface erosion and carbon and impurity deposition on graphite samples in the ASDEX divertor chamber. Papyex graphite strips were enriched to approx. 20% 13 C over the first 1000 A by ion implantation. The implanted 13 C can be clearly distinguished from 12 C of the host by Rutherford backscattering using 2.7 MeV 4 He. Carbon erosion and deposition were determined from changes in the depth distribution of the 13 C, and impurity accumulation was also monitored in the backscattering spectrum. The 13 C profiles remained sharp in spite of high heat fluxes to the samples (sufficient to melt stainless steel), and the results showed clear evidence of simultaneous erosion and deposition of carbon and impurities. During 2.4 MW neutral injection, net carbon erosion exceeded 250 A/s in the divertor throat. High carbon erosion was also observed during periods of substantial Fe and Ti impurity accumulation on the samples. The results demonstrate the applicability of the 13 C technique for measuring erosion/deposition at high-heat-flux surfaces such as limiters or divertor plates

  14. 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.

  15. Structural impact of creep in tungsten monoblock divertor target at 20 MW/m2

    Directory of Open Access Journals (Sweden)

    Muyuan Li

    2018-01-01

    Full Text Available In order to increase erosion lifetime of the divertor target, in the 2nd design phase of R&D work package ‘Divertor’ for European DEMO, armor thickness of tungsten monoblock divertor target is increased from 5 mm to 8 mm. By increasing armor thickness, surface temperature increases nearly linearly, which makes effect of creep no longer negligible at slow transients of 20 MW/m2. In this work, structural impact of creep in tungsten monoblock divertor target is for the first time quantitatively analyzed with the aid of finite element method. The numerical simulations have revealed that creep results in an increase of inelastic strain accumulation. With increasing armor thickness, tensile surface stress along x-axis (the longer edge at the plasma-facing surface of tungsten monoblock reduces, while surface stress along z-axis (axial direction of the cooling tube changes from tensile to compressive. Creep will accelerate this change. With increasing grain size, creep strain accumulation at loading surface increases due to higher creep rates, while plastic strain accumulation decreases. Creep can mitigate the risk of deep cracking by reducing the driving force for crack opening, and has a positive impact for preventing the contact between the upper parts of neighboring monoblocks in high heat flux tests.

  16. ATHENA simulations of divertor loss of heat sink transient for the GSSR - Final report with updates

    International Nuclear Information System (INIS)

    Sponton, L.L.

    2001-05-01

    The ITER-FEAT Generic Site Safety Report includes evaluations of the consequences of various types of conceivable transients that can occur during operation. The transients that have to be considered in this respect are specified in the Accident Analysis Specifications document of the safety report. For the divertor primary heat transport system the ranges of transients include amongst others a loss of heat sink at full fusion power operation. The thermal-hydraulic consequences related to the coolability of the divertor primary heat transport system components for this transient have been evaluated and summarised in the safety report and in the current report an overview of those efforts and associated outcome is provided. The analyses have been made with the ATHENA thermal-hydraulic code using a separately developed ATHENA model of the ITER-FEAT divertor cooling system. In the current report results from calculations with an updated pressurizer model and pressurizer control system are outlined. The results show that the pressurizer safety valve does not open, that the pressurizer level increase is moderate and that no temperature increases jeopardize the structure integrity

  17. Thermo-mechanical and damage analyses of EAST carbon divertor under type-I ELMy H-mode operation

    International Nuclear Information System (INIS)

    Li, W.X.; Song, Y.T.; Ye, M.Y.; Peng, X.B.; Wu, S.T.; Qian, X.Y.; Zhu, C.C.

    2016-01-01

    Highlights: • Type-I ELMy H-mode is one of the most severe operating environment in tokamak. • An actual time-history heat load has been used in thermo-mechanical analysis. • The analysis results are time-dependent during the whole discharge process. • The analysis could be very useful in evaluating the operational capability of the divertor. - Abstract: The lower carbon divertor has been used since 2008 in EAST, and many significant physical results, like the 410 s long pulse discharge and the 32 s H-mode operation, have been achieved. As the carbon divertor will still be used in the next few years while the injected auxiliary heating power would be increased gradually, it’s necessary to evaluate the operational capability of the carbon divertor under the heat loads during future operation. In this paper, an actual time-history heat load during type-I ELMy H-mode from EAST experiment, as one of the most severe operating environment in tokamak, has been used in the calculation and analysis. The finite element (FE) thermal and mechanical calculations have been carried out to analysis the stress and deformation of the carbon divertor during the heat loads. According to the results, the main impact on the overall temperature comes from the relative stable phase before and after the type-I ELMs and local peak load, and the transient thermal load such as type-I ELMy only has a significant effect on the surface temperature of the graphite tiles. The carbon divertor would work with high stress near the screw bolts in the current operational conditions, because of high preload and conservative frictional coefficient between the bolts and heatsink. For the future operation, new plasma facing materials (PFM) and divertor technology should be developed.

  18. 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

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

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka

    2005-02-01

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

  20. Studies of power exhaust and divertor design for a 1.5 GW-level fusion power DEMO

    Science.gov (United States)

    Asakura, N.; Hoshino, K.; Suzuki, S.; Tokunaga, S.; Someya, Y.; Utoh, H.; Kudo, H.; Sakamoto, Y.; Hiwatari, R.; Tobita, K.; Shimizu, K.; Ezato, K.; Seki, Y.; Ohno, N.; Ueda, Y.; Joint Special TeamDEMO Design

    2017-12-01

    Power exhaust to the divertor and the conceptual design have been investigated for a steady-state DEMO in Japan with 1.5 GW-level fusion power and the major radius of 8.5 m, where the plasma parameters were revised appropriate for the impurity seeding scenario. A system code survey for the Ar impurity seeding suggested the volume-averaged density, impurity concentration and exhaust power from the main plasma of {{P}sep ~ }   =  205-285 MW. The divertor plasma simulation (SONIC) was performed in the divertor leg length of 1.6 m with the fixed exhaust power to the edge of {{P}out}   =  250 MW and the total radiation fraction at the edge, SOL and divertor ({{P}rad}/{{P}out}   =  0.8), as a first step to investigate appropriate design of the divertor size and geometry. At the outer target, partial detachment was produced near the strike-point, and the peak heat load ({{q}target} ) at the attached region was reduced to ~5 MW m-2 with appropriate fuel and impurity puff rates. At the inner divertor target, full detachment of ion flux was produced and the peak {{q}target} was less than 10 MW m-2 mostly due to the surface-recombination. These results showed a power exhaust scenario and the divertor design concept. An integrated design of the water-cooling heat sink for the long leg divertor was proposed. Cu-ally (CuCrZr) cooling pipe was applicable as the heat sink to handle the high heat flux near the strike-point, where displacements per atom rate was estimated to be 0.5-1.5 per year by neutronics calculation. An arrangement of the coolant rooting for Cu-alloy and Reduced Activation Ferritic Martensitic (RAFM) steel (F82H) pipes in a divertor cassette was investigated, and the heat transport analysis of the W-monoblock and Cu-alloy pipe under the peak {{q}target} of 10 MWm-2 and nuclear heating was performed. The maximum temperatures on the W-surface and Cu-alloy pipe were 1021 and 331 °C. Heat flux of 16 MW m-2 was distributed in the major part

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-08-01

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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Mueller, E.R.; Steinmetz, K.; Bein, B.K.

    1984-06-01

    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 10 13 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)

  4. 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)

    Matera, R.; Merola, M.

    1992-01-01

    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

  5. Optimizing the overall configuration of a He-cooled W-alloy divertor for a power plant

    International Nuclear Information System (INIS)

    Raffray, A.R.; Malang, S.; Wang, X.

    2009-01-01

    A number of different He-cooled divertor configurations have been proposed for magnetic fusion energy (MFE) power plant application. They range in scale from a plate configuration with characteristic dimension of the order of 1 m, to the ARIES-CS T-tube configuration with characteristic dimension of the order of 10 cm, to the EU FZK finger concept with characteristic dimension of the order of 1.5 cm. All these designs utilize tungsten or tungsten alloy as structural material. This paper considers the characteristics of the different divertor configurations and proposes the possibility of optimizing the design by combining different configurations in an integrated design based on the anticipated divertor heat flux profile.

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

    DEFF Research Database (Denmark)

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

    2011-01-01

    A revised formulation of the perpendicular diffusive transport model in 2D multi-fluid edge codes is proposed. Based on theoretical predictions and experimental observations a dependence on collisionality is introduced into the transport model of EDGE2D–EIRENE. The impact on time-dependent JET 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...

  7. Interpretation of ion flux and electron temperature profiles at the JET divertor target during high recycling and detached discharges

    International Nuclear Information System (INIS)

    Monk, R.D.

    1997-01-01

    Detailed experiments have been carried out with the JET Mark I pumped divertor to characterise high recycling and detached plasma regimes. This paper presents new measurements of high resolution divertor ion flux profiles that identify the growth of additional peaks during high recycling discharges. These ion flux profiles are used in conjunction with Dα and neutral flux measurements to examine the physics of divertor detachment and compare against simple analytic models. Finally, problems are highlighted with conventional methods of single and triple probe interpretation under high recycling conditions. By assuming that the single probe behaves as an asymmetric double probe the whole characteristic may be fitted and significantly lower electron temperatures may be derived when the electron to ion saturation current ratio is reduced. The results from the asymmetric double probe fit are shown to be consistent with independent diagnostic measurements. (orig.)

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  9. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Pietro Alessandro Di Maio; Valerio Tomarchio; Giuseppe Vella; Irene Zammuto [Dipartimento di Ingegneria Nucleare, Viale delle Scienze, 90128 Palermo, (Italy); Giovanni Dell' Orco [ENEA-Brasimone, 40032 Camugnano, Bologna (Italy)

    2005-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-04-01

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

  12. Modelling of mitigation of the power divertor loading for the EU DEMO through Ar injection

    Science.gov (United States)

    Subba, Fabio; Aho-Mantila, Leena; Coster, David; Maddaluno, Giorgio; Nallo, Giuseppe F.; Sieglin, Bernard; Wenninger, Ronald; Zanino, Roberto

    2018-03-01

    In this paper we present a computational study on the divertor heat load mitigation through impurity injection for the EU DEMO. The study is performed by means of the SOLPS5.1 code. The power crossing the separatrix is considered fixed and corresponding to H-mode operation, whereas the machine operating condition is defined by the outboard mid-plane upstream electron density and the impurity level. The selected impurity for this study is Ar, based on its high radiation efficiency at SOL characteristic temperatures. We consider a conventional vertical target geometry for the EU DEMO and monitor target conditions for different operational points, considering as acceptability criteria the target electron temperature (≤5 eV to provide sufficiently low W sputtering rate) and the peak heat flux (below 5–10 MW m‑2 to guarantee safe steady-state cooling conditions). Our simulations suggest that, neglecting the radiated power deposition on the plate, it is possible to satisfy the desired constraints. However, this requires an upstream density of the order of at least 50% of the Greenwald limit and a sufficiently high argon fraction. Furthermore, if the radiated power deposition is taken into account, the peak heat flux on the outer plate could not be reduced below 15 MW m‑2 in these simulations. As these simulations do not take into account neutron loading, they strongly indicate that the vertical target divertor solution with a radiative front distributed along the divertor leg has a very marginal operational space in an EU DEMO sized reactor.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  15. Measurements of carbon and tungsten erosion/deposition in the DIII-D divertor

    International Nuclear Information System (INIS)

    Bastasz, R.; Wampler, W.R.; Cuthbertson, J.W.; Buchenauer, D.A.; Brooks, N.; Junge, R.; West, W.P.; Wong, C.P.C.

    1994-01-01

    Net erosion/deposition rates of carbon and tungsten were measured at the outer strike point of the divertor plasma on the floor of the DIII-D tokamak during deuterium H-mode operation at a peak power deposition of about 40 W/cm 2 . For carbon, net erosion rates of up to 4 nm/s were found. For a tungsten film, no appreciable erosion was detected. However, measurements of deposited tungsten on adjacent carbon surfaces indicated a net W erosion rate of 0.06 nm/s

  16. Testing candidate interlayers for an enhanced water-cooled divertor target

    Energy Technology Data Exchange (ETDEWEB)

    Hancock, David, E-mail: david.hancock@ccfe.ac.uk [CCFE, Culham Science Centre, Oxfordshire OX14 3DB (United Kingdom); Barrett, Tom; Foster, James; Fursdon, Mike; Keech, Gregory; McIntosh, Simon; Timmis, William [CCFE, Culham Science Centre, Oxfordshire OX14 3DB (United Kingdom); Rieth, Michael; Reiser, Jens [Karlsruhe Institute of Technology, IAM-AWP, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2015-10-15

    Highlights: • We introduce an optimised divertor target concept: the “Thermal Break”. • We suggest a candidate interlayer material for this concept: FeltMetal. • We describe a bespoke rig for testing the thermal conductivity of this material. • We present preliminary results for a number of samples. - Abstract: The design of a divertor target for DEMO remains one of the most challenging engineering tasks to be overcome on the path to fusion power. Under the European DEMO programme, a promising concept known as Thermal Break has been developed at CCFE. This concept is a variation of the ITER tungsten divertor in which the pure Copper interlayer between Copper Chrome Zirconium coolant pipe and Tungsten monoblock armour is replaced with a low thermal conductivity compliant interlayer, with the aim of reducing the thermal mismatch stress between the armour and structure. One candidate material for this interlayer is FeltMetal™ (Technetics Group, USA). This material consists of an amorphous matrix of fine copper wires which are sintered onto a thin copper foil, creating a sheet of approximately 1 mm thickness. FeltMetal has been successfully used for many years to provide compliant sliding electrical contacts for the MAST TF coils and on ALCATOR C-Mod and extensive material testing has therefore been undertaken to quantify thermal and mechanical properties. These tests, however, have not been performed under vacuum or DEMO-relevant conditions. A bespoke experimental test rig has therefore been designed and constructed with which to measure the interlayer thermal conductance as a function of temperature and pressure under vacuum conditions. The design of this apparatus and the results of experiments on FeltMetal as well as other candidate interlayers are presented here. In parallel, joint mockups using the candidate interlayers have been prepared and Thermal Break divertor target mockups have been manufactured, requiring the development of a dedicated

  17. Method for comparison of tokamak divertor strike point data with magnetic perturbation models

    Czech Academy of Sciences Publication Activity Database

    Cahyna, Pavel; Peterka, Matěj; Nardon, E.; Frerichs, H.; Pánek, Radomír

    2014-01-01

    Roč. 54, č. 6 (2014), 064002-064002 ISSN 0029-5515. [International Workshop on Stochasticity in Fusion Plasmas /6./. Jülich, 18.03.2013-20.03.2013] R&D Projects: GA ČR GAP205/11/2341 Institutional support: RVO:61389021 Keywords : divertor * resonant magnetic perturbation Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.062, year: 2014 http://iopscience.iop.org/0029-5515/54/6/064002/pdf/0029-5515_54_6_064002.pdf

  18. Testing candidate interlayers for an enhanced water-cooled divertor target

    International Nuclear Information System (INIS)

    Hancock, David; Barrett, Tom; Foster, James; Fursdon, Mike; Keech, Gregory; McIntosh, Simon; Timmis, William; Rieth, Michael; Reiser, Jens

    2015-01-01

    Highlights: • We introduce an optimised divertor target concept: the “Thermal Break”. • We suggest a candidate interlayer material for this concept: FeltMetal. • We describe a bespoke rig for testing the thermal conductivity of this material. • We present preliminary results for a number of samples. - Abstract: The design of a divertor target for DEMO remains one of the most challenging engineering tasks to be overcome on the path to fusion power. Under the European DEMO programme, a promising concept known as Thermal Break has been developed at CCFE. This concept is a variation of the ITER tungsten divertor in which the pure Copper interlayer between Copper Chrome Zirconium coolant pipe and Tungsten monoblock armour is replaced with a low thermal conductivity compliant interlayer, with the aim of reducing the thermal mismatch stress between the armour and structure. One candidate material for this interlayer is FeltMetal™ (Technetics Group, USA). This material consists of an amorphous matrix of fine copper wires which are sintered onto a thin copper foil, creating a sheet of approximately 1 mm thickness. FeltMetal has been successfully used for many years to provide compliant sliding electrical contacts for the MAST TF coils and on ALCATOR C-Mod and extensive material testing has therefore been undertaken to quantify thermal and mechanical properties. These tests, however, have not been performed under vacuum or DEMO-relevant conditions. A bespoke experimental test rig has therefore been designed and constructed with which to measure the interlayer thermal conductance as a function of temperature and pressure under vacuum conditions. The design of this apparatus and the results of experiments on FeltMetal as well as other candidate interlayers are presented here. In parallel, joint mockups using the candidate interlayers have been prepared and Thermal Break divertor target mockups have been manufactured, requiring the development of a dedicated

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  20. Charge state of sputtered impurity ions near a limiter or divertor in a tokamak

    International Nuclear Information System (INIS)

    Boley, C.D.; Brooks, J.N.; Kim, Y.K.

    1983-03-01

    Many impurity atoms sputtered from a limiter or divertor plate are ionized in the scrapeoff zone and return to the sputtering surface bacause of friction with incoming plasma ions. The final charge state attained by such impurities has been calculated for a variety of plasma edge conditions. The surface materials considered are tungsten, beryllium, beryllium oxide, and carbon. Estimates of the successive ionization cross sections for tungsten are developed. In all cases examined, returning impurity ions are found to be multiply ionized. This implies a significant energy gain in the sheath region, with important implications for self-sputtering of redeposited surface material

  1. Feasibility study of fast swept divertor strike point suppressing transient heat fluxes in big tokamaks.

    Czech Academy of Sciences Publication Activity Database

    Horáček, Jan; Cunningham, G.; Entler, Slavomír; Dobias, P.; Duban, R.; Imríšek, Martin; Markovič, Tomáš; Havlíček, Josef; Enikeev, R.

    2017-01-01

    Roč. 123, November (2017), s. 646-649 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] R&D Projects: GA ČR(CZ) GA16-14228S; GA MŠk(CZ) LM2015045; GA MŠk(CZ) 8D15001; GA MŠk LG14002 Institutional support: RVO:61389021 Keywords : DEMO * ELM * Divertor * Heat flux * Tokamak Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379617300376

  2. Kinetic modelling of plasma near the neutralizer plate in a tokamak divertor

    International Nuclear Information System (INIS)

    Abou-Assaleh, Z.; Marchand, R.; Matte, J.P.; Johnston, T.W.; Parbhakar, K.J.

    1990-01-01

    An electron kinetic code is used to simulate longitudinal transport and recycling near the neutralizer plate in a divertor plasma. In addition to the standard features, such as electron-electron and electron-ion Coulomb collisions, transport, ion motion, and a self-consistent electric field, the code now accounts for ionization, excitation, and recycling of hydrogen near the plate. Ions and neutrals are treated as fluids. The kinetic results are compared with those of a one-dimensional, two-temperature fluid code. Some implications of these results for recycling and impurity control in tokamaks are also discussed

  3. Electron temperature and heat load measurements in the COMPASS divertor using the new system of probes

    Czech Academy of Sciences Publication Activity Database

    Adámek, Jiří; Seidl, Jakub; Horáček, Jan; Komm, Michael; Eich, T.; Pánek, Radomír; Cavalier, J.; Devitre, A.; Peterka, Matěj; Vondráček, Petr; Stöckel, Jan; Šesták, David; Grover, Ondřej; Bílková, Petra; Böhm, Petr; Varju, Jozef; Havránek, Aleš; Weinzettl, Vladimír; Lovell, J.; Dimitrova, Miglena; Mitošinková, Klára; Dejarnac, Renaud; Hron, Martin

    2017-01-01

    Roč. 57, č. 11 (2017), č. článku 116017. ISSN 0029-5515 R&D Projects: GA ČR(CZ) GA15-10723S; GA ČR(CZ) GA16-14228S; GA MŠk(CZ) LM2015045 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : COMPASS * divertor * heat load * ELM * electron temperature * Ball-pen probe Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 3.307, year: 2016 http://iopscience.iop.org/article/10.1088/1741-4326/aa7e09

  4. 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 paper discusses a comparable simple extension of the model by introducing a non-zero characteristic velocity of the Maxwellian distributed particles. This way the experimentally observed temporal evolution as well as the in/out energy imbalance can be described. The extended model named free...

  5. Development of a helium-cooled divertor concept: design-related requirements on materials and fabrication technology

    Energy Technology Data Exchange (ETDEWEB)

    Norajitra, P. E-mail: prachai.norajitra@imf.fzk.de; Boccaccini, L.V.; Diegele, E.; Filatov, V.; Gervash, A.; Giniyatulin, R.; Gordeev, S.; Heinzel, V.; Janeschitz, G.; Konys, J.; Krauss, W.; Kruessmann, R.; Malang, S.; Mazul, I.; Moeslang, A.; Petersen, C.; Reimann, G.; Rieth, M.; Rizzi, G.; Rumyantsev, M.; Ruprecht, R.; Slobodtchouk, V

    2004-08-01

    Within the framework of the EU power plant conceptual study (PPCS), a modular He-cooled divertor concept with integrated pin array (HEMP) is being developed at the Forschungszentrum Karlsruhe. The design goal is to achieve a high heat flux of at least about 10-15 MW/m{sup 2}, which is proposed for a near-term reactor model like DEMO. The development and optimization of the divertor concept require a close link between the main issues: design, analyses, materials and fabrication technology, and experiments with feedbacks between them to be accounted for. Design-specific requirements on materials and fabrication issues will be discussed.

  6. Development of a helium-cooled divertor concept: design-related requirements on materials and fabrication technology

    International Nuclear Information System (INIS)

    Norajitra, P.; Boccaccini, L.V.; Diegele, E.; Filatov, V.; Gervash, A.; Giniyatulin, R.; Gordeev, S.; Heinzel, V.; Janeschitz, G.; Konys, J.; Krauss, W.; Kruessmann, R.; Malang, S.; Mazul, I.; Moeslang, A.; Petersen, C.; Reimann, G.; Rieth, M.; Rizzi, G.; Rumyantsev, M.; Ruprecht, R.; Slobodtchouk, V.

    2004-01-01

    Within the framework of the EU power plant conceptual study (PPCS), a modular He-cooled divertor concept with integrated pin array (HEMP) is being developed at the Forschungszentrum Karlsruhe. The design goal is to achieve a high heat flux of at least about 10-15 MW/m 2 , which is proposed for a near-term reactor model like DEMO. The development and optimization of the divertor concept require a close link between the main issues: design, analyses, materials and fabrication technology, and experiments with feedbacks between them to be accounted for. Design-specific requirements on materials and fabrication issues will be discussed

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  8. Powder Injection Molding for mass production of He-cooled divertor parts

    International Nuclear Information System (INIS)

    Antusch, S.; Norajitra, P.; Piotter, V.; Ritzhaupt-Kleissl, H.-J.

    2011-01-01

    A He-cooled divertor for future fusion power plants has been developed at KIT. Tungsten and tungsten alloys are presently considered the most promising materials for functional and structural divertor components. The advantages of tungsten materials lie, e.g. in the high melting point, and low activation, the disadvantages are high hardness and brittleness. The machinig of tungsten, e.g. milling, is very complex and cost-intensive. Powder Injection Molding (PIM) is a method for cost effective mass production of near-net-shape parts with high precision. The complete W-PIM process route is outlined and, results of product examination discussed. A binary tungsten powder feedstock with a grain size distribution in the range 0.7-1.7 μm FSSS, and a solid load of 50 vol.% was developed. After heat treatment, the successfully finished samples showed promising results, i.e. 97.6% theoretical density, a grain size of approximately 5 μm, and a hardness of 457 HV0.1.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

  10. 3D Numerical Analysis of Radiative Edge Cooling in Wendelstein 7-X Island Divertor Scenarios

    Science.gov (United States)

    Effenberg, Florian; Feng, Y.; Frerichs, H.; Schmitz, O.; Barbui, T.; Geiger, J.; Jakubowski, M.; Köenig, R.; Krychowiak, M.; Niemann, H.; Sunn Pedersen, T.; Suzuki, Y.; Wurden, G. A.; W7-X-Team Team

    2017-10-01

    Radiative edge cooling is a promising method for mitigation of high heat and particle fluxes in the 3D field geometry of Wendelstein 7-X. A new high mirror island configuration is investigated featuring a more uniform distribution of heat and particle fluxes on horizontal and vertical divertor targets. For an upstream density of nup = 2 × 1019m-3 at PECRH=8MW maximum heat loads up to qmax 7.2MWm-2 are calculated with the 3D fluid and kinetic edge transport Monte Carlo Code EMC3-EIRENE. Carbon eroded from the divertor targets is predicted to serve as effective intrinsic radiator enabling detached operational regimes at higher densities (nup > 4 × 1019m-3). The feasibility of active control of heat and particle flux levels by impurity seeding (CxHy, N2, Ne) will be discussed for the new island geometry. Impurity line radiation tends to concentrate in the islands for lower densities and causes a drop of flux levels correlated to the power loss fraction, Δq Prad/PSOL . β-effects are taken into account based on the 3D MHD-equilibrium code HINT. This work was supported by the U.S. Department of Energy (DOE) under Grant DE-SC0014210.

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    Science.gov (United States)

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

    2014-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-12-15

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  15. Coil Designs for Novel Magnetic Geometries to Cure the Divertor Heat Flux Problem for Reactors

    Science.gov (United States)

    Pekker, M.; Valanju, P.; Kotschenreuther, M.; Wiley, J. C.; Strickler, D.

    2004-11-01

    Coil designs are developed for novel magnetic divertor geometries with a second axi-symmetric x-point and flux expansion region along the separatrix. Adjacent posters describe how these lead to spreading of heat flux and the possibility of stable, complete detachment to overcome serious physics and engineering problems in reactors. The principal feasibility issue is creating, with simple coils, additional X-points on the separatrix without extensively deforming the magnetic field in the main plasma. For the spherical tokamak NSTX, we show that adding one or two poloidal coils suffices to create a divergent flux at the divertor, i.e., a new x-point. The currents and forces for the extra coils are small. We also modify ARIES ST design to show reactor feasibility. Optimized coil designs for PEGASUS, ARIES RS/AT, and a modular ITER retrofit are also being developed. For our calculations we used self consistent code FBEQ, which was used to design NSTX. We also use NCSX tools for optimization of designs with competing physics and engineering constraints.

  16. Quantitative thermal imperfection definition using non-destructive infrared thermography on an advanced DEMO divertor concept

    Science.gov (United States)

    Gallay, F.; Richou, M.; Vignal, N.; Lenci, M.; Roccella, S.; Kermouche, G.; Visca, E.; You, J. H.

    2017-12-01

    The future DEMO divertor is currently under conceptual design within the European Consortium. In this regard, several concepts have been proposed and mock-ups have been fabricated to investigate their thermo-mechanical behaviour. Indeed, as a key plasma facing component, the divertor will have to withstand extreme thermal loads (up to 20 MW m-2 during slow transient events) and will have to be able to exhaust a large amount of heat. The presence of structural defects in the component may significantly affect the thermal response and must therefore be considered. A non-destructive technique based on infrared thermography is proposed here to detect defects in mock-ups where graded material was used as an interlayer between the heatsink material and the armor material. Two methods to characterize the size and location of such defects are presented. It was shown that finite element analysis combined with experimental data from infrared thermography, provides accurate means to assess quantitatively the size and position of thermal imperfections.

  17. US assessment of free surface liquid metal divertors -- Design analysis and R and D needs

    International Nuclear Information System (INIS)

    Mattas, R.F.

    1997-01-01

    One of the objectives of the restructured US Fusion Energy Sciences Program is to identify and evaluate new high performance concepts for advanced technology with high neutron wall load capability and attractive safety and environmental features. One promising technology specified by the Advanced Technologies and Materials Working Group is liquid plasma-facing surfaces for divertors. Some of the possible advantages of using liquid surfaces in divertors, relative to conventional solid surface approaches, include higher surface heat flux capability, continuously renewable surfaces, and higher temperature operation. A planning activity has been undertaken to identify the work to be performed over approximately three years to evaluate liquid surface concepts on the basis of such factors as their compatibility with fusion plasmas, high power density handling capabilities, engineering feasibility, lifetime, safety, and R and D requirements. A group, known as the Advanced Liquid Plasma-facing Surface (ALPS) planning group, was organized to prepare a plan for the activities needed to conduct such an evaluation. This paper will summarize the work of the ALPS group including recommendations on specific activities and a tentative schedule

  18. Operational limits on WEST inertial divertor sector during the early phase experiment

    Science.gov (United States)

    Firdaouss, M.; Corre, Y.; Languille, P.; Greuner, H.; Autissier, E.; Desgranges, C.; Guilhem, D.; Gunn, J. P.; Lipa, M.; Missirlian, M.; Pascal, J.-Y.; Pocheau, C.; Richou, M.; Tsitrone, E.

    2016-02-01

    The primary goal of the WEST project is to be a test bed to characterize the fatigue and lifetime of ITER-like W divertor components subjected to relevant thermal loads. During the first phase of exploitation (S2 2016), these components (W monoblock plasma facing unit—W-PFU) will be installed in conjunction with graphite components (G-PFU). Since the G-PFU will not be actively cooled, it is necessary to ensure the expected pulse duration allows the W-PFU to reach its steady state without overheating the G-PFU assembly structure or the embedded stainless-steel diagnostics. High heat flux tests were performed at the GLADIS facility to assess the thermal behavior of the G-PFU. Some operational limits based on plasma parameters were determined. It was found that it is possible to operate at an injected power such that the maximal incident heat flux on the lower divertor is 10 MW m-2 for the required pulse length.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    International Nuclear Information System (INIS)

    Di Gironimo, G.; Carfora, D.; Esposito, G.; Labate, C.; Mozzillo, R.; Renno, F.; Lanzotti, A.; Siuko, M.

    2013-01-01

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

  1. Impurity re-distribution in the corner regions of the JET divertor

    Science.gov (United States)

    Widdowson, A.; Coad, J. P.; Alves, E.; Baron-Wiechec, A.; Barradas, N. P.; Catarino, N.; Corregidor, V.; Heinola, K.; Krat, S.; Likonen, J.; Matthews, G. F.; Mayer, M.; Petersson, P.; Rubel, M.; Contributors, JET

    2017-12-01

    The International Thermonuclear Experimental Reactor (ITER) will use a mixture of deuterium (D) and tritium (T) as the fuel to generate power. Since T is both radioactive and expensive the Joint European Torus (JET) has been at the forefront of research to discover how much T is used and where it may be retained within the main reaction chamber. Until the year 2010 the JET plasma facing components were constructed of carbon fibre composites. During the JET carbon (C) phases impurities accumulated at the corners of the divertor located towards the bottom of the chamber in regions shadowed from the plasma where they are very difficult to reach and remove. This build-up of C and the associated H-isotope (including T) retention were of particular concern for future fusion reactors therefore, in 2010 JET changed the wall protection to (mainly) Be and the divertor to tungsten (W)—the JET ITER-like wall (ILW)—the choice of materials for ITER. This paper reveals that with the JET ILW impurities are still accumulating in the shadowed regions, with Be being the majority element, though the overall quantities are very much reduced from those in the C phases. Material will be transported into the shadowed regions principally when the plasma strike points are on the corner tiles, but particles typically have about a 75% probability of reflection from line-of sight surfaces, and multiple reflection/scattering results in deposition over all surfaces.

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

    International Nuclear Information System (INIS)

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

    1997-09-01

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

  3. A practical globalization of one-shot optimization for optimal design of tokamak divertors

    Energy Technology Data Exchange (ETDEWEB)

    Blommaert, Maarten, E-mail: maarten.blommaert@kuleuven.be [Institute of Energy and Climate Research (IEK-4), FZ Jülich GmbH, D-52425 Jülich (Germany); Dekeyser, Wouter; Baelmans, Martine [KU Leuven, Department of Mechanical Engineering, 3001 Leuven (Belgium); Gauger, Nicolas R. [TU Kaiserslautern, Chair for Scientific Computing, 67663 Kaiserslautern (Germany); Reiter, Detlev [Institute of Energy and Climate Research (IEK-4), FZ Jülich GmbH, D-52425 Jülich (Germany)

    2017-01-01

    In past studies, nested optimization methods were successfully applied to design of the magnetic divertor configuration in nuclear fusion reactors. In this paper, so-called one-shot optimization methods are pursued. Due to convergence issues, a globalization strategy for the one-shot solver is sought. Whereas Griewank introduced a globalization strategy using a doubly augmented Lagrangian function that includes primal and adjoint residuals, its practical usability is limited by the necessity of second order derivatives and expensive line search iterations. In this paper, a practical alternative is offered that avoids these drawbacks by using a regular augmented Lagrangian merit function that penalizes only state residuals. Additionally, robust rank-two Hessian estimation is achieved by adaptation of Powell's damped BFGS update rule. The application of the novel one-shot approach to magnetic divertor design is considered in detail. For this purpose, the approach is adapted to be complementary with practical in parts adjoint sensitivities. Using the globalization strategy, stable convergence of the one-shot approach is achieved.

  4. Thermomechanical design evaluation and material properties requirements for net divertor elements

    Science.gov (United States)

    Zolti, E.

    1988-07-01

    The major thermomechanical problems of the divertor target plates of the Next European Torus (NET), the procedures and the material data needed for design evaluation are described. As a representative example a preliminary analysis of the divertor plates, which consist of an assembly of one-channel TZM elements, cooled with helium and protected with brazed graphite, is presented. The thermal and mechanical results show, on the one hand, the viability of this concept with respect to peak graphite temperatures, distortion limit to maintain adequate angles with the separatrix, and thermal fatigue and ratchetting of the TZM structure, for a peak surface heat flux of 5 MW/m 2. On the other hand, they contribute to the definition of the material testing programme in terms of further data needs, priorities and test parameter range. Emphasis is put on the fracture mechanics behaviour of the refractory materials under static and dynamic conditions, on the thermal and mechanical properties of intermediate joining layers, on irradiation effects on graphite and carbon fibre composites, and on the relevance of the graphite thermal conductivity in the plasma-to-coolant direction and of its strength under cyclic tensile stresses in the perpendicular direction.

  5. Thermomechanical design evaluation and material properties requirements for NET divertor elements

    International Nuclear Information System (INIS)

    Zolti, E.

    1988-01-01

    The major thermomechanical problems of the divertor target plates of the Next European Torus (NET), the procedures and the material data needed for design evaluation are described. As a representative example a preliminary analysis of the divertor plates, which consist of an assembly of one-channel TZM elements, cooled with helium and protected with brazed graphite, is presented. The thermal and mechanical results show, on the one hand, the viability of this concept with respect to peak graphite temperatures, distortion limit to maintain adequate angles with the separatrix, and thermal fatigue and ratchetting of the TZM structure, for a peak surface heat flux of 5 MW/m 2 . On the other hand, they contribute to the definition of the material testing programme in terms of further data needs, priorities and test parameter range. Emphasis is put on the fracture mechanics behavior of the refractory materials under static and dynamic conditions, on the thermal and mechanical properties of intermediate joining layers, on irradiation effects on graphite and carbon fibre composites, and on the relevance of the graphite thermal conductivity in the plasma-to-coolant direction and of its strength under cyclic tensile stresses in the perpendicular direction. (orig.)

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    Science.gov (United States)

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

    2011-10-01

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

  8. Inter-ELM and ELM-free divertor heat flux broadening induced by EHO in NSTX

    Science.gov (United States)

    Gan, Kaifu; Ahn, Joon-Wook; Maingi, Rajesh; Gray, Travis; Zweben, Stewart; Fredrickson, Eric; Canal, Gustavo; Wirth, Brian

    2016-10-01

    Recent study on multi-machine database of inter-ELM divertor heat flux indicates that the midplane Scrape-off Layer power fall-off length for ITER is expected to be very narrow, 1mm, which will induce small plasma-wetted area (Awet) and high peak heat flux (qpeak) . In NSTX, edge harmonic oscillation (EHO) was observed during certain inter-ELM and ELM-free periods of H-mode operation. The EHO is observed to significantly increase Awet, by up to a factor of 3, and decrease qpeak accordingly. The Awet increases with the amplitude of EHO. Multiple peaks appeared in the heat flux profile in the presence of EHO. An EHO-induced filament around separatrix rotating in the counter-current direction was also observed by gas puff imaging diagnostic. The increase of Awet is suspected to be caused by the rotating current filaments; the toroidally rotating filaments could have changed the edge magnetic topology and broadened heat flux profile by strike points splitting. Experimental observation of the temperal evolution of multiple peaks in heat flux profile shows consistent trend with the toroidal rotation of EHO. Inter-ELM and ELM-free divertor heat flux broadening induced by EHO in NSTX.

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

    International Nuclear Information System (INIS)

    Brossa, M.; Ciric, D.; Deksnis, E.; Falter, H.; Guerreschi, U.; Peacock, A.; Pick, M.; Rossi, M.; Shen, Y.; Zacchia, F.

    1995-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-09-01

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

  11. Electron temperature and heat load measurements in the COMPASS divertor using the new system of probes

    Science.gov (United States)

    Adamek, J.; Seidl, J.; Horacek, J.; Komm, M.; Eich, T.; Panek, R.; Cavalier, J.; Devitre, A.; Peterka, M.; Vondracek, P.; Stöckel, J.; Sestak, D.; Grover, O.; Bilkova, P.; Böhm, P.; Varju, J.; Havranek, A.; Weinzettl, V.; Lovell, J.; Dimitrova, M.; Mitosinkova, K.; Dejarnac, R.; Hron, M.; The COMPASS Team; The EUROfusion MST1 Team

    2017-11-01

    A new system of probes was recently installed in the divertor of tokamak COMPASS in order to investigate the ELM energy density with high spatial and temporal resolution. The new system consists of two arrays of rooftop-shaped Langmuir probes (LPs) used to measure the floating potential or the ion saturation current density and one array of Ball-pen probes (BPPs) used to measure the plasma potential with a spatial resolution of ~3.5 mm. The combination of floating BPPs and LPs yields the electron temperature with microsecond temporal resolution. We report on the design of the new divertor probe arrays and first results of electron temperature profile measurements in ELMy H-mode and L-mode. We also present comparative measurements of the parallel heat flux using the new probe arrays and fast infrared termography (IR) data during L-mode with excellent agreement between both techniques using a heat power transmission coefficient γ  =  7. The ELM energy density {{\\varepsilon }\\parallel } was measured during a set of NBI assisted ELMy H-mode discharges. The peak values of {{\\varepsilon }\\parallel } were compared with those predicted by model and with experimental data from JET, AUG and MAST with a good agreement.

  12. Thermal fatigue damage in monofilament reinforced copper for heat sink applications in divertor elements

    Science.gov (United States)

    Schöbel, M.; Jonke, J.; Degischer, H. P.; Paffenholz, V.; Brendel, A.; Wimpory, R. C.; Di Michiel, M.

    2011-02-01

    In fusion reactor systems extreme conditions require materials with high temperature and radiation resistance. The divertor component consists of a plasma facing W plate attached to a Cu heat sink to extract the heat from the nuclear reaction chamber coolant. The Coefficient of Thermal Expansion (CTE) mismatch between the W plate and the Cu heat sink causes interface delamination reducing the long term stability of the divertor. To avert this problem, composites are developed as interlayer materials with a high thermal conducting Cu matrix reinforced with up to 50 vol.% SiC or W monofilaments to increase the mechanical strength and to reduce the CTE mismatch. Thermal stresses are transferred from the macroscopic interface between the components into the bulk of the composite. Oscillating micro stresses may lead to fiber delamination and matrix damage during thermal cycling. Different matrix alloys, fiber materials and interface designs are investigated. In situ neutron diffraction performed during thermal cycling show the effect of bonding strength on the stress amplitudes expected under service conditions. The long term stability is tested by measurements after further ex situ cycling. Thermal fatigue damage and its propagation are visualized by in situ as well as ex situ high resolution synchrotron tomography. The combination of both methods helps to understand the strain induced damage mechanisms. Weak bonding leads to delamination of the fiber-matrix interfaces. Strong bonding causes severe matrix deformation and damage. Fiber cracks originating from sample production cause accumulating thermal fatigue damage during thermal cycling.

  13. Maximizing Heat Dissipation via Target Optimization of the Small-Angle Slot Divertor

    Science.gov (United States)

    Covele, Brent; Halpern, Federico; Casali, Livia; Canik, John; Thomas, Dan; Guo, Houyang

    2017-10-01

    The planned SAS 2 divertor uses a combination of grazing target angles and closure to direct recycling neutrals near the strike point, thus facilitating detachment onset. SAS 2 should also provide adequate pumping efficiency to be consistent with high-power steady-state scenarios on DIII-D. Initial SOLPS results indicate significantly higher neutral densities and lower electron temperatures in the SAS 2 slot, compared to a closed reference divertor model with baseline plasma profiles appropriate for high power. A systematic optimization of the parameterized SAS 2 target shape is performed in SOLPS to further reduce target heat fluxes and temperatures at lowest upstream density. To speed up the target optimization process, target neutral densities calculated by Eirene act as a performance metric by proxy for detachment facilitation. The efficacy of this proxy metric is discussed. Results are also presented from SAS 2 neutral pumping simulations in Eirene with a stationary background plasma. The feasibility of mutually satisfactory particle control and detachment control is discussed. Work supported under USDOE Cooperative Agreements DE-FC02-04ER54698 and DE-AC05-00OR22725.

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  15. Comparing scrape-off layer and divertor physics in JET pure He and D discharges

    Energy Technology Data Exchange (ETDEWEB)

    Pitts, R.A. E-mail: richard.pitts@epfl.ch; Andrew, P.; Andrew, Y.; Becoulet, M.; Coffey, I.; Coster, D.; McDonald, D.C.; Eich, T.; Erents, S.K.; Fenstermacher, M.E.; Fundamenski, W.; Haas, G.; Hermann, A.; Hidalgo, C.; Hillis, D.; Huber, A.; Ingesson, L.C.; Jachmich, S.; Kallenbach, A.; Korotkov, A.; Lawson, K.; Lomas, P.; Loarer, T.; Loarte, A.; Matthews, G.F.; McCracken, G.; Meigs, A.; Mertens, Ph.; O' Mullane, M.; Philipps, V.; Porter, G.; Pospieszczyk, A.; Rapp, J.; Reiter, D.; Riccardo, V.; Saibene, G.; Sartori, R.; Stamp, M.F.; Tsitrone, E.; Wischmeier, M.; Gafert, J

    2003-03-01

    Though helium plasmas are one option for the low activation phase of ITER, little effort has thus far been devoted to studying them in a large, diverted tokamak. A recent campaign on JET has therefore sought to address some of the important questions related to helium operation (He concentrations near 90%) in single null configurations, particularly with regard to edge and divertor physics. This contribution compiles a selection of results from these experiments, in which, in each case, discharges have been chosen to match as closely as possible previous, well characterised D plasmas in both L and ELMing H-modes. These matched pulses are used to draw conclusions regarding the principle source and location of carbon production in D plasmas, to compare and contrast the mechanisms of the density limit and the detachment process in D and He, to investigate the nature of cross-field power transport in the SOL and to gain insight into the process by which ELM energy is transported to the divertor targets.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  18. Numerical study of the connection lengths for various magnetic configurations in Wendelstein 7-X to optimize the heat load on the divertor

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, Priyanjana; Hoelbe, Hauke; Sunn Pedersen, Thomas [Max Planck Institute of Plasma Physics, Greifswald (Germany)

    2016-07-01

    Fusion has the potential to play an important role as a future energy resource. It has the capacity to produce large-scale clean energy. The two main confinement concepts are the tokamak and the stellarator. The W7-X machine is based on stellarator principle and is using special form of coils to achieve steady-state plasma confinement. Divertors are used in tokamaks and stellarator to control the exhaust of waste gases and impurities from the machine. The divertor concept of W7-X is a so-called island divertor. The island chain isolates the confinement core from regions where the plasma-wall interaction takes place. The area of the divertor that receives the main part of the heat loads, the so-called wetted area, increases with the distance along the magnetic field from the outboard midplane to the divertor target. The connection length is relatively short in tokamaks with conventional divertors. In the stellarator island divertor, the connection length can be varied significantly, which should allow for optimization of the wetted area. We present here a numerical study of the achievable connection lengths in various W7-X configurations and discuss the possibilities for running dedicated experiments to understand the physics of what sets the wetted area.

  19. Real-time control of divertor detachment in H-mode with impurity seeding using Langmuir probe feedback in JET-ITER-like wall

    Science.gov (United States)

    Guillemaut, C.; Lennholm, M.; Harrison, J.; Carvalho, I.; Valcarcel, D.; Felton, R.; Griph, S.; Hogben, C.; Lucock, R.; Matthews, G. F.; Perez Von Thun, C.; Pitts, R. A.; Wiesen, S.; contributors, JET

    2017-04-01

    Burning plasmas with 500 MW of fusion power on ITER will rely on partially detached divertor operation to keep target heat loads at manageable levels. Such divertor regimes will be maintained by a real-time control system using the seeding of radiative impurities like nitrogen (N), neon or argon as actuator and one or more diagnostic signals as sensors. Recently, real-time control of divertor detachment has been successfully achieved in Type I ELMy H-mode JET-ITER-like wall discharges by using saturation current (I sat) measurements from divertor Langmuir probes as feedback signals to control the level of N seeding. The degree of divertor detachment is calculated in real-time by comparing the outer target peak I sat measurements to the peak I sat value at the roll-over in order to control the opening of the N injection valve. Real-time control of detachment has been achieved in both fixed and swept strike point experiments. The system has been progressively improved and can now automatically drive the divertor conditions from attached through high recycling and roll-over down to a user-defined level of detachment. Such a demonstration is a successful proof of principle in the context of future operation on ITER which will be extensively equipped with divertor target probes.

  20. Quartz micro-balance results of pulse-resolved erosion/deposition in the JET-ILW divertor

    Directory of Open Access Journals (Sweden)

    G. Sergienko

    2017-08-01

    Full Text Available A set of quartz crystal microbalances (QMB was used at JET with full carbon wall to monitor mass erosion/deposition rates in the remote areas of the divertor. After introduction of the ITER- like wall (ILW in JET with beryllium main wall and tungsten divertor, strong reduction of the material deposition and accompanied fuel retention was observed. Therefore the existing QMB electronics have been modified to improve the accuracy of frequency measurements by a factor of ten down to 0.1Hz which corresponds to 1.4ngcm−2. The averaged deposition rates of 1.2–3ngcm−2s−1 and erosion rates of 5.6–8.1ngcm−2s−1 were observed in the inner divertor of JET -ILW with the inner strike point positions close to the bottom edge of vertical tile 3 and at the horizontal tile 4 respectively. The erosion with averaged rates of ≈2.1ngcm−2s−1 and ≈120ngcm−2s−1 were observed in the outer divertor for the outer strike point positions at tile 5 and tile 6 respectively.

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

    International Nuclear Information System (INIS)

    Di Gironimo, G.; Carfora, D.; Esposito, G.; Lanzotti, A.; Marzullo, D.; Siuko, M.

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    Asakura, Nobuyuki

    2007-01-01

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

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

    International Nuclear Information System (INIS)

    Chen Lei; Liu Xiang; Lian Youyun; Cai Laizhong

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    Litnovsky, A.; Rudakov, D.L.; De Temmerman, G.; Wienhold, P.; Philipps, V.; Samm, U.; McLean, A.G.; West, W.P.; Wong, C.P.C.; Brooks, N.H.; Watkins, J.G.; Wampler, W.R.; Stangeby, P.C.; Boedo, J.A.; Moyer, R.A.; Allen, S.L.; Fenstermacher, M.E.; Groth, M.; Lasnier, C.J.; Boivin, R.L.

    2008-01-01

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

  5. Numerical evaluation of heat flux and surface temperature on a misaligned JET divertor W lamella during ELMs

    Czech Academy of Sciences Publication Activity Database

    Dejarnac, Renaud; Podolník, Aleš; Komm, Michael; Arnoux, G.; Coenen, J.W.; Devaux, S.; Frassinetti, L.; Gunn, J. P.; Matthews, G. F.; Pitts, R.A.

    2014-01-01

    Roč. 54, č. 12 (2014), s. 123011-123011 ISSN 0029-5515 Institutional support: RVO:61389021 Keywords : PIC simulations * power flux * plasma–wall interactions * JET * divertor Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.062, year: 2014 http://iopscience.iop.org/0029-5515/54/12/123011/pdf/0029-5515_54_12_123011.pdf

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

    NARCIS (Netherlands)

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

    2015-01-01

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

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

    Science.gov (United States)

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

    2018-05-01

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

  8. Study on divertor particle and heat fluxes from electric probe measurements during ELMy H-modes in KSTAR

    Energy Technology Data Exchange (ETDEWEB)

    Bak, Jun-Gyo, E-mail: jgbak@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, Heung-Su [National Fusion Research Institute, Daejeon (Korea, Republic of); Bae, Min-Keun; Chung, Kyu-Sun [Hanyang University, Seoul (Korea, Republic of); Hong, Suk-Ho [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2016-11-01

    Highlights: • The characteristics of the particle and heat fluxes were investigated during ELMs in H-modes under the LSN configuration in the KSTAR tokamak.. • There was relation between the ELM amplitude and the ELM frequency as ΔW{sub ELM}/W{sub TOT} ∝ 1/f{sub ELM} in the range of f{sub ELM} ≤ 200 Hz. • The trends of the peak amplitude of the divertor flux near the OSP during ELMs due to the ELM mitigation and the plasma shaping were investigated. • The ELMs were mitigated by MP field, SMBI and ECH. The ELM mitigations due to the MP field and the SMBI were stronger than one due to the ECH. • Finally, the particle flux, evaluated at the far scrape-off layer (SOL) region, was estimated to less than 1% of the divertor particle flux. - Abstract: The characteristics of the divertor particle and heat fluxes are investigated during ELM bursts in ELMy H-mode plasmas with the lower single null (LSN) configuration in Korea Superconducting Tokamak Advanced Research (KSTAR). The particle and heat fluxes are evaluated from the electric probe measurements at the divertor region. It is found that the peak amplitude of the divertor flux during an ELM burst obtained near the outer strike point (OSP) decreases up to about 20% as the ELM frequency increases by a factor of ∼6.5 due to the ELM mitigation and the plasma shaping, which is similar to the trend of the amplitude versus the frequency of the ELM observed in other tokamaks. The ELMs are mitigated by using sev