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Sample records for experiment detached divertor

  1. 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.

  2. Divertor Heat Flux Reduction and Detachment in the National Spherical Torus eXperiment.

    Science.gov (United States)

    Soukhanovskii, Vsevolod

    2007-11-01

    Steady-state handling of the heat flux is a critical divertor issue for both the International Thermonuclear Experimental Reactor and spherical torus (ST) devices. Because of an inherently compact divertor, it was thought that ST-based devices might not be able to fully utilize radiative and dissipative divertor techniques based on induced power and momentum loss. However, initial experiments conducted in the National Spherical Torus Experiment in an open geometry horizontal carbon plate divertor using 0.8 MA 2-6 MW NBI-heated lower single null H-mode plasmas at the lower end of elongations κ=1.8-2.4 and triangularities δ=0.45-0.75 demonstrated that high divertor peak heat fluxes, up to 6-10 MW/ m^2, could be reduced by 50-75% using a high-recycling radiative divertor regime with D2 injection. Furthermore, similar reduction was obtained with a partially detached divertor (PDD) at high D2 injection rates, however, it was accompanied by an X-point MARFE that quickly led to confinement degradation. Another approach takes advantage of the ST relation between strong shaping and high performance, and utilizes the poloidal magnetic flux expansion in the divertor region. Up to 60 % reduction in divertor peak heat flux was achieved at similar levels of scrape-off layer power by varying plasma shaping and thereby increasing the outer strike point (OSP) poloidal flux expansion from 4-6 to 18-22. In recent experiments conducted in highly-shaped 1.0-1.2 MA 6 MW NBI heated H-mode plasmas with divertor D2 injection at rates up to 10^22 s-1, a PDD regime with OSP peak heat flux 0.5-1.5 MW/m^2 was obtained without noticeable confinement degradation. Calculations based on a two point scrape-off layer model with parameterized power and momentum losses show that the short parallel connection length at the OSP sets the upper limit on the radiative exhaust channel, and both the impurity radiation and large momentum sink achievable only at high divertor neutral pressures are required

  3. 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.

  4. 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   ×  \

  5. A new scaling for divertor detachment

    Science.gov (United States)

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

    2017-05-01

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

  6. Modeling detachment physics in the NSTX snowflake divertor

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

  7. Divertor characterization experiments

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

  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. Heat flux management via advanced magnetic divertor configurations and divertor detachment

    Energy Technology Data Exchange (ETDEWEB)

    Kolemen, E., E-mail: ekolemen@princeton.edu [Princeton University, Princeton, NJ 08544 (United States); Allen, S.L. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Bray, B.D. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Fenstermacher, M.E. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Humphreys, D.A.; Hyatt, A.W. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Lasnier, C.J. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Leonard, A.W. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Makowski, M.A.; McLean, A.G. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Maingi, R.; Nazikian, R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Petrie, T.W. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Unterberg, E.A. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States)

    2015-08-15

    The snowflake divertor (SFD) control and detachment control to manage the heat flux at the divertor are successfully demonstrated at DIII-D. Results of the development and implementation of these two heat flux reduction control methods are presented. The SFD control algorithm calculates the position of the two null-points in real-time and controls shaping coil currents to achieve and stabilize various snowflake configurations. Detachment control stabilizes the detachment front fixed at specified distance between the strike point and the X-point throughout the shot.

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

    International Nuclear Information System (INIS)

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

    1999-06-01

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

  12. T-12 divertor experiment

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-01-01

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

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

    Directory of Open Access Journals (Sweden)

    H.Q. Wang

    2017-08-01

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

  14. Spectroscopic investigations of divertor detachment in TCV

    Directory of Open Access Journals (Sweden)

    K. Verhaegh

    2017-08-01

    The inferred magnitude of recombination is small compared to the target ion current at the time detachment (particle flux drop starts at the target. However, recombination may be having more localized effects (to a flux tube which we cannot discern at this time. Later, at the highest densities achieved, the total recombination does reach levels similar to the particle flux.

  15. X-Divertor Geometries for Deeper Detachment Without Degrading the DIII-D H-Mode

    Science.gov (United States)

    Covele, Brent; Kotschenreuther, M. T.; Valanju, P. M.; Mahajan, S. M.; Leonard, A. W.; Hyatt, A. W.; McLean, A. G.; Thomas, D. M.; Guo, H. Y.; Watkins, J. G.; Makowski, M. A.; Hill, D. N.

    2015-11-01

    Recent DIII-D experiments comparing the standard divertor (SD) and X-Divertor (XD) geometries show heat and particle flux reduction at the divertor target plate. The XD features large poloidal flux expansion, increased connection length, and poloidal field line flaring, quantified by the Divertor Index. Both SD and XD were pushed deep into detachment with increased gas puffing, until core energy confinement and pedestal pressure were substantially reduced. As expected, outboard target heat fluxes are significantly reduced in the XD compared to the SD under similar upstream plasma conditions, even at low Greenwald fraction. The high-triangularity (floor) XD cases show larger reduction in temperature, heat, and particle flux relative to the SD in all cases, while low-triangularity (shelf) XD cases show more modest reductions over the SD. Consequently, heat flux reduction and divertor detachment may be achieved in the XD with less gas puffing and higher pedestal pressures. Further causative analysis, as well as detailed modeling with SOLPS, is underway. These initial experiments suggest the XD as a promising candidate to achieve divertor heat flux control compatible with robust H-mode operation. Work supported by US DOE under DE-FC02-04ER54698, DE-AC52-07NA27344, DE-FG02-04ER54754, and DE-FG02-04ER54742.

  16. Snowflake divertor configuration studies in National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-08-15

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

  17. 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)

  18. Versator divertor experiment: preliminary designs

    International Nuclear Information System (INIS)

    Wan, A.S.; Yang, T.F.

    1984-08-01

    The emergence of magnetic divertors as an impurity control and ash removal mechanism for future tokamak reactors bring on the need for further experimental verification of the divertor merits and their ability to operate at reactor relevant conditions, such as with auxiliary heating. This paper presents preliminary designs of a bundle and a poloidal divertor for Versator II, which can operate in conjunction with the existing 150 kW of LHRF heating or LH current drive. The bundle divertor option also features a new divertor configuration which should improve the engineering and physics results of the DITE experiment. Further design optimization in both physics and engineering designs are currently under way

  19. Numerical studies on divertor experiments

    International Nuclear Information System (INIS)

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

    1988-04-01

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

  20. Assessment of X-point target divertor configuration for power handling and detachment front control

    Directory of Open Access Journals (Sweden)

    M.V. Umansky

    2017-08-01

    Full Text Available A study of long-legged tokamak divertor configurations is performed with the edge transport code UEDGE (Rognlien et al., J. Nucl. Mater. 196, 347, 1992. The model parameters are based on the ADX tokamak concept design (LaBombard et al., Nucl. Fusion 55, 053020, 2015. Several long-legged divertor configurations are considered, in particular the X-point target configuration proposed for ADX, and compared with a standard divertor. For otherwise identical conditions, a scan of the input power from the core plasma is performed. It is found that as the power is reduced to a threshold value, the plasma in the outer leg transitions to a fully detached state which defines the upper limit on the power for detached divertor operation. Reducing the power further results in the detachment front shifting upstream but remaining stable. At low power the detachment front eventually moves to the primary X-point, which is usually associated with degradation of the core plasma, and this defines the lower limit on the power for the detached divertor operation. For the studied parameters, the operation window for a detached divertor in the standard divertor configuration is very small, or even non-existent; under the same conditions for long-legged divertors the detached operation window is quite large, in particular for the X-point target configuration, allowing a factor of 5–10 variation in the input power. These modeling results point to possibility of stable fully detached divertor operation for a tokamak with extended divertor legs.

  1. 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

  2. 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.)

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

    Science.gov (United States)

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

    2018-05-01

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

  4. 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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  6. Advanced divertor experiments on DIII-D

    International Nuclear Information System (INIS)

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

    1991-04-01

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

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

    Directory of Open Access Journals (Sweden)

    T.D. Rognlien

    2017-08-01

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

  8. 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.

  9. 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.

  10. Effect of transport on MAR in detached divertor plasma

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  11. First results of closed helical divertor experiment in LHD

    International Nuclear Information System (INIS)

    Morisaki, T.; Masuzaki, S.; Kobayashi, M.

    2012-11-01

    The baffle-structured closed Helical Divertor (CHD) is being constructed in LHD to actively control the edge plasma, which consists of ten discrete modules installed on inboard side of the torus. At this stage, two of ten modules have been constructed. In the initial experiments, performance of CHD was experimentally investigated, comparing with numerical expectations. During the continuous gas puffing discharge, it was observed the neutral pressure in the CHD was more than 10 times higher than that in the open HD, which agrees well with the numerical simulation. In the high density regime, indication of the divertor detachment was observed in CHD, which was caused by the high recycling and high density state in CHD. With a Penning discharge diagnostics, the neutral particle behaviour with different species was investigated. Little difference between hydrogen and helium was observed in transport property. (author)

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  13. The role of ''momentum removal'' in divertor detachment

    Energy Technology Data Exchange (ETDEWEB)

    Kukushkin, A.S. [Kurchatov Institute, Moscow (Russian Federation); NRNU MEPhI, Moscow (Russian Federation); Pacher, H.D. [INRS-EMT, Varennes, Quebec (Canada)

    2016-08-15

    The role of ''momentum removal'' (the drag force on the plasma ion flow) in divertor detachment is considered and analysed in detail. This analysis of the 2D modelling results shows that the drag force cannot reduce the power and particle flux to the target directly. However, it is essential for creating the conditions for efficient radiation and volumetric plasma recombination, which in turn do the job. (copyright 2015 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  15. 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

  16. Snowflake divertor experiments on TCV

    International Nuclear Information System (INIS)

    Piras, F; Coda, S; Duval, B P; Labit, B; Marki, J; Moret, J-M; Pitzschke, A; Sauter, O; Medvedev, S Yu

    2010-01-01

    An ELMy H-mode 'snowflake' (SF) divertor is established and studied for the first time in the TCV tokamak. The H-mode access and the edge localized mode (ELM) dynamics are compared with a conventional single-null diverted configuration. The SF configuration exhibits 15% higher confinement and 2-3 times lower ELM frequency. Ideal MHD stability analysis suggests enhanced stability of the SF H-mode pedestal to mid- to high-toroidal-mode-number modes. The capability of the SF to redistribute the edge power on the additional strike points has been confirmed experimentally.

  17. 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

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

    International Nuclear Information System (INIS)

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

    1996-10-01

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

  19. Electron beam irradiation experiments of monoblock divertor mock-up

    International Nuclear Information System (INIS)

    Satoh, Kazuyoshi; Akiba, Masato; Araki, Masanori; Suzuki, Satoshi; Yokoyama, Kenji; Smid, I.; Cardella, A.; Duwe, R.; Di Pietro, E.

    1993-03-01

    It is one of the key issues for ITER to develop the divertor plate. Electron beam irradiation tests were carried out on a NET divertor mock-up using JEBIS at JAERI under a collaboration between The NET team, JAERI and KFA Juelich. Screening tests (maximum heat flux of 23 MW/m 2 ) and thermal cycling tests (18 MW/m 2 , 30s, 1000cycle) were carried out. As a result of the screening tests, the erosion caused by sublimation of C/C was observed on the surface of armor tile. No serious damage such as cracks or detachments, however, were found. As a result of the thermal cycling tests, no major damage was detected on the C/C surface. However cooling time constant of the divertor mock-up increased over 600cycle. Therefore it implies that some defects would occur at the brazing interface of the divertor mock-up. (author)

  20. 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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  2. 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

  3. 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

  4. Modelling the detachment dependence on strike point location in the small angle slot divertor (SAS) with SOLPS

    Science.gov (United States)

    Casali, Livia; Covele, Brent; Guo, Houyang

    2017-10-01

    The new Small Angle Slot (SAS) divertor in DIII-D is characterized by a shallow-angle target enclosed by a slot structure about the strike point (SP). SOLPS modelling results of SAS have demonstrated divertor closure's utility in widening the range of acceptable densities for adequate heat handling. An extensive database of runs has been built to study the detachment dependence on SP location in SAS. Density scans show that lower Te at lower upstream density occur when the SP is at the critical location in the slot. The cooling front spreads across the entire target at higher densities, in agreement with experimental Langmuir probe measurements. A localized increase of the atomic and molecular density takes place near the SP, which reduces the target incident power density and facilitates detachment at lower upstream density. Systematic scans of variables such as power, transport, and viscosity have been carried out to assess the detachment sensitivity. Therein, a positive role of the viscosity is found. This work supported by DOE Contract Number DE-FC02-04ER54698.

  5. 2D statistical analysis of Non-Diffusive transport under attached and detached plasma conditions of the linear divertor simulator

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  6. Status of National Spherical Torus Experiment Liquid Lithium Divertor

    Science.gov (United States)

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

    2009-11-01

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

  7. 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

  8. 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.

  9. Electron beam facility for divertor target experiments

    International Nuclear Information System (INIS)

    Anisimov, A.; Gagen-Torn, V.; Giniyatulin, R.N.

    1994-01-01

    To test different concepts of divertor targets and bumpers an electron beam facility was assembled in Efremov Institute. It consists of a vacuum chamber (3m 3 ), vacuum pump, electron beam gun, manipulator to place and remove the samples, water loop and liquid metal loop. The following diagnostics of mock-ups is stipulated: (1) temperature distribution on the mock-up working surface (scanning pyrometer and infra-red imager); (2) temperature distribution over mocked-up thickness in 3 typical cross-sections (thermo-couples); (3) cracking dynamics during thermal cycling (acoustic-emission method), (4) defects in the mock-up before and after tests (ultra-sonic diagnostics, electron and optical microscopes). Carbon-based and beryllium mock-ups are made for experimental feasibility study of water and liquid-metal-cooled divertor/bumper concepts

  10. Divertor experiment for impurity control in DIVA

    International Nuclear Information System (INIS)

    Nagami, Masayuki

    1979-04-01

    Divertor actions of controlling the impurities and the transport of impurity ions in the plasma have been investigated in the DIVA device. Following are the results: (1) The radial transport of impurity ions is not described only by neoclassical theory, but it is strongly influenced by anomalous process. Radial diffusion of impurity ions across the whole minor radius is well described by a neoclassical diffusion superposed by the anomalous diffusion for protons. Due to this anomalous process, which spreads the radial density profile of impurity ions, 80 to 90% of the impurity flux in the plasma outer edge is shielded even in a nondiverted discharge. (2) The divertor reduces the impurity flux entering the main plasma by a factor of 2 to 4. The impurity ions shielded by the scrape-off plasma are rapidly guided into the burial chamber with a poloidal excursion time roughly equal to that of the scrape-off plasma. (3) The divertor reduces the impurity ion flux onto the main vacuum chamber by guiding the impurity ions diffusing from the main plasma into the burial chamber, thereby reducing the plasma-wall interaction caused by diffusing impurity ions at the main vacuum chamber. The impurity ions produced in the burial chamber may flow back to the main plasma through the scrape-off layer. However, roughly only 0.3% of the impurity flux into the scrape-off plasma in the burial chamber penetrates into the main plasma due to the impurity backflow. (4) A slight cooling of the scrape-off plasma with light-impurity injection effectively reduces the metal impurity production at the first wall by reducing the potential difference between the plasma and the wall, thereby reducing the accumulation of the metal impurity in the discharge. Radiation cooling by low-Z impurities in the plasma outer edge, which may become an important feature in future large tokamaks both with and without divertor, is numerically evaluated for carbon, oxygen and neon. (author)

  11. 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

  12. 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)

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  14. Divertor heat flux mitigation in the National Spherical Torus Experimenta)

    Science.gov (United States)

    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.; NSTX Team

    2009-02-01

    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 4-6MWm-2to0.5-2MWm-2 in small-ELM 0.8-1.0MA, 4-6MW neutral beam injection-heated H-mode discharges. A self-consistent picture of the 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.

  15. 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

  16. Engineering of the divertor injection tokamak experiment (DITE)

    International Nuclear Information System (INIS)

    Plummer, K.M.; Bayes, D.V.; Bell, D.; Burt, J.; Galloway, F.; Sanders, B.C.; Skelton, D.E.; Varley, G.L.

    1976-01-01

    The DITE assembly has been constructed to study the effect of powerful neutral injection and the use of magnetic divertors in Tokamak systems. In addition, the plasma is stabilized by a position controlled feed-back vertical field system developed from results on the CLEO experiment, and added to DITE later in the design stage. The machine is designed for an ultimate plasma current of 340 kA, having a minor radius of 23 cm at q = 2, on a major radius of 113 cm. The 28 kG Bphi field, from 16 liquid nitrogen cooled coils has a 2% ripple at the edge of the plasma. The divertor is a ''bundle'' type, the present design of which is limited to operating in a Bphi field of 18 kG. Neutral Injection, initially by two, and ultimately by four injectors, is intended to supply about 1,500 kW of beam power. The engineering is now complete and the machine commissioned; this paper describes the up-to-date design of the machine and includes some of our experiences during design, construction and commissioning

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

    International Nuclear Information System (INIS)

    Nakamura, Hiroo; Tsuji, Shunji; Nagami, Masayuki

    1989-08-01

    In JT-60, hydrogen H-mode experiments with outer and lower divertors were performed. In the outer divertor, H-mode were obtained, similar to the ones observed in the other lower/upper divertors. Its threshold absorbed power and electron density were 16 MW and 1.8 x 10 19 m -3 . In the two combined heatings with NB+ICRF and NB+LHRF, H-mode discharges are also obtained. Moreover, in new configuration of lower divertor, H-mode phases without and with ELM are obtained. Typical results of the lower divertor are shown to compare the H-mode characteristics between the two configurations. Improvement of the energy confinement time in the two divertors was limited to 10 %. Analyses on ballooning/interchange instabilities were carried out with precise equlibria of JT-60. These results showed that the both modes were enough stable. (author)

  18. Divertor scenario development for NSTX Upgrade

    Science.gov (United States)

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

    2012-10-01

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

  19. Advanced divertor configurations with large flux expansion

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-15

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

  20. 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

  1. Plasma decontamination during ergodic divertor experiments in TORE SUPRA

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  2. 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

  3. 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.

  4. Theory of Advanced Magnetic Divertors

    Science.gov (United States)

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

    2013-10-01

    The magnetic field structure in the SOL is the most important determinant of divertor physics. A comprehensive analytical and numerical methodology is developed to investigate SOL magnetic fields in the backdrop of two advanced divertor geometries- the X-divertor (XD) proposed and discussed in 2004, and the snowflake divertor (SFD) of 2007-2010. The analysis shows that XD and SFD represent very distinct and readily distinguishable magnetic geometries, epitomized through a differentiating metric, the Divertor Index (DI). In terms of this simple metric, the XD (DI > 1) and the SFD (DI XD flux surfaces are less convergent, in fact, divergent (flaring). These different SOL magnetics imply different physics, particularly with respect to detachment dynamics. It is also shown that some experiments on NSTX and DIII-D match both the prescription and the predictions of the 2004 XD paper. Work supported under US-DOE projects DE-FG02-04ER54742 and DE-FG02-04ER54754.

  5. 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.)

  6. DOE FES FY2017 Joint Research Target Fourth Quarter Milestone Report for theNational Spherical Torus Experiment Upgrade.

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-09-13

    A successful high-performance plasma operation with a radiative divertor has been demonstrated on many tokamak devices, however, significant uncertainty remains in accurately modeling detachment thresholds, and in how detachment depends on divertor geometry. Whereas it was originally planned to perform dedicated divertor experiments on the National Spherical Tokamak Upgrade to address critical detachment and divertor geometry questions for this milestone, the experiments were deferred due to technical difficulties. Instead, existing NSTX divertor data was summarized and re-analyzed where applicable, and additional simulations were performed.

  7. Modelling of island divertor physics and comparison to W7-AS experimental results

    International Nuclear Information System (INIS)

    Feng, Y.; Sardei, F.; Grigull, P.; McCormick, K.; Giannone, L.; Kisslinger, J.; Reiter, D.; Igitkhanov, Y.; Wenzel, U.

    2003-01-01

    Extensive parameter studies have been carried out with the EMC3-EIRENE code. Major code predictions, namely the absence of high recycling prior to detachment, additional momentum losses associated with the specific island divertor geometry and the jump of the radiation at detachment transition have been verified by the W7-AS divertor experiments. Measurements and simulations are compared for high density, high power W7-AS divertor discharges and the physics related to rollover and detachment is discussed in detail. Local comparisons with the W7-AS experiment have been started with a new code version accounting for the real open-island geometry. Specifically, the observed asymmetric power unloading of the target plates at detachment transition could now be reproduced and explained. Agreement with the experiment was also found for the unexpected spatial structure of particle deposition by including classical ExB drifts into the code

  8. 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).

  9. Overview of the modification to the poloidal divertor experiment (PDX) to produce the Princeton beta experiment (PBX)

    International Nuclear Information System (INIS)

    Knutson, D.

    1984-01-01

    The Poloidal Divertor Experiment at the Princeton Plasma Physics Laboratory has been recently transformed into the Princeton Beta Experiment. The purpose of the modification is to produce a bean-shaped plasma with beta values in excess of 10%, which is substantially above those achieved with more conventional plasma shapes. (author)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  11. Neutral particle retention in the JET MK I divertor

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  12. Plasma flow in the DIII-D divertor

    International Nuclear Information System (INIS)

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

    1998-07-01

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

  13. Scrape-off layer radiation and heat load to the ASDEX Upgrade LYRA divertor

    International Nuclear Information System (INIS)

    Kallenbach, A.; Kaufmann, M.; Coster, D.P.

    1999-01-01

    In 1997 the new 'LYRA' divertor went into operation at ASDEX Upgrade and, in parallel, the neutral beam heating power was increased to 20 MW by installation of a second injector leading to a P/R value of 12 MW/m. Experiments have shown that the ASDEX Upgrade LYRA divertor is capable of handling such high heating powers. There is an overall reduction of the maximum heat flux in the LYRA divertor by about a factor of 2 compared with the previous open divertor Div I. This reduction is mainly due to increased radiative losses inside the divertor region, which are caused by an effective reflection of hydrogen neutrals into the hot separatrix region. The main channel of radiative loss is carbon radiation, which cools the divertor plasma down to a few electronvolts, where hydrogen radiation losses become significant. The radiative losses preferentially reduce the power flux at the separatrix, leading to early detachment around the strike point position. With increasing density, the detached region extends upwards on the vertical target. The power fraction radiated in the LYRA divertor is around 45% and nearly independent of the heating power. This value is a factor of 2 higher than the typical radiation fraction in Div I. B2-EIRENE modelling of the performed experiments supports the experimental finding and refines the understanding of loss processes in the divertor region. (author)

  14. Snowflake Divertor Configuration in NSTX

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  15. 'Snowflake' divertor configuration in NSTX

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  16. "Snowflake" divertor configuration in NSTX

    Science.gov (United States)

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

    2011-08-01

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

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

    International Nuclear Information System (INIS)

    Strait, E.J.

    1979-09-01

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

  18. 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)

  19. An analytical erosion model for divertor plates and limiter experiments in CHS

    International Nuclear Information System (INIS)

    Sagara, A.; Noda, N.; Akiyama, R.; Arimoto, H.; Idei, H.; Iguchi, H.; Kaneko, O.; Kohmoto, T.; Kubo, S.; Matsuoka, K.; Morita, S.; Motojima, O.; Nishimura, K.; Okamura, S.; Takahasi, C.; Takita, Y.; Yamada, I.; Matsunami, N.; Rice, J.; Yamada, H.; Shoji, T.; Ueda, M.

    1992-01-01

    A self-consistent analytical solution for net erosion of a divertor plate which is set perpendicular to magnetic field lines is presented. The primary flux profile of hydrogen and impurities except redepositing particles is externally given as well as the return ratio of sputtered atoms to the plate. In the direction along the divertor trace, all conditions are uniform. The ionization mean free path is assumed constant to simplify equations. The analytical solution is compared with net erosion experiments carried out in compact helical system (CHS) by exposing a graphite target to a neutral beam heated plasma column introduced perpendicularly to the target along the magnetic field lines through a 2 cm slit opend on a graphite limiter. After exposure to 98 discharges, the target surface is analyzed with Rutherford backscattering method. Deposition profiles of Ti and O impurities are very well explained with the analytical predictions. (orig.)

  20. Physics design requirements for the National Spherical Torus Experiment liquid lithium divertor

    International Nuclear Information System (INIS)

    Kugel, H.; Bell, M.; Berzak, L.; Brooks, A.; Ellis, R.; Gerhardt, S.P.; Kaita, R.; Kallman, J.; Majeski, R.; Mansfield, D.K.; Menard, J.; Stotler, D.; Zakharov, L.E.; Maingi, Rajesh; Nygren, R.E.; Soukhanovskii, V.; Wakeland, P.

    2009-01-01

    Recent National Spherical Tokamak Experiment (NSTX) high-power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components (PFCs) 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 (2) in the H-mode, to test the ability to operate at significantly lower density (e.g., ne/nGW = 0.25), for future reactor designs based on the Spherical Tokamak, and eventually to investigate high heat-flux power handling (10 MW/m2) with long pulse discharges (>1.5 s). 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.

  1. 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

  2. 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

  3. 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.

  4. Physics Design Requirements for the National Spherical Torus Experiment Liquid Lithium Divertor

    International Nuclear Information System (INIS)

    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.

    2008-01-01

    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 long pulse 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

  5. Fast time resolution charge-exchange measurements during the fishbone instability in the poloidal divertor experiment

    International Nuclear Information System (INIS)

    Beiersdorfer, P.; Kaita, R.; Goldston, R.J.

    1984-01-01

    Measurements of fast ion losses due to the fishbone instability during high β/sub T/q neutral beam heated discharges in the Poloidal Divertor Experiment have been made using two new vertical-viewing charge-exchange analyzers. The measurements show that the instability has an n=1 toroidal mode number, and that it ejects beam ions in a toroidally rotating beacon directed outward along a major radius. Observations of ejected ions with energies up to twice the beam injection energy at R approx. = R 0 + a indicate the presence of a non-μ-conserving acceleration mechanism

  6. Simulation of experimentally achieved detached plasmas using the UEDGE code

    International Nuclear Information System (INIS)

    Porter, G.D.; Allen, S.; Fenstermacher, M.

    1995-01-01

    The introduction of a divertor Thomson scattering system in DIII-D has enabled accurate determination of the plasma properties in the divertor region. We identify two plasma regimes; detached and attached. The electron temperature in the detached regime is about 2 eV, much lower than 5 to 10 eV determined earlier. We show that fluid models of the DIII-D scrape-off layer plasma are able to reproduce many of the features of these two plasma regimes, including the boundaries for transition between them. Detailed comparison between the results obtained from the fluid models and experiment suggest the models underestimate the spatial extent of the low temperature region associated the detached plasma mode. We suggest that atomic physics processes at the low electron temperatures reported here may account for this discrepancy

  7. 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

  8. 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

  9. Examining Innovative Divertor and Main Chamber Options for a National Divertor Test Tokamak

    Science.gov (United States)

    Labombard, B.; Umansky, M.; Brunner, D.; Kuang, A. Q.; Marmar, E.; Wallace, G.; Whyte, D.; Wukitch, S.

    2016-10-01

    The US fusion community has identified a compelling need for a National Divertor Test Tokamak. The 2015 Community Planning Workshop on PMI called for a national working group to develop options. Important elements of a NDTT, adopted from the ADX concept, include the ability to explore long-leg divertor `solutions for power exhaust and particle control' (Priority Research Direction B) and to employ inside-launch RF actuators combined with double-null topologies as `plasma solution for main chamber wall components, including tools for controllable sustained operation' (PRD-C). Here we examine new information on these ideas. The projected performance of super-X and X-point target long-leg divertors is looking very promising; a stable fully-detached divertor condition handling an order-of-magnitude increase in power handling over conventional divertors may be possible. New experiments on Alcator C-Mod are addressing issues of high-field side versus low-field side heat flux sharing in double-null topologies and the screening of impurities that might originate from RF actuators placed in the high-field side - both with favorable results. Supported by USDoE Awards DE-FC02-99ER54512 and DE-AC52-07NA27344.

  10. 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.

  11. 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

  12. Onyx embolization with the Apollo detachable tip microcatheter: A single-center experience.

    Science.gov (United States)

    Miller, Timothy R; Giacon, Luciano; Kole, Matthew J; Chen, Rong; Jindal, Gaurav; Gandhi, Dheeraj

    2018-06-01

    Purpose The Apollo Onyx Delivery Microcatheter (Ev3, Irvine, CA) is a detachable-tip microcatheter that was developed to reduce the risk of microcatheter entrapment during ethylene-vinyl alcohol copolymer (Onyx) embolizations. We report our experience with the microcatheter in a variety of neurointerventional procedures. Methods We retrospectively reviewed all Onyx embolizations performed in the head, neck, and spine using the Apollo Onyx Delivery Microcatheter from its introduction at our institution in July 2014 to August 2016. Information regarding patient diagnoses, procedural details, as well as clinical outcomes were obtained from the electronic medical record, procedure reports, and relevant angiographic imaging. Results A total of 58 arterial pedicle Onyx embolizations were performed in 37 patients. There were no cases of microcatheter entrapment, early/inadvertent tip detachment, or vessel injury upon removal of the device. There were two instances (3.5%) of leakage of Onyx from the microcatheter detachment site during embolization, which did not result in adverse sequelae. Clinical outcomes were excellent, with nearly all embolizations achieving the intended goal. In multivariate analysis, length of Onyx reflux along the microcatheter tip and utilization of a higher viscosity agent, Onyx 34, were significantly associated with tip detachment. Conclusion The use of the Apollo Microcatheter is both safe and effective during neurointerventional embolizations using Onyx. Leakage of liquid embolic agent from the detachment site is an infrequent technical complication that may be encountered with the device.

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

    International Nuclear Information System (INIS)

    Heidbrink, W.W.; Beiersdorfer, P.

    1986-11-01

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

  14. 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)

  15. Models for poloidal divertors

    International Nuclear Information System (INIS)

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

    1982-07-01

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

  16. Models for poloidal divertors

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-07-01

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

  17. 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.

  18. Overview of the modification to the Poloidal Divertor Experiment (PDX) to produce the Princeton Beta Experiment (PBX)

    International Nuclear Information System (INIS)

    Kuntson, D.

    1985-01-01

    The Poloidal Divertor Experiment at the Princeton Plasma Physics Laboratory has been recently transformed into the Princeton Beta Experiment. The purpose of the modification is to produce a bean-shaped plasma with beta values in excess of 10%, which is substantially above those achieved with more conventional plasma shapes. This transformation is accomplished by relocating several of the existing coils within the vacuum vessel, without a major disassembly of the device. One of the former PDX divertor coils is relocated on the mid-plane to be used as a ''pusher'' coil to create the plasma indentation. The ''pusher'' coil is protected from neutral beam impingement by watercooled graphite armor. The remaining internal PDX poloidal field coils are moved vertically to optimize the new configuration. The major new component is the set of passive stabilization coils. These coils are fabricated in segments and installed inside of the vacuum vessel. The purpose of the passive coils is to dampen the vertical instability of the bean-shaped plasma. The conversion to PBX also required reworking of internal and external poloidal coil bus leads, and the fabrication of new mechanical support structure

  19. Dissipative divertor operation in the Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    Lipschultz, B.; Goetz, J.; LaBombard, B.; McCracken, G.M.; Terry, J.L.; Graf, M.; Granetz, R.S.; Jablonski, D.; Kurz, C.; Niemczewski, A.; Snipes, J.

    1995-01-01

    The achievement of large volumetric power losses (dissipation) in the Alcator C-Mod divertor region is demonstrated in two operational modes: radiative divertor and detached divertor. During radiative divertor operation, the fraction of SOL power lost by radiation is P R /P SOL ∼0.8 with single null plasmas, n e 20 m -3 and I p e,div ≤6x10 20 m -3 . As the divertor radiation and density increase, the plasma eventually detaches abruptly from the divertor plates: I SAT drops at the target and the divertor radiation peak moves to the X-point region. Probe measurements at the divertor plate show that the transition occurs when T e ∼5 eV. The critical n e for detachment depends linearly on the input power. This abrupt divertor detachment is preceded by a comparatively long period ( similar 1-200 ms) where a partial detachment is observed to grow at the outer divertor plate. ((orig.))

  20. 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)

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

    International Nuclear Information System (INIS)

    Suzuki, Satoshi

    2011-01-01

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

  2. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  5. The design of the poloidal divertor experiment tokamak wall armor and inner limiter system

    International Nuclear Information System (INIS)

    Kugel, H.W.; Ulrickson, M.

    1982-01-01

    The inner wall protective plates for the Poloidal Divertor Experiment Tokamak are designed to absorb 8 MW of neutral deuterium beam power at maximum power densities of 3 kW/cm 2 for pulse lengths of 0.5 s. Preliminary studies indicate that the design could survive several pulses of l-s duration. The design consists of a tile and mounting plate structure. The mounting plates are water cooled to allow short duty cycles and beam calorimetry. The temperature and flow of the coolant are measured to obtain the injected power. A thermocouple array on the tiles provides beam position and power density profiles. Several material combinations for the tiles were subjected to thermal tests using both electron and neutral beams, and titanium-carbidecoated graphite was selected as the tile material. The heat transfer coefficient of the tile backing plate structure was measured to determine the maximum pulse rate allowable. The design of the armor system allows the structure to be used as a neutral beam power diagnostic and as an inner plasma limiter. The electrical and cooling systems external to the vacuum vessel are discussed

  6. 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.

  7. Radiation power profiles and density limit with a divertor in the W7-AS stellarator

    International Nuclear Information System (INIS)

    Giannone, L.; Burhenn, R.; McCormick, K.; Brakel, R.; Feng, Y.; Grigull, P.; Igitkhanov, Y.

    2002-01-01

    The addition of a divertor into the W7-AS stellarator has allowed access to a high density regime where the radiation profiles reach a steady state. In earlier limiter discharges, the plasma suffered a radiative collapse at high densities. In contrast to limiter experiments, where the impurity confinement time measured by Al laser blow-off increased with increasing line integrated density, in divertor discharges, above a density threshold, the impurity confinement time decreased with increasing line integrated density. The observation that the divertor plasma radiates mainly at the plasma edge rather than the plasma centre is a further indication that changes to the impurity transport coefficients at these high densities are the basis for the achievement of steady state discharges in the divertor configuration of W7-AS. The maximum line integrated density reached with a divertor is compared to that reached with a limiter. The previously derived scaling law for the density limit with a limiter shows that the achieved densities do not exceed those predicted when the higher deposited power is taken into account. In a divertor the radiated power is located at the plasma edge and increasing the density, cooling the plasma edge and radiating sufficient power to cause plasma detachment determines the density limit. (author)

  8. Evaluating Stellarator Divertor Designs with EMC3

    Science.gov (United States)

    Bader, Aaron; Anderson, D. T.; Feng, Y.; Hegna, C. C.; Talmadge, J. N.

    2013-10-01

    In this paper various improvements of stellarator divertor design are explored. Next step stellarator devices require innovative divertor solutions to handle heat flux loads and impurity control. One avenue is to enhance magnetic flux expansion near strike points, somewhat akin to the X-Divertor concept in Tokamaks. The effect of judiciously placed external coils on flux deposition is calculated for configurations based on the HSX stellarator. In addition, we attempt to optimize divertor plate location to facilitate the external coil placement. Alternate areas of focus involve altering edge island size to elucidate the driving physics in the edge. The 3-D nature of stellarators complicates design and necessitates analysis of new divertor structures with appropriate simulation tools. We evaluate the various configurations with the coupled codes EMC3-EIRENE, allowing us to benchmark configurations based on target heat flux, impurity behavior, radiated power, and transitions to high recycling and detached regimes. Work supported by DOE-SC0006103.

  9. OEDGE modeling of DIII-D density scan discharges leading to detachment

    Science.gov (United States)

    Elder, J. D.; Stangeby, P. C.; Bray, B. D.; Brooks, N.; Leonard, A. W.; McLean, A. G.; Unterberg, E. A.; Watkins, J. G.

    2015-08-01

    The OEDGE code is used to model the outer divertor plasma for discharges from a density scan experiment on DIII-D with the objective of assessing EIRENE and ADAS hydrogenic emission atomic physics data for Dα, Dβ and Dγ for values of Te and ne characteristic of the range of divertor plasma conditions from attached to weakly detached. Confidence in these values is essential to spectroscopic interpretation of any experiment or modeling effort. Good agreement between experiment and calculated emissions is found for both EIRENE and ADAS calculated emission profiles, confirming their reliability for plasma conditions down to ∼1 eV. For the cold dense plasma conditions characteristic of detachment, it is found that the calculated emissions are especially sensitive to Te.

  10. OEDGE modeling of DIII-D density scan discharges leading to detachment

    Energy Technology Data Exchange (ETDEWEB)

    Elder, J. D. [Univ. of Toronto, ON (Canada); Stangeby, P. C. [Univ. of Toronto, ON (Canada); General Atomics, San Diego, CA (United States); Bray, B. D. [General Atomics, San Diego, CA (United States); Brooks, N. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Leonard, A. W. [General Atomics, San Diego, CA (United States); McLean, A. G. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Unterberg, Ezekial A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Watkins, J. G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-09-30

    Here, we study the OEDGE code that is used to model the outer divertor plasma for discharges from a density scan experiment on DIII-D with the objective of assessing EIRENE and ADAS hydrogenic emission atomic physics data for Dα, Dβ and Dγ for values of Te and ne characteristic of the range of divertor plasma conditions from attached to weakly detached. Confidence in these values is essential to spectroscopic interpretation of any experiment or modeling effort. Good agreement between experiment and calculated emissions is found for both EIRENE and ADAS calculated emission profiles, confirming their reliability for plasma conditions down to ~1 eV. Lastly, for the cold dense plasma conditions characteristic of detachment, it is found that the calculated emissions are especially sensitive to Te.

  11. OEDGE modeling of DIII-D density scan discharges leading to detachment

    Energy Technology Data Exchange (ETDEWEB)

    Elder, J.D., E-mail: david@starfire.utias.utoronto.ca [University of Toronto Institute for Aerospace Studies, Toronto M3H 5T6 (Canada); Stangeby, P.C. [University of Toronto Institute for Aerospace Studies, Toronto M3H 5T6 (Canada); General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Bray, B.D. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Brooks, N. [Lawrence Livermore National Laboratory, PO Box 808, Livermore, CA 94550 (United States); Leonard, A.W. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, PO Box 808, Livermore, CA 94550 (United States); Unterberg, E.A. [Oak Ridge National Laboratories, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Watkins, J.G. [Sandia National Laboratories, PO Box 5800, Albuquerque, NM 87185 (United States)

    2015-08-15

    The OEDGE code is used to model the outer divertor plasma for discharges from a density scan experiment on DIII-D with the objective of assessing EIRENE and ADAS hydrogenic emission atomic physics data for D{sub α}, D{sub β} and D{sub γ} for values of T{sub e} and n{sub e} characteristic of the range of divertor plasma conditions from attached to weakly detached. Confidence in these values is essential to spectroscopic interpretation of any experiment or modeling effort. Good agreement between experiment and calculated emissions is found for both EIRENE and ADAS calculated emission profiles, confirming their reliability for plasma conditions down to ∼1 eV. For the cold dense plasma conditions characteristic of detachment, it is found that the calculated emissions are especially sensitive to T{sub e}.

  12. A review of direct experimental measurements of detachment

    Science.gov (United States)

    Boedo, J.; McLean, A. G.; Rudakov, D. L.; Watkins, J. G.

    2018-04-01

    Detached divertor plasmas feature strong radial and parallel gradients of density, temperature, electric fields and flow over the divertor volume and therefore, sampling the divertor plasma directly provides crucial knowledge to the interpretation and modeling efforts. We review the contribution of diagnostics that directly sample the plasma to the advancement of knowledge of the physics of detachment and detached divertors, such as the characteristics of the various regimes, discovery and quantification of drifts and identification of convection of heat and particles. We focus on wall probes, scanning probes, retarding field analyzers and Thomson scattering in the divertor region and also include the contribution of measurements away from the divertor that provide insight on how divertor detachment affects core, edge or pedestal conditions. Wall probes are critical as they can be installed in closed volumes of difficult access to other diagnostics and measure plasma parameters at the divertor structures, which define the plasma boundary conditions and where detachment effects are more likely to be strongest.

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

    International Nuclear Information System (INIS)

    1982-03-01

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

  14. 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.

  15. Towards a physics-integrated view on divertor pumping

    International Nuclear Information System (INIS)

    Day, Chr.; Gleason-González, C.; Hauer, V.; Igitkhanov, Y.; Kalupin, D.; Varoutis, S.

    2014-01-01

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

  16. 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

  17. 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.

  18. Model of detached plasmas

    International Nuclear Information System (INIS)

    Yoshikawa, S.; Chance, M.

    1986-07-01

    Recently a tokamak plasma was observed in TFTR that was not limited by a limiter or a divertor. A model is proposed to explain this equilibrium, which is called a detached plasma. The model consists of (1) the core plasma where ohmic heating power is lost by anomalous heat conduction and (2) the shell plasma where the heat from the core plasma is radiated away by the atomic processes of impurity ions. A simple scaling law is proposed to test the validity of this model

  19. A review of progress towards radiative divertor

    International Nuclear Information System (INIS)

    Zagorski, Roman

    1997-07-01

    A solution of the problem of the power and particle exhaust from the next step tokamaks, will require new techniques which redistribute the power entering the SOL onto much larger surface area than conventional divertor design permits, while maintaining good impurity retention in divertor volume and allowing for efficient helium pumping. Progress made in developing such techniques is discussed. Status of the modelling studies of dynamic gas target divertor and impurity seeded radiating divertors is presented. Recent results of experiments on radiative and gas target divertors are reviewed

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

    Science.gov (United States)

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

    2017-04-01

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

  1. Erosion/redeposition analysis : status of modeling and code validation for semi-detached tokamak edge plasmas

    International Nuclear Information System (INIS)

    Brooks, J. N.

    1999-01-01

    We are analyzing erosion and tritium codeposition for ITER, DIII-D, and other devices with a focus on carbon divertor and metallic wall sputtering, for detached and semi-detached edge plasmas. Carbon chemical-sputtering hydrocarbon-transport is computed in detail using upgraded models for sputtering yields, species, and atomic and molecular processes. For the DIII-D analysis this includes proton impact and dissociative recombination for the full methane and higher hydrocarbon chains. Several mixed material (Si-C doping and Be/C) effects on erosion are examined. A semi-detached reactor plasma regime yields peak net wall erosion rates of ∼1.0 (Be), ∼0.3 (Fe), and ∼0.01 (W) cm/burn-yr, and ∼50 cm/burn-yr for a carbon divertor. Net carbon erosion is dominated by chemical sputtering in the ∼1-3 eV detached plasma zone. Tritium codeposition in divertor-sputtered redeposited carbon is high (∼10-20 g-T/1000 s ). Silicon and beryllium mixing tends to reduce carbon erosion. Initial hydrocarbon transport calculations for the DIII-D DiMES-73 detached plasma experiment show a broad spectrum of redeposited molecules with ∼90% redeposition fraction

  2. 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)

  3. Numerical experiment to estimate the validity of negative ion diagnostic using photo-detachment combined with Langmuir probing

    Energy Technology Data Exchange (ETDEWEB)

    Oudini, N. [Laboratoire des plasmas de décharges, Centre de Développement des Technologies Avancées, Cité du 20 Aout BP 17 Baba Hassen, 16081 Algiers (Algeria); Sirse, N.; Ellingboe, A. R. [Plasma Research Laboratory, School of Physical Sciences and NCPST, Dublin City University, Dublin 9 (Ireland); Benallal, R. [Unité de Recherche Matériaux et Energies Renouvelables, BP 119, Université Abou Bekr Belkaïd, Tlemcen 13000 (Algeria); Taccogna, F. [Istituto di Metodologie Inorganiche e di Plasmi, CNR, via Amendola 122/D, 70126 Bari (Italy); Aanesland, A. [Laboratoire de Physique des Plasmas, (CNRS, Ecole Polytechnique, Sorbonne Universités, UPMC Univ Paris 06, Univ Paris-Sud), École Polytechnique, 91128 Palaiseau Cedex (France); Bendib, A. [Laboratoire d' Electronique Quantique, Faculté de Physique, USTHB, El Alia BP 32, Bab Ezzouar, 16111 Algiers (Algeria)

    2015-07-15

    This paper presents a critical assessment of the theory of photo-detachment diagnostic method used to probe the negative ion density and electronegativity α = n{sub -}/n{sub e}. In this method, a laser pulse is used to photo-detach all negative ions located within the electropositive channel (laser spot region). The negative ion density is estimated based on the assumption that the increase of the current collected by an electrostatic probe biased positively to the plasma is a result of only the creation of photo-detached electrons. In parallel, the background electron density and temperature are considered as constants during this diagnostics. While the numerical experiments performed here show that the background electron density and temperature increase due to the formation of an electrostatic potential barrier around the electropositive channel. The time scale of potential barrier rise is about 2 ns, which is comparable to the time required to completely photo-detach the negative ions in the electropositive channel (∼3 ns). We find that neglecting the effect of the potential barrier on the background plasma leads to an erroneous determination of the negative ion density. Moreover, the background electron velocity distribution function within the electropositive channel is not Maxwellian. This is due to the acceleration of these electrons through the electrostatic potential barrier. In this work, the validity of the photo-detachment diagnostic assumptions is questioned and our results illustrate the weakness of these assumptions.

  4. The effect of density on divertor conditions in ASDEX-Upgrade

    International Nuclear Information System (INIS)

    Pitcher, C.S.; Bosch, H.-S.; Buechl, K.; Field, A.; Fuchs, C.; Haas, G.; Junker, W.; Neu, R.; Neuhauser, J.; Wenzel, U.

    1995-01-01

    Detailed experimental divertor data are presented on the profiles of density and temperature in the inner and outer divertor fans, the radiated power distribution, the gas pressure and the spectroscopically derived particle fluxes, all as a function of the discharge density. At low and medium density, the inner divertor is cold and dense compared to the outer divertor. At high density, strong X-point MARFE and separatrix radiation partially detaches the inner divertor. Probe measurements which penetrate into the X-point MARFE at the outer divertor are presented. ((orig.))

  5. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

  7. Operational limits on WEST inertial divertor sector during the early phase experiment

    International Nuclear Information System (INIS)

    Firdaouss, M; Corre, Y; Languille, P; Autissier, E; Desgranges, C; Guilhem, D; Gunn, J P; Lipa, M; Missirlian, M; Pascal, J-Y; Pocheau, C; Richou, M; Tsitrone, E; Greuner, H

    2016-01-01

    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. (paper)

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

    International Nuclear Information System (INIS)

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

    1995-10-01

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

  9. 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.)

  10. [Functional results of cryosurgical procedures in rhegmatogenous retinal detachment including macula region - our experience].

    Science.gov (United States)

    Chrapek, O; Sín, M; Jirková, B; Jarkovský, J; Rehák, J

    2013-10-01

    Aim of this study is to evaluate retrospectively functional results of cryosurgical treatment of uncomplicated, idiopathic rhegmatogenous retinal detachment including macula region in phakic patients operated on at the Department of Ophthalmology, Faculty Hospital, Palacký University, Olomouc, Czech Republic, E.U., during the period 2002 -2013, and to evaluate the significance of the macula detachment duration for the final visual acuity. In the study group were included 56 eyes of 56 patients operated in the years 2003 - 2012 at the Department of Ophthalmology, Faculty Hospital, Palacký University, Olomouc. All patients were phakic and in all of them, the retinal detachment including the macula region was diagnosed. The mean follow-up period of the patients was 8,75 months. The initial and final visual acuity testing were performed. Comparing the initial and final visual acuity we rated the level of the visual acuity change. The result was stated as improved, if the visual acuity improved by 1 or more lines on the ETDRS chart. The result was rated as stabilized, if the visual acuity remained the same or it changed by 1 line of the ETDRS chart only. The result was evaluated as worsened, if the visual acuity decreased by 1 or more lines of the ETDRS chart. In the followed-up group, the authors compared visual acuity levels in patients with the macula detachment duration 10 days and 11 days. For the statistical evaluation of achieved results, the Mann - Whitney U test was used. The visual acuity improved in 49 (87 %), did not changed in 5 (9 %) and worsened in 2 (4 %) patients. The patients with macula detachment duration 10 days achieved statistically significant better visual acuity than patients with macula detachment duration 11 days. Patients with macula detachment duration 10 days have better prognosis for functional result than patients with macula detachment duration 11 days.

  11. Divertor development for ITER

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  12. 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

  13. Divertor radiation in the ASDEX upgrade tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-07-01

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

  14. 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)

  15. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

  17. 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

  18. 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)

  19. 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

  20. Charge exchange measurements of MHD activity during neutral beam injection in the Princeton Large Torus and the Poloidal Divertor Experiment

    International Nuclear Information System (INIS)

    Goldston, R.J.; Kaita, R.; Beiersdorfer, P.; Gammel, G.; Herndon, D.L.; McCune, D.C.; Meyerhofer, D.D.

    1987-01-01

    The horizontally scanning, multi-angle charge exchange analysers on the Princeton Large Torus (PLT) and the Poloidal Divertor Experiment (PDX) were used to study the effects of MHD activity on the background ion distribution function and on the beam ion slowing-down process during high power neutral injection. Sawtooth oscillations were observed in the fast ion flux on PLT and PDX, and measurements with neutral beams providing local neutral density enhancement indicated that the ions were transported radially when these events occurred. With near-perpendicular injection in PDX, at the lower toroidal fields necessary to maximize the plasma beta, repetitive bursts of greatly enhanced charge exchange flux were observed. These were associated with the 'fishbone' MHD instability, and a substantial depletion of the perpendicular slowing-down spectrum below the injection energy was seen. A simple phenomenological model for this loss mechanism was developed, and its use in simulation codes has been successful in providing good agreement with the experimental data. The behaviour and characteristics of this model are well matched by direct theoretical calculations. (author)

  1. 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

  2. Retinal Detachment

    Science.gov (United States)

    ... to your brain. It provides the sharp, central vision needed for reading, driving, and seeing fine detail. A retinal detachment lifts or pulls the retina from its normal position. It can occur at ...

  3. 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)

  4. The lithium vapor box divertor

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  5. 3D nonlinear numerical simulation of the current-convective instability in detached diverter plasma

    Science.gov (United States)

    Stepanenko, Alexander; Krasheninnikov, Sergei

    2017-10-01

    One of the possible mechanisms responsible for strong radiation fluctuations observed in the recent experiments with detached plasmas at ASDEX Upgrade [Potzel et al., Nuclear Fusion, 2014] can be related to the onset of the current-convective instability (CCI) driven by strong asymmetry of detachment in the inner and outer tokamak divertors [Krasheninnikov and Smolyakov, PoP, 2016]. In this study we present the first results of 3D nonlinear numerical simulations of the CCI in divertor plasma for the conditions relevant to the AUG experiment. The general physical model used to simulate the CCI, qualitative estimates for the instability characteristic growth rate and transverse wavelengths derived for plasma, which is spatially inhomogeneous both across and along the magnetic field lines, are presented. The simulation results, demonstrating nonlinear dynamics of the CCI, provide the frequency spectra of turbulent divertor plasma fluctuations showing good agreement with the available experimental data. This material is based upon the work supported by the U.S. Department of Energy under Award No. DE-FG02-04ER54739 at UCSD and by the Russian Ministry of Education and Science Grant No. 14.Y26.31.0008 at MEPhI.

  6. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  8. Pediatric retinal detachment in the Eastern Province of Saudi Arabia: experience of a tertiary care hospital.

    Science.gov (United States)

    Cheema, Rizwan A; Al-Khars, Wajeeha; Al-Askar, Essam; Amin, Yasir M

    2009-01-01

    Because no previous studies have addressed the issue, we describe clinical characteristics and surgical outcome of patients with rhegmatogenous retinal detachment (RRD) in a pediatric population of the Eastern province of Saudi Arabia. We conducted a retrospective review of all consecutive cases of pediatric RRD (0-18 years) patients presenting at Dhahran Eye Specialist Hospital, a tertiary care hospital, in the Eastern Province of Saudi Arabia over a period of 3 years. Twenty patients were included in the study, accounting for 9.4% of all retinal detachment surgery cases performed over a period of 3 years (January 2006 to December 2008). The median age was 11.0 years, (range, birth to 18 years). Trauma, (45%) myopia/vitreoretinal degeneration (10%) and prior ocular surgery (25%) were significant risk factors for RRD. Proliferative vitreoretinopathy (PVR) more than grade C was present in 14/20 (70%) of cases. Most patients (15/20, 75%) were treated with pars plana vitrectomy and placement of an encircling buckle, while silicone oil or gas was used as tamponade in 13/20 (65%) patients. Surgery was successful in 17/20 (85%) cases in achieving retinal re-attachment. Visual acuity improved significantly following surgery (Mean preop 2.146 LogMAR, Mean postop 1.497 LogMAR) ( P= .014). Longer duration of RRD ( P =.007) and macular involvement ( P =.05) were associated with worse anatomical outcomes following surgery. Pediatric RRD in the Eastern province is often associated with predisposing pathology. Surgery is successful in achieving anatomical reattachment of the retina in a majority of cases with improvement of visual acuity.

  9. Scaling of divertor heat flux profile widths in DIII-D

    International Nuclear Information System (INIS)

    Lasnier, C.J.; Makowski, M.A.; Boedo, J.A.; Allen, S.L.; Brooks, N.H.; Hill, D.N.; Leonard, A.W.; Watkins, J.G.; West, W.P.

    2011-01-01

    New scalings of the dependence of divertor heat flux peak and profile width, important parameters for the design of future large tokamaks, have been obtained from recent DIII-D experiments. We find the peak heat flux depends linearly on input power, decreases linearly with increasing density, and increases linearly with plasma current. The profile width has a weak dependence on input power, is independent of density up to the onset of detachment, and is inversely proportional to the plasma current. We compare these results with previously published scalings, and present mathematical expressions incorporating these results.

  10. 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.

  11. 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)

  12. Erosion of divertor materials in simulation experiments of ITEP thermonuclear reactor operation

    International Nuclear Information System (INIS)

    Guseva, M.I.; Gureev, V.M.; Danelyan, L.S.; Kolbasov, B.N.; Korshunov, S.N.; Martynenko, Yu.V.; Petrov, V.B.; Stolyarova, V.G.; Khripunov, B.I.; Vasil'ev, V.I.; Strunnikov, V.M.

    2004-01-01

    The tungsten sputtering by the deuterium ions with the energy of 5 eV at the temperature of 1470 K with the sputtering coefficient of 1.5 x 10 -4 at/ion is identified. Comparison of the erosion degree of various W types and W + C coatings on W is carried out in the experiments on simulating the plasma break-away; W drops distribution by the diameters, depending on the collector drops position, is studied. W(111) tungsten is characterized by the minimal erosion and W-10% Re tungsten - by the maximum one [ru

  13. 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.

  14. Transition in plasma fluctuation between attached and detached plasmas

    International Nuclear Information System (INIS)

    Okazaki, Katsuya; Ohno, Noriyasu; Kajita, Shin; Tanaka, Hirohiko

    2012-01-01

    The static and dynamic behaviors of detached plasmas have received considerable attention because the use of a detached divertor is thought to provide a promising method for reducing the heat flux to plasma-facing components. In this study, fluctuations were measured with an electrostatic probe as the plasma was changed from attached to detached states by increasing the neutral gas pressure. The transition from an attached plasma to a detached plasma was found to change the phase relation between the density and the potential. (author)

  15. Self-sustained detachment observed in LHD and comparison with detachment and Marfe in W7-AS

    International Nuclear Information System (INIS)

    Miyazawa, J.; Masuzaki, S.; Yamada, H.; Sakamoto, R.; Peterson, B. J.; Shoji, M.; Ohyabu, N.; Komori, A.; Motojima, O.

    2005-01-01

    One of the crucial issues for a magnetically confined fusion reactor [1-3] is the reduction of the divertor heat load. Edge plasma cooling by increasing the density or introducing impurities such as neon as radiators is effective for reducing the divertor heat load by detaching the plasma from the divertor plates. In high-density tokamak plasmas, the so-called high-recycling regime is predicted by two-point model [4,5], where the divertor temperature (density) nonlinearly decreases (increases) with the main plasma density. Detachment takes place when the density is increased further. Then, the pressure in the flux tube is no longer conserved due to the increased radiation loss, charge exchange loss, and volume recombination. Eventually, the discharge is terminated by disruption often via rapid formation of a radiation condensation Marfe [5,6]. This scenario well explains the experimental results of tokamaks [5,7]. A similar kind of conventional detachment has been also intensively studied in a stellarator, W7-AS, after modification to the island divertor configuration [8,9]. Detachment in W7-AS is achieved by increasing the density beyond the density threshold for the high-density H-mode (HDH) [10]. As the density is increased further to the operational density limit, complete detachment takes place and, in some cases, the Marfe is formed on (or inside of) the last-closed-flux-surface (LCFS) [11-13]. These are reviewed in Section 4. Recently, a new state of self-sustained detachment has been found in LHD and named the Serpens mode (Self-regulated plasma edge neath the last-closed-flux-surface) [15,16]. In the Serpens mode, the hot plasma boundary is shrinking to equal 90 % of the LCFS radius. Large fluctuations appear in the divertor flux, H alpha and Cm signals, together with a rotating radiation belt named the serpent [17]. Detachment in LHD is observed in a significant decrease in the ion saturation current, Isat, measured on the divertor tiles [18,19]. In this

  16. Design of DIII-D advanced divertor

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  17. Design of DIII-D Advanced Divertor

    International Nuclear Information System (INIS)

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

    1989-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

  19. Effect of magnetic island on three-dimensional structure of edge radiation and its consequences for detachment in LHD (EX-D)

    International Nuclear Information System (INIS)

    Drapiko, E.A.; Peterson, B.J.; Kobayashi, M.

    2010-11-01

    Detached plasmas represent an important operational regime for a fusion reactor whereby the heat load to the divertor can be reduced through enhanced radiation to ensure sustainable steady-state discharges. Normally, in toroidal devices there exists a density threshold above which detachment occurs, but in LHD the plasma commonly experiences radiative collapse before this threshold is reached. Recent work on LHD has shown the addition of an n/m=1/1 magnetic island (MI) enhances the detachment process by lowering its density threshold. In this paper the effects of the MI on the 3D radiation structure in attached and detached plasmas as predicted by EMC3-EIRENE are clearly seen in the imaging bolometer (IRVB) data, experimentally confirming the role that the MI plays in the detachment process. With the addition of the MI the carbon radiation profile from the code in a poloidal cross-section becomes more localized near the helical divertor (HD) x-points (X). This is reflected in the focussing of the radiation patterns corresponding to the HDX in both the IRVB and code data in images corresponding to the IRVB field of view (FOV). Detachment results in a more asymmetric radiation profile in the poloidal cross-section code data with localized peaks near the HDX and MIX. The radiation from the MIX is reflected in strong radiation from the corresponding location in the IRVB FOV from both code and IRVB data. However the relative increase in the radiation from the MIX is greater in the code data than in the IRVB data for reasons which are so far unknown. Also similar discharges with and without the MI show detachment with the MI, albeit at a lower density than the discharge without the MI. This work confirms the previous conclusion that the MI enhances the localization of the radiation and is conducive to achieving and sustaining the detachment. (author)

  20. 2D imaging of helium ion velocity in the DIII-D divertor

    Science.gov (United States)

    Samuell, C. M.; Porter, G. D.; Meyer, W. H.; Rognlien, T. D.; Allen, S. L.; Briesemeister, A.; Mclean, A. G.; Zeng, L.; Jaervinen, A. E.; Howard, J.

    2018-05-01

    Two-dimensional imaging of parallel ion velocities is compared to fluid modeling simulations to understand the role of ions in determining divertor conditions and benchmark the UEDGE fluid modeling code. Pure helium discharges are used so that spectroscopic He+ measurements represent the main-ion population at small electron temperatures. Electron temperatures and densities in the divertor match simulated values to within about 20%-30%, establishing the experiment/model match as being at least as good as those normally obtained in the more regularly simulated deuterium plasmas. He+ brightness (HeII) comparison indicates that the degree of detachment is captured well by UEDGE, principally due to the inclusion of E ×B drifts. Tomographically inverted Coherence Imaging Spectroscopy measurements are used to determine the He+ parallel velocities which display excellent agreement between the model and the experiment near the divertor target where He+ is predicted to be the main-ion species and where electron-dominated physics dictates the parallel momentum balance. Upstream near the X-point where He+ is a minority species and ion-dominated physics plays a more important role, there is an underestimation of the flow velocity magnitude by a factor of 2-3. These results indicate that more effort is required to be able to correctly predict ion momentum in these challenging regimes.

  1. Detached Statements

    Directory of Open Access Journals (Sweden)

    Mark McBride

    2018-04-01

    Full Text Available Joseph Raz has introduced an interesting class of statements —detached statements— into the philosophical lexicon. In brief, such statements are (informa- tive normative statements, yet the speaker does not, in so uttering them, express or convey acceptance of the point of view of the hearer to whom they are addressed (as contrasted with committed statements, where the speaker does express or convey such acceptance. I propose to offer a novel analysis of such statements (and to clear away some confusions about them. In brief, such statements will be analysed as wide-scope normative conditionals.

  2. Retinal Detachment

    Directory of Open Access Journals (Sweden)

    Adnan Riaz, MD

    2018-04-01

    Full Text Available History of present illness: A 58-year-old female presented to the emergency department reporting six days of progressive, atraumatic left eye vision loss. Her symptoms started with the appearance of dark spots and “spider webs,” and then progressed to darkening of vision in her left eye. She reports mild pain since yesterday. Her review of symptoms was otherwise negative. Ocular physical examination revealed normal external appearance, intact extraocular movements, and visual acuities of 20/25 OD and light/dark sensitivity OS. Fluorescein uptake was negative and slit lamp exam was unremarkable. Significant findings: Bedside ocular ultrasound revealed a serpentine, hyperechoic membrane that appeared tethered to the optic disc posteriorly with hyperechoic material underneath. These findings are consistent with retinal detachment (RD and associated retinal hemorrhage. Discussion: The retina is a layer of organized neurons that line the posterior portion of the posterior chamber of the eye. RD occurs when this layer separates from the underlying epithelium, resulting in ischemia and progressive photoreceptor degeneration, with potentially rapid and permanent vision loss if left untreated.1 Risk factors include advanced age, male sex (60%, race (Asians and Jews, and myopia and lattice degeneration.2 Bedside ultrasound (US performed by emergency physicians provides a valuable tool that has been used by ophthalmologists for decades to evaluate intraocular disease.1,3 Findings on bedside ultrasound consistent with RD include a hyperechoic membrane floating in the posterior chamber. RD usuallyremain tethered to the optic disc posteriorly and do not cross midline, a feature distinguishing them from posterior vitreous detachments. Associated retinal hemorrhage, seen as hyperechoic material under the retinal flap, can often be seen.1,2 US can also distinguish between “mac-on” and “mac-off” detachments. If the retina is still attached to the

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

    International Nuclear Information System (INIS)

    Asakura, Nobuyuki

    2007-01-01

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

  4. 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

  5. Particle recirculation in the ergodic divertor of Tore Supra

    International Nuclear Information System (INIS)

    Gunn, J.P.; Azeoual, A.; Becoulet, M.

    1999-01-01

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

  6. 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.))

  7. Energy efficiency in existing detached housing. Danish experiences with different policy instruments

    Energy Technology Data Exchange (ETDEWEB)

    Gram-Hanssen, K.; Haunstrup Christensen, T. (Aalborg Univ., Danish Building Research Institute, Hoersholm (Denmark))

    2011-07-01

    This report contains a memo written as an input to the German project Enef-haus on energy-efficient restoration of single-family houses in Germany. The memo contains a summary of the Danish experiences divided into three main sections: first is a short historic overview of the Danish energy policy indicating when different relevant instruments have been introduced to increase the energy efficiency of privately owned single-family houses. Second is a short introduction to the Danish housing sector and its energy supplies. The third and main part of the report is an examination of the most recent and relevant instruments concluding both on the results concerning the impact of the instruments especially on owners of single-family houses and on more general experiences with their implementation. Finally the memo concludes on the general lessons that can be learned from the Danish experiences. (Author)

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  9. Searching for Heavy Photons with Detached Verices in the Heavy Photon Search Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Szumila-Vance, Holly [Old Dominion Univ., Norfolk, VA (United States); Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States)

    2017-08-01

    The Jefferson Lab Heavy Photon Search (HPS) experiment is searching for a hypothetical massive particle called the heavy photon which could mediate a dark electromagnetic-type force. If heavy photons kinetically mix with Standard Model photons, they may be radiated by electrons scattering from a heavy nucleus and then decay to e+e- pairs. HPS uniquely searches for heavy photons that either decay at the target or a measurable distance after. The experiment utilizes a silicon vertex tracker (SVT) for momentum and vertex reconstruction, together with an electromagnetic calorimeter for measuring particle energies and triggering events. The HPS experiment took its first data during the spring 2015 engineering run using a 1 GeV electron beam incident on a tungsten target and its second data in the spring of 2016 at a beam energy of 2.3 GeV. The 2015 run obtained two days of production data that was used for the first physics results. The analysis of the data was conducted as a blinded analysis by tuning cuts on 10% of the data. This dissertation discusses the displaced vertex search for heavy photons in the 2015 engineering run. It describes the theoretical motivation for looking for heavy photons and provides an overview of the HPS experimental design and performance. The performance details of the experiment are primarily derived from the 2015 engineering run with some discussion from the higher energy running in 2016. This dissertation further discusses the cuts used to optimize the displaced vertex search and the results of the search. The displaced vertex search did not set a limit on the heavy photon but did validate the methodology for conducting the search. Finally, we used the full data set to make projections and guide future analyses.

  10. A 60-GHz interferometer with a local oscillator integrated antenna array for divertor simulation experiments on GAMMA 10/PDX

    Energy Technology Data Exchange (ETDEWEB)

    Kohagura, J., E-mail: kohagura@prc.tsukuba.ac.jp; Yoshikawa, M.; Shima, Y.; Nojiri, K.; Sakamoto, M.; Nakashima, Y. [Plasma Research Center, University of Tsukuba, Tsukuba, Ibaraki 305-8577 (Japan); Wang, X. [Saitama University, Saitama 338-8570 (Japan); Kuwahara, D. [Tokyo University of Agriculture and Technology, Koganei, Tokyo 184-8588 (Japan); Ito, N. [National Institute of Technology, Ube College, Ube, Yamaguchi 755-8555 (Japan); Nagayama, Y. [National Institute of Fusion Science, Toki, Gifu 509-5292 (Japan); Mase, A. [Kyushu University, Kasuga, Fukuoka 816-8580 (Japan)

    2016-11-15

    In conventional multichannel/imaging microwave diagnostics of interferometry, reflectometry, and electron cyclotron emission measurements, a local oscillator (LO) signal is commonly supplied to a receiver array via irradiation using LO optics. In this work, we present a 60-GHz interferometer with a new eight-channel receiver array, called a local oscillator integrated antenna array (LIA). An outstanding feature of LIA is that it incorporates a frequency quadrupler integrated circuit for LO supply to each channel. This enables simple and uniform LO supply to the receiver array using only a 15-GHz LO source and a coaxial cable transmission line instead of using an expensive 60-GHz source, LO optics, and a waveguide transmission line. The new interferometer system is first applied to measure electron line-averaged density inside the divertor simulation experimental module (D-module) on GAMMA 10/PDX tandem mirror device.

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  12. Plasma structure change and intermittent fluctuation near magnetic island X-point under detached plasma condition in LHD

    International Nuclear Information System (INIS)

    Ohno, N.; Tsuji, Y.; Tanaka, H.; Masuzaki, S.; Kobayashi, M.; Akiyama, T.; Morisaki, T.; Motojima, G.; Narushima, Y.

    2014-10-01

    Plasma profiles and intermittent fluctuations near the helical divertor X-point and on a divertor plate were investigated using a fast scanning Langmuir probe and a probe array embedded on a divertor plate in detached divertor condition that was sustained by applying a resonant magnetic perturbation (RMP) field in LHD. When the RMP induced magnetic island X-point (n/m = 1/1) is located near the helical divertor X-point, the reduction of particle flux accompanied by the plasma detachment occurred near the helical divertor X-point (n/m = 2/10), which leads to the reduction of the particle flux at the strike point on the divertor plate. We also found that when the divertor plasma turned to be the detached condition, the enhanced plasma fluctuations were confirmed between the helical divertor X-point and ergodic region, which exhibited a dynamic behavior having a large amount of positive-spike components with highly intermittent property. (author)

  13. 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)

  14. 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)

  15. Textor bundle divertor

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  16. TEXTOR bundle divertor

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  17. NSTX Tangential Divertor Camera

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  18. Divertor cooling device

    International Nuclear Information System (INIS)

    Nakayama, Tadakazu; Hayashi, Katsumi; Handa, Hiroyuki

    1993-01-01

    Cooling water for a divertor cooling system cools the divertor, thereafter, passes through pipelines connecting the exit pipelines of the divertor cooling system and the inlet pipelines of a blanket cooling system and is introduced to the blanket cooling system in a vacuum vessel. It undergoes emission of neutrons, and cooling water in the divertor cooling system containing a great amount of N-16 which is generated by radioactivation of O-16 is introduced to the blanket cooling system in the vacuum vessel by way of pipelines, and after cooling, passes through exit pipelines of the blanket cooling system and is introduced to the outside of the vacuum vessel. Radiation of N-16 in the cooling water is decayed sufficiently with passage of time during cooling of the blanket, thereby enabling to decrease the amount of shielding materials such as facilities and pipelines, and ensure spaces. (N.H.)

  19. '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

  20. The contribution of radio-frequency rectification to field-aligned losses of high-harmonic fast wave power to the divertor in the National Spherical Torus eXperiment

    Energy Technology Data Exchange (ETDEWEB)

    Perkins, R. J., E-mail: rperkins@pppl.gov; Hosea, J. C.; Jaworski, M. A.; Diallo, A.; Bell, R. E.; Bertelli, N.; Gerhardt, S.; Kramer, G. J.; LeBlanc, B. P.; Phillips, C. K.; Podestà, M.; Roquemore, L.; Taylor, G.; Wilson, J. R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Ahn, J.-W.; Gray, T. K. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); McLean, A. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Sabbagh, S. [Department of Applied Physics and Applied Mathematics, Columbia University, New York, New York 10027 (United States)

    2015-04-15

    The National Spherical Torus eXperiment (NSTX) can exhibit a major loss of high-harmonic fast wave (HHFW) power along scrape-off layer (SOL) field lines passing in front of the antenna, resulting in bright and hot spirals on both the upper and lower divertor regions. One possible mechanism for this loss is RF sheaths forming at the divertors. Here, we demonstrate that swept-voltage Langmuir probe characteristics for probes under the spiral are shifted relative to those not under the spiral in a manner consistent with RF rectification. We estimate both the magnitude of the RF voltage across the sheath and the sheath heat flux transmission coefficient in the presence of the RF field. Although precise comparison between the computed heat flux and infrared (IR) thermography cannot yet be made, the computed heat deposition compares favorably with the projections from IR camera measurements. The RF sheath losses are significant and contribute substantially to the total SOL losses of HHFW power to the divertor for the cases studied. This work will guide future experimentation on NSTX-U, where a wide-angle IR camera and a dedicated set of coaxial Langmuir probes for measuring the RF sheath voltage directly will quantify the contribution of RF sheath rectification to the heat deposition from the SOL to the divertor.

  1. 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

  2. Production and control of an edge radiating layer with the ergodic divertor on TORE SUPRA

    International Nuclear Information System (INIS)

    Poutchy, L.; Vallet, J.C.; Michelis, C. de; Grosman, A.; Hess, W.; Mattioli, M.; Monier-Garbet, P.

    1992-01-01

    We have recently defined on Tore Supra a discharge piloting strategy to prevent disruptions at the density limit, based on the property of the Ergodic Divertor to stabilize MHD activity. This strategy allows plasma studies close to the density limit without disruptions, and has been successfully used on Tore Supra ohmicly heated plasmas. The Ergodic Divertor allows also the stabilization of detached plasmas which appear spontaneously near the density limit. Edge impurities are shown to play a fundamental role in this stabilization. Detached plasmas have been thereby controlled during 10 seconds

  3. 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)

  4. Divertor characteristics and control on the W-shaped divertor with pump of JT-60U

    International Nuclear Information System (INIS)

    Hosogane, N.; Kubo, H.; Higashijima, S.

    1999-01-01

    Roles of the inner leg pumping and the private dome, which are special features of the W-shaped divertor of JT-60U, have been investigated. The following observations were made: The inner leg pumping functions well in attached states or partially detached states with weak X-point MARFE where the inner particle recycling is enhanced. A combination of main gas puff and inner leg pump is effective in reduction of intrinsic carbon impurity. Geometrical effects of the private dome on transport of hydrocarbons in the private flux region was confirmed by spectroscopic measurements of CD-band intensity profile and impurity transport simulation code using experimental data. (author)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  6. 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.

  7. Divertor scaling laws for tokamaks

    International Nuclear Information System (INIS)

    Catto, P.J.; Krasheninnikov, S.I.; Connor, J.W.

    1997-01-01

    The breakdown of two body scaling laws is illustrated by using the two dimensional plasma code UEDGE coupled to an advanced Navier-Stokes neutrals transport package to model attached and detached regimes in a simplified geometry. Two body similarity scalings are used as benchmarks for runs retaining non-two body modifications due to the effects of (i) multi-step processes altering ionization and radiation via the excited states of atomic hydrogen and (ii) three body recombination. Preliminary investigations indicate that two body scaling interpretations of experimental data fail due to (i) multi-step processes when a significant region of the plasma exceeds a plasma density of 10 19 m -3 , or (ii) three body recombination when there is a significant region in which the temperature is ≤1 eV while the plasma density is ≥10 20 m -3 . These studies demonstrate that two body scaling arguments are often inappropriate in the divertor and the first results for alternate scalings are presented. (orig.)

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  9. Physics design and experimental study of tokamak divertor

    International Nuclear Information System (INIS)

    Yan Jiancheng; Gao Qingdi; Yan Longwen; Wang Mingxu; Deng Baiquan; Zhang Fu; Zhang Nianman; Ran Hong; Cheng Fayin; Tang Yiwu; Chen Xiaoping

    2007-06-01

    The divertor configuration of HL-2A tokamak is optimized, and the plasma performance in divertor is simulated with B2-code. The effects of collisionality on plasma-wall transition in the scrape-off layer of divertor are investigated, high performances of the divertor plasma in HL-2A are simulated, and a quasi- stationary RS operation mode is established with the plasma controlled by LHCD and NBI. HL-2A tokamak has been successfully operated in divertor configuration. The major parameters: plasma current I p =320 kA, toroidal field B t =2.2 T, plasma discharger duration T d =1580 ms ware achieved at the end of 2004. The preliminary experimental researches of advanced diverter have been carried out. Design studies of divertor target plate for high power density fusion reactor have been carried out, especially, the physical processes on the surface of flowing liquid lithium target plate. The exploration research of improving divertor ash removal efficiency and reducing tritium inventory resulting from applying the RF ponderomotive force potential is studied. The optimization structure design studies of FEB-E reactor divertor are performed. High flux thermal shock experiments were carried on tungsten and carbon based materials. Hot Isostatic Press (HIP) method was employed to bond tungsten to copper alloys. Electron beam simulated thermal fatigue tests were also carried out to W/Cu bondings. Thermal desorption and surface modification of He + implanted into tungsten have been studied. (authors)

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

    International Nuclear Information System (INIS)

    Kallenbach, A.; Bosch, H.-S.; De Pena Hempel, S.; Dux, R.; Kaufmann, M.; Mertens, V.; Neuhauser, J.; Suttrop, W.; Zohm, H.

    1997-01-01

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

  11. 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.

  12. A study of X-divertor in NSTX-U with SOLPS simulations

    Science.gov (United States)

    Chen, Zhong-Ping; Kotschenreuther, Mike; Mahajan, Swadesh; Gerhardt, Stefan

    2018-03-01

    The X-divertor (XD) geometry in NSTX-U is demonstrated, via SOLPS simulations, to perform better than the standard divertor (SD); in particular, it allows detachment at a lower upstream density and stabilizes the detachment front near the target, away from the main X-point. Consequently a stable detached operation becomes possible—the localization near the plate allows a vast reduction of heat fluxes without degrading the core plasma. Indeed, it is confirmed by our simulation that at similar states of detachment the XD outperforms the SD by reducing the heat fluxes to the target and maintaining higher upstream temperatures, resulting in scrape-off layers that are more favorable for advanced tokamak operation. These advantages are attributed to the unique geometric characteristics of XD—poloidal flaring near the target.

  13. DiMES Studies of Temperature Dependence of Carbon Erosion and Re-Deposition in the DIII-D Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Rudakov, D L; Jacob, W; Krieger, K; Litnovsky, A; Philipps, V; West, W P; Wong, C C; Allen, S L; Bastasz, R J; Boedo, J A; Brooks, N H; Boivin, R L; De Temmerman, G; Fenstermacher, M E; Groth, M; Hollmann, E M; Lasnier, C J; McLean, A G; Moyer, R A; Stangeby, P C; Wampler, W R; Watkins, J G; Wienhold, P; Whaley, J

    2007-03-15

    A strong effect of a moderately elevated surface temperature on net carbon deposition and deuterium co-deposition in the DIII-D divertor was observed under detached conditions. A DiMES sample with a gap 2 mm wide and 18 mm deep was exposed to lower-single-null (LSN) L-mode plasmas first at room temperature, and then at 200 C. At the elevated temperature, deuterium co-deposition in the gap was reduced by an order of magnitude. At the plasma-facing surface of the heated sample net carbon erosion was measured at a rate of 3 nm/s, whereas without heating net deposition is normally observed under detachment. In a related experiment three sets of molybdenum mirrors recessed 2 cm below the divertor floor were exposed to identical LSN ELMy H-mode discharges. The first set of mirrors exposed at ambient temperature exhibited net carbon deposition at a rate of up to 3.7 nm/s and suffered a significant drop in reflectivity. In contrast, two other mirror sets exposed at elevated temperatures between 90 C and 175 C exhibited practically no carbon deposition.

  14. Charge-exchange measurements of MHD activity during neutral beam injection in the Princeton Large Torus and the Poloidal Divertor Experiment

    International Nuclear Information System (INIS)

    Goldston, R.J.; Kaita, R.; Beiersdorfer, P.; Gammel, G.; Herndon, D.L.; McCune, D.C.; Meyerhofer, D.D.

    1986-07-01

    The horizontally scanning, multiangle charge-exchange analyzers on the Princeton Large Torus (PLT) and the Poloidal Divertor Experiment (PDX) were used to study the effects of MHD activity on the background ion distribution function and on the beam ion slowing-down process during high-power neutral injection. Sawtooth oscillations were observed in the fast ion flux on PLT and PDX, and measurements with neutral beams providing local neutral density enhancement indicate that ions are transported radially when these events occur. With near-perpendicular injection in PDX, at the lower toroidal fields necessary to maximize beta, rapid, repetitive bursts of greatly enhanced charge-exchange flux were observed. These are associated with the ''fishbone'' MHD instability, and a substantial depletion of the perpendicular slowing-down spectrum below the injection energy was seen. A simple phenomenological model for this loss mechanism was developed, and its use in simulation codes has been successful in providing good agreement with the data. The behavior and characteristics of this model are well matched by the direct theoretical calculations

  15. Characterisation of detached plasmas on the MAST tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, J.R., E-mail: james.harrison@ccfe.ac.uk [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom); Department of Physics, University of York, Heslington, York, YO10 5DD (United Kingdom); Lisgo, S.W. [ITER Organization, Route de Vinon-sur-Verdon, St.Paul-lez-Durance, Cedex (France); Gibson, K.J. [Department of Physics, University of York, Heslington, York, YO10 5DD (United Kingdom); Tamain, P. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Dowling, J. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom)

    2011-08-01

    Divertor detachment is an attractive operating regime for the next generation of tokamak devices, as it offers a means of mitigating the steady-state heat flux to plasma facing components. In order to clarify the dominant physical mechanisms that govern detachment, high quality data from several diagnostics are required to constrain theoretical models. To that end, high spatial ({approx}3 mm) and temporal (5 kHz) resolution measurements have been made of the intensity of deuterium Balmer and carbon emission lines during the onset and evolution of detachment of the lower inner strike point in MAST L-mode discharges. Furthermore, spatially-resolved measurements of the shapes and intensities of high-n Balmer lines have been recorded to infer plasma conditions during the detached phase.

  16. Fluctuation characteristics in detached recombining plasmas

    International Nuclear Information System (INIS)

    Ohno, Noriyasu; Tanaka, Naoyuki; Takamura, Shuichi; Budaev, Viatcheslav

    2002-01-01

    Fluctuation in detached recombining plasmas has been investigated experimentally in the linear divertor plasma simulator, NAGDIS-II. As increasing neutral gas pressure, floating potential fluctuation of the target plate installed at the end of the NADIS-II device becomes larger and bursty negative spikes are observed in the signal associated with a transition from attached to detached a plasmas. The fluctuation property has been analyzed by using Fast Fourier Transform (FFT), probability distribution function (PDF) and wavelet transform. The PDF of the floating potential fluctuation in the attached plasma condition obeys the Gaussian distribution function, on the other hand, the PDF in detached plasma shows a strong deviation from the Gaussian distribution function, which can be characterized by flatness and skewness. Comparison of the fluctuation properties between the floating potential and the optical emission from the detached plasma has been done based on the wavelet transform to show that a strong correlation between them, which could indicate bursty transport of energetic electrons from upstream to downstream region along the magnetic field. (author)

  17. The ITER divertor concept

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Kaufmann, M.; Bosch, H.S.; Herrmann, A.

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Kaufmann, M.; Bosch, H.-S.; Herrmann, A.

    2001-01-01

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

  20. Operating windows of pebble divertor

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  1. Recombination dominated hydrogenic emission from the detached plasmas in W7-AS

    International Nuclear Information System (INIS)

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

    2003-01-01

    Beyond a certain threshold average density in the High-Density H-Mode the island divertor plasma in the stellarator W7-AS undergoes partial detachment. The tomographic reconstruction of the radiated power density from the detached pulses show that the radiation profile in the triangular plane is also asymmetric. In the detached phase, the spectrometer viewing tangentially to the target tiles in the top divertor region manifests that the impurity radiation layer is close to the X-points. The spectral analysis also demonstrates the presence of a hydrogen radiation zone dominated by recombination emission close to the target tiles. This papers presents the emission from the deeply detached locations including the volume recombination in a stable discharge. (orig.)

  2. [Surgical managment of retinal detachment].

    Science.gov (United States)

    Haritoglou, C; Wolf, A

    2015-05-01

    The detachment of the neurosensory retina from the underlying retinal pigment epithelium can be related to breaks of the retina allowing vitreous fluid to gain access to the subretinal space, to exudative changes of the choroid such as tumours or inflammatory diseases or to excessive tractional forces exerted by interactions of the collagenous vitreous and the retina. Tractional retinal detachment is usually treated by vitrectomy and exudative detachment can be addressed by treatment of the underlying condition in many cases. In rhegmatogenous retinal detachment two different surgical procedures, vitrectomy and scleral buckling, can be applied for functional and anatomic rehabilitation of our patients. The choice of the surgical procedure is not really standardised and often depends on the experience of the surgeon and other more ocular factors including lens status, the number of retinal breaks, the extent of the detachment and the amount of preexisting PVR. Using both techniques, anatomic success rates of over 90 % can be achieved. Especially in young phakic patients scleral buckling offers the true advantage to prevent the progression of cataract formation requiring cataract extraction and intraocular lens implantation. Therefore, scleral buckling should be considered in selected cases as an alternative surgical option in spite of the very important technical refinements in modern vitrectomy techniques. Georg Thieme Verlag KG Stuttgart · New York.

  3. Visualization of intermittent blobby plasma transport in attached and detached plasmas of the NAGDIS-II

    International Nuclear Information System (INIS)

    Ohno, Noriyasu; Furuta, Katsuhiro; Takamura, Shuichi

    2004-01-01

    We investigated the intermittent convective plasma transport in a attached and/or detached plasma condition of the linear divertor plasma simulator, NAGDIS-II. Images taken by a fast-imaging camera clearly show that in attached plasmas, blobs are peeled off the bulk plasma, and propagate outward with an azimuthal motion. In detached plasmas, plasma turbulence observed near the plasma recombining region drives strong intermittent radial plasma transport, which could broaden the radial density profile. (author)

  4. Conceptual design of a divertor Thomson scattering diagnostic for NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    McLean, A. G., E-mail: mclean@fusion.gat.com; Soukhanovskii, V. A.; Allen, S. L. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94550 (United States); Carlstrom, T. N. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); LeBlanc, B. P.; Ono, M.; Stratton, B. C. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

    2014-11-15

    A conceptual design for a divertor Thomson scattering (DTS) diagnostic has been developed for the NSTX-U device to operate in parallel with the existing multipoint Thomson scattering system. Higher projected peak heat flux in NSTX-U will necessitate application of advanced magnetics geometries and divertor detachment. Interpretation and modeling of these divertor scenarios will depend heavily on local measurement of electron temperature, T{sub e}, and density, n{sub e}, which DTS provides in a passive manner. The DTS design for NSTX-U adopts major elements from the successful DIII-D DTS system including 7-channel polychromators measuring T{sub e} to 0.5 eV. If implemented on NSTX-U, the divertor TS system would provide an invaluable diagnostic for the boundary program to characterize the edge plasma.

  5. The JET divertor coil

    International Nuclear Information System (INIS)

    Last, J.R.; Froger, C.; Sborchia, C.

    1989-01-01

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

  6. CIT divertor conceptual design

    International Nuclear Information System (INIS)

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

    1988-06-01

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

  7. Retinal detachment following endophthalmitis.

    Science.gov (United States)

    Nelsen, P T; Marcus, D A; Bovino, J A

    1985-08-01

    Fifty-five consecutive patients with a clinical diagnosis of bacterial endophthalmitis were reviewed. All patients were treated with systemic, periocular, topical, and intravitreal antibiotics. In addition, 33 of the patients underwent a pars plana vitrectomy. Nine retinal detachments occurred within six months of initial diagnosis. The higher frequency of retinal detachment in the vitrectomy group (21%) as compared to those patients managed without vitrectomy (9%) may be explained by a combination of surgical complications and the increased severity of endophthalmitis in the vitrectomy group. The two patients who developed retinal detachment during vitrectomy surgery rapidly progressed to no light perception. Conversely, the repair of retinal detachments diagnosed postoperatively had a good prognosis.

  8. 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

  9. 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

  10. Divertor pumping system with NBI cryopump for JT-60

    International Nuclear Information System (INIS)

    Akino, Noboru; Kuriyama, Masaaki; Ohga, Tokumichi; Seki, Hiroshi; Tanai, Yutaka

    1998-08-01

    The pumping system for JT-60 W-shape divertor with the NBI cryopump have been developed. The pumping speed achieved in the divertor region was 13-15 m 3 /s for deuterium gas with three units of the NBI cryopumps. In a simulation experiment of helium ash exhaust through the divertor, pumping of a mixed gas of helium and deuterium has been demonstrated using the NBI cryosorption pumps covered with an argon condensed layer. Control of neutral particle pressure in the divertor region became possible by having remodeled an aperture of the existing fast shutter, which is installed between the JT-60 vacuum vessel and NBI beam-line, to be regulated. (author)

  11. Divertor remote handling for DEMO: Concept design and preliminary FMECA studies

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

    Highlights: • Concept design of the RH system for the DEMO fusion power plant. • Divertor mover: hydraulic telescopic boom concept design. • An alternative solution to ITER rack and pinion divertor mover (CMM). • Divertor cassettes end effector studies. • FMECA studies started on the DEMO divertor mover. - Abstract: The paper describes a concept design of a remote handling (RH) system for replacing divertor cassettes and cooling pipes in future DEMO fusion power plant. In DEMO reactor design important considerations are the reactor availability and reliable maintenance operations. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel. Two alternative designs of the end effector to grip and manipulate the divertor cassette are presented in this work. Both concepts are hydraulically actuated, based on ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate. Taking advantage of the ITER RH background and experience, the proposed hydraulic RH system is compared with the rack and pinion system currently designed for ITER and is an object of simulations at Divertor Test Platform (DTP2) in VTT's Labs of Tampere, Finland. Pros and cons will be put in evidence.

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

    International Nuclear Information System (INIS)

    Bertolini, E.

    1995-01-01

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

  13. Development of a radiative divertor for DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Allen, S.L. [Lawrence Livermore National Lab., CA (United States); Brooks, N.H. [General Atomics, San Diego, CA (United States); Campbell, R.B. [Sandia National Labs., Albuquerque, NM (United States); Fenstermacher, M.E. [Lawrence Livermore National Lab., CA (United States); Hill, D.N. [Lawrence Livermore National Lab., CA (United States); Hyatt, A.W. [General Atomics, San Diego, CA (United States); Knoll, D.; Lasnier, C.J. [Lawrence Livermore National Lab., CA (United States); Lazarus, E.A. [Oak Ridge National Lab., TN (United States); Leonard, A.W. [General Atomics, San Diego, CA (United States); Lippmann, S.I. [General Atomics, San Diego, CA (United States); Mahdavi, M.A. [General Atomics, San Diego, CA (United States); Maingi, R. [Oak Ridge National Lab., TN (United States); Meyer, W. [Lawrence Livermore National Lab., CA (United States); Moyer, R.A. [California Univ., Los Angeles, CA (United States); Petrie, T.W. [General Atomics, San Diego, CA (United States); Porter, G.D. [Lawrence Livermore National Lab., CA (United States); Rensink, M.E. [Lawrence Livermore National Lab., CA (United States); Rognlien, T.D. [Lawrence Livermore National Lab., CA (United States); Schaffer, M.J. [General Atomics, San Diego, CA (United States); Smith, J.P. [General Atomics, San Diego, CA (United States); Staebler, G.M. [General Atomics, San Diego, CA (United States); Stambaugh, R.D. [General Atomics, San Diego, CA (United States); West, W.P. [General Atomics, San Diego, CA (United States); Wood, R.D. [Lawrence Livermore National Lab., CA (United States)

    1995-04-01

    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 {tau}{sub E} remains similar 2 times ITER-89P scaling. However, n{sub e} increases with D{sub 2} puffing, and Z{sub 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 ({delta}{approx}0.8) is important for high {tau}{sub E} VH-modes. Results from engineering feasibility studies, including diagnostic access, will be presented. ((orig.)).

  14. 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.)

  15. The remote exchange of the JET divertor

    International Nuclear Information System (INIS)

    Pick, M.

    1999-01-01

    In 1997 a series of experiments were performed in the JET machine using deuterium-tritium (D-T) mixtures and resulting in discharges with record breaking fusion power and fusion energy. The experiments demonstrated a key technology required for fusion, namely the on-line operation of a tritium fuel re-processing plant. These experiments left the inside of the JET vessel inaccessible to manned access for approximately one year. During this time, the complete Mark IIA divertor, a major system within the torus, was successfully removed and replaced with a new divertor design, the Mark II Gas Box divertor, using only remote handling techniques. This was the first application of the JET remote handling system and a demonstration of a further key ITER technology. The paper explains the methodology and operational approach taken to achieve the results using the remote handling system developed at JET. It describes the remote handling equipment including the force-reflecting servo-manipulator, the specialised tools designed, the facilities needed, and the trials, planning and training carried out to ensure the safe, reliable and rapid completion of the remote handling tasks. The planned tasks are outlined including the execution of the novel procedure for a remote, sub-millimetre precision, dimensional survey of the divertor support structure using digital photogrammetry. Furthermore the paper shows how the adaptability of the system was used to successfully undertake a large number of unplanned tasks including the removal of damaged tiles, a damaged diagnostic system and the vacuum cleaning of diagnostic windows. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Clever, Meike

    2010-07-01

    observed. Its absence can be explained using an extended two point model including heat convection applied to the region dominated by parallel transport (laminar region). The radial penetration depth of the neutral hydrogen particles ({lambda}{sub n} {approx} 3-4 cm) estimated from spectroscopic measurements was found to be often larger than the varying radial extent of this laminar region (few mm up to 6 cm) which finally leads to convective heat transport reducing parallel temperature gradients. Increasing the radial extent of the laminar region especially in front of the divertor strike points could lead to an improvement in this respect and provide access to a high recycling regime. The radiation instability developing at high plasma densities in the helical divertor in TEXTOR is preceded by a transient partial detachment of the plasma from the divertor target plates and leads to the formation of a poloidally structured and helically inclined radiating belt, a helical divertor MARFE. While typically leading to a density limit disruption, this MARFE has been stabilised using a feedback system and could provide some divertor functionality such as low target temperature, increased neutral density and increased radiation within the stochastic boundary. Simulations using two different cross-field transport coefficients showed, that an agreement is only found at a certain level of cross-field transport (D {sub perpendicular} {sub to} =1 m{sup 2}s{sup -1}). The inclusion of carbon impurities in the simulations results in the experimentally observed reduction of the recycling flux. (orig.)

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

    International Nuclear Information System (INIS)

    Clever, Meike

    2010-01-01

    observed. Its absence can be explained using an extended two point model including heat convection applied to the region dominated by parallel transport (laminar region). The radial penetration depth of the neutral hydrogen particles (λ n ∼ 3-4 cm) estimated from spectroscopic measurements was found to be often larger than the varying radial extent of this laminar region (few mm up to 6 cm) which finally leads to convective heat transport reducing parallel temperature gradients. Increasing the radial extent of the laminar region especially in front of the divertor strike points could lead to an improvement in this respect and provide access to a high recycling regime. The radiation instability developing at high plasma densities in the helical divertor in TEXTOR is preceded by a transient partial detachment of the plasma from the divertor target plates and leads to the formation of a poloidally structured and helically inclined radiating belt, a helical divertor MARFE. While typically leading to a density limit disruption, this MARFE has been stabilised using a feedback system and could provide some divertor functionality such as low target temperature, increased neutral density and increased radiation within the stochastic boundary. Simulations using two different cross-field transport coefficients showed, that an agreement is only found at a certain level of cross-field transport (D perpendicular to =1 m 2 s -1 ). The inclusion of carbon impurities in the simulations results in the experimentally observed reduction of the recycling flux. (orig.)

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

    International Nuclear Information System (INIS)

    Yamamoto, Shin

    1979-03-01

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

  19. Radiation loss and global energy balance of ohmically heated divertor discharge in JT-60 tokamak

    International Nuclear Information System (INIS)

    Koide, Yoshihiko; Yamada, Kimio; Yoshida, Hidetoshi; Nakamura, Hiroo; Niikura, Setsuo; Tsuji, Shunji

    1986-03-01

    Divertor experiment in JT-60 with a small divertor chamber has been successfully performed up to 1.6 MA discharge. Several divertor effects were experimentally confirmed as follows. Radiation loss in main plasma saturates with the increase of plasma current and its ratio to the input power is about 20 % at 1.5 MA. The rest of input power is exhausted into the divertor chamber and a half of it is dissipated as the radiation loss. Impurity accumulation is not observed during a few sec without internal MHD activity and gross impurity confinement time is several hundred msec. (author)

  20. NSTX Plasma Response to Lithium Coated Divertor

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-01-21

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

  1. 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

  2. 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.

  3. 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)

  4. 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)

  5. 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)

  6. 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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  8. 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)

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

    International Nuclear Information System (INIS)

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

    2007-11-01

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

  11. Negative ion detachment processes

    International Nuclear Information System (INIS)

    Champion, R.L.; Doverspike, L.D.

    1990-10-01

    This paper discusses the following topics: H - and D - collisions with atomic hydrogen; collisional decomposition of SF 6 - ; two-electron loss processes in negative ion collisions; associative electron detachment; and negative ion desorption from surfaces

  12. Retinal detachment repair

    Science.gov (United States)

    ... medicines Problems breathing You may not recover full vision. ... detachments can be repaired. Failure to repair the retina always results in loss of vision to some degree. After surgery, the quality of ...

  13. Serous Macular Detachments

    Directory of Open Access Journals (Sweden)

    Hakan Özdemir

    2012-03-01

    Full Text Available Serous macular detachment has only recently been recognized to occur in a significant number of eyes with macular pathology including diabetic retinopathy, retinal vein occlusion, Behçet disease, Irvine-Gass syndrome and pars planitis. These serous retinal detachments associated with retinal vascular leakage are not suspected clinically or angiographically but can be diagnosed with optical coherence tomography (OCT beneath the edematous neurosensory retina. The detection of shallow foveal detachment may be helpful in better understanding the pathogenesis of these disorders. In addition, the detection of serous macular detachment may also help to better guide and assess the results of therapy in the future. (Turk J Oph thal mol 2012; 42: 146-9

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

    Science.gov (United States)

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

    2013-10-01

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

  15. THERMAL HYDRAULIC ANALYSIS OF FIRE DIVERTOR

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  16. Kinematic Fitting of Detached Vertices

    Energy Technology Data Exchange (ETDEWEB)

    Mattione, Paul [Rice Univ., Houston, TX (United States)

    2007-05-01

    The eg3 experiment at the Jefferson Lab CLAS detector aims to determine the existence of the $\\Xi_{5}$ pentaquarks and investigate the excited $\\Xi$ states. Specifically, the exotic $\\Xi_{5}^{--}$ pentaquark will be sought by first reconstructing the $\\Xi^{-}$ particle through its weak decays, $\\Xi^{-}\\to\\pi^{-}\\Lambda$ and $\\Lambda\\to\\pi^{-}$. A kinematic fitting routine was developed to reconstruct the detached vertices of these decays, where confidence level cuts on the fits are used to remove background events. Prior to fitting these decays, the exclusive reaction $\\gamma D\\rightarrow pp\\pi^{-}$ was studied in order to correct the track measurements and covariance matrices of the charged particles. The $\\Lambda\\rightarrow p\\pi^{-}$ and $\\Xi^{-}\\to\\pi^{-}\\Lambda$ decays were then investigated to demonstrate that the kinematic fitting routine reconstructs the decaying particles and their detached vertices correctly.

  17. High Mach flow associated with plasma detachment in JT-60U

    International Nuclear Information System (INIS)

    Hatayama, A.; Hoshino, K.; Miyamoto, K.

    2003-01-01

    Recent new results of the high Mach flows associated with plasma detachment are presented on the basis of numerical simulations by a 2-D edge simulation code (the B2-Eirene code) and their comparisons with experiments in JT-60U W-shaped divertor plasma. High Mach flows appear near the ionization front away from the target plate. The plasma static pressure rapidly drops, while the total pressure is kept almost constant near the ionization front, because the ionization front near the X-point is clearly separated from the momentum loss region near the target plate. Redistribution from static to dynamic pressure without a large momentum loss is confirmed to be a possible mechanism of the high Mach flows. It has been also shown that the radial structure of the high Mach flow near the X point away from the target plate has a strong correlation with the DOD (Degree of Detachment) at the target plate. Also, we have made systematic analyses on the high Mach flows for both the 'Open' geometry and the 'W-shaped' geometry of JT-60U in order to clarify the geometric effects on the flows. (author)

  18. The divertor remote maintenance project

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  19. The divertor remote maintenance project

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  20. Current surgery of retinal detachment recurrence. Review

    Directory of Open Access Journals (Sweden)

    V. D. Zakharov

    2012-01-01

    Full Text Available this review presents a detailed analysis and an experience of surgical treatment of retinal detachment recurrence associated with light silicone oil tamponade of vitreous cavity. Approaches and variants of treatment were described in the historical aspect and till now. there are considered general and particular issues in case of retinal detachment recurrence appearance, expediency and volume of intraoperative manipulations, time of operation and choice of temporary substitute of vitreous body for a purpose of postoperative tamponade of vitreous cavity.

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

    NARCIS (Netherlands)

    Ďuran, I.; Ješko, K.; Fuchs, V.; Groth, M.; Guillemaut, C.; Gunn, J.P.; Horáček, J.; Pitts, R.A.; Tskhakaya, D.

    2015-01-01

    Abstract Higher than expected electron temperatures (Te) are often measured by divertor Langmuir probes (LP) in high recycling and detached regimes in JET and other tokamaks. As a possible mechanism to explain this discrepancy, we investigate the effect of penetration of fast, almost collisionless

  2. Module of lithium divertor for KTM tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Meslin, B

    1998-04-30

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

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

    Science.gov (United States)

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

    2017-10-01

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

  5. Variation of Particle Control with Changes in Divertor Geometry

    International Nuclear Information System (INIS)

    Petrie, T W; Allen, S L; Brooks, N H; Fenstermacher, M E; Ferron, J R; Greenfield, C M; Groth, M; Hyatt, A W; Leonard, A W; Luce, T C; Mahdavi, M A; Murakami, M; Porter, G D; Rensink, M E; Schaffer, M J; Wade, M R; Watkins, J G; West, W P; Wolf, N S

    2004-01-01

    Recent experiments on DIII-D point to the importance of two factors in determining how effectively the deuterium particle inventory in a tokamak plasma can be controlled through pumping at the divertor target(s): (1) the divertor magnetic balance, i.e., the degree to which the divertor topology is single-null (SN) or double-null (DN), and (2) the direction of the of Bx(divergent)B ion drift with respect to the X-point(s). Changes in divertor magnetic balance near the DN shape have a much stronger effect on the particle exhaust rate at the inner divertor target(s) than on the particle exhaust rate at the outer divertor target(s). The particle exhaust rate for the DN shape is strongest at the outer strike point opposite the Bx(divergent)B ion particle drift direction. Our data suggests that the presence of Bx(divergent)B and ExB ion particle drifts in the scrapeoff layer (SOL) and divertors play an important role in the particle exhaust rates of DN and near-DN plasmas. Particle exhaust rates are shown to depend strongly on the edge (pedestal) density n e,PED . In the lower range of densities considered in this study, i.e., n e,PED / n GREENWALD <0.4, particle exhaust rates are also found to be approximately proportional to the deuterium recycling intensity in front of the respective plenum entrance. Our results are shown to have implications for particle control in ITER and other future tokamaks

  6. Variation of particle control with changes in divertor geometry

    International Nuclear Information System (INIS)

    Petrie, T.W.; Allen, S.L.; Brooks, N.H.; Fenstermacher, M.E.; Groth, M.; Porter, G.D.; Rensink, M.E.; Wolf, N.S.; Ferron, J.R.; Greenfield, C.M.; Hyatt, A.W.; Leonard, A.W.; Luce, T.C.; Mahdavi, M.A.; Schaffer, M.J.; West, W.P.; Murakami, M.; Wade, M.R.; Watkins, J.G.

    2005-01-01

    Recent experiments on DIII-D point to the importance of two factors in determining how effectively the deuterium particle inventory in a tokamak plasma can be controlled through pumping at the divertor target(s): (1) the divertor magnetic balance, i.e., the degree to which the divertor topology is single-null (SN) or double-null (DN), and (2) the direction of the of Bx∇B ion drift with respect to the X-point(s). Changes in divertor magnetic balance near the DN shape have a much stronger effect on the particle exhaust rate at the inner divertor target(s) than on the particle exhaust rate at the outer divertor target(s). The particle exhaust rate for the DN shape is strongest at the outer strike point opposite the Bx∇B ion particle drift direction. Our data suggests that the presence of Bx∇B and ExB ion particle drifts in the scrapeoff layer (SOL) and divertors play an important role in the particle exhaust rates of DN and near-DN plasmas. Particle exhaust rates are shown to depend strongly on the edge (pedestal) density n e,PED . In the lower range of densities considered in this study, i.e., n e,PED /n GREENWALD <0.4, particle exhaust rates are also found to be approximately proportional to the deuterium recycling intensity in front of the respective plenum entrance. Our results are shown to have implications for particle control in ITER and other future tokamaks. (author)

  7. A computational study of operating regimes for poloidal divertors

    International Nuclear Information System (INIS)

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

    1982-01-01

    We have identified three theoretical operating regimes for poloidal divertors. These regimes are determined by the geometry of the divertor and the input energy and particle fluxes, and are characterized by the divertor plasma density and temperature. A fully self-consistent two-dimensional model for the plasma and neutral atom and molecule transport was used to study poloidal divertor operation. Extensions of our previous calculations important to this study were the inclusion of parallel electron and ion thermal conduction. We find that the key physics in divertor operation is the neutral recycling near the neutralizer plate. This can be parametrized by R = GAMMAsub(P)/GAMMAsub(O), the ratio of particle flux striking the neutralizer plate to the particle flux entering the divertor. Values of R approx. equal to 1 can be produced by large pumping rates near the neutralizer plates resulting in low neutral recycling and a high temperature, low density divertor plasma. By decreasing the pumping near the neutralizer plate, R can be raised to an intermediate value of 5-10, the plasma temperature lowered by the same factor, and the density raised by a factor of 10-30. In this regime, escape of the neutrals back to the main plasma is virtually blocked. By further restricting the pumping, R can be raised to twenty or more, thereby lowering the temperature by a factor of twenty or more and raising the density by a factor of ninety or more. Such high density regimes have been observed on D-III and appear to offer the most promise for impurity control and particle control on large reactor experiments such as INTOR or FED. In this paper, we explore the range 3 < R < 16. (orig.)

  8. Effects of discharge operation regimes and magnetic field geometry on the in-out divertor asymmetry in EAST

    International Nuclear Information System (INIS)

    Du, Hailong; Sang, Chaofeng; Wang, Liang; Bonnin, Xavier; Sun, Jizhong; Wang, Dezhen

    2016-01-01

    Highlights: • The in-out divertor asymmetry is studied using SOLPS. • The discharge operation and the magnetic filed have a great influence on the divertor asymmetry. • The asymmetry is not obvious in low recycling regime as that in high recycling regime. - Abstract: This paper aims to investigate the reason why the divertor in-out asymmetry was not obvious as experimentally observed in EAST only considering the classical drifts from previous simulations (Guo et al., J. Nucl. Mater. 438 (2013) 280; Du et al., J. Nucl. Mater. 463 (2015) 485). With consideration of the classical drifts, a series of different typical discharge scenarios in EAST in different magnetic field geometries were simulated by using the SOLPS5.2 code package. The simulated results reveal that the classical drifts make a major contribution to the in-out divertor asymmetry in the high recycling regime (HRR) and partial detachment (one divertor target begins to detach, while the other divertor remains attached) regime. In comparison, in low recycling regime the classical drifts play a much smaller role in the contributions to the in-out divertor asymmetry, which can explain reasonably the reason for it in Guo et al. (J. Nucl. Mater. 438 (2013) 280). In addition, the magnetic field geometry also has a great impact on the classical drifts inducing the asymmetry; it is found that for lower single-null, upper single-null and connected double-null topologies, in HRR the classical drifts play an dominant role in the contribution to the in-out divertor asymmetry, while for a disconnected double null magnetic field configuration, they play a minor role, which is the reason why the in-out asymmetry was unobvious by considering the drifts in Du et al. (J. Nucl. Mater. 463 (2015) 485).

  9. Effects of discharge operation regimes and magnetic field geometry on the in-out divertor asymmetry in EAST

    Energy Technology Data Exchange (ETDEWEB)

    Du, Hailong; Sang, Chaofeng [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China); Wang, Liang [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Bonnin, Xavier [LSPM-CNRS, Université Paris 13, Sorbonne Paris Cité, F-93430 Villetaneuse (France); Sun, Jizhong [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China); Wang, Dezhen, E-mail: wangdez@dlut.edu.cn [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China)

    2016-11-01

    Highlights: • The in-out divertor asymmetry is studied using SOLPS. • The discharge operation and the magnetic filed have a great influence on the divertor asymmetry. • The asymmetry is not obvious in low recycling regime as that in high recycling regime. - Abstract: This paper aims to investigate the reason why the divertor in-out asymmetry was not obvious as experimentally observed in EAST only considering the classical drifts from previous simulations (Guo et al., J. Nucl. Mater. 438 (2013) 280; Du et al., J. Nucl. Mater. 463 (2015) 485). With consideration of the classical drifts, a series of different typical discharge scenarios in EAST in different magnetic field geometries were simulated by using the SOLPS5.2 code package. The simulated results reveal that the classical drifts make a major contribution to the in-out divertor asymmetry in the high recycling regime (HRR) and partial detachment (one divertor target begins to detach, while the other divertor remains attached) regime. In comparison, in low recycling regime the classical drifts play a much smaller role in the contributions to the in-out divertor asymmetry, which can explain reasonably the reason for it in Guo et al. (J. Nucl. Mater. 438 (2013) 280). In addition, the magnetic field geometry also has a great impact on the classical drifts inducing the asymmetry; it is found that for lower single-null, upper single-null and connected double-null topologies, in HRR the classical drifts play an dominant role in the contribution to the in-out divertor asymmetry, while for a disconnected double null magnetic field configuration, they play a minor role, which is the reason why the in-out asymmetry was unobvious by considering the drifts in Du et al. (J. Nucl. Mater. 463 (2015) 485).

  10. Improved performance of the W7-AS stellarator with the new island divertor

    International Nuclear Information System (INIS)

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

    2003-01-01

    The island divertor concept has successfully been realized at W7-AS. The divertor gives access to a new NBI-heated high density regime with densities up to 4x1020 m-3 and energy confinement well above customary scalings (high density H-mode, HDH). In parallel, impurity confinement dramatically decreases. This prevents impurity accumulation, and allows for quasi-stationary high density discharges with excellent confinement and with radiation profiles peaked at the edge. At the highest densities partial detachment occurs with a radiated power fraction up to 90% at a tolerable mitigation of plasma energy. The HDH-mode appears promising with respect to the requirements of both confinement and exhaust. Major experimental results, such as the lack of a high recycling phase preceding detachment, are also predicted by the EMC3/EIRENE code. (author)

  11. TCV divertor upgrade for alternative magnetic configurations

    Directory of Open Access Journals (Sweden)

    H. Reimerdes

    2017-08-01

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

  12. Rhegmatogenous Retinal Detachment Associated with Choroidal Detachment

    Directory of Open Access Journals (Sweden)

    L. Sh. Bilandarli

    2017-01-01

    Full Text Available Review describes the theme of rhegmatogenous retinal dеtаchment associated with choroidal separation. It is rare, but quite severe eye pathology. In most cases it has a very poor prognosis. Most authors consider the retinal detachment as a primary pathogenetic part, which decompensates the production of aqueous humor by increasing the absorptive surface of the retinal pigment epithelium. Dilatation of choroidal arterioles occurs in hypotension, it leads to extravasation of protein-rich fluid in the choroidal and the suprachoroidal space. This helps to further swelling and separation of the ciliary body and the choroid with reduced production of aqueous humor and progressive hypotension. There is a high risk of developing “retino-choroidal” separation in patients with macular rupture due to localization of retinal separation and rupture rear hyaloid membrane. The protein level in aqueous humor can be increased to 70 times. It may be a result of reflux of suprachoroidal proteins through uveoscleral route and / or venous proteins through the trabecular network. In addition, the diffusion of proteins from the posterior camera and vitreous cavity is possible. This creates favorable conditions for cell proliferation that can lead to postoperative proliferative vitreoretinopathy. Typically patients have a pronounced signs of inflammation, pain, and “red eye”, which is accompanied with vision decrement. Rhegmatogenous retinal reparationcan be associated with such clinical symptoms as severe panuveit, deepening of the anterior camera and the inflammatory response in the moisture, concentric wrinkles and sagging back of the iris, posterior synechia, iridofakodenez, blurred vitreous detachment of the ciliary body, hypotension, and choroidal and retinal detachment in addition. Debatableness of etiopathogenesis and a clinical picture, which is similar to other eye diseases create significant difficulties in early diagnosis and proper treatment of

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

    Science.gov (United States)

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

    2012-10-01

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

  14. 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

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

    Science.gov (United States)

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

    2016-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  18. 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

  19. Divertor design through shape optimization

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  20. Snowflake divertor plasmas on TCV

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  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. Features and Initial Results of the DIII-D Advanced Tokamak Radiative Divertor

    International Nuclear Information System (INIS)

    R.C. O'Neill; A.S. Bozek; M.E. Friend; C.B. Baxi; E.E. Reis; M.A. Mahdavi; D.G. Nilson; S.L. Allen; W.P. West

    1999-01-01

    The Radiative Divertor Program of DIII-D is in its final phase with the installation of the cryopump and baffle structure (Phase 1B Divertor) in the upper inner radius of the DIII-D vacuum vessel at the end of this calendar year. This divertor, in conjunction with the Advanced Divertor and the Phase 1A Divertor, located in the lower and upper outer radius of the DIII-D vacuum vessel respectively, provides pumping for density control of the plasma while minimizing the effects on the core confinement. Each divertor consists of a cryobelium cooling ring and a shielded protective structure. The cryo/helium-cooled pumps of all three diverters exhaust helium from the plasma. The protective shielded structure or baffle structure, in the case of the diverters located at the top of the vacuum vessel, provides baffling of neutral charged particles and minimize the flow of impurities back into the core of the plasma. The baffles, which consist of water-cooled panels that allow for the attachment of tiles of various sizes and shapes, house gas puff systems. The intent of the puffing systems is to inject gas in and around the divertor to minimize the heat flux on specific areas on the divertor and its components. The reduction of the heat flux on the divertor minimizes the impurities that are generated from excess heat on divertor components, specifically tiles. Experiments involving the gas puff systems and the divertor structures have shown the heat flux can be spread over a large area of the divertor, reducing the peak heat flux in specific areas. The three diverters also incorporate a variety of diagnostic tools such as halo current monitors, magnetic probes and thermocouples to monitor certain plasma characteristics as well as determine the effectiveness of the cryopumps and baffle configurations. The diverters were designed to optimize pumping performance and to withstand the electromagnetic loads from both halo currents and toroidal induced currents. Incorporated also

  3. Self-sustained detachment in the Large Helical Device

    International Nuclear Information System (INIS)

    Miyazawa, J.; Masuzaki, S.; Sakamoto, R.; Arimoto, H.; Kondo, K.; Tamura, N.; Shoji, M.; Nishiura, M.; Murakami, S.; Funaba, H.; Peterson, B.J.; Sakakibara, S.; Kobayashi, M.; Tanaka, K.; Narihara, K.; Yamada, I.; Morita, S.; Goto, M.; Osakabe, M.; Ashikawa, N.; Morisaki, T.; Nishimura, K.; Yamada, H.; Ohyabu, N.; Komori, A.; Motojima, O.

    2006-01-01

    Self-sustained detachment has been obtained in the Large Helical Device (LHD). Strong hydrogen gas puffing of ∼200 Pa m 3 s -1 after a density feedback phase detaches the plasma from the divertor plates with high reproducibility. High electron density of over 1 x 10 20 m -3 is sustained without gas puffing until the heating beam stops and a high-density flat top for 2 s has been demonstrated. Throughout the self-sustained detachment phase, the minor radius of the hot plasma column shrinks to ∼90% of the last-closed-flux-surface, which corresponds to the ι-bar ι/2π = 1/q =1 rational surface. This new state has been named the 'Serpens mode', for self-regulated plasma edge 'neath the last-closed-flux-surface. Global energy confinement of the Serpens mode is compared with the international stellarator scaling 1995 (ISS95) and the recently established scaling for high-density LHD plasmas (HD scaling), where shrinking confinement volume and shallow penetration of the heating beams are taken into account. Although the energy confinement of the Serpens mode seems deteriorated compared with ISS95, as in the case of high-density attached plasmas, it is consistent with the HD scaling. This suggests that the energy confinement properties of detached plasmas in LHD are similar to those in high-density attached plasmas

  4. FEM investigation and thermo-mechanic tests of the new solid tungsten divertor tile for ASDEX Upgrade

    International Nuclear Information System (INIS)

    Jaksic, Nikola; Greuner, Henri; Herrmann, Albrecht

    2013-01-01

    Highlights: • New solid tungsten divertor for fusion experiment ASDEX Upgrade. • Design validation in the high heat flux (HHF) test facility GLADIS (Garching Large Divertor Sample Test Facility). • FEA simulation. -- Abstract: A new solid tungsten divertor for the fusion experiment ASDEX Upgrade is under construction at present. A new divertor tile design has been developed to improve the thermal performance of the current divertor made of tungsten coated fine grain graphite. Compared to thin tungsten coatings, divertor tiles made of massive tungsten allow to extend the operational range and to study the plasma material interaction of tungsten in more detail. The improved design for the solid tungsten divertor was tested on different full scale prototypes with a hydrogen ion beam. The influence of a possible material degradation due to thermal cracking or recrystallization can be studied. Furthermore, intensive Finite Element Method (FEM) numerical analysis with the respective test parameters has been performed. The elastic–plastic calculation was applied to analyze thermal stress and the observed elastic and plastic deformation during the heat loading. Additionally, the knowledge gained by the tests and especially by the numerical analysis has been used to optimize the shape of the divertor tiles and the accompanying divertor support structure. This paper discusses the main results of the high heat flux tests and their numerical simulations. In addition, results from some special structural mechanic analysis by means of FEM tools are presented. Finally, first results from the numerical lifecycle analysis of the current tungsten tiles will be reported

  5. Safety detaching hook specification.

    CSIR Research Space (South Africa)

    Roux, JD

    1999-05-01

    Full Text Available hydraulic tensioning system The impactor must subject the safety detaching hook to an impact energy of 150 kJ minimum. A high speed digital imaging system, capable of recording at least 2000 full frames per second, shall be used to record video images... document compiled by the Contractor, detailing all major events in the production phase, including but not necessarily limited to inspection requirements, test procedures and acceptance/rejection criteria, sampling plans and equipment to be employed...

  6. Divertor development for a future fusion power plant

    International Nuclear Information System (INIS)

    Norajitra, Prachai

    2011-01-01

    Nuclear fusion is considered as a future source of sustainable energy supply. In the first chapter, the physical principle of magnetic plasma confinement, and the function of a tokamak are described. Since the discovery of the H-mode in ASDEX experiment ''Divertor I'' in 1982, the divertor has been an integral part of all modern tokamaks and stellarators, not least the ITER machine. The goal of this work is to develop a feasible divertor design for a fusion power plant to be built after ITER. This task is particularly challenging because a fusion power plant formulates much greater demands on the structural material and the design than ITER in terms of neutron wall load and radiation. First several divertor concepts proposed in the literature e.g. the Power Plant Conceptual Study (PPCS) using different coolants are reviewed and analyzed with respect to their performance. As a result helium cooled divertor concept exhibited the best potential to come up to the highest safety requirements and therefore has been chosen for the design process. From the third chapter the necessary steps towards this goal are described. First, the boundary conditions for the arrangement of a divertor with respect to the fusion plasma are discussed, as this determines the main thermal and neutronic load parameters. Based on the loads material selection criteria are inherently formulated. In the next step, the reference design is defined in accordance with the established functional design specifications. The developed concept is of modular nature and consists of cooling fingers of tungsten using an impingement cooling in order to achieve a heat dissipation of 10 MW/m 2 . In the next step, the design was subjected to the thermal-hydraulic and thermo-mechanical calculations in order to analyze and improve the performance and the manufacturing technologies. Based on these results, a prototype was produced and experimentally tested on their cooling capacity, their thermo-cyclic loading

  7. 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.

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

    International Nuclear Information System (INIS)

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

    1997-06-01

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

  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. Between Involvement and Detachment

    DEFF Research Database (Denmark)

    Thomasen, Gry

    that the Department of State attempted a ‘political bargain’, with which the allies would be given political consultation and a détente in return for re-commitment to integration, whereas the Acheson Committee proposed a détente and deterrence formula in NATO to the overcome this perceived alliance disintegration....... Thus the US proposed the Harmel formula before Harmel. In general, the developments in Western Europe put the Johnson administration in a state of alarm, and the European allies therefore had a larger impact on America’s policies, except in the essentially detached nuclear policy, which...

  11. Fabrication of divertor cassette for ITER

    International Nuclear Information System (INIS)

    Sanguinetti, G.P.

    2008-01-01

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

  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. Plasma shape control calculations for BPX divertor design

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  15. Divertor plate for thermonuclear reactor

    International Nuclear Information System (INIS)

    Yamazaki, Seiichiro; Sato, Keisuke; Nishio, Satoshi.

    1993-01-01

    In a divertor plate for a thermonuclear reactor, adjacent cooling pipes are electrically insulated from each other and pipes made of a gradient functional material prepared by compositing ceramics having an insulation property and metals are metallurgically joined to at least one portion of each of the cooling pipes. Electric current caused upon occurrence of plasma disruption is interrupted by the insulation portion, so that a large circuit is not formed and electromagnetic force is decreased to such a extent that the divertor plate is not ruptured. Since a header of the cooling pipes can be installed at any optional position, the installation space can be reduced. Further, since inlet and exit collection headers can be disposed on both ends of the cooling pipes, it is possible to shorten the length of the cooling pipe of the divertor plate corresponded to high heat fluxes and reduce the pressure loss on the side of coolants to about 1/2. Further, turn back portions of small radius of curvature of the cooling pipes are eliminated to reduce the cost and extend the lifetime and, in addition, protection tiles can be attached easily. (N.H.)

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

    International Nuclear Information System (INIS)

    Boedo, J.A.; Lehmer, R.; Moyer, R.A.; Watkins, J.G.; Porter, G.D.; Evans, T.E.; Leonard, A.W.; Schaffer, M.J.

    1998-06-01

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

  17. 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.

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

    International Nuclear Information System (INIS)

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

    1991-02-01

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

  19. 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

  20. Engineering conceptual design of CFETR divertor

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  1. A large divertor manipulator for ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  2. Steady-state exhaust of helium ash in the W-shaped divertor of JT-60U

    International Nuclear Information System (INIS)

    Sakasai, A.; Takenaga, H.; Hosogane, N.

    2001-01-01

    By injecting a neutral beam of 60 keV helium (He) atoms as central fueling of helium into the ELMy H-mode plasmas, helium exhaust has been studied in the W-shaped pumped divertor on JT-60U. Efficient He exhaust was realized by He pumping using argon frosted cryopumps in the JT-60U new divertor. In steady state, good He exhaust capability (τ He */τ E =4 and high enrichment factor, where τ He * is a global particle confinement time of helium and τ E is the energy confinement time) was successfully demonstrated in attached ELMy H-mode plasmas. Good He exhaust capability was also obtained in detached ELMy H-mode plasmas, which was comparable to one in attached plasmas. This result of the helium exhaust is sufficient to support a detached divertor operation on ITER. After the divertor modification, helium exhaust in reversed shear plasmas has been investigated using He gas puff. Helium removal inside the internal transport barrier (ITB) is about two times as difficult as that outside the ITB in reversed shear discharges. (author)

  3. 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

  4. [Prophylactic treatment of retinal detachment].

    Science.gov (United States)

    Binder, S; Riss, B

    1981-08-01

    The indications for and results of prophylactic treatment of retinal detachment during a period of five years are reported and compared with the results in the literature. Half of the cases (3 out of 6 eyes) which developed a retinal detachment had been horse-shoe tears combined with a vitreous hemorrhage. For this reason a small buckle operation is recommended in these cases, to prevent further traction. Lattice degeneration should rather be observed than treated, except in special cases: This includes eyes where the fellow eye had a detachment from a lattice degeneration, cases in which one eye is blind from an uncured detachment or has no useful visual acuity, and eyes whose fellow eye has giant tears. In aphakic eyes treatment of lattice degeneration is recommended, because the incidence of detachment from these areas is high, especially in young aphakic cases. In one aphakic eye which had been photocoagulated several times the formation of a preretinal membrane was observed.

  5. Moving Divertor Plates in a Tokamak

    International Nuclear Information System (INIS)

    Zweben, S.J.; Zhang, H.

    2009-01-01

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

  6. 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

  7. Moving Divertor Plates in a Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S.J. Zweben, H. Zhang

    2009-02-12

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

  8. Divertor power and particle fluxes between and during type-I ELMs in the ASDEX Upgrade

    Science.gov (United States)

    Kallenbach, A.; Dux, R.; Eich, T.; Fischer, R.; Giannone, L.; Harhausen, J.; Herrmann, A.; Müller, H. W.; Pautasso, G.; Wischmeier, M.; ASDEX Upgrade Team

    2008-08-01

    Particle, electric charge and power fluxes for type-I ELMy H-modes are measured in the divertor of the ASDEX Upgrade tokamak by triple Langmuir probes, shunts, infrared (IR) thermography and spectroscopy. The discharges are in the medium to high density range, resulting in predominantly convective edge localized modes (ELMs) with moderate fractional stored energy losses of 2% or below. Time resolved data over ELM cycles are obtained by coherent averaging of typically one hundred similar ELMs, spatial profiles from the flush-mounted Langmuir probes are obtained by strike point sweeps. The application of simple physics models is used to compare different diagnostics and to make consistency checks, e.g. the standard sheath model applied to the Langmuir probes yields power fluxes which are compared with the thermographic measurements. In between ELMs, Langmuir probe and thermography power loads appear consistent in the outer divertor, taking into account additional load due to radiation and charge exchange neutrals measured by thermography. The inner divertor is completely detached and no significant power flow by charged particles is measured. During ELMs, quite similar power flux profiles are found in the outer divertor by thermography and probes, albeit larger uncertainties in Langmuir probe evaluation during ELMs have to be taken into account. In the inner divertor, ELM power fluxes from thermography are a factor 10 larger than those derived from probes using the standard sheath model. This deviation is too large to be caused by deficiencies of probe analysis. The total ELM energy deposition from IR is about a factor 2 higher in the inner divertor compared with the outer divertor. Spectroscopic measurements suggest a quite moderate contribution of radiation to the target power load. Shunt measurements reveal a significant positive charge flow into the inner target during ELMs. The net number of elementary charges correlates well with the total core particle loss

  9. Rapidly Moving Divertor Plates In A Tokamak

    International Nuclear Information System (INIS)

    Zweben, S.

    2011-01-01

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

  10. 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

  11. 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)

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

    Directory of Open Access Journals (Sweden)

    W. Dekeyser

    2017-08-01

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

  13. 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)

  14. 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

  15. 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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Brunner, D.; Umansky, M.V.; LaBombard, B.; Rognlien, T.D.

    2013-01-01

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

  19. Multiple equilibria of divertor plasmas

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  20. Shape Evolution of Detached Bridgman Crystals Grown in Microgravity

    Science.gov (United States)

    Volz, M. P.; Mazuruk, K.

    2015-01-01

    A theory describing the shape evolution of detached Bridgman crystals in microgravity has been developed. A starting crystal of initial radius r0 will evolve to one of the following states: Stable detached gap; Attachment to the crucible wall; Meniscus collapse. Only crystals where alpha plus omega is great than 180 degrees will achieve stable detached growth in microgravity. Results of the crystal shape evolution theory are consistent with predictions of the dynamic stability of crystallization (Tatarchenko, Shaped Crystal Growth, Kluwer, 1993). Tests of transient crystal evolution are planned for ICESAGE, a series of Ge and GeSi crystal growth experiments planned to be conducted on the International Space Station (ISS).

  1. Divertor Design and Physics Issues of Huge Power Handling for SlimCS Demo Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Asakura, N.; Hoshino, K.; Tobita, K.; Someya, Y.; Utoh, H.; Nakamura, M., E-mail: asakura.nobuyuki@jaea.go.jp [Japan Atomic Energy Agency, Rokkasho (Japan); Shimizu, K. [Japan Atomic Energy Agency, Naka (Japan); Takizuka, T. [Osaka University, Osaka (Japan)

    2012-09-15

    Full text: Power exhaust scenario for a 3 GW class fusion reactor with the ITER-size plasma has been developed with enhancing the radiation loss from seeding impurity. Transport of plasma, impurity and neutrals was simulated self-consistently, for the first time, under the Demo divertor condition using an integrated divertor simulation code SONIC. The total heat load, q{sub target}, was evaluated including radiation power load and neutral load, in addition to the plasma heat load. It was found that heat and particle diffusion coefficients significantly affect the plasma detachment. For the case of increasing the coefficients by the factor of two, peak q{sub target} is reduced from 18 MW/m{sup 2} to below the engineering design level of 10 MW/m{sup 2}, while the characteristic width of the heat flux at the midplane SOL increases slightly from 2.2 to 2.7 mm. It was also found that that enhancement of the local {chi} and D at the outer SOL affects a reduction in the peak q{sub target} near the separatrix. Effects of the divertor geometry such as the divertor leg were investigated. Outer divertor leg length was extended to 2.7 m, while the magnetic flux expansion at the target is reduced to a half compared to the reference case of 1.8 m. Large radiation volume is shifted further upstream from the target due to a reduction in T{sub e}. The peak q{sub target} decreases to 10 MW/m{sup 2} due to reduction in both the plasma heat load and the radiation power load. (author)

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  3. Investigation of electron parallel pressure balance in the scrape-off layer of deuterium-based radiative divertor discharges IN DIII-D

    International Nuclear Information System (INIS)

    Petrie, T.W.; Carlstrom, T.N.; Allen, S.L.

    1996-10-01

    Electron density, temperature, and parallel pressure measurements at several locations along field lines connecting the midplane scrapeoff layer (SOL) with the outer divertor are presented for both attached and partially-detached divertor cases: I p = 1.4 MA, q 95 = 4.2, and P input ∼ 6.7 MW under ELMing H-mode conditions. At the onset of the Partially Detached Divertor (PDD), a high density, low temperature plasma forms in the divertor SOL (divertor MARFE). The electron pressure drops by a factor of ∼ 2 between the midplane separatrix and the X-point, and then an additional ∼3--5 times between the X-point and the outboard separatrix strike point. These results are in contrast to the attached (non-PDD) case, where electron pressure in the SOL is reduced by, at most, a factor of two between the midplane and the divertor target. Divertor MARFEs generally have only marginal adverse impact on important H-mode characteristics, such as confinement time. In fact, PDD discharges at low input power maintains good H-mode characteristics until a high density, low temperature plasma abruptly forms inside the separatrix near the X-point (X-point MARFE). Concurrent with the appearance of this X-point MARFE is a degradation in both energy confinement and the plasma fueling rate, and an increase in the carbon impurity concentration inside the core plasma. The formation of the X-point MARFE is consistent with a thermal instability resulting from the temperature dependence of the carbon radiative cooling rate in the range ∼ 7--30 eV

  4. Outcome of surgery in post-cytomegalovirus retinal detachment: Experience before and in the era of highly active anti-retroviral therapy in Indian eyes

    Directory of Open Access Journals (Sweden)

    Ramandeep Singh

    2013-01-01

    Full Text Available Purpose: To evaluate the outcome of surgery for cytomegalovirus associated retinal detachment (CMVRD in human immunodeficiency virus (HIV-infected patients in pre-highly active antiretroviral therapy (HAART and HAART era in Indian eyes. Materials and Methods: Retrospective, we reviewed medical records of all consecutive HIV patients, who underwent surgical repair for CMVRD from July 1998 to June 2011. We divided patients into two groups, i.e. group 1, pre HAART era and group 2, HAART era. We compared two groups for various parameters like visual outcome, surgical success, additional procedures, follow-up, etc., Results: Twenty-eight eyes of 26 patients were included; 12 eyes of the 11 patients in group 1 and 16 eyes of the 15 patients in group 2. Significant visual acuity improvement was seen in both groups. Complete anatomic success was seen in 11 eyes in group 1 and 15 eyes in group 2. One additional procedure in group 1 and 29 additional procedures were done in group 2. A mean follow-up was 16 months in group 1 and 41 months in group 2. Conclusion: There was no difference in outcome in pre-HAART and HAART group, except for longer follow-up and additional surgical procedures in HAART group.

  5. An Asdex-type divertor for ITER

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1989-01-01

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

  6. Comparison of kinetic and fluid neutral models for attached and detached state

    International Nuclear Information System (INIS)

    Furubayashi, M.; Hoshino, K.; Toma, M.; Hatayama, A.; Coster, D.; Schneider, R.; Bonnin, X.; Kawashima, H.; Asakura, N.; Suzuki, Y.

    2009-01-01

    Neutral behavior has an important role in the transport simulations of the edge plasma. Most of the edge plasma transport codes treat neutral particles by a simple fluid model or a kinetic model. The fluid model allows faster calculations. However, the applicability of the fluid model is limited. In this study, simulation results of JT-60U from kinetic neutral model and fluid neutral model are compared under the attached and detached state, using the 2D edge plasma code package, SOLPS5.0. In the SOL region, no significant differences are observed in the upstream plasma profiles between kinetic and fluid neutral models. However, in the divertor region, large differences are observed in plasma and neutral profiles. Therefore, further optimization of the fluid neutral model should be performed. Otherwise kinetic neutral model should be used to analyze the divertor region.

  7. NSTX plasma operation with a Liquid Lithium Divertor

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  8. Overview of HL-2A experiment results

    International Nuclear Information System (INIS)

    Yang, Q.W.; Yong Liu; Ding, X.T.

    2007-01-01

    Recent experiment results from the HL-2A tokamak are presented in this paper. Supersonic molecular beam injection (SMBI) with liquid nitrogen temperature propellant is used. Low temperature SMBI can form hydrogen clusters that penetrate into the plasma more deeply and efficiently. Particle diffusion coefficient and convection velocity (D = 0.5-1.5 m 2 s -1 and V conv -1 , respectively) are obtained at the plasma periphery using modulated SMBI. Multi-probe measurements reveal the m = 0-1, n = 0 symmetries of directly measured low frequency (7-9 kHz) electric potential and field are simultaneously observed for the first time. Impurity transport is determined with the laser blow-off system and transport code. A disruption predictor has been derived based on MHD activity observations and statistical analysis. Sawtooth characteristics during ECRH are investigated and coupling between m = 1 and m/n = 2/1 modes is studied. Detachment features of HL-2A divertor are numerically and experimentally studied using the code SOLPS5.0 and measured data. The long divertor legs and thin divertor throats in HL-2A pose MHD shaping problems resulting in momentum losses even at low densities and strongly enhanced main chamber losses

  9. Divertor Heat Flux Reduction by Resonant Magnetic Perturbations in the LHD-Type Helical DEMO Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yanagi, N.; Sagara, A.; Goto, T.; Masuzaki, S.; Miyazawa, J., E-mail: yanagi@lhd.nifs.ac.jp [National Institute for Fusion Science, Toki (Japan)

    2012-09-15

    Full text: The conceptual design studies of the LHD-type helical fusion DEMO reactor, FFHR-d1, are progressing steadfastly. The LHD-type heliotron magnetic configuration equipped with the built- in helical divertors has a potential to realize low divertor heat flux in spatial average. However, the toroidal asymmetry may give more than a couple of times higher peak heat flux at some locations, as has been experimentally observed in LHD and confirmed by magnetic field-line tracing. By providing radiation dispersion accompanied with a plasma detachment, the heat flux may decrease significantly though the compatibility with a good core plasma confinement is an important issue to be explored. Whereas the engineering difficulties for developing materials to be used under the neutron environment require even further decrease of the heat flux (even though the heliotron is a unique configuration that divertor plates be largely shielded from the direct irradiation of neutrons by breeder blankets). In this respect, we proposed, in the last IAEA FEC, a new strike point sweeping scheme using a set of auxiliary helical coils, termed helical divertor (HD) coils. The HD coils carrying a few percent of the current amplitude of the main helical coils sweep the divertor strike points without altering the core plasma. Though this scheme is effective in dispersing the heat flux in the poloidal direction, the toroidal asymmetry still remains. The AC operation may also give unforeseen engineering difficulties. We here propose that the peak heat flux be mitigated using RMP fields in steady-state. The magnetic field-lines are numerically traced in the vacuum configuration and their footprints coming to the divertor regions are counted. Their fraction plotted as a function of the toroidal angle indicates that the peak heat flux be mitigated to {approx} 20 MW per square meters at 3 GW fusion power generation without having radiation dispersion when an RMP field is applied. We note that the

  10. 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)

  11. Divertor cassette movers prototypes for ITER

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  12. A solid tungsten divertor for ASDEX Upgrade

    International Nuclear Information System (INIS)

    Herrmann, A; Greuner, H; Jaksic, N; Böswirth, B; Maier, H; Neu, R; Vorbrugg, S

    2011-01-01

    The conceptual design of a solid tungsten divertor for ASDEX Upgrade (AUG) is presented. The Div-III design is compatible with the existing divertor structure. It re-establishes the energy and heat receiving capability of a graphite divertor and overcomes the limitations of tungsten coatings. In addition, a solid tungsten divertor allows us to investigate erosion and bulk deuterium retention as well as test castellation and target tilting. The design criteria as well as calculations of forces due to halo and eddy currents are presented. The thermal properties of the proposed sandwich structure are calculated with finite element method models. After extensive testing of a target tile in the high heat flux test facility GLADIS, two solid tungsten tiles were installed in AUG for in-situ testing.

  13. Stochasticity about a poloidal divertor separatrix

    International Nuclear Information System (INIS)

    Skinner, D.A.; Osborne, T.H.; Prager, S.C.; Park, W.

    1986-10-01

    The stochasticization of the magnetic separatrix due to the presence of a helical perturbation in a poloidal divertor tokamak is illustrated by a numerical computation which traces magnetic field lines

  14. 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

  15. ARIES-III divertor engineering design

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.C.; Schultz, K.R. [General Atomics, San Diego, CA (United States); Cheng, E.T. [TSI Research, Solana Beach, CA (United States); Grotz, S.; Hasan, M.A.; Najmabadi, F.; Sharafat, S. [California Univ., Los Angeles, CA (United States). Dept. of Mechanical, Aerospace and Nuclear Engineering; Brooks, J.N.; Ehst, D.A.; Sze, D.K. [Argonne National Lab., IL (United States); Herring, J.S. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Valenti, M.; Steiner, D. [Rensselaer Polytechnic Inst., Troy, NY (United States). Plasma Dynamics Lab.

    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{sup 2}, a conventional divertor design can be used. The divertor surface is contoured to have a constant heat flux of 5.4 MW/m{sup 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.

  16. The WEST programme: Minimizing technology and operational risks of a full actively cooled tungsten divertor on ITER

    Energy Technology Data Exchange (ETDEWEB)

    Grosman, André, E-mail: andre.grosman@cea.fr [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Bucalossi, Jérôme; Doceul, Louis [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Escourbiac, Frédéric [ITER Organization, Cadarache, 13115 St. Paul-lez-Durance (France); Lipa, Manfred [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Merola, Mario [ITER Organization, Cadarache, 13115 St. Paul-lez-Durance (France); Missirlian, Marc [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Pitts, Richard A. [ITER Organization, Cadarache, 13115 St. Paul-lez-Durance (France); Samaille, Franck; Tsitrone, Emmanuelle [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France)

    2013-10-15

    Highlights: ► The WEST programme is a unique opportunity to experience the industrial scale manufacture of tungsten plasma-facing components similar to the ITER divertor ones. ► In Tore Supra, it will bring important know how for actively cooled W divertor operation. ► This can be done by a reasonable modification of the Tore Supra tokamak. ► A fast implementation of the project would make this information available in due time. ► This allows a significant contribution to the W ITER divertor risk minimization in its manufacturing and operation phase. -- Abstract: The WEST programme consists in transforming the Tore Supra tokamak into an X point divertor device, while taking advantage of its long discharge capability. This is obtained by inserting in vessel coils to create the X point while adapting the in-vessel elements to this new geometry. This will allow the full tungsten divertor technology to be used on ITER to be tested in anticipation of its use on ITER under relevant heat loading conditions and pulse duration. The early manufacturing of a significant industrial series of ITER-similar W plasma-facing units will contribute to the ITER divertor manufacturing risk mitigation and to that associated with early W divertor plasma operation on ITER.

  17. Disappearance of a detached vapor mass in subcooled water

    International Nuclear Information System (INIS)

    Inada, Shigeaki; Miyasaka, Yoshiki; Izumi, Ryotaro.

    1986-01-01

    Experiments on pool transition boiling of water under atmospheric pressure on a heated surface 10 mm in diameter were conducted for subcooling 15 - 50 K. The mass flux of condensation of a detached coalescent vapor bubble was experimentally estimated by a mathematical model based on the mass transfer mechanism of condensation. As a result, it is clarified that the mass flux of condensation of the detached bubble was influenced by the initial growing velocity of a vapor bubble immediately following the detached bubble. The disappearance velocity of the detached bubble defined as a ratio of the bubble diameter at the departure to the time required until the disappearance, is in the range 0.2 to 2.0 m/sec. The disappearance velocity is proportional to the initial growing velocity of the bubble, to the square of the heat flux of the heated surface and to the cube of the wall superheat, separately. (author)

  18. 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

  19. Radiative and three-body recombination in the Alcator C-Mod divertor

    International Nuclear Information System (INIS)

    Lumma, D.; Terry, J.L.; Lipschultz, B.

    1997-01-01

    Significant recombination of the majority ion species has been observed in the divertor region of Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] under detached conditions. This determination is made by analysis of the visible spectrum from the divertor, in particular the Balmer series line emission and the observed recombination continuum, including an apparent recombination edge at ∼375 nm. The analysis shows that the electron temperature in the recombining plasma is 0.8 endash 1.5 eV. The measured volume recombination rate is comparable to the rate of ion collection at the divertor plates. The dominant recombination mechanism is three-body recombination into excited states (e+e+D + Right-arrow D 0 +e), although radiative recombination (e+D + Right-arrow D 0 +hν) contributes ∼5% to the total rate. Analysis of the Balmer series line intensities (from n upper =3 through 10) shows that the upper levels of these transitions are populated primarily by recombination. Thus the brightnesses of the Balmer series (and Lyman series) are directly related to the recombination rate. copyright 1997 American Institute of Physics

  20. 3D modelling of the island divertor for W7-AS

    International Nuclear Information System (INIS)

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

    1996-01-01

    Island divertors in low-shear stellarators exhibit the same basic topology (X-point diversion of field lines towards target plates) as tokamak divertors. However, the geometry is different. For island divertors, the small distance between the target and the LCFS (∼5cm for W7-AS and 8cm for W7-X) requires higher plasma densities than in comparable tokamaks to effectively decouple the target plasma and the neutrals from the core. These are basic prerequisites to realize high recycling and detachment conditions necessary for exaust. On the other hand, the island SOL can be used to confine recycling particles outside the LCFS, which may result in a density rise inside the islands, and hence in an improved screening of the neutrals. Nonlinear 3D effects are introduced in the transport equations by the non-axisymmetry of the configuration and by the segmentation of the target plates. The resulting toroidal inhomogeneities (variable connection lengths, toroidally localized recycling, poor parallel equilibration at low T) can hardly be approximated by an averaging 2D model. (orig.)

  1. Design and analysis of the W7-X divertor scraper element

    International Nuclear Information System (INIS)

    Lumsdaine, A.; Tipton, J.; Lore, J.; McGinnis, D.; Canik, J.; Harris, J.; Peacock, A.; Boscary, J.; Tretter, J.; Andreeva, T.

    2013-01-01

    Highlights: • A high heat flux actively cooled divertor component is thermally modeled with CFD. • CFC monoblocks are analyzed to verify peak steady-state temperatures do not exceed 1200 °C. • A field line diffusion code is developed to determine the heat flux on the divertor components. • Iteration is required to develop a surface that meets the criteria and fits into the limited space. -- Abstract: Thehigh heat-flux divertor of the Wendelstein 7-X large stellarator experiment consists of 10 divertor units which are designed to carry a steady-state heat flux of 10 MW/m 2 . However, the edge elements of this divertor are limited to only 5 MW/m 2 , and may be overloaded in certain plasma scenarios. It is proposed to reduce this heat by placing an additional “scraper element” in each of the ten divertor locations. It will be constructed using carbon fiber composite (CFC) monoblock technology. The design of the monoblocks and the path of the cooling tubes must be optimized in order to survive the significant steady-state heat loads, provide adequate coverage for the existing divertor, be located within sub-millimeter accuracy, and take into account the boundaries to other in vessel components, all at a minimum cost. Computational fluid dynamics modeling has been performed to examine the thermal transfer through the monoblock swirl tube channels for the design of the monoblock orientation. An iterative physics modeling and computer aided design process is being performed to optimize the placement of the scraper element within the severe spatial restrictions

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

    International Nuclear Information System (INIS)

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

    2016-01-01

    In the divertor structure of ITER and EAST with mono-block module, tungsten plays not only a role of armor material but also a role of structural material, because electromagnetic (EM) impact will be exerted on tungsten components in VDEs or CQ. The EM loads can reach to 100 MN, which would cause high strain rates. In addition, directly exposed to high-temperature plasma, the temperature regime of divertor components is complex. Aiming at studying dynamical response of tungsten divertors under EM loads, an experiment on tungsten employed in EAST divertors was performed using a Kolsky bar system. The testing strain rates and temperatures is derived from actual working conditions, which makes the constitutive equation concluded by using John-Cook model and testing data very accurate and practical. The work would give a guidance to estimate the dynamical response, fatigue life and damage evolution of tungsten divertor components under EM impact loads. - Graphical abstract: From the comparison between the experimental curves and the predicted curves calculated by adopting the corrected m, it is very clear that the new model is of great capability to explain the deformation behavior of the tungsten material under dynamic compression at high temperatures. (EC, PC and PCM refers to experimental curve, predicted curve and predicted curve with a corrected m. Different colors represent different scenarios.). - Highlights: • Test research on dynamic properties of tungsten at working temperature range and strain rate range of divertors. • Constitutive equation descrbing strain hardening, strain rate hardening and temperature softening. • A guidance to estimate dynamical response and damage evolution of tungsten divertor components under impact.

  3. Recent DIII-D high power heating and current drive experiments

    International Nuclear Information System (INIS)

    Simonen, T.C.; Jackson, G.L.; Lazarus, E.A.; Mahdavi, M.A.; Petrie, T.W.; Politzer, P.A.; Taylor, T.S.

    1995-01-01

    This paper describes recent DIII-D high power heating and current drive experiments. Described are experiments with improved wall conditioning, divertor particle pumping, radiative divertor experiments, studies of plasma shape and high poloidal β. ((orig.))

  4. Recent DIII-D high power heating and current drive experiments

    International Nuclear Information System (INIS)

    Simonen, T.C.; Jackson, G.L.; Mahdavi, M.A.; Petrie, T.W.; Politzer, P.A.; Taylor, T.S.; Lazarus, E.A.

    1994-02-01

    This paper describes recent DIII-D high power heating and current drive experiments. Describes are experiments with improved wall conditioning, divertor particle pumping, radiative divertor experiments, studies of plasma shape and high poloidal beta

  5. Recent DIII-D high power heating and current drive experiments

    Energy Technology Data Exchange (ETDEWEB)

    Simonen, T.C. [General Atomics, San Diego, CA (United States); Jackson, G.L. [General Atomics, San Diego, CA (United States); Lazarus, E.A. [Oak Ridge National Lab., TN (United States); Mahdavi, M.A. [General Atomics, San Diego, CA (United States); Petrie, T.W. [General Atomics, San Diego, CA (United States); Politzer, P.A. [General Atomics, San Diego, CA (United States); Taylor, T.S. [General Atomics, San Diego, CA (United States); DIII-D Team

    1995-01-01

    This paper describes recent DIII-D high power heating and current drive experiments. Described are experiments with improved wall conditioning, divertor particle pumping, radiative divertor experiments, studies of plasma shape and high poloidal {beta}. ((orig.)).

  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. 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.

  8. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-01-01

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

  10. The edge plasma and divertor in TIBER

    Energy Technology Data Exchange (ETDEWEB)

    Barr, W.L.

    1987-10-16

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

  11. The edge plasma and divertor in TIBER

    International Nuclear Information System (INIS)

    Barr, W.L.

    1987-01-01

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

  12. First results from the dynamic ergodic divertor at TEXTOR

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  13. Stability and propagation of the high field side high density front in the fluctuating state of detachment in ASDEX Upgrade

    Directory of Open Access Journals (Sweden)

    P. Manz

    2017-08-01

    Full Text Available During detachment a structure of strongly enhanced density develops close to the inner target. Its dynamics is approximated by those of radiative fluctuations appearing at a similar position and studied by means of a time-delay-estimation technique in the ASDEX Upgrade tokamak. Compared to theory the dynamics can be described as follows: at increasing density the ionization front moves upstream to reduce ionization radiation in order to balance the increased recombination radiation. The recombination zone stays close to the target strike point. The parallel motion of the ionization front is determined by the perpendicular neutral motion. The divertor nose constitutes an obstacle for the perpendicular neutral flux from the target to the region above the X-point. Passing into this shadow the neutral flux above the X-point is strongly reduced, the ionization front fades away and the heat flux from upstream can increase the temperature in the recombination region, subsequently reducing recombination and reforming an ionization front below the X-point. A cyclic reformation of the ionization front propagating from below to above the X-point occurs leading to a fluctuation as observed in the experiment.

  14. 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.

  15. 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

  16. EU R and D on divertor components

    International Nuclear Information System (INIS)

    Merola, M.; Daenner, W.; Pick, M.

    2005-01-01

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

  17. 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.

  18. Retinal detachment in paediatric patients

    International Nuclear Information System (INIS)

    Zafar, S. N.; Qureshi, N.; Azad, N.; Khan, A.

    2013-01-01

    Objective: To assess the causes of retinal detachment in children and the various operative procedures requiring vitreoretinal surgical intervention for the same. Study Design: Case series. Place and Duration of Study: Department of Ophthalmology, Al-Shifa Trust Eye Hospital, Rawalpindi, from January 2006 to May 2009. Methodology: A total of 281 eyes of 258 patients, (aged 0 - 18 years) who underwent vitreo-retinal surgical intervention for retinal detachment were included. Surgical log was searched for the type of retinal detachment and its causes. Frequencies of various interventions done in these patients viz. vitrectomy, scleral buckle, use of tamponading agents, laser photocoagulation and cryotherapy were noted. Results were described as descriptive statistics. Results: Myopia was the cause in 62 (22.1%) and trauma in 51 (18.1%) of the eyes. Total retinal detachment (RD) was treated in 94 (33.5%) eyes, sub total RD in 36 (12.8%), recurrent RD in 32 (11.4%), giant retinal tear in 28 (10%), tractional RD in 15 (5.3%) and exudative RD in 2 (0.7%). Prophylactic laser or cryotherapy was applied in 74 (26.3%) of the eyes. Pars plana vitrectomy (PPV) was carried out in 159 (56.6%) eyes while scleral buckle procedure was done in 129 (45.9%) eyes. Silicon oil was used in 149 (53%), perfluorocarbon liquid in 32 (11.4%) and gas tamponade in 20 (7.1%) eyes. Conclusion: The most common cause of retinal detachment in paediatric patients was myopia, followed by trauma. Total RD was more common as compared to the other types. The most common procedure adopted was pars plana vitrectomy followed by scleral buckle procedure. (author)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  1. The fellow eye of patients with phakic rhegmatogenous retinal detachment from atrophic holes of lattice degeneration without posterior vitreous detachment.

    Science.gov (United States)

    Gonzales, C R; Gupta, A; Schwartz, S D; Kreiger, A E

    2004-11-01

    Primary phakic rhegmatogenous retinal detachment (RRD) without posterior vitreous detachment (PVD) represents a unique clinical entity that behaves differently from RRD associated with PVD. While previous studies have reported the long term findings in the fellow eye of patients with RRD and PVD, the outcome of the fellow eye of patients with RRD without PVD is not known. Consecutive patients with RRD not associated with PVD were studied retrospectively. The authors evaluated the fellow eye for retinal detachment or other vision threatening pathology. 27 patients (mean age 32 years) were studied with follow up of between 9 and 326 months (mean 111 months). 24 (89%) were myopic. Bilateral retinal detachment occurred in eight patients (30%). On initial examination, 17 patients (63%) had retinal findings (including lattice degeneration, atrophic holes, and/or cystic retinal tufts) in the fellow eye that might predispose them to retinal detachment. 14 vision threatening events or diagnoses occurred (nine of which were rhegmatogenous in nature) in the fellow eye including eight retinal detachments, one traumatic PVD without retinal tears, one retinal tear after PVD, one diagnosis of pigmentary glaucoma needing trabeculectomy, two visually significant cataracts, and one diagnosis of chorioretinitis. 23 patients (85%) maintained visual acuity better than 20/50, with most retaining 20/20 vision in the fellow eye. Patients who experience RRD without PVD are at risk of developing vision threatening events in the contralateral eye and, as such, the fellow eye should be followed carefully.

  2. Analysis of particle transport in a gas target divertor

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

  3. 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

  4. 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.)

  5. 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)

  6. Modelling of radial electric fields and currents during divertor plate biasing on TdeV

    International Nuclear Information System (INIS)

    Lachambre, J.L.; Quirion, B.; Boucher, C.

    1994-01-01

    A simple model based on non-ambipolar radial transport and planar sheath physics is used to describe the generation of radial electric fields and currents in the scrape-off layer of the Tokamak de Varennes (TdeV) during divertor plate biasing. In general, the calculated predictions compare favourably with TdeV results over a variety of plasma conditions and divertor magnetic configurations. Validated by the experiment, the model is used to study the scaling laws of perpendicular ion mobility and to test existing related theories. Finally, the model is proposed as a useful tool for the design and upgrade of biased divertors through optimization of the plate and throat geometry. (author). 35 refs, 16 figs, 1 tab

  7. Thermographic observation of the divertor target plates in the stellarators W7-AS and W7-X

    International Nuclear Information System (INIS)

    Hildebrandt, D.; Gadelmeier, F.; Grigull, P.; McCormick, K.; Naujoks, D.; Suender, D.

    2003-01-01

    Thermography is applied on the stellarator W7-AS to monitor the thermal load of the recently installed divertor targets. A three dimensional numerical code was developed to evaluate power fluxes arriving at the targets from the measured temporal evolution of the surface temperature distribution. Values of the thermal conductivity of the used CFC-target material for all three directions are required for this evaluation and determined by observing the propagation of controlled heat pulses applied by an infrared laser. The evaluation of the thermographic measurements during plasma operation shows characteristic spatial and temporal features of the arrived heat fluxes. Significant features in high density regimes like plasma detachment from the divertor target plates or strongly enhanced localised plasma radiation (MARFE) has been observed by the installed infrared cameras. The implications of these observations for the thermographic system for W7-X are shortly addressed

  8. Molecular deuterium behaviour in tungsten divertor on JET

    Energy Technology Data Exchange (ETDEWEB)

    Sergienko, G., E-mail: g.sergienko@fz-juelich.de [Institute of Energy and Climate Research –Plasma Physics, Forschungszentrum Jülich, EURATOM Association, Trilateral Euregio Cluster, D-52425 Jülich (Germany); Arnoux, G. [Euratom-CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Brezinsek, S.; Clever, M.; Huber, A.; Kruezi, U. [Institute of Energy and Climate Research –Plasma Physics, Forschungszentrum Jülich, EURATOM Association, Trilateral Euregio Cluster, D-52425 Jülich (Germany); Meigs, A.G. [Euratom-CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Mertens, Ph.; Samm, U. [Institute of Energy and Climate Research –Plasma Physics, Forschungszentrum Jülich, EURATOM Association, Trilateral Euregio Cluster, D-52425 Jülich (Germany); Stamp, M. [Euratom-CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)

    2013-07-15

    Molecular spectroscopy was used to observe molecular deuterium at the outer strike point of the new bulk tungsten JET divertor. The rotational and vibrational populations of the deuterium molecules in the ground state were determined from the deuterium Q-branches of Fulcher-α band emission (d{sup 3}Π{sub u}{sup -}→a{sup 3}Σ{sub g}{sup +}) in the 600–640 nm spectral range. For L-mode plasmas in the low recycling regime the molecular emission maximum is located in the vicinity of the strike point. The spatial profile of the emission was strongly modified during plasma detachment in both L- and H-mode plasmas. The rotational temperature of excited molecules reached 2760 K in L-mode. The vibrational population has a peculiarity: a remarkably high population of the d{sup 3}Π{sub u}{sup -}(v = 0) vibrational level indicating a non-Boltzmann vibrational distribution of D{sub 2} in tungsten environment.

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

    International Nuclear Information System (INIS)

    Heijnen, S.H.; Pol, M.J. van de.

    1994-09-01

    In this report, the feasibility of a pulsed radar system for measuring the electron density profile in the divertor region of JET is studied. Some dedicated experiments are performed with a four-channel system, which was designed for the Rijnhuizen Tokamak Project. To simulate divertor plasmas the measurements are performed in ECRH induced plasmas without current. The parameters of these kinds of plasmas are: n e 19 m -3 , T e <100 eV, and a diameter of ∼30 cm. (HSI)

  10. Liquid metal cooled divertor for ARIES

    International Nuclear Information System (INIS)

    Muraviev, E.

    1995-01-01

    A liquid metal, Ga-cooled divertor design was completed for the double null ARIES-II divertor design. The design analysis indicated a surface heat flux removal capability of up to 15 MW/m 2 , and its relative easy maintenance. Design issues of configuration, thermal hydraulics, thermal stresses, liquid metal loop and safety effects were evaluated. For coolant flow control, it was found that it is necessary to use some part of the blanket cooling ducts for the draining of liquid metal from the top divertor. In order to minimize the inventory of Ga, it was recommended that the liquid metal loop equipment should be located as close to the torus as possible. More detailed analysis of transient conditions especially under accident conditions was identified as an issue that will need to be addressed

  11. Constrained ripple optimization of Tokamak bundle divertors

    International Nuclear Information System (INIS)

    Hively, L.M.; Rome, J.A.; Lynch, V.E.; Lyon, J.F.; Fowler, R.H.; Peng, Y-K.M.; Dory, R.A.

    1983-02-01

    Magnetic field ripple from a tokamak bundle divertor is localized to a small toroidal sector and must be treated differently from the usual (distributed) toroidal field (TF) coil ripple. Generally, in a tokamak with an unoptimized divertor design, all of the banana-trapped fast ions are quickly lost due to banana drift diffusion or to trapping between the 1/R variation in absolute value vector B ω B and local field maxima due to the divertor. A computer code has been written to optimize automatically on-axis ripple subject to these constraints, while varying up to nine design parameters. Optimum configurations have low on-axis ripple ( 0 ) are lost. However, because finite-sized TF coils have not been used in this study, the flux bundle is not expanded

  12. Design integration of liquid surface divertors

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  13. 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

  14. Control of divertor configuration in JT-60

    International Nuclear Information System (INIS)

    Yoshino, R.; Kukuchi, M.; Ninomiya, H.; Yoshida, H.; Tsuji, S.; Hosogane, N.; Seki, S.

    1985-01-01

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

  15. 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.))

  16. 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.

  17. Overview of the U.S. programme on particle control and divertors

    International Nuclear Information System (INIS)

    Oktay, E.

    1981-01-01

    In this overview, the particle control and divertor issues are reviewed briefly, the status of the U.S. experimental programs relevant to this topic are mentioned, and the future direction of the program is discussed. The results of experiments will be presented by the participants from the U.S. laboratories. (orig.)

  18. Negative ion detachment cross sections

    International Nuclear Information System (INIS)

    Champion, R.L.; Doverspike, L.D.

    1992-10-01

    The authors have measured absolute cross sections for electron detachment and charge exchange for collision of O and S with atomic hydrogen, have investigated the sputtering and photodesorption of negative ions from gas covered surfaces, and have begun an investigation of photon-induced field emission of electrons from exotic structures. Brief descriptions of these activities as well as future plans for these projects are given below

  19. Variations of posterior vitreous detachment

    OpenAIRE

    Kakehashi, A.; Kado, M.; Akiba, J.; Hirokawa, H.

    1997-01-01

    AIMS—To identify variations in posterior vitreous detachment (PVD) and establish a clinical classification system for PVD.
METHODS—400 consecutive eyes were examined using biomicroscopy and vitreous photography and classified the PVD variations—complete PVD with collapse, complete PVD without collapse, partial PVD with thickened posterior vitreous cortex (TPVC), or partial PVD without TPVC.
RESULTS—In each PVD type, the most frequently seen ocular pathologies were as follows: in complete PVD ...

  20. Classification of posterior vitreous detachment

    OpenAIRE

    Kakehashi, Akihiro; Takezawa, Mikiko; Akiba, Jun

    2013-01-01

    Akihiro Kakehashi,1 Mikiko Takezawa,1 Jun Akiba21Department of Ophthalmology, Jichi Medical University, Saitama Medical Center, Saitama, 2Kanjodori Eye Clinic, Asahikawa, JapanAbstract: Diagnosing a posterior vitreous detachment (PVD) is important for predicting the prognosis and determining the indication for vitreoretinal surgery in many vitreoretinal diseases. This article presents both classifications of a PVD by slit-lamp biomicroscopy and of a shallow PVD by optical coherence tomography...

  1. 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.

  2. 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.

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

    International Nuclear Information System (INIS)

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

    1997-11-01

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

  4. Implications of steady-state operation on divertor design

    International Nuclear Information System (INIS)

    Sevier, D.L.; Reis, E.E.; Baxi, C.B.; Silke, G.W.; Wong, C.P.C.; Hill, D.N.

    1996-01-01

    As fusion experiments progress towards long pulse or steady state operation, plasma facing components are undergoing a significant change in their design. This change represents the transition from inertially cooled pulsed systems to steady state designs of significant power handling capacity. A limited number of Plasma Facing Component (PFC) systems are in operation or planning to address this steady state challenge at low heat flux. However in most divertor designs components are required to operate at heat fluxes at 5 MW/m 2 or above. The need for data in this area has resulted in a significant amount of thermal/hydraulic and thermal fatigue testing being done on prototypical elements. Short pulse design solutions are not adequate for longer pulse experiments and the areas of thermal design, structural design, material selection, maintainability, and lifetime prediction are undergoing significant changes. A prudent engineering approach will guide us through the transitional phase of divertor design to steady-state power plant components. This paper reviews the design implications in this transition to steady state machines and the status of the community efforts to meet evolving design requirements. 54 refs., 5 figs., 2 tabs

  5. 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.)

  6. ITER divertor, design issues and research and development

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-11-01

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

  7. An X-point ergodic divertor

    International Nuclear Information System (INIS)

    Chu, M.S.; Jensen, T.H.; La Haye, R.J.; Taylor, T.S.; Evans, T.E.

    1991-10-01

    A new ergodic divertor is proposed. It utilizes a system of external (n = 3) coils arranged to generate overlapping magnetic islands in the edge region of a diverted tokamak and connect the randomized field lines to the external (cold) divertor plate. The novel feature in the configuration is the placement of the external coils close to the X-point. A realistic design of the external coil set is studied by using the field line tracing method for a low aspect ratio (A ≅ 3) tokamak. Two types of effects are observed. First, by placing the coils close to the X-point, where the poloidal magnetic field is weak and the rational surfaces are closely packed only a moderate amount of current in the external coils is needed to ergodize the edge region. This ergodized edge enhances the edge transport in the X-point region and leads to the potential of edge profile control and the avoidance of edge localized modes (ELMs). Furthermore, the trajectories of the field lines close to the X-point are modified by the external coil set, causing the hit points on the external divertor plates to be randomized and spread out in the major radius direction. A time-dependent modulation of the currents in the external (n = 3) coils can potentially spread the heat flux more uniformly on the divertor plate avoiding high concentration of the heat flux. 10 refs., 9 figs

  8. 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

  9. Compact poloidal divertor reference design for TNS

    International Nuclear Information System (INIS)

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

    1977-01-01

    A compact poloidal divertor concept has been developed for TNS tokamaks and its feasibility has been demonstrated by sufficient detailed magnetic, thermal, mechanical and vacuum analyses. This particular divertor is formed by a pair of opposing coil sets which define a magnetic flux slot where the particle burial chamber is located. The magnetic flux in the space between the coil sets is compressed vertically to limit the height and to expand the horizontal width of the particle and energy burial chamber. The intensity of the poloidal field is increased to make the pitch angle of the flux lines very large so that the diverted particles can be intercepted by a large number of panels oriented at a small angle with respect to the flux lines. Large collecting surface areas can be obtained so that the thermal load and particle flux are reduced to a practical level. Flowing lithium film and solid metal panels have been considered as the particle collector and the latter is preferred. This divertor allows for most economical use of the available space inside the TF coils and thus has minor impact on the overall size of the tokamak. The divertor design is essentially independent of the tokamak system, although analyses were performed based on TNS

  10. Posterior vitreous detachment and retinal detachment after cataract surgery.

    Science.gov (United States)

    Ripandelli, Guido; Coppé, Andrea Maria; Parisi, Vincenzo; Olzi, Diego; Scassa, Cecilia; Chiaravalloti, Adele; Stirpe, Mario

    2007-04-01

    To evaluate possible changes of vitreous status in emmetropic eyes after uneventful phacoemulsification surgery, and possible related complications such as the onset of retinal detachment (RD). Retrospective case series. Four hundred fifty-three emmetropic eyes from 453 patients (mean age, 62.03+/-5.57 years) subjected to uneventful phacoemulsification with intraocular lens implantation in the capsular bag were considered in the study. They had a refractive error within +/-0.5 diopters (mean, -0.21+/-0.08). Eyes with peripheral retinal lattice degeneration were included only if asymptomatic and only if the degeneration involved one retinal quadrant. After cataract surgery, the 453 eyes were evaluated preoperatively at days 1, 15, and 30 and months 3, 6, 12, 18, 24, 36, 48, and 60. The whole period of follow-up was 5 years. Evaluation of vitreous status by biomicroscopic examination, indirect binocular ophthalmoscopy, and B-scan ultrasonography. Postoperative onset of posterior vitreous detachment (PVD) and RD. After cataract surgery, a PVD occurred in 107 of 141 (75.88%) eyes without preoperative PVD or lattice degeneration. Posterior vitreous detachment occurred in 41 of 47 eyes (87.23%) with preoperative lattice degeneration and no PVD. Eyes with preoperative lattice degeneration and postoperative PVD showed a higher incidence of RD after cataract surgery (21.27%) than eyes without preoperative PVD or lattice degeneration (0.70%). In all patients with lattice degeneration, RD originated from horseshoe retinal tears on lattice areas located on the superior quadrants. No correlation was observed between the development of RD and age. Our results suggest that the onset of postoperative PVD should be considered an important risk factor for the development of RD after cataract surgery, particularly in eyes with lattice areas.

  11. Spectroscopic measurement of target plate erosion in the ASDEX Upgrade divertor

    Energy Technology Data Exchange (ETDEWEB)

    Filed, A R; Garcia-Rosales, C; Lieder, G; Pitcher, C S; Radtke, R [Association Euratom-Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); ASDEX Upgrade Team

    1996-02-01

    The erosion of the graphite divertor plates in the ASDEX Upgrade tokamak is measured spectroscopically. Spatial profiles of the D{sup 0} and C{sup +} influxes across the outer target plate are determined from measured absolute line intensities. Plasma parameters (n{sub e}, T{sub e}) at the target, which are required to determine the appropriate photon emission efficiencies for these lines, are obtained from an in-vessel reciprocating Langmuir probe above the target plate. Yields for the erosion of the graphite by the incident D{sup +} flux are determined from the ratio of the measured C{sup +} to D{sup 0} fluxes. Over a range of moderate densities the measured yields of {<=} 4% are explicable in terms of physical sputtering alone. Chemical sputtering by low energy Franck-Condon neutrals probably contributes, however, to the total erosion. At higher densities detachment of the plasma from the targets occurs owing to formation of a MARFE near the X point. Under these conditions localized physical sputtering of the targets ceases. The impurity level (Z{sub eff}) is, however, maintained following detachment, indicating a corresponding maintenance of carbon influx, perhaps due to chemical erosion of the total graphite surface and/or an improvement in particle confinement in the detached state. (author). 26 refs, 14 figs, 1 tab.

  12. Peripheral retinal degenerations and the risk of retinal detachment.

    Science.gov (United States)

    Lewis, Hilel

    2003-07-01

    To review the degenerative diseases of the peripheral retina in relationship with the risk to develop a rhegmatogenous retinal detachment and to present recommendations for use in eyes at increased risk of developing a retinal detachment. Focused literature review and author's clinical experience. Retinal degenerations are common lesions involving the peripheral retina, and most of them are clinically insignificant. Lattice degeneration, degenerative retinoschisis, cystic retinal tufts, and, rarely, zonular traction tufts, can result in a rhegmatogenous retinal detachment. Therefore, these lesions have been considered for prophylactic therapy; however, adequate studies have not been performed to date. Well-designed, prospective, randomized clinical studies are necessary to determine the benefit-risk ratio of prophylactic treatment. In the meantime, the evidence available suggests that most of the peripheral retinal degenerations should not be treated except in rare, high-risk situations.

  13. 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

  14. Classification of posterior vitreous detachment

    Science.gov (United States)

    Kakehashi, Akihiro; Takezawa, Mikiko; Akiba, Jun

    2014-01-01

    Diagnosing a posterior vitreous detachment (PVD) is important for predicting the prognosis and determining the indication for vitreoretinal surgery in many vitreoretinal diseases. This article presents both classifications of a PVD by slit-lamp biomicroscopy and of a shallow PVD by optical coherence tomography (OCT). By biomicroscopy, the vitreous condition is determined based on the presence or absence of a PVD. The PVD then is classified as either a complete posterior vitreous detachment (C-PVD) or a partial posterior vitreous detachment (P-PVD). A C-PVD is further divided into a C-PVD with collapse and a C-PVD without collapse, while a P-PVD is divided into a P-PVD with shrinkage of the posterior hyaloid membrane (P-PVD with shrinkage) and a P-PVD without shrinkage of the posterior hyaloid membrane (P-PVD without shrinkage). A P-PVD without shrinkage has a subtype characterized by vitreous gel attachment through the premacular hole in a posterior hyaloid membrane to the macula (P-PVD without shrinkage [M]). By OCT, a shallow PVD is classified as the absence of a shallow PVD or as a shallow PVD. A shallow PVD is then subclassified as a shallow PVD without shrinkage of the posterior vitreous cortex, a shallow PVD with shrinkage of the posterior vitreous cortex, and a peripheral shallow PVD. A shallow PVD without shrinkage of the posterior vitreous cortex has two subtypes: an age-related shallow PVD and a perifoveal PVD associated with a macular hole. PMID:24376338

  15. Outcomes in bullous retinal detachment

    Directory of Open Access Journals (Sweden)

    Sarah P. Read

    2017-06-01

    Conclusions and importance: GRTs are an uncommon cause of retinal detachment. While pars plana vitrectomy with tamponade is standard in GRT management, there is variability in the use of scleral buckling and PFO in these cases. This is in contrast to retinal dialysis where scleral buckle alone can yield favorable results. Though a baseball ocular trauma is common, retinal involvement is rare compared to other sports injuries such as those occurring with tennis, soccer and golf. Sports trauma remains an important cause of retinal injury and patients should be counseled on the need for eye protection.

  16. Cellular volume regulation and substrate stiffness modulate the detachment dynamics of adherent cells

    Science.gov (United States)

    Yang, Yuehua; Jiang, Hongyuan

    2018-03-01

    Quantitative characterizations of cell detachment are vital for understanding the fundamental mechanisms of cell adhesion. Experiments have found that cell detachment shows strong rate dependence, which is mostly attributed to the binding-unbinding kinetics of receptor-ligand bond. However, our recent study showed that the cellular volume regulation can significantly regulate the dynamics of adherent cell and cell detachment. How this cellular volume regulation contributes to the rate dependence of cell detachment remains elusive. Here, we systematically study the role of cellular volume regulation in the rate dependence of cell detachment by investigating the cell detachments of nonspecific adhesion and specific adhesion. We find that the cellular volume regulation and the bond kinetics dominate the rate dependence of cell detachment at different time scales. We further test the validity of the traditional Johnson-Kendall-Roberts (JKR) contact model and the detachment model developed by Wyart and Gennes et al (W-G model). When the cell volume is changeable, the JKR model is not appropriate for both the detachments of convex cells and concave cells. The W-G model is valid for the detachment of convex cells but is no longer applicable for the detachment of concave cells. Finally, we show that the rupture force of adherent cells is also highly sensitive to substrate stiffness, since an increase in substrate stiffness will lead to more associated bonds. These findings can provide insight into the critical role of cell volume in cell detachment and might have profound implications for other adhesion-related physiological processes.

  17. ALPS - advanced limiter-divertor plasma-facing systems

    International Nuclear Information System (INIS)

    Allain, J. P.; Bastasz, R.; Brooks, J. N.; Evans, T.; Hassanein, A.; Luckhardt, S.; Maingi, R.; Mattas, R. F.; McCarthy, K.; Mioduszewski, P.; Mogahed, E.; Moir, R.; Molokov, S.; Morely, N.; Nygren, R.; Reed, C.; Rognlien, T.; Ruzic, D.; Sviatoslavsky, I.; Sze, D.; Tillack, M.; Ulrickson, M.; Wade, P. M.; Wong, C.; Wooley, R.

    1999-01-01

    The Advanced Limiter-divertor Plasma-facing Systems (ALPS) program was initiated in order to evaluate the potential for improved performance and lifetime for plasma-facing systems. The main goal of the program is to demonstrate the advantages of advanced limiter/divertor systems over conventional systems in terms of power density capability, component lifetime, and power conversion efficiency, while providing for safe operation and minimizing impurity concerns for the plasma. Most of the work to date has been applied to free surface liquids. A multi-disciplinary team from several institutions has been organized to address the key issues associated with these systems. The main performance goals for advanced limiters and diverters are a peak heat flux of >50 MW/m 2 ,elimination of a lifetime limit for erosion, and the ability to extract useful heat at high power conversion efficiency (approximately40%). The evaluation of various options is being conducted through a combination of laboratory experiments, modeling of key processes, and conceptual design studies. The current emphasis for the work is on the effects of free surface liquids on plasma edge performance

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

    International Nuclear Information System (INIS)

    Fantz, U.

    2002-01-01

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

  19. Self-sustained oscillations in the divertor plasma

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  20. 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)

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

    International Nuclear Information System (INIS)

    FENSTERMACHER, ME; PORTER, GD; LEONARD, AW; BROOKS, NH; BOEDO, JA; COLCHIN, RJ; GRAY, DS; GROEBNER, RJ; GROTH, M; HOGAN, JT; HOLLMANN, EM; LASNIER, CJ; OSBORNE, TH; PETRIE, TW; RUDAKOV, DL; SNYDER, PB; TAKAHASHI, H; WATKINS, JG; ZENG, L; DIII-D TEAM

    2003-01-01

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

  3. Detached Bridgman Growth of Germanium and Germanium-Silicon Alloy Crystals

    Science.gov (United States)

    Szofran, F. R.; Volz, M. P.; Schweizer, M.; Cobb, S. D.; Motakef, S.; Croell, A.; Dold, P.; Curreri, Peter A. (Technical Monitor)

    2002-01-01

    Earth based experiments on the science of detached crystal growth are being conducted on germanium and germanium-silicon alloys (2 at% Si average composition) in preparation for a series of experiments aboard the International Space Station (ISS). The purpose of the microgravity experiments includes differentiating among proposed mechanisms contributing to detachment, and confirming or refining our understanding of the detachment mechanism. Because large contact angle are critical to detachment, sessile drop measurements were used to determine the contact angles as a function of temperature and composition for a large number of substrates made of potential ampoule materials. Growth experiments have used pyrolytic boron nitride (pBN) and fused silica ampoules with the majority of the detached results occurring predictably in the pBN. The contact angles were 173 deg (Ge) and 165 deg (GeSi) for pBN. For fused silica, the contact angle decreases from 150 deg to an equilibrium value of 117 deg (Ge) or from 129 deg to an equilibrium value of 100 deg (GeSi) over the duration of the experiment. The nature and extent of detachment is determined by using profilometry in conjunction with optical and electron microscopy. The stability of detachment has been analyzed, and an empirical model for the conditions necessary to achieve sufficient stability to maintain detached growth for extended periods has been developed. Results in this presentation will show that we have established the effects on detachment of ampoule material, pressure difference above and below the melt, and silicon concentration; samples that are nearly completely detached can be grown repeatedly in pBN.

  4. Risk factor profile in retinal detachment

    Directory of Open Access Journals (Sweden)

    Azad Raj

    1988-01-01

    Full Text Available 150 cases of retinal detachment comprising 50 patients each of bilateral retinal detachment, unilateral retinal detachment without any retinal lesions in the fellow eve and unilateral retinal detachment with retinal lesions in the fellow eye were studied and the various associated risk factors were statistically analysed. The findings are discussed in relation to their aetiological and prognostic significance in the different types of retinal detachment. Based on these observations certain guidelines are offered which may be of value in decision making, in prophylactic detachment surgery. Tractional breaks in the superior temporal quadrant especially when symptomatic. mandate prophylactic treatment. Urgency is enhanced it′ the patient is aphakic. Associated myopia adds to the urgency. The higher incidence of initial right e′ e involvement in all groups suggests a vascular original possibly ischaemic.

  5. DETACHMENT OF BACTERIOPHAGE FROM ITS CARRIER PARTICLES.

    Science.gov (United States)

    Hetler, D M; Bronfenbrenner, J

    1931-05-20

    The active substance (phage) present in the lytic broth filtrate is distributed through the medium in the form of particles. These particles vary in size within broad limits. The average size of these particles as calculated on the basis of the rate of diffusion approximates 4.4 mmicro in radius. Fractionation by means of ultrafiltration permits partial separation of particles of different sizes. Under conditions of experiments here reported the particles varied in the radius size from 0.6 mmicro to 11.4 mmicro. The active agent apparently is not intimately identified with these particles. It is merely carried by them by adsorption, and under suitable experimental conditions it can be detached from the larger particles and redistributed on smaller particles of the medium.

  6. 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.

  7. Characteristics of the Secondary Divertor on DIII-D

    Science.gov (United States)

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

    2009-11-01

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

  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. Rovibrationally Resolved Time-Dependent Collisional-Radiative Model of Molecular Hydrogen and Its Application to a Fusion Detached Plasma

    Directory of Open Access Journals (Sweden)

    Keiji Sawada

    2016-12-01

    Full Text Available A novel rovibrationally resolved collisional-radiative model of molecular hydrogen that includes 4,133 rovibrational levels for electronic states whose united atom principal quantum number is below six is developed. The rovibrational X 1 Σ g + population distribution in a SlimCS fusion demo detached divertor plasma is investigated by solving the model time dependently with an initial 300 K Boltzmann distribution. The effective reaction rate coefficients of molecular assisted recombination and of other processes in which atomic hydrogen is produced are calculated using the obtained time-dependent population distribution.

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  12. Effect of separatrix magnetic geometry on divertor behavior in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Petrie, T.W., E-mail: petrie@fusion.gat.com [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Canik, J.M. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Lasnier, C.J. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Leonard, A.W.; Mahdavi, M.A. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Watkins, J.G. [Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); Fenstermacher, M.E. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Ferron, J.R.; Groebner, R.J. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Hill, D.N. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Hyatt, A.W. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Holcomb, C.T. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Luce, T.C. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Moyer, R.A. [University of California–San Diego, La Jolla, CA 92093-0417 (United States); Stangeby, P.C. [University of Toronto Institute of Aerospace Studies, Toronto (Canada)

    2013-07-15

    We report on recent experiments on DIII-D that examined the effects that variations in the parallel connection length in the scrape-off layer (SOL), L{sub ||}, and the radial location of the outer divertor target, R{sub TAR}, have on divertor plasma properties. Two-point modeling of the SOL plasma predicts that larger values of L{sub ||} and R{sub TAR} should lower temperature and raise density at the outer divertor target for fixed upstream separatrix density and temperature, i.e., n{sub TAR} ∝ [R{sub TAR}]{sup 2}[L{sub ||}]{sup 6/7} and T{sub TAR} ∝ [R{sub TAR}]{sup −2}[L{sub ||}]{sup −4/7}. The dependence of n{sub TAR} and T{sub TAR} on L{sub ||} was consistent with our data, but the dependence of n{sub TAR} and T{sub TAR} on R{sub TAR} was not. The surprising result that the divertor plasma parameters did not depend on R{sub TAR} in the predicted way may be due to convected heat flux, driven by escaping neutrals, in the more open configuration of the larger R{sub TAR} cases. Modeling results using the SOLPS code support this postulate.

  13. 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

  14. 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)

  15. SLAC divertor channel entrance thermal stress analysis

    International Nuclear Information System (INIS)

    Johnson, G.L.; Stein, W.; Lu, S.C.; Riddle, R.A.

    1985-01-01

    X-ray beams emerging from the new SLAC electron-positron storage ring (PEP) impinge on the entrance to tangential divertor channels causing highly localized heating in the channel structure. Analyses were completed to determine the temperatures and thermally-induced stresses due to this heating. These parts are cooled with water flowing axially over them at 30 0 C. The current design and operating conditions should result in the entrance to the new divertor channel operating at a peak temperature of 123 0 C with a peak thermal stress at 91% of yield. Any micro-cracks that form due to thermally-induced stresses should not propagate to the coolant wall nor form a path for the coolant to leak into the storage ring vacuum. 34 figs., 4 tabs

  16. Thermal effects of divertor sweeping in ITER

    International Nuclear Information System (INIS)

    Wesley, J.C.

    1992-01-01

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

  17. 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)

  18. 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.

  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. Thermo-mechanical and damage analyses of EAST carbon divertor under type-I ELMy H-mode operation

    Energy Technology Data Exchange (ETDEWEB)

    Li, W.X. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Song, Y.T. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Ye, M.Y. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Peng, X.B., E-mail: pengxb@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Wu, S.T. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Qian, X.Y.; Zhu, C.C. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China)

    2016-04-15

    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.

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

    International Nuclear Information System (INIS)

    Ruchti, C.

    1981-01-01

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

  2. Influence of stray light for divertor spectroscopy in ITER

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  3. Technological development of the Monobloc Divertor Concept

    International Nuclear Information System (INIS)

    DiPietro, E.; Brossa, M.; Guerreschi, U.; Suresh, D.; Cardella, A.

    1992-01-01

    This paper reports on a technological program devoted to the assessment of the feasibility and the qualification of the Monobloc Divertor Concept for the divertor of the NET/ITER Machine which has been developed with the joint collaboration between ENEA, the NET Team, Ansaldo DNT and Metallwerk Plansee. The basic idea guiding the development of the monobloc divertor consists in obtaining a component suitable to sustain the operation thermal loads, attaining peak values in the range of 15 MW/2 in steady state conditions, by a proper arrangement of refractory tiles (acting as an armour) directly brazed to the cooling pipes. In the first phase the main activities have been devoted to find a reliable joint between the armour and the cooling pipes. A number of candidate armour materials have been investigated chosen among the most promising CFC currently available in combination with molybdenum alloys (T2M and Mo41Re) and dispersion strengthened copper. The most relevant results of the test activity including the comparison of different brazing alloys and techniques and the evaluation of suitable NDE techniques are reported

  4. Westinghouse compact poloidal divertor reference design

    International Nuclear Information System (INIS)

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

    1977-08-01

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

  5. 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.)

  6. Rhegmatogenous retinal detachment and uveitis.

    Science.gov (United States)

    Kerkhoff, Frank T; Lamberts, Querin J; van den Biesen, Pieter R; Rothova, Aniki

    2003-02-01

    To evaluate the frequency, high-risk factors, and visual prognosis of rhegmatogenous retinal detachment (RRD) in patients with uveitis. Retrospective case-control study. We included 1387 consecutive patients with uveitis who consulted our uveitis clinic from January 1990 through December 1997 of whom 43 patients (46 eyes) with RRD were identified. The retinal detachment (RD) controls were 212 consecutive patients with RRD (221 eyes, first occurrence of RD, not associated with uveitis) who were admitted for surgery in the period from April 1999 to April 2000. The uveitis control group consisted of 150 age-matched patients (210 eyes) selected from the entire uveitis series. Retrospective analysis of clinical data. The presence of RRD and eventual risk factors for RRD, such as myopia, retinal lattice degeneration, prior intraocular surgery, anatomic location of uveitis, its specific diagnosis, and clinical manifestations. Furthermore, the surgical and nonsurgical outcomes of RRD, as well as the results of various treatment regimens, were analyzed. RRD was identified in 3.1% of the patients with uveitis. RRD was most frequently associated with panuveitis (6.6%). RRD was associated more frequently with infectious (7.6%) than noninfectious uveitis (2.1%). At the onset of RRD, uveitis was active in most (46%) affected eyes. Proliferative vitreoretinopathy was present in 30% of the uveitic RRD eyes at presentation in contrast to 12% of the RRD control eyes. In uveitic RRD, the retina was reattached in 59% of eyes with a single operation; the final anatomic reattachment rate was 88%. Finally, a visual acuity of less than 20/200 was present in 71% of the uveitic RRD eyes, 10% of which had no light perception. We discovered a high prevalence of RRD in patients with active panuveitis and infectious uveitis and document that uveitis in itself is a risk factor for the development of RRD. The visual prognosis of RRD in uveitis was poor because of the uveitis itself and the

  7. 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

  8. Experimental results from detached plasmas in TFTR

    International Nuclear Information System (INIS)

    Strachan, J.D.; Boody, F.P.; Bush, C.E.

    1986-10-01

    Detached plasmas are formed in TFTR which have the principal property of the boundary to the high temperature plasma core being defined by a radiating layer. This paper documents the properties of TFTR ohmic-detached plasmas with a range of plasma densities at two different plasma currents

  9. Endovascular therapy of arteriovenous fistulae with electrolytically detachable coils

    Energy Technology Data Exchange (ETDEWEB)

    Jansen, O.; Doerfler, A.; Forsting, M.; Hartmann, M.; Kummer, R. von; Tronnier, V.; Sartor, K. [Dept. of Neuroradiology, University of Heidelberg Medical School (Germany)

    1999-12-01

    We report our experience in using Guglielmi electrolytically detachable coils (GDC) alone or in combination with other materials in the treatment of intracranial or cervical high-flow fistulae. We treated 14 patients with arteriovenous fistulae on brain-supplying vessels - three involving the external carotid or the vertebral artery, five the cavernous sinus and six the dural sinuses - by endovascular occlusion using electrolytically detachable platinum coils. The fistula was caused by trauma in six cases. In one case Ehlers-Danlos syndrome was the underlying disease, and in the remaining seven cases no aetiology could be found. Fistulae of the external carotid and vertebral arteries and caroticocavernous fistulae were reached via the transarterial route, while in all dural fistulae a combined transarterial-transvenous approach was chosen. All fistulae were treated using electrolytically detachable coils. While small fistulae could be occluded with electrolytically detachable coils alone, large fistulae were treated by using coils to build a stable basket for other types of coil or balloons. In 11 of the 14 patients, endovascular treatment resulted in complete occlusion of the fistula; in the remaining three occlusion was subtotal. Symptoms and signs were completely abolished by this treatment in 12 patients and reduced in 2. On clinical and neuroradiological follow-up (mean 16 months) no reappearance of symptoms was recorded. (orig.)

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  11. Intraretinal proliferation induced by retinal detachment

    International Nuclear Information System (INIS)

    Fisher, S.K.; Erickson, P.A.; Lewis, G.P.; Anderson, D.H.

    1991-01-01

    Cellular proliferation after retinal detachment was studied by 3 H-thymidine light microscopic autoradiography in cats that had experimental detachments of 0.5-180 days duration. The animals underwent labeling 2 hr before death with an intraocular injection of 200 microCi of 3 H-thymidine. The number of labeled nuclei were counted in 1-micron thick tissue sections in regions of detachment, in regions of the experimental eyes that remained attached, and in control eyes that had no detachments. In the normal eye, in one that had only the lens and vitreous removed, and in the eyes with 0.5- and 1-day detachments, the number of labeled nuclei ranged from 0/mm (0.5-day detachment) to 0.38/mm (lens and vitreous removed only). By 2 days postdetachment, the number of labeled nuclei increased to 2.09/mm. The highest levels of labeling occurred in two animals with detachments of 3 (7.86/mm) and 4 (7.09/mm) days. Thereafter, the numbers declined steadily until near-baseline counts were obtained at 14 days. The number of labeled nuclei was slightly elevated in the attached regions of two animals with 3-day detachments. Labeled cell types included: Mueller cells, astrocytes, pericytes, and endothelial cells of the retinal vasculature, and both resident (microglial cells) and invading macrophages. In an earlier study RPE cells were also shown to proliferate in response to detachment. Thus, these data show that proliferation is a rapid response to detachment, reaching a maximum within 4 days, and that virtually every nonneuronal cell type in the retina can participate in this response. The data suggest that events leading to such clinical manifestations as proliferative vitreoretinopathy and subretinal fibrosis may have their beginnings in this very early proliferative response

  12. Paediatric retinal detachment: aetiology, characteristics and outcomes

    Directory of Open Access Journals (Sweden)

    Elizabeth McElnea

    2018-02-01

    Full Text Available AIM: To provide contemporary data on the aetiology, clinical features and outcomes of paediatric retinal detachment. METHODS: A retrospective review of all those under 16y who underwent surgical repair for retinal detachment at a single centre between the years 2008 and 2015 inclusive was performed. In each case the cause of retinal detachment, the type of detachment, the presence or absence of macular involvement, the number and form of reparative surgeries undertaken, and the surgical outcome achieved was recorded. RESULTS: Twenty-eight eyes of 24 patients, 15 (62.5% of whom were male and 9 (37.5% of whom were female, their mean age being 11.6y and range 2-16y developed retinal detachment over the eight year period studied. Trauma featured in the development of retinal detachment in 14 (50.0% cases. Retinal detachment was associated with other ocular and/or systemic conditions in 11 (39.3% cases. A mean of 3.0 procedures with a range of 1-9 procedures per patient were undertaken in the management of retinal detachment. Complex vitrectomy combined with scleral buckling or complex vitrectomy alone were those most frequently performed. Mean postoperative visual acuity was 1.2 logMAR with range 0.0-3.0 logMAR. In 22 of 26 (84.6% cases which underwent surgical repair the retina was attached at last follow-up. CONCLUSION: Aggressive management of paediatric retinal detachment including re-operation increases the likelihood of anatomical success. In cases where the retinal detachment can be repaired by an external approach alone there is a more favourable visual outcome.

  13. Paediatric retinal detachment: aetiology, characteristics and outcomes.

    Science.gov (United States)

    McElnea, Elizabeth; Stephenson, Kirk; Gilmore, Sarah; O'Keefe, Michael; Keegan, David

    2018-01-01

    To provide contemporary data on the aetiology, clinical features and outcomes of paediatric retinal detachment. A retrospective review of all those under 16y who underwent surgical repair for retinal detachment at a single centre between the years 2008 and 2015 inclusive was performed. In each case the cause of retinal detachment, the type of detachment, the presence or absence of macular involvement, the number and form of reparative surgeries undertaken, and the surgical outcome achieved was recorded. Twenty-eight eyes of 24 patients, 15 (62.5%) of whom were male and 9 (37.5%) of whom were female, their mean age being 11.6y and range 2-16y developed retinal detachment over the eight year period studied. Trauma featured in the development of retinal detachment in 14 (50.0%) cases. Retinal detachment was associated with other ocular and/or systemic conditions in 11 (39.3%) cases. A mean of 3.0 procedures with a range of 1-9 procedures per patient were undertaken in the management of retinal detachment. Complex vitrectomy combined with scleral buckling or complex vitrectomy alone were those most frequently performed. Mean postoperative visual acuity was 1.2 logMAR with range 0.0-3.0 logMAR. In 22 of 26 (84.6%) cases which underwent surgical repair the retina was attached at last follow-up. Aggressive management of paediatric retinal detachment including re-operation increases the likelihood of anatomical success. In cases where the retinal detachment can be repaired by an external approach alone there is a more favourable visual outcome.

  14. 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.

  15. 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.

  16. 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

  17. Divertor impurity monitor for the International Thermonuclear Experimental Reactor

    Science.gov (United States)

    Sugie, T.; Ogawa, H.; Nishitani, T.; Kasai, S.; Katsunuma, J.; Maruo, M.; Ebisawa, K.; Ando, T.; Kita, Y.

    1999-01-01

    The divertor impurity monitoring system of the International Thermonuclear Experimental Reactor 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 200-1000 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 λthe 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.

  18. 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)

  19. Particle control in the DIII-D advanced divertor

    International Nuclear Information System (INIS)

    Schaffer, M.J.; Lippmann, S.I.; Mahdavi, M.A.; Petrie, T.W.; Stambaugh, R.D.; Hogan, J.; Klepper, C.C.; Mioduszewski, P.; Owen, L.; Hill, D.N.; Rensink, M.; Buchenauer, D.

    1991-11-01

    A new, electrically biasable, semi-closed divertor was installed and operated in the D3-D lower outside divertor location. The semi-closed divertor has yielded static gas pressure buildups in the pumping plenum in excess of 10 mtorr. (The planned cryogenic pumping is not yet installed). Electrical bias controls the distribution of particle recycle between the inner and outer divertors by rvec E x rvec B drifts. Depending on sign, bias increases or decreases the plenum gas pressure. Bias greatly reduce the sensitivity of plenum pressure to separatrix position. In particular, rvec E x rvec B drifts in the D3-D geometry can direct plasma across a divertor target and then optimally into the pumping aperture. Bias, even without active pumping, has also demonstrated a limited control of ELMing H-mode plasma density. 5 refs., 8 figs

  20. Divertor design for the TITAN reversed-field-pinch reactor

    International Nuclear Information System (INIS)

    Cooke, P.I.H.; Bathke, C.G.; Blanchard, J.P.; Creedon, R.L.; Grotz, S.P.; Hasan, M.Z.; Orient, G.; Sharafat, S.; Werley, K.A.

    1987-01-01

    The design of the toroidal-field divertor for the TITAN high-power-density reversed-field-pinch reactor is described. The heat flux on the divertor target is limited to acceptable levels (≤ 10 MW/m 2 ) for liquid-lithium cooling by use of an open divertor geometry, strong radiation from the core and edge plasma, and careful shaping of the target surface. The divertor coils are based on the Integrated-Blanket-Coil approach to minimize the loss in breeding-blanket coverage due to the divertor. A tungsten-rhenium armour plate, chosen for reasons of sputtering resistance, and good thermal and mechanical properties, protects the vanadium-alloy coolant tubes

  1. Structural analysis of the ITER Divertor toroidal rails

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  2. Simulation of the ASDEX divertor performance after hardening

    International Nuclear Information System (INIS)

    Schneider, W.; Lackner, K.; Neuhauser, J.; Wunderlich, R.

    1985-05-01

    Two combined computer models - a fluid description of the plasma scrape-off layer (SOLID) and a Monte-Carlo code for the neutral gas dynamics (DEGAS) - are used to assess changes in the divertor performance expected from the modifications in geometry needed for hardening the ASDEX divertor chamber for long-pulse, high-power heating. Stand-alone DEGAS calculations with assumed fixed scrape-off plasma parameters predict a doubling of the neutral escape probability, which, however, still remains so low, that achievement of the high divertor recycling regime can be expected over roughly the same operational regime as before modifications. This conclusion is also supported by fully self-consistent calculations with the combined model. Due to the reduced divertor, a significant reduction is predicted in the divertor time constant, which is expected to affect transient phenomena. (orig.)

  3. 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)

  4. 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

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

    International Nuclear Information System (INIS)

    Coenen, J.W.; Clever, M.; Knaup, M.; Arnoux, G.; Matthews, G.F.; Balboa, I.; Meigs, A.; Bazylev, B.; Autricque, A.; Dejarnac, R.; Horacek, J.; Komm, M.; Coffey, I.; Corre, Y.; Gauthier, E.; Devaux, S.; Krieger, K.; Frassinetti, L.; Jachmich, S.; Marsen, S.

    2015-01-01

    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 I P  = 3.0 MA/B T  = 2.9 T H-mode pulses with an input power of P IN  = 23 MW, a stored energy of ∼6 MJ and regular type I ELMs at ΔW ELM  = 0.3 MJ and f ELM  ∼ 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 mm 3 ) 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

  6. BTeV detached vertex trigger

    International Nuclear Information System (INIS)

    Gottschalk, E.E.

    2001-01-01

    BTeV is a collider experiment that has been approved to run in the Tevatron at Fermilab. The experiment will conduct precision studies of CP violation using a forward-geometry detector. The detector will be optimized for high-rate detection of beauty and charm particles produced in collisions between protons and anti-protons. BTeV will trigger on beauty and charm events by taking advantage of the main difference between these heavy quark events and more typical hadronic events - the presence of detached beauty and charm decay vertices. The first stage of the BTeV trigger will receive data from a pixel vertex detector at a rate of 100 gb s -1 , reconstruct tracks and vertices for every beam crossing, reject 99% of beam crossings that do not produce beauty or charm particles, and trigger on beauty events with high efficiency. An overview of the trigger design and its influence on the design of the pixel vertex detector is presented

  7. The influence of refractive error and lattice degeneration on the incidence of retinal detachment.

    Science.gov (United States)

    Burton, T C

    1989-01-01

    This study indicates the feasibility of stratifying the general population into various risk pools for retinal detachment depending on a person's age, refractive status, and the presence of lattice degeneration. At first impression the risks seem at variance with the fine clinical studies of Byer, who has shown a very low detachment rate in the population with lattice degeneration. In all likelihood the vast majority of his patients were emmetropic or mildly myopic, so that very few would be expected to develop detachments during their entire lifetimes, let along during intervals of only 10 to 20 years. This study shows the futility of following, or treating prophylactically, young emmetropic individuals with lattice degeneration. Assuming that prophylaxis is actually effective, one would have to treat 1000 emmetropic lattice patients in the 30 to 39 year age group to prevent a single detachment over a 10-year period. Lattice patients with low to moderate degrees of myopia tend to develop detachments between 40 and 60 years of age caused by premature posterior vitreous separation and tractional tears. Clearly prophylaxis for this group is not warranted, since only 5% to 10% of these individuals will experience detachments in their lifetimes. On the other hand this study has verified the previous suspicions that persons with myopia exceeding -5.0 D accompanied by lattice degeneration have an extraordinarily high risk of detachment during their lifetimes. Detachments in this group tend to cluster in the second, third, and fourth decades, are typically caused by atrophic holes, are slowly progressive, and are often simultaneously bilateral. Enhanced vigilance is certainly appropriate during this time and perhaps consideration should be given to prophylactically treating this group. This would be no small task, since within a population of 1 million persons there would be about 1150 aged 10 to 39 years with myopia exceeding -5.0 D and lattice degeneration. Only 4

  8. Design analysis of the ITER divertor

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  9. Design analysis of the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  11. Measurements of poloidal and toroidal energy deposition asymmetries in the ASDEX divertors

    International Nuclear Information System (INIS)

    Evans, T.E.

    1991-03-01

    Energy deposition characteristics in the ASDEX divertors have been analyzed over a wide range of discharges and wall conditions during ohmically heated, additionally heated, or lower hybrid current drive experiments. Changes in discharge operating parameters with high power additional heating produce a diversity of effects on the magnitudes and distributions of the energy absorbed in the divertors. Poloidally and toroidally resolved energy deposition patterns are particularly sensitive to changes in the edge safety factor, the type and power level of additional heating used, and the vertical position of the plasma. In most additionally heated discharges, a large fraction of the incremental divertor loading is found on only one or two target rings. Poloidal in-out asymmetries, which typically favor the low-field side by a factor of 2.5 in ohmic discharges, commonly range between a factor of 2.5 and 4.5 in additionally heated experiments and in extreme cases can be as large as a factor of 5.6. At the same time, toroidal asymmetries on individual target rings are found to range between a factor of 1.4 and 3.8 in typical ICRH and NBI cases with extreme LHCD cases of 4.3. A model, proposed to explain the cause of discharge asymmetries, is compared with the experimental observations. Under some conditions, for example during LHCD experiments, the model is in good agreement with the data. A method is proposed for supressing discharge asymmetries which may generally improve the divertor performance as well. (orig./AH)

  12. Rhegmatogenous retinal detachment following intravitreal ocriplasmin

    NARCIS (Netherlands)

    Madi, Haifa A.; Haynes, Richard J.; Depla, Diana; de la Cour, Morten D.; Lesnik-Oberstein, Sarit; Muqit, Mahi M. K.; Patton, Niall; Price, Nick; Steel, David H. W.

    2016-01-01

    To describe the characteristics and outcomes of patients presenting with rhegmatogenous retinal detachment (RRD) after ocriplasmin (OCP) injection. Retrospective, multi-centre, observational case series with case note review. Eight patients with symptomatic vitreomacular traction (six with

  13. A Meta-Analysis on Antecedents and Outcomes of Detachment from Work.

    Science.gov (United States)

    Wendsche, Johannes; Lohmann-Haislah, Andrea

    2016-01-01

    Detachment from work has been proposed as an important non-work experience helping employees to recover from work demands. This meta-analysis (86 publications, k = 91 independent study samples, N = 38,124 employees) examined core antecedents and outcomes of detachment in employee samples. With regard to outcomes, results indicated average positive correlations between detachment and self-reported mental (i.e., less exhaustion, higher life satisfaction, more well-being, better sleep) and physical (i.e., lower physical discomfort) health, state well-being (i.e., less fatigue, higher positive affect, more intensive state of recovery), and task performance (small to medium sized effects). However, average relationships between detachment and physiological stress indicators and work motivation were not significant while associations with contextual performance and creativity were significant, but negative. Concerning work characteristics, as expected, job demands were negatively related and job resources were positively related to detachment (small sized effects). Further, analyses revealed that person characteristics such as negative affectivity/neuroticism (small sized effect) and heavy work investment (medium sized effect) were negatively related to detachment whereas detachment and demographic variables (i.e., age and gender) were not related. Moreover, we found a medium sized average negative relationship between engagement in work-related activities during non-work time and detachment. For most of the examined relationships heterogeneity of effect sizes was moderate to high. We identified study design, samples' gender distribution, and affective valence of work-related thoughts as moderators for some of these aforementioned relationships. The results of this meta-analysis point to detachment as a non-work (recovery) experience that is influenced by work-related and personal characteristics which in turn is relevant for a range of employee outcomes.

  14. A Meta-Analysis on Antecedents and Outcomes of Detachment from Work

    Science.gov (United States)

    Wendsche, Johannes; Lohmann-Haislah, Andrea

    2017-01-01

    Detachment from work has been proposed as an important non-work experience helping employees to recover from work demands. This meta-analysis (86 publications, k = 91 independent study samples, N = 38,124 employees) examined core antecedents and outcomes of detachment in employee samples. With regard to outcomes, results indicated average positive correlations between detachment and self-reported mental (i.e., less exhaustion, higher life satisfaction, more well-being, better sleep) and physical (i.e., lower physical discomfort) health, state well-being (i.e., less fatigue, higher positive affect, more intensive state of recovery), and task performance (small to medium sized effects). However, average relationships between detachment and physiological stress indicators and work motivation were not significant while associations with contextual performance and creativity were significant, but negative. Concerning work characteristics, as expected, job demands were negatively related and job resources were positively related to detachment (small sized effects). Further, analyses revealed that person characteristics such as negative affectivity/neuroticism (small sized effect) and heavy work investment (medium sized effect) were negatively related to detachment whereas detachment and demographic variables (i.e., age and gender) were not related. Moreover, we found a medium sized average negative relationship between engagement in work-related activities during non-work time and detachment. For most of the examined relationships heterogeneity of effect sizes was moderate to high. We identified study design, samples' gender distribution, and affective valence of work-related thoughts as moderators for some of these aforementioned relationships. The results of this meta-analysis point to detachment as a non-work (recovery) experience that is influenced by work-related and personal characteristics which in turn is relevant for a range of employee outcomes. PMID:28133454

  15. 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)

  16. Evaluation of divertor conceptual designs for a fusion power plant

    International Nuclear Information System (INIS)

    Ferrari, M.; Giancarli, L.; Kleefeldt, K.; Nardi, C.; Roedig, M.; Reimann, J.; Salavy, J.F.

    2001-01-01

    In the frame of the preliminary study of plants suitable for the energy production from the fusion power, particular emphasis has been given on the divertor studies. Since a significant percentage of the power generated from the fusion process is absorbed in the divertor, the thermal efficiency of the power conversion cycle requires a high coolant outlet temperature of the divertor, leading to solutions that are different from those adopted for the present experimental fusion plants. Therefore, copper alloys having extremely high thermal conductivity, cannot be used as structural material for this kind of devices. The most suitable coolants to be used in the divertor are water, helium and liquid metals. A conceptual design study has been developed for each of these three fluids, with the aim to evaluate the maximum allowable thermal flux at the divertor target plate and the R and D requirements for each solution. While a water-cooled divertor can be designed with a limited R and D effort, the development of helium or liquid metal cooled divertors requires a more engaging R and D program

  17. 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.)

  18. Mitigation of divertor heat loads by strike point sweeping in high power JET discharges

    Science.gov (United States)

    Silburn, S. A.; Matthews, G. F.; Challis, C. D.; Frigione, D.; Graves, J. P.; Mantsinen, M. J.; Belonohy, E.; Hobirk, J.; Iglesias, D.; Keeling, D. L.; King, D.; Kirov, K.; Lennholm, M.; Lomas, P. J.; Moradi, S.; Sips, A. C. C.; Tsalas, M.; Contributors, JET

    2017-12-01

    Deliberate periodic movement (sweeping) of the high heat flux divertor strike lines in tokamak plasmas can be used to manage the heat fluxes experienced by exhaust handling plasma facing components, by spreading the heat loads over a larger surface area. Sweeping has recently been adopted as a routine part of the main high performance plasma configurations used on JET, and has enabled pulses with 30 MW plasma heating power and 10 MW radiation to run for 5 s without overheating the divertor tiles. We present analysis of the effectiveness of sweeping for divertor temperature control on JET, using infrared camera data and comparison with a simple 2D heat diffusion model. Around 50% reduction in tile temperature rise is obtained with 5.4 cm sweeping compared to the un-swept case, and the temperature reduction is found to scale slower than linearly with sweeping amplitude in both experiments and modelling. Compatibility of sweeping with high fusion performance is demonstrated, and effects of sweeping on the edge-localised mode behaviour of the plasma are reported and discussed. The prospects of using sweeping in future JET experiments with up to 40 MW heating power are investigated using a model validated against existing experimental data.

  19. Mitigation of divertor heat loads by strike point sweeping in high power JET discharges

    International Nuclear Information System (INIS)

    Silburn, S A; Matthews, G F; Challis, C D; Belonohy, E; Iglesias, D; Keeling, D L; King, D; Kirov, K; Lomas, P J; Frigione, D; Graves, J P; Mantsinen, M J; Hobirk, J; Lennholm, M; Moradi, S; Sips, A C C; Tsalas, M

    2017-01-01

    Deliberate periodic movement (sweeping) of the high heat flux divertor strike lines in tokamak plasmas can be used to manage the heat fluxes experienced by exhaust handling plasma facing components, by spreading the heat loads over a larger surface area. Sweeping has recently been adopted as a routine part of the main high performance plasma configurations used on JET, and has enabled pulses with 30 MW plasma heating power and 10 MW radiation to run for 5 s without overheating the divertor tiles. We present analysis of the effectiveness of sweeping for divertor temperature control on JET, using infrared camera data and comparison with a simple 2D heat diffusion model. Around 50% reduction in tile temperature rise is obtained with 5.4 cm sweeping compared to the un-swept case, and the temperature reduction is found to scale slower than linearly with sweeping amplitude in both experiments and modelling. Compatibility of sweeping with high fusion performance is demonstrated, and effects of sweeping on the edge-localised mode behaviour of the plasma are reported and discussed. The prospects of using sweeping in future JET experiments with up to 40 MW heating power are investigated using a model validated against existing experimental data. (paper)

  20. Fruit Detachment and Classification Method for Strawberry Harvesting Robot

    Directory of Open Access Journals (Sweden)

    Guo Feng

    2008-11-01

    Full Text Available Fruit detachment and on-line classification is important for the development of harvesting robot. With the specific requriements of robot used for harvesting strawberries growing on the ground, a fruit detachment and classification method is introduced in this paper. OHTA color spaces based image segmentation algorithm is utilized to extract strawberry from background; Principal inertia axis of binary strawberry blob is calculated to give the pose information of fruit. Strawberry is picked selectively according to its ripeness and classified according to its shape feature. Histogram matching based method for fruit shape judgment is introduced firstly. Experiment results show that this method can achieve 93% accuracy of strawberry's stem detection, 90% above accuracy of ripeness and shape quality judgment on black and white background. With the improvement of harvesting mechanism design, this method has application potential in the field operation.

  1. Fruit Detachment and Classification Method for Strawberry Harvesting Robot

    Directory of Open Access Journals (Sweden)

    Guo Feng

    2008-03-01

    Full Text Available Fruit detachment and on-line classification is important for the development of harvesting robot. With the specific requriements of robot used for harvesting strawberries growing on the ground, a fruit detachment and classification method is introduced in this paper. OHTA color spaces based image segmentation algorithm is utilized to extract strawberry from background; Principal inertia axis of binary strawberry blob is calculated to give the pose information of fruit. Strawberry is picked selectively according to its ripeness and classified according to its shape feature. Histogram matching based method for fruit shape judgment is introduced firstly. Experiment results show that this method can achieve 93% accuracy of strawberry's stem detection, 90% above accuracy of ripeness and shape quality judgment on black and white background. With the improvement of harvesting mechanism design, this method has application potential in the field operation.

  2. Disruption characteristics in PDX with limiter and divertor discharges

    International Nuclear Information System (INIS)

    Couture, P.; McGuire, K.

    1986-09-01

    A comparison has been made between the characteristics of disruptions with limiter and divertor configurations in PDX. A large data base on disruptions has been collected over four years of machine operation, and a total of 15,000 discharges are contained in the data file. It was found that divertor discharges have less disruptions during ramp up and flattop of the plasma current. However, for divertor discharges a large number of fast, low current disruptions take place during the current ramp down. These disruptions are probably caused by the deformation of the plasma shape

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  4. 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

  5. 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.

  6. ELM induced divertor heat loads on TCV

    Science.gov (United States)

    Marki, J.; Pitts, R. A.; Horacek, J.; Tskhakaya, D.; TCV Team

    2009-06-01

    Results are presented for heat loads at the TCV outer divertor target during ELMing H-mode using a fast IR camera. Benefitting from a recent surface cleaning of the entire first wall graphite armour, a comparison of the transient thermal response of freshly cleaned and untreated tile surfaces (coated with thick co-deposited layers) has been performed. The latter routinely exhibit temperature transients exceeding those of the clean ones by a factor ˜3, even if co-deposition throughout the first days of operation following the cleaning process leads to the steady regrowth of thin layers. Filaments are occasionally observed during the ELM heat flux rise phase, showing a spatial structure consistent with energy release at discrete toroidal locations in the outer midplane vicinity and with individual filaments carrying ˜1% of the total ELM energy. The temporal waveform of the ELM heat load is found to be in good agreement with the collisionless free streaming particle model.

  7. ELM induced divertor heat loads on TCV

    Energy Technology Data Exchange (ETDEWEB)

    Marki, J., E-mail: janos.marki@epfl.c [Centre de Recherches en Physique des Plasmas (CRPP), Ecole Polytechnique Federale de Lausanne (EPFL), Association Euratom - Confederation Suisse, CH-1015 Lausanne (Switzerland); Pitts, R.A. [Centre de Recherches en Physique des Plasmas (CRPP), Ecole Polytechnique Federale de Lausanne (EPFL), Association Euratom - Confederation Suisse, CH-1015 Lausanne (Switzerland); Horacek, J. [Institute of Plasma Physics, Association EUROATOM-IPP.CR, Za Slovankou 3, 182 00 Prague 8 (Czech Republic); Tskhakaya, D. [Association EURATOM-OAW, Institut fuer Theoretische Physik, A-6020 Innsbruck (Austria)

    2009-06-15

    Results are presented for heat loads at the TCV outer divertor target during ELMing H-mode using a fast IR camera. Benefitting from a recent surface cleaning of the entire first wall graphite armour, a comparison of the transient thermal response of freshly cleaned and untreated tile surfaces (coated with thick co-deposited layers) has been performed. The latter routinely exhibit temperature transients exceeding those of the clean ones by a factor approx3, even if co-deposition throughout the first days of operation following the cleaning process leads to the steady regrowth of thin layers. Filaments are occasionally observed during the ELM heat flux rise phase, showing a spatial structure consistent with energy release at discrete toroidal locations in the outer midplane vicinity and with individual filaments carrying approx1% of the total ELM energy. The temporal waveform of the ELM heat load is found to be in good agreement with the collisionless free streaming particle model.

  8. 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.

  9. 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)

  10. 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.

  11. Powder Injection Molding - An innovative manufacturing method for He-cooled DEMO divertor components

    International Nuclear Information System (INIS)

    Antusch, Steffen; Norajitra, Prachai; Piotter, Volker; Ritzhaupt-Kleissl, Hans-Joachim; Spatafora, Luigi

    2011-01-01

    At Karlsruhe Institute of Technology (KIT), a He-cooled divertor design for future fusion power plants has been developed. This concept is based on the use of modular cooling fingers made from tungsten and tungsten alloy, which are presently considered the most promising divertor materials to withstand the specific heat load of 10 MW/m 2 . Since a large number of the finger modules (n > 250,000) are needed for the whole reactor, developing a mass-oriented manufacturing method is indispensable. In this regard, an innovative manufacturing technology, Powder Injection Molding (PIM), has been adapted to W 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 complete technological PIM process for tungsten materials and its application on manufacturing of real divertor components, including the design of a new PIM tool is outlined and, results of the examination of the finished product after heat-treatment are discussed. A binary tungsten powder feedstock with a solid load of 50 vol.% was developed and successfully tested in molding experiments. After design, simulation and manufacturing of a new PIM tool, real divertor parts are produced. After heat-treatment (pre-sintering and HIP) the successful finished samples showed a sintered density of approximately 99%, a hardness of 457 HV0.1, a grain size of approximately 5 μm and a microstructure without cracks and porosity.

  12. 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

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

    International Nuclear Information System (INIS)

    Ando, K.; Ejima, S.; Davis, S.; Hawryluk, R.; Hsuan, H.; Meade, D.; Okabayaski, M.; Sauthoff, N.; Schmidt, J.; Sinnis, J.

    1974-10-01

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

  14. Two-dimensional divertor modeling and scaling laws

    International Nuclear Information System (INIS)

    Catto, P.J.; Connor, J.W.; Knoll, D.A.

    1996-01-01

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

  15. 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

  16. 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.

  17. 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.

  18. 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.)

  19. TCV experiments towards the development of a plasma exhaust solution

    Science.gov (United States)

    Reimerdes, H.; Duval, B. P.; Harrison, J. R.; Labit, B.; Lipschultz, B.; Lunt, T.; Theiler, C.; Tsui, C. K.; Verhaegh, K.; Vijvers, W. A. J.; Boedo, J. A.; Calabro, G.; Crisanti, F.; Innocente, P.; Maurizio, R.; Pericoli, V.; Sheikh, U.; Spolare, M.; Vianello, N.; the TCV Team; the EUROfusion MST1 Team

    2017-12-01

    Research towards a plasma exhaust solution for a fusion power plant aims at validating edge physics models, strengthening predictive capabilities and improving the divertor configuration. The TCV tokamak is extensively used to investigate the extent that geometric configuration modifications can affect plasma exhaust performance. Recent TCV experiments continue previous detachment studies of Ohmically heated L-mode plasmas in standard single-null configurations, benefitting from a range of improved diagnostic capabilities. Studies were extended to nitrogen seeding and an entire suite of alternative magnetic configurations, including flux flaring towards the target (X divertor), increasing the outer target radius (Super-X) and movement of a secondary x-point inside the vessel (X-point target) as well as the entire range of snowflake configurations. Nitrogen seeding into a snowflake minus configuration demonstrated a regime with strong radiation in the large region between the two x-points, confirming EMC3-Eirene simulations, and opening a promising path towards highly radiating regimes with limited adverse effects on core performance.

  20. 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.

  1. 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.

  2. Non-ambipolar divertor flows in heliotron E

    International Nuclear Information System (INIS)

    Chechkin, V.V.; Voitsenya, V.S.; Smirnova, M.S.; Sorokovoj, E.L.; Mizuuchi, T.; Nagasaki, K.; Okada, H.; Funaba, H.; Hamada, T.; Sano, F.; Zushi, H.; Nakasuga, M.; Kondo, K.; Masuzaki, S.; Motojima, O.

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Smirnov, R.D.; Pigarov, A.Yu.; Krasheninnikov, S.I.; Rognlien, T.D.; Soukhanovskii, V.A.; Rensink, M.E.; Maingi, R.; Skinner, C.H.; Stotler, D.P.; Bell, R.E.; Kugel, H.W.

    2010-01-01

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

  4. Multi-Fluid Modeling of Low-Recycling Divertor Regimes

    International Nuclear Information System (INIS)

    Smirnov, R.D.; Pigarov, A.Y.; Krasheninnikov, S.I.; Rognlien, T.D.; Soukhanovskii, V.A.; Rensink, M.E.; Maingi, R.; Skinner, C.H.; Stotler, D.P.; Bell, R.E.; Kugel, H.W.

    2010-01-01

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

  5. Structural design of the DIII-D radiative divertor

    International Nuclear Information System (INIS)

    Reis, E.E.; Smith, J.P.; Baxi, C.B.; Bozek, A.S.; Chin, E.; Hollerbach, M.A.; Laughon, G.J.; Sevier, D.L.

    1996-10-01

    The divertor of the DIII-D tokamak is being modified to operate as a slot type, dissipative divertor. This modification, called the Radiative Divertor Program (RDP) is being carried out in two phases. The design and analysis is complete and hardware is being fabricated for the first phase. This first phase consists of an upper divertor baffle and cryopump to provide some density control for high triangularity, single or double null discharges. Installation of the first phase is scheduled to start in October, 1996. The second phase provides pumping at all four divertor strike points of double null high triangularity discharges and baffling of the neutral particles from transport back to the core plasma. Studies of the effects of varying the slot length and width of the divertor can be easily accomplished with the design of RDP hardware. Static and dynamic analyses of the baffle structures, new cryopumps, and feedlines were performed during the preliminary and final design phases. Disruption loads and differential thermal displacements must be accommodated in the design of these components. With the full RDP hardware installed, the plasma current in DIII-D will be a maximum of 3.0 MA. Plasma disruptions induce toroidal currents in the cryopump, producing complex dynamic loads. Simultaneously, the vacuum vessel vibrations impose a sinusoidal base excitation to the supports for the cryopump. Static and dynamic analyses of the cryopump demonstrate that the stresses due to disruption and thermal loadings satisfy the stress and deflection criteria

  6. [Paediatric retinal detachment and hereditary vitreoretinal disorders].

    Science.gov (United States)

    Meier, P

    2013-09-01

    The number of retinal detachments in children is very low in comparison to the number in adults. One predisposing factor for development of paediatric retinal detachment is suffering from hereditary vitreoretinal degeneration (e.g., Stickler syndrome, Wagner syndrome, Kniest dysplasia, familial exudative vitreoretinopathy, congenital X-linked retinoschisis, Knobloch syndrome, incontinentia pigmenti, Norrie disease). Hereditary vitreoretinopathies are characterised by an abnormal-appearing vitreous gel with associated retinal changes. In most of these eyes further ocular abnormalities can be diagnosed. A group of hereditary disorders is associated with characteristic systemic abnormalities. Allied conditions should be considered in the clinical diagnosis. Vitreoretinopathies are the most common cause of inherited retinal detachment. In most eyes primary vitrectomy is necessary, and disease-specific surgical treatment is discussed. Georg Thieme Verlag KG Stuttgart · New York.

  7. Testing of improved CFC/Cu bondings for the W7-X divertor targets

    International Nuclear Information System (INIS)

    Greuner, H.; Buswirth, B.; Boscary, J.; Tivey, R.; Plankensteiner, A.; Schedler, B.

    2007-01-01

    Full text of publication follows: Extensive high heat flux (HHF) testing of pre-series divertor targets was performed to establish the industrial process for the manufacturing of 890 targets, which will be needed for the installation of the Wendelstein 7-X (W7-X) divertor. The target design consists of flat tiles of CFC NB31 as plasma facing material bonded by an Active Meta] Casting copper (AMC) interlayer onto a water-cooled CuCrZr structure. This design is required by the specific geometrical requirements of the W7-X divertor. The heat removal capability of this target concept has been demonstrated for the envisaged operational power load of 10 MW/m 2 in previous test series of more than 30 full-scale elements. No large detachment or loss of CFC tiles occurred during cyclic loading tests at 10.5 and 13 MW/m 2 , but growing local de-bonded zones at the free edges of several CFC tiles were observed. Therefore a detailed analysis of the system of CFC/Cu bonding was carried out with respect to a further reduction of the stress at the CFC/Cu interface. Based on the results of the 3/D non-linear thermomechanical FEM analysis of the CFC/Cu interface a set of 17 additional pre-series elements was manufactured by PLANSEE SE. Three types of design variations have been investigated: - adopting an additional plastically compliant Cu interlayer between the cooling structure and the AMC region, - reduced size of CFC tiles, - arrangement of tiles with 90 deg. rotation of the CFC fibre plane. HHF tests were performed in the ion beam test facility GLADIS at IPP Garching with up to 3000 cycles at 10.5 MW/m 2 on this elements. The aim of these tests is to investigate the crack propagation between CFC/Cu and to define the acceptable defect size after 100 HHF cycles as an acceptance criterion for the series manufacturing. The applied criterion should allow the selection of elements for W7-X expected to achieve a suitable operational life time. Finally, the design variant with the

  8. Dynamic divertor control using resonant mixed toroidal harmonic magnetic fields during ELM suppression in DIII-D

    Science.gov (United States)

    Jia, M.; Sun, Y.; Paz-Soldan, C.; Nazikian, R.; Gu, S.; Liu, Y. Q.; Abrams, T.; Bykov, I.; Cui, L.; Evans, T.; Garofalo, A.; Guo, W.; Gong, X.; Lasnier, C.; Logan, N. C.; Makowski, M.; Orlov, D.; Wang, H. H.

    2018-05-01

    Experiments using Resonant Magnetic Perturbations (RMPs), with a rotating n = 2 toroidal harmonic combined with a stationary n = 3 toroidal harmonic, have validated predictions that divertor heat and particle flux can be dynamically controlled while maintaining Edge Localized Mode (ELM) suppression in the DIII-D tokamak. Here, n is the toroidal mode number. ELM suppression over one full cycle of a rotating n = 2 RMP that was mixed with a static n = 3 RMP field has been achieved. Prominent heat flux splitting on the outer divertor has been observed during ELM suppression by RMPs in low collisionality regime in DIII-D. Strong changes in the three dimensional heat and particle flux footprint in the divertor were observed during the application of the mixed toroidal harmonic magnetic perturbations. These results agree well with modeling of the edge magnetic field structure using the TOP2D code, which takes into account the plasma response from the MARS-F code. These results expand the potential effectiveness of the RMP ELM suppression technique for the simultaneous control of divertor heat and particle load required in ITER.

  9. From detached to attached buildup complexes

    DEFF Research Database (Denmark)

    Rafaelsen, B.; Elvebakk, G.; Andreassen, K.

    2008-01-01

    -like ridges and possibly areas with restricted circulation. Warm-water carbonate buildups, forming ridges and isolated mounds, occur in the Gipsdalen Group (latest Serpukhovian-mid-Sakmarian), where they initially grew in a detached platform setting. The carbonate buildups are several tens of kilometres long...... deposition and buildup growth bridged the detached platform with the attached platform. In the Bjarmeland Group (Lower Permian) 0.35-4.8 km wide, 1.5-27 km long and 60-420 m thick cool-water bryozoan-dominated straight, sinuous and continuous carbonate ridges or atoll-like ridges are located on top...

  10. Raised intraocular pressure and recurrence of retinal detachment as complications of external retinal detachment surgery

    International Nuclear Information System (INIS)

    Jawwad, M.; Khan, B.; Shah, M.A.; Qayyum, I.; Aftab, M.; Qayyum, I.

    2015-01-01

    Patients with Rhegmatogenous retinal detachment may develop raised intraocular pressure and recurrence of retinal detachment when they undergo external retinal detachment surgery. The present study was conducted to determine the postoperative rise in intraocular pressure (IOP) and recurrence of retinal detachment. Methods: The present descriptive study was conducted at Eye department of Lady Reading Hospital, Peshawar on 25 patients of both genders from August 2012 to July 2014. Results: Of the 25 patients, 18 (72%) developed raised IOP in the immediate postoperative period; this figure decreased to 12 (48%) at one week. Following medical or surgical intervention in these 12 cases, there was only 1 (4%) case with mildly raised IOP at two weeks postoperative. Five (20%) cases developed recurrent retinal detachment which later resolved with treatment. There were no significant differences by age or gender. Conclusion: External Retinal Detachment Surgery raised intraocular pressure postoperatively and caused recurrence of retinal detachment. These complications were treated medically and surgically with resolution within two weeks. (author)

  11. Electron detachment by H- + H- collisions

    International Nuclear Information System (INIS)

    Chibisov, M.I.; Yavlinskii, Yu.N.

    1994-01-01

    The autoionization channels for electrons detachment in collisions of two negative hydrogen ions in the energy range 300 - 500 eV and > 500 eV are investigated. In small energy range at large inter-ion distances the asymptotic approximation is used; at high energies - perturbation theory and classical motion are used. 18 refs., 3 figs

  12. Retinal detachment in black South Africans

    African Journals Online (AJOL)

    low incidence of retinal detachment in black patients is not known. ... a retinal break. Predisposing factors include peripheral retinal degenerations, myopia, aphakia and trauma. Delay in presentation increases the difficulty in achieving adequate surgical ... On examination, note was taken of the visual acuity in both eyes, the ...

  13. Detached Eddy Simulations of Hypersonic Transition

    Science.gov (United States)

    Yoon, S.; Barnhardt, M.; Candler, G.

    2010-01-01

    This slide presentation reviews the use of Detached Eddy Simulation (DES) of hypersonic transistion. The objective of the study was to investigate the feasibility of using CFD in general, DES in particular, for prediction of roughness-induced boundary layer transition to turbulence and the resulting increase in heat transfer.

  14. Shoreline response to detached breakwaters in prototype

    NARCIS (Netherlands)

    Khuong, T.C.

    2016-01-01

    An accurate prediction of shoreline changes behind detached breakwaters is, in regard to the adjustment to the environmental impact, still a challenge for designers and coastal managers. This research is expected to fill the gaps in the estimation of shoreline changes by developing new and

  15. Electron detachment in ion-atom collisions

    International Nuclear Information System (INIS)

    Vreugd, C. de.

    1980-01-01

    The electron detachment process that occurs in negative ion-atom collisions is investigated. Differential cross sections were measured for the collisions of F - , Cl - , Br - , I - on He, Ne, Ar, Kr, Xe, Na and K. Electron energy distributions were obtained for some of the systems. (Auth.)

  16. Detached Eddy Simulations of an Airfoil in Turbulent Inflow

    DEFF Research Database (Denmark)

    Gilling, Lasse; Sørensen, Niels; Davidson, Lars

    2009-01-01

    The effect of resolving inflow turbulence in detached eddy simulations of airfoil flows is studied. Synthetic turbulence is used for inflow boundary condition. The generated turbulence fields are shown to decay according to experimental data as they are convected through the domain with the free...... stream velocity. The subsonic flow around a NACA 0015 airfoil is studied at Reynolds number 1.6 × 106 and at various angles of attack before and after stall. Simulations with turbulent inflow are compared to experiments and to simulations without turbulent inflow. The results show that the flow...

  17. 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

  18. First measurements of the ion energy distribution at the divertor strike point during DIII-D disruptions

    International Nuclear Information System (INIS)

    Parks, P.B.; Brooks, N.H.; West, W.P.; Wong, C.P.C.; Bastasz, R.; Wampler, W.R.; Whyte, D.

    1995-12-01

    Plasma/wall interaction studies are being carried out using the Divertor Materials Exposure System (DiMES) on DIII-D. The objective of the experiment is to determine the kinetic energy and flux of deuterium ions reaching the divertor target during argon-induced radiative disruptions. The experiment utilizes a special slotted ion analyzer mounted over a Si sample to collect the fast charge-exchange (CX) deuterium neutrals emitted within the recycled cold neutral layer (CNL) which serves as a CX target for the incident ions. A theoretical interpretation of the experiment reveals a strong forward pitch-angle dependence in the approaching ion distribution function. The depth distribution of the trapped D in the Si sample was measured using low-energy direct recoil spectroscopy. Comparison with the TRIM code using monoenergetic ions indicated that the best fit to the data was obtained for an ion energy of 100 eV

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

    Science.gov (United States)

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

    2016-02-01

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

  20. Investigation of limiter recycling in the divertor tokamak ASDEX

    International Nuclear Information System (INIS)

    Wagner, F.

    1981-08-01

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

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

    International Nuclear Information System (INIS)

    Kukushkin, A.S.

    1994-01-01

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

  2. Effects of ELMs on ITER divertor armour materials

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-06-15

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

  3. 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.

  4. Acceptance criteria for the ITER divertor vertical target

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  5. 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

    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...

  6. Comparison study of toroidal-field divertors for a compact reversed-field pinch reactor

    International Nuclear Information System (INIS)

    Bathke, C.G.; Krakowski, R.A.; Miller, R.L.

    1985-01-01

    Two divertor configurations for the Compact Reversed-Field Pinch Reactor (CRFPR) based on diverting the minority (toroidal) field have been reported. A critical factor in evaluating the performance of both poloidally symmetric and bundle divertor configurations is the accurate determination of the divertor connection length and the monitoring of magnetic islands introduced by the divertors, the latter being a three-dimensional effect. To this end the poloidal-field, toroidal-field, and divertor coils and the plasma currents are simulated in three dimensions for field-line tracings in both the divertor channel and the plasma-edge regions. The results of this analysis indicate a clear preference for the poloidally symmetric toroidal-field divertor. Design modifications to the limiter-based CRFPR design that accommodate this divertor are presented

  7. Two-dimensional impurity transport calculations for a high recycling divertor

    International Nuclear Information System (INIS)

    Brooks, J.N.

    1986-04-01

    Two dimensional analysis of impurity transport in a high recycling divertor shows asymmetric particle fluxes to the divertor plate, low helium pumping efficiency, and high scrapeoff zone shielding for sputtered impurities

  8. Work stressors and partner social undermining: Comparing negative affect and psychological detachment as mechanisms.

    Science.gov (United States)

    Meier, Laurenz L; Cho, Eunae

    2018-05-14

    With the mounting evidence that employees' work experiences spill over into the family domain and cross over to family members, it is important to understand the underlying mechanism through which work experiences affect the family domain and what factors may alleviate the adverse impact of work stress. Expanding previous research that mainly focused on the affect-based mechanism (negative affect), the present research investigated a resource-based mechanism (psychological detachment from work) in the relationship linking two work stressors (high workload and workplace incivility) with social undermining toward the partner at home. We also explored the relative strength of the mediating effects of the two mechanisms. In addition, we tested whether relationship satisfaction moderates the proposed effect of detachment on partner undermining. We tested these research questions using two studies with differing designs: a five-wave longitudinal study (N = 470) and a multisource study (N = 131). The results suggest that stressful work experiences affect the family domain via lack of detachment as well as negative affect, that the two pathways have comparable strength, and that high relationship satisfaction mitigates the negative effect of lack of detachment on partner undermining. In sum, this research extends the spillover-crossover model by establishing that poor psychological detachment from work during leisure time is an additional mechanism that links work and family. (PsycINFO Database Record (c) 2018 APA, all rights reserved).