WorldWideScience

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. Plasma detachment in divertor tokamaks

    Science.gov (United States)

    Leonard, A. W.

    2018-04-01

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

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

    Science.gov (United States)

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

    2017-10-01

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

  6. 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.

  7. Plans of LHD divertor experiment

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-08-04

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

  9. 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.

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  11. Numerical studies on divertor experiments

    International Nuclear Information System (INIS)

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

    1988-04-01

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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  13. 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

  14. 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.)

  15. 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

  16. Real-time control of divertor detachment in H-mode with impurity seeding using Langmuir probe feedback in JET-ITER-like wall

    Science.gov (United States)

    Guillemaut, C.; Lennholm, M.; Harrison, J.; Carvalho, I.; Valcarcel, D.; Felton, R.; Griph, S.; Hogben, C.; Lucock, R.; Matthews, G. F.; Perez Von Thun, C.; Pitts, R. A.; Wiesen, S.; contributors, JET

    2017-04-01

    Burning plasmas with 500 MW of fusion power on ITER will rely on partially detached divertor operation to keep target heat loads at manageable levels. Such divertor regimes will be maintained by a real-time control system using the seeding of radiative impurities like nitrogen (N), neon or argon as actuator and one or more diagnostic signals as sensors. Recently, real-time control of divertor detachment has been successfully achieved in Type I ELMy H-mode JET-ITER-like wall discharges by using saturation current (I sat) measurements from divertor Langmuir probes as feedback signals to control the level of N seeding. The degree of divertor detachment is calculated in real-time by comparing the outer target peak I sat measurements to the peak I sat value at the roll-over in order to control the opening of the N injection valve. Real-time control of detachment has been achieved in both fixed and swept strike point experiments. The system has been progressively improved and can now automatically drive the divertor conditions from attached through high recycling and roll-over down to a user-defined level of detachment. Such a demonstration is a successful proof of principle in the context of future operation on ITER which will be extensively equipped with divertor target probes.

  17. 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.

  18. 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.

  19. 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

  20. Local island divertor experiments on LHD

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-03-01

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

  1. 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

  2. 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

  3. Time-dependent modeling of dust injection in semi-detached ITER divertor plasma

    Science.gov (United States)

    Smirnov, Roman; Krasheninnikov, Sergei

    2017-10-01

    At present, it is generally understood that dust related issues will play important role in operation of the next step fusion devices, i.e. ITER, and in the development of future fusion reactors. Recent progress in research on dust in magnetic fusion devises has outlined several topics of particular concern: a) degradation of fusion plasma performance; b) impairment of in-vessel diagnostic instruments; and c) safety issues related to dust reactivity and tritium retention. In addition, observed dust events in fusion edge plasmas are highly irregular and require consideration of temporal evolution of both the dust and the fusion plasma. In order to address the dust-related fusion performance issues, we have coupled the dust transport code DUSTT and the edge plasma transport code UEDGE in time-dependent manner, allowing modeling of transient dust-induced phenomena in fusion edge plasmas. Using the coupled codes we simulate burst-like injection of tungsten dust into ITER divertor plasma in semi-detached regime, which is considered as preferable ITER divertor operational mode based on the plasma and heat load control restrictions. Analysis of transport of the dust and the dust-produced impurities, and of dynamics of the ITER divertor and edge plasma in response to the dust injection will be presented. This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, under Award Number DE-FG02-06ER54852.

  4. 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

  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

    A revised formulation of the perpendicular diffusive transport model in 2D multi-fluid edge codes is proposed. Based on theoretical predictions and experimental observations a dependence on collisionality is introduced into the transport model of EDGE2D–EIRENE. The impact on time-dependent JET gas...... fuelled ramp-up scenario modelling of the full transient from attached divertor into the high-recycling regime, following a target flux roll over into divertor detachment, ultimately ending in a density limit is presented. A strong dependence on divertor geometry is observed which can mask features...... of the new transport model: a smoothly decaying target recycling flux roll over, an asymmetric drop of temperature and pressure along the field lines as well as macroscopic power dependent plasma oscillations near the density limit which had been previously observed also experimentally. The latter effect...

  6. 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

  7. 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.

  8. 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

  9. 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.

  10. 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

  11. 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.

  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. 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

  14. 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.)

  15. Characterization of divertor footprints and the pedestal plasmas in the presence of applied n = 3 fields for the attached and detached conditions in NSTX

    International Nuclear Information System (INIS)

    Ahn, J-W; Canik, J M; Lore, J D; Maingi, R; Gray, T K; Scotti, F; Kim, K; Bell, R E; Diallo, A; Gerhardt, S P; Kaye, S M; LeBlanc, B P; McLean, A G; Soukhanovskii, V A; Tritz, K

    2014-01-01

    Recent progress in the study of 3D field effects on the divertor and pedestal plasmas is reported with the use of a new set of diagnostics. A wide angle visible camera provides 2D data of lower divertor surface covering almost the full range of radius (r) and toroidal angle (Φ), a significant advantage over the conventional 1D radial profile in examining non-axisymmetric effects of 3D fields on the divertor footprints. The spatial distribution of connection lengths (L c ) calculated by vacuum field line tracing in the presence of 3D fields (n = 3) agrees with the footprint pattern observed in the 2D wide angle camera images. The full (r, Φ) image data with high temporal resolution revealed that the spatial structure of modified divertor footprints is maintained even during the edge-localized modes (ELMs) triggered by applied n = 3 fields, when the ELM size is sufficiently small, i.e. the ELMs are ‘phase locked’ to the imposed perturbation field structure. This phase-lock is lost during the ELM rise time for ELMs with large energy loss, e.g. ΔW ELM /W MHD  > 4–5%. Divertor gas puff was used to create detached divertor condition and the effect of 3D fields on the detachment was investigated. The divertor remains partially detached with the 3D field application when a sufficient amount of gas is injected into the divertor region, which is accompanied by a noticeable drop of pedestal electron temperature (T e ). However, with a lower gas puff, the divertor plasma re-attaches, when 3D fields were applied to the detached plasma, and the pedestal T e rises back up. There observed no other change in the pedestal profile associated with the re-attachment, indicating that this is likely to be dominated by a change in the electron thermal transport processes. A TRANSP analysis shows that the drop of pedestal electron heat diffusivity (χ e ) is responsible for this change but the source of this reduction is yet unclear. (paper)

  16. 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

  17. 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

  18. Edge and divertor plasma: detachment, stability, and plasma-wall interactions

    Science.gov (United States)

    Krasheninnikov, S. I.; Kukushkin, A. S.; Lee, Wonjae; Phsenov, A. A.; Smirnov, R. D.; Smolyakov, A. I.; Stepanenko, A. A.; Zhang, Yanzeng

    2017-10-01

    The paper presents an overview of the results of studies on a wide range of the edge plasma related issues. The rollover of the plasma flux to the target during progressing detachment process is shown to be caused by the increase of the impurity radiation loss and volumetric plasma recombination, whereas the ion-neutral friction, although important for establishing the necessary edge plasma conditions, does not contribute per se to the rollover of the plasma flux to the target. The processes limiting the power loss by impurity radiation are discussed and a simple estimate of this limit is obtained. Different mechanisms of meso-scale thermal instabilities driven by impurity radiation and resulting in self-sustained oscillations in the edge plasma are identified. An impact of sheared magnetic field on the dynamics of the blobs and ELM filaments playing an important role in the edge and SOL plasma transport is discussed. Trapping of He, which is an intrinsic impurity for the fusion plasmas, in the plasma-facing tungsten material is considered. A newly developed model, accounting for the generation of additional He traps caused by He bubble growth, fits all the available experimental data on the layer of nano-bubbles observed in W under irradiation by low energy He plasma.

  19. Magnetic divertors

    International Nuclear Information System (INIS)

    Keilhacker, M.

    1978-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-01

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

  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. DiMES divertor erosion experiments on DIII-D

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  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. 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).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-09-01

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

  8. 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.

  9. 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)

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

    Science.gov (United States)

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

    2008-11-01

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

  11. Design and construction of a lithium vapor box divertor similarity experiment

    Science.gov (United States)

    Schwartz, J. A.; Cohen, R. A.; Emdee, E. D.; Jaworski, M. A.; Goldston, R. J.

    2017-10-01

    Future fusion devices will require handling extreme heat fluxes. The lithium vapor box divertor is a concept to manage this heat flux. The divertor plasma impinges on a dense cloud of lithium vapor, leading to volumetric cooling, radiation, and recombination. The vapor is localized by baffles and condensation on the divertor slot walls upstream of the target, limiting the lithium reaching the main chamber. A series of test stand experiments will study vapor confinement and plasma plugging in a simplified baffled-pipe geometry. A first experiment without plasma will validate a DSMC model for evaporation, flow, and condensation of lithium vapor. Three stainless steel cylindrical cans will be heated to 550C, 600C, and 650C respectively inside a vacuum chamber. Lithium flow will be measured by weighing the cans before and after heating and by calorimetry of the latent heat of the vapor. Progress on the experiment will be presented. This work supported by DOE Contract No. DE-AC02-09CH11466.

  12. 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

  13. 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.

  14. 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

  15. 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

  16. 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

  17. Divertor erosion in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-08-01

    Net erosion rates of carbon target plates have been measured in situ for the DIII-D lower divertor. The principal method of obtaining this data is the DiMES sample probe. Recent experiments have focused on erosion at the outer strike-point (OSP) of two divertor plasma conditions: attached (T{sub e} > 40 eV) ELMing plasmas, and detached (T{sub e} < 2 eV) ELMing plasmas. For the attached cases, the erosion rates exceed 10 cm/exposure-year, even with incident heat flux < 1 MW/m{sup 2}. In this case, measurements and modeling agree for both gross and net carbon erosion, showing the near-surface transport and redeposition of the carbon is well understood. In the attached cases, physical sputtering (with enhancement from self-sputtering and oblique incidence) is dominant, and the effective sputtering yield, Y, is greater than 10%. In ELM-free discharges, the total OSP net erosion rate is equal to the rate of carbon accumulation in the core plasma. For the detached divertor cases, the cold incident plasma eliminates physical sputtering. Attempts to measure chemically eroded hydrocarbon molecules spectroscopically indicate an upper limit of Y {le} 0.1% for the chemical sputtering yield. Net erosion is suppressed at the outer strike-point, which becomes a region of net redeposition ({approximately} 4 cm/exposure-year). The private flux wall is measured to be a region of net redeposition with dense, high neutral pressure, attached divertor plasmas. Leading edges intercepting parallel heat flux ({approximately} 50 MW/m{sup 2}) have very high net erosion rates at the OSP of an attached plasma ({approximately} 10 {micro}m/s > 1,000x erosion rate of aligned surfaces). Leading edge erosion, and subsequent carbon redeposition, caused by tile gaps can account for half of the deuterium codeposition in the DIII-D divertor.

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

    International Nuclear Information System (INIS)

    Costanzo, L.

    2001-10-01

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

  19. 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.

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  1. 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

  2. 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)

  3. 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)

  4. 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

  5. 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

  6. 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

  7. 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.

  8. 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

  9. 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.

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  15. 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

  16. 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

  17. 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.

  18. 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.)

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

    Science.gov (United States)

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

    2012-10-01

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

  20. Septum assessment of the JET gas box divertor

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-09-15

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

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

    International Nuclear Information System (INIS)

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

    2014-10-01

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

  2. Charge-exchange measurements of beam ion thermalization in MHD-quiescent plasmas in the Poloidal Divertor Experiment

    International Nuclear Information System (INIS)

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

    1984-10-01

    The horizontally scanning, multiangle charge-exchange analyzer on the Poloidal Divertor Experiment (PDX) was used to study the beam ion slowing-down process with high-power perpendicular injection. Measurements were made over a wide range in toroidal field (8 kG < B(T) < 22 kG), plasma current (200 kA < I(p) < 500 kA), and beam power (1 MW < P/sub B/ < 7 MW). In MHD-quiescent plasmas, good agreement is found between the measured slowing-down spectra and theoretical predictions as a function of both angle and energy. Classes of prompt orbit losses are observed with both co- and counter-injection which have been understood and applied to plasma diagnostics. The effects of MHD activity on fast ion thermalization will be the subject of a companion paper

  3. 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.

  4. 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.

  5. 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.

  6. 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}.

  7. 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

  8. 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)

  9. 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.

  10. Simulations of material damage to divertor and first wall armour under ITER transient loads by modelling and experiments

    International Nuclear Information System (INIS)

    Bazylev, B.

    2008-01-01

    Operation of ITER at high fusion gain is assumed to be the H-mode. A characteristic feature of this regime is the transient energy release (TE) from the confined plasma onto plasma facing components (PFCs), which can play a determining role in lifetime of these components. The expected fluxes on the ITER PFCs during transients are: Type I ELM Q = 0.5 - 4 MJ/m 2 in timescales t = 0.3 - 0.6 ms, and thermal quench Q = 2 - 13 MJ/m 2 with t = 1 - 3 ms. CFC and tungsten macrobrush armour are foreseen as PFCs for ITER divertor and Be - as FW armour. During the intense TE in ITER the evaporation (CFC, W, Be) and surface melting and melt splashing (W and Be) are seen as the main mechanisms of PFC erosion. A noticeable erosion of CFC PAN fibres and rather intense crack formation for the W targets were observed in plasma gun experiments at rather small heat loads at which the melt damage to W armour is not substantial. The expected erosion of the ITER PFCs TE can be properly estimated by numerical simulations validated against erosion experiments at the plasma gun facilities QSPA-T, MK- 200UG and QSPA-Kh50. Within collaboration between EU fusion programme and Russian Federation, CFC and W macrobrush targets manufactured in EU were exposed to multiple ITER TE-like loads with Q = 0.5 - 2.2 MJ/m 2 and t = 0 .5 ms at the QSPA-T. The measured erosion was used to validate the modelling codes developed in FZK (PEGASUS, MEMOS, and others), which are then applied to model the erosion of the divertor and main chamber ITER PFCs under expected transient loads in ITER. Numerical simulations performed for the expected ITER-like loads predicted: a significant erosion of the CFC target for Q > 0.5 MJ/m 2 was caused by the inhomogeneous structure of the CFC; the W macrobrush structure is effective in preventing gross melt layer displacement. Optimization of macrobrush geometry to minimize melt splashing is done. Different mechanisms of melt splashing are compared with the results obtained in

  11. 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

  12. 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)

  13. 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)

  14. 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

  15. 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

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

    Science.gov (United States)

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

    2017-07-01

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

  17. Divertor retention for recycling impurities

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  18. Retinal Detachment Vision Simulator

    Science.gov (United States)

    ... What Is a Torn or Detached Retina? Retinal Detachment: Torn or Detached Retina Causes Retinal Detachment: Who Is At Risk for a Torn or Detached Retina? Retinal Detachment: Torn or Detached Retina Symptoms Retinal Detachment: Torn or Detached Retina Diagnosis ...

  19. 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.

  20. 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)

  1. Actively convected liquid metal divertor

    Science.gov (United States)

    Shimada, Michiya; Hirooka, Yoshi

    2014-12-01

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

  2. 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.

  3. 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.

  4. 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.

  5. Particle and impurity transport in the Axial Symmetric Divertor Experiment Upgrade and the Joint European Torus, experimental observations and theoretical understanding

    DEFF Research Database (Denmark)

    Angioni, C.; Carraro, L.; Dannert, T.

    2007-01-01

    Experimental observations on core particle and impurity transport from the Axial Symmetric Divertor Experiment Upgrade [O. Gruber, H.-S. Bosch, S. Gunter , Nucl Fusion 39, 1321 (1999)] and the Joint European Torus [J. Pamela, E. R. Solano, and JET EFDA Contributors, Nucl. Fusion 43, 1540 (2003......)] tokamaks are reviewed and compared. Robust general experimental behaviors observed in both the devices and related parametric dependences are identified. The experimental observations are compared with the most recent theoretical results in the field of core particle transport. (C) 2007 American Institute...

  6. [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.

  7. Retinal Detachment

    Science.gov (United States)

    ... specific questions. Retinal Detachment Defined What is retinal detachment? The retina is the light-sensitive layer of tissue that ... the most common. Tractional —In this type of detachment, scar tissue on the retina’s surface contracts and causes the retina to separate ...

  8. 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.))

  9. 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.

  10. 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

  11. 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

  12. Retinal Detachment: Torn or Detached Retina Symptoms

    Science.gov (United States)

    ... Health / Eye Health A-Z Detached or Torn Retina Sections Retinal Detachment: What Is a Torn or ... Detachment Vision Simulator Retinal Detachment: Torn or Detached Retina Symptoms Leer en Español: Síntomas de desgarramiento o ...

  13. Thermal stress analysis of FIRE divertor

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  14. Divertor Materials Evaluation System (DiMES)

    International Nuclear Information System (INIS)

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

    1997-11-01

    The mission of the Divertor Materials Evaluation System (DiMES) in DIII-D is to establish an integrated data base from measurements in the divertor of a tokamak in order to address some of the ITER and fusion power reactor plasma material interaction issues. Carbon and metal coatings of Be, W, V, and Mo were exposed to the steady-state outer strike point on DIII-D for 4-18 s. These short exposure times ensure controlled exposure conditions, and the extensive arrays of DIII-D divertor diagnostics provide a well-characterized plasma for modeling efforts. Postexposure analysis provides a direct measure of surface material erosion rates and the amount of retained deuterium. For carbon, these results match closely with the results of accumulated carbon deposition and erosion, and the corresponding deuterium retention of long term exposure tiles in DIII-D. Deuterium retention of different materials was measured using the 3 He(d,p) 4 He nuclear reaction. For carbon, these measurements showed peak deuterium areal density of about 8 x 10 18 D/cm 2 in a co-deposited layer about 6 microm deep, mainly at the usually detached inboard divertor leg. That layer of carbon near the inner divertor strike point has an atomic saturation concentration of D/C ∼ 0.25, which is not significantly lower than the laboratory-measured saturation retention of 0.4. Under the carbon contaminated background plasma of DIII-D, metal coatings of Be, V, Mo, and W were exposed to the steady state outer strike point under ELMing and ELM-free H-mode discharges. The rate of material erosion and tritium retention were measured. As expected, W shows the lowest erosion rate at 0.1 nm/s and the lowest deuterium uptake

  15. 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)

  16. 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)

  17. Divertor coil device

    International Nuclear Information System (INIS)

    Hanai, Satoru.

    1990-01-01

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

  18. 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

  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. Hydrogen molecules in the divertor of ASDEX Upgrade

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

    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

  2. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

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

    International Nuclear Information System (INIS)

    Cheema Rizwan A; AlKhars Wajeeha; AlAskar, 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. (author)

  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. 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.

  8. Retinal Detachment: Torn or Detached Retina Diagnosis

    Science.gov (United States)

    ... Health / Eye Health A-Z Detached or Torn Retina Sections Retinal Detachment: What Is a Torn or ... Detachment Vision Simulator Retinal Detachment: Torn or Detached Retina Diagnosis Leer en Español: Diagnóstico de un desgarramiento ...

  9. 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)

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  11. 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)

  12. 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

  13. 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

  14. Comparing scrape-off layer and divertor physics in JET pure He and D discharges

    Energy Technology Data Exchange (ETDEWEB)

    Pitts, R.A. E-mail: richard.pitts@epfl.ch; Andrew, P.; Andrew, Y.; Becoulet, M.; Coffey, I.; Coster, D.; McDonald, D.C.; Eich, T.; Erents, S.K.; Fenstermacher, M.E.; Fundamenski, W.; Haas, G.; Hermann, A.; Hidalgo, C.; Hillis, D.; Huber, A.; Ingesson, L.C.; Jachmich, S.; Kallenbach, A.; Korotkov, A.; Lawson, K.; Lomas, P.; Loarer, T.; Loarte, A.; Matthews, G.F.; McCracken, G.; Meigs, A.; Mertens, Ph.; O' Mullane, M.; Philipps, V.; Porter, G.; Pospieszczyk, A.; Rapp, J.; Reiter, D.; Riccardo, V.; Saibene, G.; Sartori, R.; Stamp, M.F.; Tsitrone, E.; Wischmeier, M.; Gafert, J

    2003-03-01

    Though helium plasmas are one option for the low activation phase of ITER, little effort has thus far been devoted to studying them in a large, diverted tokamak. A recent campaign on JET has therefore sought to address some of the important questions related to helium operation (He concentrations near 90%) in single null configurations, particularly with regard to edge and divertor physics. This contribution compiles a selection of results from these experiments, in which, in each case, discharges have been chosen to match as closely as possible previous, well characterised D plasmas in both L and ELMing H-modes. These matched pulses are used to draw conclusions regarding the principle source and location of carbon production in D plasmas, to compare and contrast the mechanisms of the density limit and the detachment process in D and He, to investigate the nature of cross-field power transport in the SOL and to gain insight into the process by which ELM energy is transported to the divertor targets.

  15. 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.

  16. 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

  17. Behaviour of the W7-AS island divertor

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

    1999-10-01

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

  19. Basic physical processes and reduced models for plasma detachment

    Science.gov (United States)

    Stangeby, P. C.

    2018-04-01

    The divertor of a tokamak reactor will have to satisfy a number of critical constraints, the first of which is that the divertor targets not fail due to excessive heating or sputter-erosion. This paramount constraint of target survival defines the operating window for the principal plasma properties at the divertor target, the density n t and temperature, T t. In particular T et level of radiative cooling in the divertor, and (b) the ion flux to the target in the presence of volumetric loss of particles, momentum and power in the divertor. The 2 Point Model, 2PM, is a widely used analytic model for relating (T t, n t) to the controlling upstream conditions. The 2PM is derived here for various levels of complexity regarding the effects included. Analytic models of divertor detachment provide valuable insight and useful approximations, but more complete modeling requires the use of edge codes such as EDGE2D, SOLPS, SONIC, UEDGE, etc. Edge codes have grown to become quite sophisticated and now constitute, in effect, ‘code-experiments’ that—just as for actual experiments—can benefit from interpretation in terms of simple conceptual frameworks. 2 Point Model Formatting, 2PMF, of edge code output can provide such a conceptual framework. Methods of applying 2PMF are illustrated here with some examples.

  20. 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

  1. Retinal detachment repair

    Science.gov (United States)

    ... area (the macula). This can help prevent further detachment of the retina. It also will increase the chance of preserving ... buckling; Vitrectomy; Pneumatic retinopexy; Laser retinopexy; Rhegmatogenous retinal detachment repair Images ... detachment repair - series References Connolly BP, Regillo ...

  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. 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

  4. 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.

  5. 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)

  6. 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)

  7. 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.)

  8. 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.)

  9. 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)

  10. Plasma-Surface Interaction Studies on DIII-D and Their Implications for Next-Step Fusion Experiments

    International Nuclear Information System (INIS)

    Whyte, D.G.

    2005-01-01

    Unique diagnostic and access features of the DIII-D tokamak, including a sample exposure system, have been used to carry out controlled and well-diagnosed plasma-surface interactions (PSI) experiments. An important contribution of the experiments has been the ability to link a given plasma exposure condition to a measured response of the plasma-facing surface and to thus understand the interaction. This has allowed for benchmarking certain aspects of erosion models, particularly near-surface particle transport. DIII-D has empirically quantified some of the PSI effects that will limit the operation availability and lifetime of future fusion devices, namely, net erosion limiting divertor plate lifetime and hydrogenic fuel retention in deposit layers. Cold divertor plasmas obtained with detachment can suppress net carbon divertor erosion, but many low-temperature divertor PSI phenomena remain poorly understood: nondivertor erosion sources, long-range particle transport, global erosion/deposition patterns, the enhancement of carbon erosion with neon impurity seeding, the sputtered carbon velocity distribution, and the apparent suppression of carbon chemical erosion in detachment. Long-term particle and energy fluences have reduced the chemical erosion yield of lower-divertor tiles. Plasma-caused modification of a material's erosion properties, including material mixing, will occur quickly and be important in long-pulse fusion devices, making prediction of PSI difficult in future devices

  11. 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.)

  12. 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)

  13. 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)

  14. 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)

  15. 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

  16. 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.)

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

    International Nuclear Information System (INIS)

    Kumagai, Akira

    1999-03-01

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

  18. 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)

  19. 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

  20. An Assessment of Molten Metal Detachment Hazards During Electron Beam Welding in the Space Shuttle Bay at LEO for the International Space Welding Experiment

    Science.gov (United States)

    Fragomeni, James M.

    1996-01-01

    In 1997, the United States [NASA] and the Paton Electric Welding Institute are scheduled to cooperate in a flight demonstration on the U.S. Space Shuttle to demonstrate the feasibility of welding in space for a possible repair option for the International Space Station Alpha. This endeavor, known as the International Space Welding Experiment (ISWE), will involve astronauts performing various welding exercises such as brazing, cutting, welding, and coating using an electron beam space welding system that was developed by the E.O. Paton Electric Welding Institute (PWI), Kiev Ukraine. This electron beam welding system known as the "Universal Weld System" consists of hand tools capable of brazing, cutting, autogeneous welding, and coating using an 8 kV (8000 volts) electron beam. The electron beam hand tools have also been developed by the Paton Welding Institute with greater capabilities than the original hand tool, including filler wire feeding, to be used with the Universal Weld System on the U.S. Space Shuttle Bay as part of ISWE. The hand tool(s) known as the Ukrainian Universal Hand [Electron Beam Welding] Tool (UHT) will be utilized for the ISWE Space Shuttle flight welding exercises to perform welding on various metal alloy samples. A total of 61 metal alloy samples, which include 304 stainless steel, Ti-6AI-4V, 2219 aluminum, and 5456 aluminum alloys, have been provided by NASA for the ISWE electron beam welding exercises using the UHT. These samples were chosen to replicate both the U.S. and Russian module materials. The ISWE requires extravehicular activity (EVA) of two astronauts to perform the space shuttle electron beam welding operations of the 61 alloy samples. This study was undertaken to determine if a hazard could exist with ISWE during the electron beam welding exercises in the Space Shuttle Bay using the Ukrainian Universal Weld System with the UHT. The safety issue has been raised with regard to molten metal detachments as a result of several

  1. 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

  2. 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

  3. Diagnostic measurements of the pumped divertor plasma

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  4. 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.

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

    Science.gov (United States)

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

    2017-08-01

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

  6. Dipole Map For Divertor Tokamaks

    International Nuclear Information System (INIS)

    Ali, Halima; Punjabi, Alkesh; Boozer, Allen

    2003-01-01

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

  7. Retinal Detachment: Torn or Detached Retina Treatment

    Science.gov (United States)

    ... in special positions for a time. Do not fly in an airplane or travel at high altitudes until you are ... With an oil bubble, it is safe to fly on an airplane. Most retinal detachment surgeries (80 to 90 percent) ...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    Science.gov (United States)

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

    2018-01-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  13. 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)

  14. 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)

  15. 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

  16. 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

  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. 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)

  19. 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

  20. Studies of power exhaust and divertor design for a 1.5 GW-level fusion power DEMO

    Science.gov (United States)

    Asakura, N.; Hoshino, K.; Suzuki, S.; Tokunaga, S.; Someya, Y.; Utoh, H.; Kudo, H.; Sakamoto, Y.; Hiwatari, R.; Tobita, K.; Shimizu, K.; Ezato, K.; Seki, Y.; Ohno, N.; Ueda, Y.; Joint Special TeamDEMO Design

    2017-12-01

    Power exhaust to the divertor and the conceptual design have been investigated for a steady-state DEMO in Japan with 1.5 GW-level fusion power and the major radius of 8.5 m, where the plasma parameters were revised appropriate for the impurity seeding scenario. A system code survey for the Ar impurity seeding suggested the volume-averaged density, impurity concentration and exhaust power from the main plasma of {{P}sep ~ }   =  205-285 MW. The divertor plasma simulation (SONIC) was performed in the divertor leg length of 1.6 m with the fixed exhaust power to the edge of {{P}out}   =  250 MW and the total radiation fraction at the edge, SOL and divertor ({{P}rad}/{{P}out}   =  0.8), as a first step to investigate appropriate design of the divertor size and geometry. At the outer target, partial detachment was produced near the strike-point, and the peak heat load ({{q}target} ) at the attached region was reduced to ~5 MW m-2 with appropriate fuel and impurity puff rates. At the inner divertor target, full detachment of ion flux was produced and the peak {{q}target} was less than 10 MW m-2 mostly due to the surface-recombination. These results showed a power exhaust scenario and the divertor design concept. An integrated design of the water-cooling heat sink for the long leg divertor was proposed. Cu-ally (CuCrZr) cooling pipe was applicable as the heat sink to handle the high heat flux near the strike-point, where displacements per atom rate was estimated to be 0.5-1.5 per year by neutronics calculation. An arrangement of the coolant rooting for Cu-alloy and Reduced Activation Ferritic Martensitic (RAFM) steel (F82H) pipes in a divertor cassette was investigated, and the heat transport analysis of the W-monoblock and Cu-alloy pipe under the peak {{q}target} of 10 MWm-2 and nuclear heating was performed. The maximum temperatures on the W-surface and Cu-alloy pipe were 1021 and 331 °C. Heat flux of 16 MW m-2 was distributed in the major part

  1. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

  3. 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

  4. 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.

  5. 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)

  6. Maximizing Heat Dissipation via Target Optimization of the Small-Angle Slot Divertor

    Science.gov (United States)

    Covele, Brent; Halpern, Federico; Casali, Livia; Canik, John; Thomas, Dan; Guo, Houyang

    2017-10-01

    The planned SAS 2 divertor uses a combination of grazing target angles and closure to direct recycling neutrals near the strike point, thus facilitating detachment onset. SAS 2 should also provide adequate pumping efficiency to be consistent with high-power steady-state scenarios on DIII-D. Initial SOLPS results indicate significantly higher neutral densities and lower electron temperatures in the SAS 2 slot, compared to a closed reference divertor model with baseline plasma profiles appropriate for high power. A systematic optimization of the parameterized SAS 2 target shape is performed in SOLPS to further reduce target heat fluxes and temperatures at lowest upstream density. To speed up the target optimization process, target neutral densities calculated by Eirene act as a performance metric by proxy for detachment facilitation. The efficacy of this proxy metric is discussed. Results are also presented from SAS 2 neutral pumping simulations in Eirene with a stationary background plasma. The feasibility of mutually satisfactory particle control and detachment control is discussed. Work supported under USDOE Cooperative Agreements DE-FC02-04ER54698 and DE-AC05-00OR22725.

  7. Recent developments towards ITER 2001 divertor maintenance

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    Science.gov (United States)

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

    2017-08-01

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

  9. 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)

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

    Science.gov (United States)

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

    2016-12-01

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

  11. Retinal Detachment in Preeclampsia

    Directory of Open Access Journals (Sweden)

    Prado Renata Silva do

    2002-01-01

    Full Text Available Preeclampsia is an obstetric disease of unknown cause that affects approximately 5% of pregnant women. The visual system may be affected with variable intensity, being the retinal detachment a rare complication. The retinal detachment in preeclampsia is usually bilateral and serous, and its pathogenesis is related to the choroidal ischemia secondary to an intense arteriolar vasospasm. The majority of patients have complete recovery of vision with clinical management, and surgery is unnecessary. This is a case report of a 27 year old patient who developed the severe form of preeclampsia on her first pregnancy. She had progressive blurred vision, until she could see only shadows. Ophthalmic examination diagnosed spread and bilateral retinal detachment. With blood pressure control at postpartum, the patient had her retina reattached, and recovery of vision.

  12. ARIES-III divertor engineering design

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  13. Safety detaching hook specification.

    CSIR Research Space (South Africa)

    Roux, JD

    1999-05-01

    Full Text Available Single-sided Impact Tests 14 Appendix C Record of Amendments 15 5. REFERENCES 17 4 1. INTRODUCTION 1.1 Subiect The subject of this specification is safety detaching hooks for mine shaft conveyances. According to South African Law1 it is a legal... after a full detachment in the catch plate or bell. Engineer The Mine Engineer, or his authorised sub-ordinate, legally appointed in terms of an Act to be responsible for shaft equipment. NDE Non-Destructive Examination. OEM Original Equipment...

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

    International Nuclear Information System (INIS)

    O'NEIL, RC; STAMBAUGH, RD

    2002-01-01

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

  15. Detachable caster adapter

    Science.gov (United States)

    Mohr, R. J.

    1969-01-01

    Detachable caster adapter moves heavy welding tables when fork lift trucks are not practical. A support saddle on the adapter, connected to the caster platform by means of a hinge, fits the leg of the welding table, but can be modified to fit other leg configurations.

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

    International Nuclear Information System (INIS)

    Ou Jing; Gan Chunyun; Ye Lei

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-01

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

  18. 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.

  19. 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.

  20. 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.

  1. Modification of SOL profiles and fluctuations with line-average density and divertor flux expansion in TCV

    Science.gov (United States)

    Vianello, N.; Tsui, C.; Theiler, C.; Allan, S.; Boedo, J.; Labit, B.; Reimerdes, H.; Verhaegh, K.; Vijvers, W. A. J.; Walkden, N.; Costea, S.; Kovacic, J.; Ionita, C.; Naulin, V.; Nielsen, A. H.; Rasmussen, J. Juul; Schneider, B.; Schrittwieser, R.; Spolaore, M.; Carralero, D.; Madsen, J.; Lipschultz, B.; Militello, F.; The TCV Team; The EUROfusion MST1 Team

    2017-11-01

    A set of Ohmic density ramp experiments addressing the role of parallel connection length in modifying scrape off layer (SOL) properties has been performed on the TCV tokamak. The parallel connection length has been modified by varying the poloidal flux expansion f x . It will be shown that this modification does not influence neither the detachment density threshold, nor the development of a flat SOL density profile which instead depends strongly on the increase of the core line average density. The modification of the SOL upstream profile, with the appearance of what is generally called a density shoulder, has been related to the properties of filamentary blobs. Blob size increases with density, without any dependence on the parallel connection length both in the near and far SOL. The increase of the density decay length, corresponding to a profile flattening, has been related to the variation of the divertor normalized collisionality Λ_div (Myra et al 2006 Phys. Plasmas 13 112502, Carralero et al, ASDEX Upgrade Team, JET Contributors and EUROfusion MST1 Team 2015 Phys. Rev. Let. 115 215002), showing that in TCV the increase of Λ_div is not sufficient to guarantee the SOL upstream profile flattening.

  2. 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.

  3. 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

  4. 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

  5. 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.

  6. 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

  7. 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.

  8. 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.)

  9. 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

  10. The roles of heritage vs thermal state of the lithosphere in the localization of detachment zones : insights from Mediterranean Core Complexes and numerical experiments.

    Science.gov (United States)

    Labrousse, L.; Huet, B.; Le Pourhiet, L.; Burov, E.; Jolivet, L.

    2012-04-01

    The most enigmatic features of metamorphic core complexes (MCC) refer to localized shallow dipping normal detachment shear zones, and preservation of almost flat Moho below the extended crust. Since the seminal work of R. Buck (1991), it is accepted that MCC form during extension of thermally relaxed hot, hence rheologically weak continental lithosphere. Initial Moho temperatures higher than 800°C are indeed predicted by many numerical models, and migmatites found in MCC cores also imply high temperature for the exhumed lower crust. A systematic review of tectonostratigraphies of the described-so-far Mediterranean MCCs shows that the detachment zones did not all develop on top of high-temperature metamorphic domes but some of them formed under much colder thermal conditions. This diversity can be described within a multi-parameter (P,T, strength) domain bound by 3 end-member cases: (1) high temperature core end-member (HT-MCC), representing most studied MCCs, and two cold end-member cases, one defined by (2) localization of crustal detachment in or on top of a preserved metasedimentary high-pressure metamorphic unit (HP-MCC), and (3) another one where the detachment is localized at the base of a high-strength upper unit, such as an obducted mafic sequence (HSU-MCC). Natural cases scatter within this triangular system, with pure HT-MCC cases (such as the Kabylian detachment, Algeria), pure HP-MCC cases (such as the Filabres detachment in the Betics, Spain), while HSU tectonostratigraphy is always coeval with a high-temperature core (eg Nigde, Anatolia) or a high-pressure nappe (in Corsica for instance). The largest core-complex systems, such as Menderes (Turkey), Rhodope (Greece and Bulgaria), and Cyclades (Greece), relate to the three end-member cases. We run thermo-mechanically coupled numerical models of extension of multi-layered lithosphere. In these models we primarily varied the rheological strength of crustal layers and initial thermal conditions to explore

  11. 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)

  12. Pneumatic retinopexy for retinal detachment associated with severe choroidal detachment.

    Science.gov (United States)

    Yeung, Ling; Kokame, Gregg T; Brod, Roy D; Lightman, David A; Lai, James C

    2011-01-01

    The purpose of this study was to evaluate the role of pneumatic retinopexy as an initial management of retinal detachment associated with hypotony, severe choroidal detachment, and vitritis. Retrospective, interventional, noncomparative case series. We included nine eyes from nine patients (six women and three men) with retinal detachment associated with hypotony, severe choroidal detachment, and vitritis managed with pneumatic retinopexy (either SF6 or C3F8) as their initial management between January 1, 1992, and December 31, 2007. Hypotony and choroidal detachment were rapidly and significantly improved 1 to 3 days after pneumatic retinopexy in all patients. The extent of retinal detachment was decreased in five patients. After vitreoretinal surgery for these five patients, all had attached retina. Complete reattachment of the retina was noted in four patients after pneumatic retinopexy. Two of these patients did not require further surgery because the entire retina remained attached at 6 months and 16 months postoperatively. Pneumatic retinopexy is a useful initial procedure in managing retinal detachment associated with hypotony, severe choroidal detachment, and vitritis. By rapidly resolving the hypotony and choroidal detachments, it facilitates subsequent surgical repair of this complicated retinal detachment. In addition, complete retinal reattachment after pneumatic retinopexy alone was initially achieved in 33% of eyes.

  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. 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.))

  15. 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.

  16. 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.

  17. 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.

  18. Between Involvement and Detachment

    DEFF Research Database (Denmark)

    Thomasen, Gry

    stood between the current ‘balanced’ Western Europe and the Europe of the pre-War period. In addition the administration held the opinion that the German problem and the Western European détente tampered with the US unilateralism in its relations with the Soviet Union, and its position as the leader....... 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...

  19. Liquid metal cooled divertor for ARIES

    International Nuclear Information System (INIS)

    Muraviev, E.

    1994-01-01

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

  20. 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

  1. 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.)

  2. 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.

  3. 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

  4. Dissociative detachment relates to psychotic symptoms and personality decompensation.

    Science.gov (United States)

    Allen, J G; Coyne, L; Console, D A

    1997-01-01

    Previous studies have addressed the prominence of psychotic symptoms in conjunction with multiple personality disorder (now dissociative identity disorder). The present study examines the relation between psychotic symptoms and a more pervasive form of dissociative disturbance, namely dissociative detachment. Two hundred sixty-six women in inpatient treatment for severe trauma-related disorders completed the Dissociative Experiences Scale (DES), and 102 of these patients also completed the Millon Clinical Multiaxial Inventory (MCMI-III). A factor analysis of the DES yielded two dimensions of dissociative detachment: detachment from one's own actions and detachment from the self and the environment. Each of these DES dimensions relates strongly to the thought disorder and schizotypal personality disorder scales of the MCMI-III. We propose that severe dissociative detachment, by virtue of loosening the moorings in inner and outer reality, is conducive to psychotic symptoms and personality decompensation.

  5. 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

  6. 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)

  7. 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.

  8. Paediatric retinal detachment: a review

    Directory of Open Access Journals (Sweden)

    Raffaele Nuzzi

    2017-10-01

    Full Text Available Paediatric retinal detachment (PRD is an uncommon and challenging disease; it differs from adult detachments in etiology, anatomical characteristics, management and prognosis. PRDs can be particularly challenging, even for the most expert paediatric surgeons due to the higher prevalence of total retinal detachments, late diagnosis and bilateral involvement with respect to those which occur in adulthood. Moreover, the anatomical success, when achieved, is frequently not related to a functional recover. Postsurgical adverse events, refractive errors and amblyopia may additionally undermine the final outcome. Up to date there are few reviews regarding the approach of retinal detachment in children, mainly dealing with rhegmatogenous retinal detachment. In this review, rhegmatogenous, retinopathy of prematurity-related and Coats’-related PRDs were considered. The available literature from the last decades were reviewed and summarized. Epidemiology, etiology and clinical presentation, together with therapeutic approaches and outcomes have been reviewed and discussed.

  9. [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.

  10. 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.

  11. 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.

  12. 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.)

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  14. 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

  15. 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)

  16. 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.

  17. 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

  18. ELM induced divertor heat loads on TCV

    Czech Academy of Sciences Publication Activity Database

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

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

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

    International Nuclear Information System (INIS)

    Li, W.X.; Song, Y.T.; Ye, M.Y.; Peng, X.B.; Wu, S.T.; Qian, X.Y.; Zhu, C.C.

    2016-01-01

    Highlights: • Type-I ELMy H-mode is one of the most severe operating environment in tokamak. • An actual time-history heat load has been used in thermo-mechanical analysis. • The analysis results are time-dependent during the whole discharge process. • The analysis could be very useful in evaluating the operational capability of the divertor. - Abstract: The lower carbon divertor has been used since 2008 in EAST, and many significant physical results, like the 410 s long pulse discharge and the 32 s H-mode operation, have been achieved. As the carbon divertor will still be used in the next few years while the injected auxiliary heating power would be increased gradually, it’s necessary to evaluate the operational capability of the carbon divertor under the heat loads during future operation. In this paper, an actual time-history heat load during type-I ELMy H-mode from EAST experiment, as one of the most severe operating environment in tokamak, has been used in the calculation and analysis. The finite element (FE) thermal and mechanical calculations have been carried out to analysis the stress and deformation of the carbon divertor during the heat loads. According to the results, the main impact on the overall temperature comes from the relative stable phase before and after the type-I ELMs and local peak load, and the transient thermal load such as type-I ELMy only has a significant effect on the surface temperature of the graphite tiles. The carbon divertor would work with high stress near the screw bolts in the current operational conditions, because of high preload and conservative frictional coefficient between the bolts and heatsink. For the future operation, new plasma facing materials (PFM) and divertor technology should be developed.

  20. 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.)

  1. Drusenoid retinal pigment epithelium detachments

    Directory of Open Access Journals (Sweden)

    Miguel Hage Amaro

    2015-10-01

    Full Text Available ABSTRACT The authors make a review of drusenoid retinal pigment epithelium detachments(DPDs, a form of retinal pigment epithelium detachment(PED that evolves from confluent and large soft drusen.Drusenoidretinal pigment epithelial detachments are a recognized element of the "dry" AMD. Until now, no treatment is indicated in drusenoid PEDs. The authors describe the clinical characteristics of drusenoid retinal pigment epithelium detachments (DPEDs and make a review of the DPEDs related in the international literature. We related in this revision paper the multimodal advanced image exams in two cases of dusenoid retinal pigment epithelium detachments (DPEDs and the general characteristics of thisfinding associated with Dry Macular degeneration.Upon examination of the ocular fundusDPEDs emerge as well-circumscribed yellow or yellow–white elevations of the RPE that are usually found within the macula.They may show scalloped borders and a slightly irregular surface. When visualized using fluorescein angiography (FA,DPEDs are typically described as faint hyper-fluorescent in the early phase followed by a slow increase in fluorescence throughout the transit stage of the study without late leakage. With optical coherence tomography (OCT, drusenoid PEDs usually show a smooth contour of the detached hyperreflective RPE band that may have an undulating appearance.Drusenoid PEDs encompass far above the ground possibility type of "dry" AMD that develops in relationship with large confluent soft drusen.At this point no treatment is utilized in drusenoid retinal pigment epithelium detachment(DPEDs.

  2. 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).

  3. Shock detachment from curved wedges

    Science.gov (United States)

    Mölder, S.

    2017-09-01

    Curved shock theory is used to show that the flow behind attached shocks on doubly curved wedges can have either positive or negative post-shock pressure gradients depending on the freestream Mach number, the wedge angle and the two wedge curvatures. Given enough wedge length, the flow near the leading edge can choke to force the shock to detach from the wedge. This local choking can preempt both the maximum deflection and the sonic criteria for shock detachment. Analytical predictions for detachment by local choking are supported by CFD results.

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

    Science.gov (United States)

    Crosatti, Lorenzo

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

  5. 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

  6. 3D Numerical Analysis of Radiative Edge Cooling in Wendelstein 7-X Island Divertor Scenarios

    Science.gov (United States)

    Effenberg, Florian; Feng, Y.; Frerichs, H.; Schmitz, O.; Barbui, T.; Geiger, J.; Jakubowski, M.; Köenig, R.; Krychowiak, M.; Niemann, H.; Sunn Pedersen, T.; Suzuki, Y.; Wurden, G. A.; W7-X-Team Team

    2017-10-01

    Radiative edge cooling is a promising method for mitigation of high heat and particle fluxes in the 3D field geometry of Wendelstein 7-X. A new high mirror island configuration is investigated featuring a more uniform distribution of heat and particle fluxes on horizontal and vertical divertor targets. For an upstream density of nup = 2 × 1019m-3 at PECRH=8MW maximum heat loads up to qmax 7.2MWm-2 are calculated with the 3D fluid and kinetic edge transport Monte Carlo Code EMC3-EIRENE. Carbon eroded from the divertor targets is predicted to serve as effective intrinsic radiator enabling detached operational regimes at higher densities (nup > 4 × 1019m-3). The feasibility of active control of heat and particle flux levels by impurity seeding (CxHy, N2, Ne) will be discussed for the new island geometry. Impurity line radiation tends to concentrate in the islands for lower densities and causes a drop of flux levels correlated to the power loss fraction, Δq Prad/PSOL . β-effects are taken into account based on the 3D MHD-equilibrium code HINT. This work was supported by the U.S. Department of Energy (DOE) under Grant DE-SC0014210.

  7. Coil Designs for Novel Magnetic Geometries to Cure the Divertor Heat Flux Problem for Reactors

    Science.gov (United States)

    Pekker, M.; Valanju, P.; Kotschenreuther, M.; Wiley, J. C.; Strickler, D.

    2004-11-01

    Coil designs are developed for novel magnetic divertor geometries with a second axi-symmetric x-point and flux expansion region along the separatrix. Adjacent posters describe how these lead to spreading of heat flux and the possibility of stable, complete detachment to overcome serious physics and engineering problems in reactors. The principal feasibility issue is creating, with simple coils, additional X-points on the separatrix without extensively deforming the magnetic field in the main plasma. For the spherical tokamak NSTX, we show that adding one or two poloidal coils suffices to create a divergent flux at the divertor, i.e., a new x-point. The currents and forces for the extra coils are small. We also modify ARIES ST design to show reactor feasibility. Optimized coil designs for PEGASUS, ARIES RS/AT, and a modular ITER retrofit are also being developed. For our calculations we used self consistent code FBEQ, which was used to design NSTX. We also use NCSX tools for optimization of designs with competing physics and engineering constraints.

  8. Ergodic divertor impact on Tore Supra plasma edge

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    Directory of Open Access Journals (Sweden)

    R.A. Pitts

    2017-08-01

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

  10. 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.

  11. 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

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

    Science.gov (United States)

    Nührenberg, C.

    2016-07-01

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

  13. 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)

  14. Development of a helium-cooled divertor concept: design-related requirements on materials and fabrication technology

    Energy Technology Data Exchange (ETDEWEB)

    Norajitra, P. E-mail: prachai.norajitra@imf.fzk.de; Boccaccini, L.V.; Diegele, E.; Filatov, V.; Gervash, A.; Giniyatulin, R.; Gordeev, S.; Heinzel, V.; Janeschitz, G.; Konys, J.; Krauss, W.; Kruessmann, R.; Malang, S.; Mazul, I.; Moeslang, A.; Petersen, C.; Reimann, G.; Rieth, M.; Rizzi, G.; Rumyantsev, M.; Ruprecht, R.; Slobodtchouk, V

    2004-08-01

    Within the framework of the EU power plant conceptual study (PPCS), a modular He-cooled divertor concept with integrated pin array (HEMP) is being developed at the Forschungszentrum Karlsruhe. The design goal is to achieve a high heat flux of at least about 10-15 MW/m{sup 2}, which is proposed for a near-term reactor model like DEMO. The development and optimization of the divertor concept require a close link between the main issues: design, analyses, materials and fabrication technology, and experiments with feedbacks between them to be accounted for. Design-specific requirements on materials and fabrication issues will be discussed.

  15. Development of a helium-cooled divertor concept: design-related requirements on materials and fabrication technology

    International Nuclear Information System (INIS)

    Norajitra, P.; Boccaccini, L.V.; Diegele, E.; Filatov, V.; Gervash, A.; Giniyatulin, R.; Gordeev, S.; Heinzel, V.; Janeschitz, G.; Konys, J.; Krauss, W.; Kruessmann, R.; Malang, S.; Mazul, I.; Moeslang, A.; Petersen, C.; Reimann, G.; Rieth, M.; Rizzi, G.; Rumyantsev, M.; Ruprecht, R.; Slobodtchouk, V.

    2004-01-01

    Within the framework of the EU power plant conceptual study (PPCS), a modular He-cooled divertor concept with integrated pin array (HEMP) is being developed at the Forschungszentrum Karlsruhe. The design goal is to achieve a high heat flux of at least about 10-15 MW/m 2 , which is proposed for a near-term reactor model like DEMO. The development and optimization of the divertor concept require a close link between the main issues: design, analyses, materials and fabrication technology, and experiments with feedbacks between them to be accounted for. Design-specific requirements on materials and fabrication issues will be discussed

  16. 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)

  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. 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

  19. 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

  20. Retinal Detachment Associated With Atopic Dermatitis.

    Science.gov (United States)

    Kothari, Nikisha; Young, Ryan C; Read, Sarah P; Tutiven, Jacqueline; Perez, Victor L; Flynn, Harry W; Berrocal, Audina M

    2017-06-01

    Ocular manifestations related to atopic dermatitis include keratoconus, keratoconjunctivitis, cataract, and retinal detachment. The authors report three cases of retinal detachment associated with atopic dermatitis. Although the pathogenesis is poorly understood, chronic blunt trauma may play a role in the development of retinal detachment. In addition, retinal detachments associated with atopic dermatitis may have lower rates of successful retinal detachment repair. [Ophthalmic Surg Lasers Imaging Retina. 2017;48:513-517.]. Copyright 2017, SLACK Incorporated.

  1. Reversal of plasma flow in tokamak divertors

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

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

    Science.gov (United States)

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

    2017-12-01

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

  4. 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)

  5. 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.)

  6. 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.

  7. The mechanics of retinal detachment

    Science.gov (United States)

    Chou, Tom; Siegel, Michael

    2013-03-01

    We present a model of the mechanical and fluid forces associated with exudative retinal detachments where the retinal photoreceptor cells separate typically from the underlying retinal pigment epithelium (RPE). By computing the total fluid volume flow arising from transretinal, vascular, and retinal pigment epithelium (RPE) pump currents, we determine the conditions under which the subretinal fluid pressure exceeds the maximum yield stress holding the retina and RPE together, giving rise to an irreversible, extended retinal delamination. We also investigate localized, blister-like retinal detachments by balancing mechanical tension in the retina with both the retina-RPE adhesion energy and the hydraulic pressure jump across the retina. For detachments induced by traction forces, we find a critical radius beyond which the blister is unstable to growth. Growth of a detached blister can also be driven by inflamed tissue within which e.g., the hydraulic conductivities of the retina or choroid increase, the RPE pumps fail, or the adhesion properties change. We determine the parameter regimes in which the blister either becomes unstable to growth, remains stable and finite-sized, or shrinks, allowing possible healing. This work supported by the Army Research Office through grant 58386MA

  8. 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)

  9. 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.

  10. 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

  11. 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

  12. 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.

  13. 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.

  14. 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.

  15. 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)

  16. 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

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

    International Nuclear Information System (INIS)

    Sugie, Tatsuo; Ogawa, Hiroaki; Ebisawa, Katsuyuki; Ando, Toshiro; Kasai, Satoshi; Katsunuma, Atsushi; Maruo, Mitsumasa; Kita, Yoshio

    1998-11-01

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

  18. Design of divertor impurity monitoring system for ITER

    International Nuclear Information System (INIS)

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

    1996-12-01

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

  19. 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.

  20. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Parks, P.B.; Brooks, N.H.; West, W.P.; Wong, C.P.C. [General Atomics, San Diego, CA (United States); Bastasz, R.; Wampler, W.R. [Sandia National Labs., Albuquerque, NM (United States); Whyte, D. [Inst. National de la Recherche Scientifique, Varennes, Quebec (Canada)

    1996-03-01

    Plasma disruptions are a serious concern in tokamak design because of the high impulsive heat loads which can cause strong erosion of divertor materials due to enhanced sputtering, or melting/ablation in the most severe cases. Predictions of net erosion rates and hence component lifetimes are very difficult and are highly dependent on the plasma conditions over the divertor target. It is therefore necessary to characterize the properties of the scrape-off plasma near the divertor target plate under these special conditions. Here, 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. An estimate of the CNL thickness {integral}nd{ell} indicates that during disruptions the CNL cushion is thick enough to reduce the local ion heat load by {approximately}30% due to CX refluxing.

  2. 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.

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

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

    1996-10-01

    We have obtained the first measurements of n e and T e in the DIII-D divertor region with a multi-pulse (20 Hz) Divertor Thomson Scattering (DTS) system. Eight measurement locations are distributed vertically up to 21 cm above the divertor plate. Two-dimensional distributions have been obtained by sweeping the divertor plasma across the DTS measurement location. Several operating modes have been studied, including ohmic, L-mode, Elming H-mode, and Radiative Divertor operation with puffing of D 2 and impurities. Mapping of the data to either the (L pol , φ) or (R, Z) planes with the EFIT equilibrium is used to analyze the 2D profiles. We find that in ELMing H-mode: n e , T e , and P e are relatively constant along field lines from the X-point to the divertor plate, especially near the separatrix field line. With D 2 puffing, the DTS profiles indicate that T e in a large part of divertor region below the X-point is dramatically reduced from ∼30-40 eV in ELMing H-mode to 1-2 eV. This results in a fairly uniform low-T e divertor, with an increased electron density in the range of 2 to 4 x 10 20 m -3 . Detailed comparisons of the spatial profiles of n e , T e , and electron pressure P e , are presented for several operating modes. In addition, these data are compared with initial calculations from the UEDGE fluid code

  5. 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.

  6. 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.)

  7. 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

  8. 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 ...

  9. 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...

  10. Mediterranean detachment zones : thermicity vs heritage.

    Science.gov (United States)

    Labrousse, Loic; Huet, Benjamin; Le Pourhiet, Laetitia; Jolivet, Laurent; Burov, Evgenii

    2017-04-01

    Even if the seminal comprehensive descriptions of Metamorphic Core Complexes (MCCs) in the American Cordillera mentionned lower plates constituted of gneiss and intruded by granites (e. g. Snake Range, Whipple Mountains), the actual definition of MCCs : « Cordilleran metamorphic core complexes appear to be bodies from the middle crust that have been dragged out from beneath fracturing and extending upper crustal rocks, and exposed beneath shallow-dipping (normal slip) faults of large areal extent » {as in Lister & Davis, 1989, Journal of Structural Geology, v. 11, pp. 65-94} refers to rocks exhumed from the middle crust whatever their thermal history. The fundamental property of this middle crust resides in its ability fo flow lateraly toward the forming dome, to accommodate stretching of the upper plate and preserve a relatively flat moho. Even though thermal reequilibration can induce weakening of the lower crust, a similar strength profile can also be inherited from pre-extension evolution of the continental crust and promote development of the original structure of MCCs : their detachment. In order to unravel the rheological meaning of detachments, we propose here a review of extensional shear zones described as detachments in the Mediterranean realm, and establish a three end-members typology with « hot MCCs » as one end-member, and two cold MCC end-members with a weak middle crust due to stacking of high pressure metapelitic nappes or a strong upper crust responsible for the strength contrast exaggeration between the upper and lower crust. New fully coupled thermo-mechanical modeling experiments together with a review of comparable published results allow to test this three end-member typology and determine the critical strength constrast for the perennial development of a detachment zone. A 1000 ratio between the strength at the brittle-ductile transition and the strength at the base of the crust seems a boundary value between localized extensional modes

  11. 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

  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. 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

  14. 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.

  15. 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.

  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. Efficiency of water coolant for DEMO divertor

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  18. Efficiency of water coolant for DEMO divertor

    International Nuclear Information System (INIS)

    Fetzer, Renate; Igitkhanov, Yuri; Bazylev, Boris

    2015-01-01

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

  19. Analysis of sweeping heat loads on divertor plate materials

    International Nuclear Information System (INIS)

    Hassanein, A.

    1991-01-01

    The heat flux on the divertor plate of a fusion reactor is probably one of the most limiting constraints on its lifetime. The current heat flux profile on the outer divertor plate of a device like ITER is highly peaked with narrow profile. The peak heat flux can be as high as 30--40 MW/m 2 with full width at half maximum (FWHM) is in the order of a few centimeters. Sweeping the separatrix along the divertor plate is one of the options proposed to reduce the thermomechanical effects of this highly peaked narrow profile distribution. The effectiveness of the sweeping process is investigated parametrically for various design values. The optimum sweeping parameters of a particular heat load will depend on the design of the divertor plate as well as on the profile of such a heat load. In general, moving a highly peaked heat load results in substantial reduction of the thermomechanical effects on the divertor plate. 3 refs., 8 figs

  20. Toroidal asymmetries in divertor impurity influxes in NSTX

    Directory of Open Access Journals (Sweden)

    F. Scotti

    2017-08-01

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  2. 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

  3. Fluorine negative ion detachment kinetics

    Science.gov (United States)

    Burke, R. R.; Miller, W. J.; Gould, R. K.

    1971-01-01

    A study of the rate of F(-) detachment by O and H atoms via the reactions F(-) + O yields FO + e and F(-) + H yields FH+ e was undertaken using a drift tube to produce F(-) ions at various drift velocities and therefore different ion temperatures. Preliminary mobility measurements of F(-) ions in Ar were made, indicating that ion temperatures in the 300 K to 5000 K range could be achieved; however due to numerous difficulties experienced in obtaining a reliable F(-) ion source, the study could not be completed.

  4. 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.

  5. Detached Solidification of Germanium-Silicon Crystals on the ISS

    Science.gov (United States)

    Volz, M. P.; Mazuruk, K.; Croell, A.

    2016-01-01

    A series of Ge(sub 1-x) Si(sub x) crystal growth experiments are planned to be conducted in the Low Gradient Furnace (LGF) onboard the International Space Station. The primary objective of the research is to determine the influence of containment on the processing-induced defects and impurity incorporation in germanium-silicon alloy crystals. A comparison will be made between crystals grown by the normal and "detached" Bridgman methods and the ground-based float zone technique. Crystals grown without being in contact with a container have superior quality to otherwise similar crystals grown in direct contact with a container, especially with respect to impurity incorporation, formation of dislocations, and residual stress in crystals. "Detached" or "dewetted" Bridgman growth is similar to regular Bridgman growth in that most of the melt is in contact with the crucible wall, but the crystal is separated from the wall by a small gap, typically of the order of 10-100 microns. Long duration reduced gravity is essential to test the proposed theory of detached growth. Detached growth requires the establishment of a meniscus between the crystal and the ampoule wall. The existence of this meniscus depends on the ratio of the strength of gravity to capillary forces. On Earth, this ratio is large and stable detached growth can only be obtained over limited conditions. Crystals grown detached on the ground exhibited superior structural quality as evidenced by measurements of etch pit density, synchrotron white beam X-ray topography and double axis X-ray diffraction.

  6. 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.

  7. 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.

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

    International Nuclear Information System (INIS)

    Marshall, T.D.; Watson, R.D.; McDonald, J.M.

    1995-01-01

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

  9. Stability of the plasma in a bundle divertor

    International Nuclear Information System (INIS)

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

    1979-02-01

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

  10. Plasma transport in a simulated magnetic-divertor configuration

    Energy Technology Data Exchange (ETDEWEB)

    Strawitch, C. M.

    1981-03-01

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  13. Radiative divertor plasmas with convection in DIII-D

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  14. Divertor plate concept with carbon based armour for NET

    International Nuclear Information System (INIS)

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

    1989-01-01

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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  18. Rhematogenous retinal detachment complicated by severe intraocular inflammation, hypotony, and choroidal detachment.

    OpenAIRE

    Jarrett, W H

    1981-01-01

    An unusual type of rhegmatogenous retinal detachment is described and compared with a control group of patients with detached retina. Features of the condition, in addition to retinal detachment, include severe anterior and posterior uveitis, choroidal detachment, hypotony, deepened anterior chamber, posterior synechiae, iridophakodonesis, and a poor surgical and visual prognosis due to massive periretinal proliferation. The disease occurs in a disproportionately high ratio in blacks. These c...

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

    Science.gov (United States)

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

    2018-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-02-01

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  2. 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

  3. 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.)

  4. 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

  5. Retinal detachment in black South Africans

    African Journals Online (AJOL)

    Rhegmatogenous retinal detachments seen in black patients attending King Edward VIII Hospital. Ophthalmology Clinic over a 5-year period from January. 1987 to December 1991 were reviewed. Penetrating trauma and diabetic retinopathy were excluded. There were 114 detachments in 112 patients, which gave.

  6. An Assessment of Molten Metal Detachment Hazards During Electron Beam Welding in Space

    Science.gov (United States)

    Fragomeni, James M.; Nunes, Arthur C., Jr.

    1998-01-01

    the detached metal drops. It was not particularly easy to generate the detachments for this experiment. This document presents the details of the theoretical modeling effort and a summary of the experimental effort to measure molten metal drop detachments from terrestrial electron beam welding in the enclosed vacuum chamber. The results of the experimental effort have shown that molten metal detachments can occur from the sample/weld plate only if a sufficiently large impact force is applied to the weld plate. A "weld pool detachment parameter" was determined to indicate whether detachment would occur. Detachment can be either full or partial (dripping), Partial detachment means that the weld pool detached from one side of the liquid-solid boundary so as to leave a hole at the puddle site but remained attached over part of the liquid-solid boundary and dripped down the plate with no fully detached material detected. Full detachment, however, does not necessarily mean that the whole pool fully detached; in some cases only a smaller portion of the pool detached, the remainder dripping down the plate. The weld pool detachment parameter according to theory and according to the empirical data allows a determination of whether full detachments might occur. Theoretical calculations indicated titanium alloy would be the most difficult from which to detach molten metal droplets followed by stainless steel and then by aluminum. The experimental results were for the most part consistent with the theoretical analysis and predictions. The above theory is applicable to other situations as desired for assessing the potential for molten metal detachments.

  7. Enhancing the DEMO divertor target by interlayer engineering

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  8. Enhancing the DEMO divertor target by interlayer engineering

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  9. Visible spectroscopy in the DIII-D divertor

    International Nuclear Information System (INIS)

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

    1996-06-01

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

  10. Visible spectroscopy in the DIII endash D divertor

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    Czech Academy of Sciences Publication Activity Database

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

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

  12. Modeling results for a linear simulator of a divertor

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-06-23

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

  13. Modeling results for a linear simulator of a divertor

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    1993-07-01

    The properties of materials for the carbon armored ITER divertor were evaluated from literature and manufacturers' documentation. Most of these data, however, have been not known or not published yet. We have evaluated an optimum data set of the candidate materials of the ITER divertor, which were needed for finite element analyses (FEM). The materials evaluated are as follows; MFC-1, CX2002U, SEP-N112, P-130, IG-430U for the carbon based materials, and Oxygen Free Copper (OFCu), Dispersion Strengthened Copper (DSCu), TZM, W5Re and W-Cu as a heat sink material. It should be noted that W-Cu is first proposed for a heat sink application of the ITER divertor plate. The finite element analyses were performed for the residual stress induced by brazing, thermal response and thermal stresses under a uniform heat flux of 15 MW/m 2 to the plasma facing surface. The stress free temperature of 750degC is assumed for the residual stress by brazing. Ten different geometries of the divertor were considered in the analyses including possible material combinations. The FEM results show that the material combinations of MFC-1 and W-30Cu or DSUc in the flat-plate geometry satisfy the presently accepted ITER requirements. The combinations of CX2002U and TZM or W5Re is considered a good choice in terms of residual and thermal stresses, whereas the surface temperature exceeds the ITER requirements. (author) 106 refs

  15. Modular He-cooled divertor for power plant application

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

  17. Manufacturing and joining technologies for helium cooled divertors

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  18. Intraretinal proliferation induced by retinal detachment

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, S.K.; Erickson, P.A.; Lewis, G.P.; Anderson, D.H. (Univ. of California, Santa Barbara (USA))

    1991-05-01

    Cellular proliferation after retinal detachment was studied by {sup 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 {sup 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.

  19. 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

  20. Testing candidate interlayers for an enhanced water-cooled divertor target

    Energy Technology Data Exchange (ETDEWEB)

    Hancock, David, E-mail: david.hancock@ccfe.ac.uk [CCFE, Culham Science Centre, Oxfordshire OX14 3DB (United Kingdom); Barrett, Tom; Foster, James; Fursdon, Mike; Keech, Gregory; McIntosh, Simon; Timmis, William [CCFE, Culham Science Centre, Oxfordshire OX14 3DB (United Kingdom); Rieth, Michael; Reiser, Jens [Karlsruhe Institute of Technology, IAM-AWP, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2015-10-15

    Highlights: • We introduce an optimised divertor target concept: the “Thermal Break”. • We suggest a candidate interlayer material for this concept: FeltMetal. • We describe a bespoke rig for testing the thermal conductivity of this material. • We present preliminary results for a number of samples. - Abstract: The design of a divertor target for DEMO remains one of the most challenging engineering tasks to be overcome on the path to fusion power. Under the European DEMO programme, a promising concept known as Thermal Break has been developed at CCFE. This concept is a variation of the ITER tungsten divertor in which the pure Copper interlayer between Copper Chrome Zirconium coolant pipe and Tungsten monoblock armour is replaced with a low thermal conductivity compliant interlayer, with the aim of reducing the thermal mismatch stress between the armour and structure. One candidate material for this interlayer is FeltMetal™ (Technetics Group, USA). This material consists of an amorphous matrix of fine copper wires which are sintered onto a thin copper foil, creating a sheet of approximately 1 mm thickness. FeltMetal has been successfully used for many years to provide compliant sliding electrical contacts for the MAST TF coils and on ALCATOR C-Mod and extensive material testing has therefore been undertaken to quantify thermal and mechanical properties. These tests, however, have not been performed under vacuum or DEMO-relevant conditions. A bespoke experimental test rig has therefore been designed and constructed with which to measure the interlayer thermal conductance as a function of temperature and pressure under vacuum conditions. The design of this apparatus and the results of experiments on FeltMetal as well as other candidate interlayers are presented here. In parallel, joint mockups using the candidate interlayers have been prepared and Thermal Break divertor target mockups have been manufactured, requiring the development of a dedicated

  1. Testing candidate interlayers for an enhanced water-cooled divertor target

    International Nuclear Information System (INIS)

    Hancock, David; Barrett, Tom; Foster, James; Fursdon, Mike; Keech, Gregory; McIntosh, Simon; Timmis, William; Rieth, Michael; Reiser, Jens

    2015-01-01

    Highlights: • We introduce an optimised divertor target concept: the “Thermal Break”. • We suggest a candidate interlayer material for this concept: FeltMetal. • We describe a bespoke rig for testing the thermal conductivity of this material. • We present preliminary results for a number of samples. - Abstract: The design of a divertor target for DEMO remains one of the most challenging engineering tasks to be overcome on the path to fusion power. Under the European DEMO programme, a promising concept known as Thermal Break has been developed at CCFE. This concept is a variation of the ITER tungsten divertor in which the pure Copper interlayer between Copper Chrome Zirconium coolant pipe and Tungsten monoblock armour is replaced with a low thermal conductivity compliant interlayer, with the aim of reducing the thermal mismatch stress between the armour and structure. One candidate material for this interlayer is FeltMetal™ (Technetics Group, USA). This material consists of an amorphous matrix of fine copper wires which are sintered onto a thin copper foil, creating a sheet of approximately 1 mm thickness. FeltMetal has been successfully used for many years to provide compliant sliding electrical contacts for the MAST TF coils and on ALCATOR C-Mod and extensive material testing has therefore been undertaken to quantify thermal and mechanical properties. These tests, however, have not been performed under vacuum or DEMO-relevant conditions. A bespoke experimental test rig has therefore been designed and constructed with which to measure the interlayer thermal conductance as a function of temperature and pressure under vacuum conditions. The design of this apparatus and the results of experiments on FeltMetal as well as other candidate interlayers are presented here. In parallel, joint mockups using the candidate interlayers have been prepared and Thermal Break divertor target mockups have been manufactured, requiring the development of a dedicated

  2. Plasma facing components integration studies for the WEST divertor

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  3. Optimization and limitations of known DEMO divertor concepts

    International Nuclear Information System (INIS)

    Reiser, Jens; Rieth, Michael

    2012-01-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  5. 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

  6. [Surgical treatment of very advanced rhegmatogenous retinal detachment].

    Science.gov (United States)

    Hejsek, L; Ernest, J; Němec, P; Rejmont, L; Manethová, K; Stepanov, A; Rozsíval, P

    2013-12-01

    indication for the procedure, or to the prognostics and therefore we can not use this technique to distinguish between operable and inoperable findings. Based on our clinical experience, we recommend carefully consider the suitability of surgical treatment of rhegmatogenous retinal detachment with low vision than the hand movement with the correct projection of light, when the duration is longer than three months and the anatomic findings is contractile anterior proliferative vitreoretinopathy.

  7. The development of divertor and first wall armour parts at JAERI, Sandia N.L. and KFA Juelich

    International Nuclear Information System (INIS)

    Akiba, M.; Bolt, H.; Watson, R.; Kneringer, G.; Linke, J.

    1991-01-01

    The development of new armour materials, and fabrication and testings of the divertor and first wall mock-ups have worldwidely been carried out during the Conceptual Design Activites (CDA) of ITER. This paper is a review of the activities on the divertor and first wall armour components which has been performed by JAERI, Sandia National Laboratory, and KFA Juelich. The design requirements have instantly been reflected in material development. For instance, carbon fiber composites (CFCs) have already been developed to have a thermal conductivity as high as copper at room temperature. Further modification of CFC's has been made. Based on the extensive progress in armour materials, the fabrication and testings of mock-ups have been started. Divertor mock-ups which are able to endure a stationary heat flux of 8 to 15 MW/m 2 have already been developed. Some new high heat flux test facilities have been constructed and are able to simulate a heat load of plasma disruption. Extensive understanding on disruption erosion of the armour materials has been obtained by experiments with these facilities. Some mock-up tests and disruption simulating tests have been performed under international collaboration. (orig.)

  8. Retinal detachment associated with traumatic chorioretinal rupture.

    Science.gov (United States)

    Papakostas, Thanos D; Yonekawa, Yoshihiro; Wu, David; Miller, John B; Veldman, Peter B; Chee, Yewlin E; Husain, Deeba; Eliott, Dean

    2014-01-01

    Traumatic chorioretinal rupture, also known as sclopetaria, is a full-thickness break of the choroid and retina caused by a high-velocity projectile striking or passing adjacent to, but not penetrating, the globe. Previous reports have emphasized that retinal detachment seldom occurs, and observation alone has been the recommended management strategy. However, the authors present herein a series of consecutive patients with retinal detachment associated with sclopetaria and provide a literature review of the topic. They recommend that patients with traumatic chorioretinal rupture be monitored closely for the development of retinal detachment during the first few weeks after the injury. Copyright 2014, SLACK Incorporated.

  9. Impact of the ITER-like wall on divertor detachment and on the density limit in the JET tokamak

    Czech Academy of Sciences Publication Activity Database

    Huber, A.; Brezinsek, S.; Groth, M.; de Vries, P.C.; Riccardo, V.; van Rooij, G.; Sergienko, G.; Arnoux, G.; Boboc, A.; Bílková, Petra; Calabrò, G.; Clever, M.; Coenen, J.W.; Beurskens, M.N.A.; Eich, E.; Jachmich, S.; Lehnen, M.; Lerche, E.; Marsen, S.; Matthews, G. F.; McCormick, K.; Meigs, A.G.; Mertens, Ph.; Philipps, V.; Rapp, J.; Samm, U.; Stamp, M.; Wischmeier, M.; Wiesen, S.

    2013-01-01

    Roč. 438, suppl (2013), S139-S147 ISSN 0022-3115. [International Conference on Plasma-Surface Interactions in Controlled Fusion Devices/20./. Aachen, 21.05.2012-25.05.2012] R&D Projects: GA ČR GAP205/10/2055; GA MŠk(CZ) LG11018 Institutional support: RVO:61389021 Keywords : plasma * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.016, year: 2013 http://www.sciencedirect.com/science/article/pii/S0022311513000305#

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-12-15

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  12. 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.

  13. MRt in the recurrent retinal detachment diagnosis after intraocular tamponade media injection: state of the art

    International Nuclear Information System (INIS)

    Manfre, L.; Midiri, M.; Casto, A.; Angileri, T.; Cardinale, A.

    1994-01-01

    Inraocular Silicon (SO), Fluorosilicon (FSO) oil or Perfluorocarbon fluid (PFCL) injection is a new succesfull surgical technique in the treatment detachment. After personal casu istic review, we report our experience in 37 patients, who underwent pars plana vitrectomy with intraocular SO, FSO or PFCL in injection for retinal detachment, monitored with Magnetic Resonance Imaging controls. MRI, showing no significant oil-related artifcats, revealed as a confident, non-invasive imaging modality in evaluating patient undergone tamponade media intraocular injection

  14. [Intraocular hypertension after retinal detachment surgery].

    Science.gov (United States)

    Muşat, O; Cristescu, R; Coman, Corina; Asandi, R

    2012-01-01

    This papers presents a case of a patient with retinal detachment, 3 days ago operated (posterior vitrectomy, internal tamponament with silicon oil 1000) who developed increased ocular pressure following silicon oil output in the anterior chamber.

  15. 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

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

    International Nuclear Information System (INIS)

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

    1999-11-01

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

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

    DEFF Research Database (Denmark)

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

    2016-01-01

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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-07-01

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

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  3. Alignment systems for pumped divertor installation at JET

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    International Nuclear Information System (INIS)

    Pestchanyi, S.

    2009-01-01

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

  6. Thermal and structural design study of divertor collector plates

    International Nuclear Information System (INIS)

    Kitamura, Kazunori; Iida, Hiromasa; Sako, Kiyoshi

    1982-01-01

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

  7. Plasma Parameters in the COMPASS Divertor During Ohmic Plasmas

    Czech Academy of Sciences Publication Activity Database

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

    2014-01-01

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

  8. Surface heat loads on the ITER divertor vertical targets

    Czech Academy of Sciences Publication Activity Database

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

    2017-01-01

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

  9. Thermal and structural analysis of the TPX divertor

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  10. Remote bolting tools for the JET divertor exchange

    International Nuclear Information System (INIS)

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

    1998-01-01

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

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

    Czech Academy of Sciences Publication Activity Database

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

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    Science.gov (United States)

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

    2010-11-01

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

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

    International Nuclear Information System (INIS)

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

    1986-07-01

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

  17. Magnum-PSI: A new plasma-wall interaction experiment

    International Nuclear Information System (INIS)

    Koppers, W.; Eck, H. van; Scholten, J.

    2006-01-01

    The FOM-Institute for Plasma Physics Rijnhuizen is preparing the construction of Magnum-PSI, a magnetized (3 T), steady-state, large area (diameter 10 cm), high-flux plasma (10 24 ions m -2 s -1 generator. The aim of the linear plasma device Magnum-PSI is to provide a controlled, highly accessible laboratory experiment in which the interaction of a magnetized plasma with different surfaces can be studied in detail. Plasma parameters can be varied over a wide range, in particular covering the high-density, low-temperature conditions expected for the detached divertor plasma of ITER. The target set-up will be extremely flexible allowing the investigation of different materials under a large variety of conditions (temperatures, inclination, biasing, coatings, etc.). A range of target materials will be used, including carbon, tungsten and other metals, and mixed materials. Because of the large plasma beam of 10 cm diameter and spacious vacuum tank, even the test of whole plasma-facing component mock-ups will be possible. Dedicated diagnostics will be installed to allow for detailed studies of the fundamental physics and chemistry of plasma-surface interaction, such as erosion and deposition, hydrogen recycling, retention and removal, dust and layer formation, plasma sheath physics and heat loads (steady-state or transient). Magnum-PSI will be a unique experiment to address the ITER divertor physics which will essentially differ from present day Tokamak and/or linear plasma generator physics. In this contribution, we will present the pre-design of the Magnum-PSI experiment. We will discuss the requirements on the vacuum system, 3T superconducting magnet, plasma source, target manipulator and additional plasma heating. In addition, we will briefly introduce the plasma and surface diagnostics that will be used in the Magnum-PSI experiment. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  1. Automated magnetic divertor design for optimal power exhaust

    International Nuclear Information System (INIS)

    Blommaert, Maarten

    2017-01-01

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

  2. Automated magnetic divertor design for optimal power exhaust

    Energy Technology Data Exchange (ETDEWEB)

    Blommaert, Maarten

    2017-07-01

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

  3. A mechanical model of retinal detachment

    International Nuclear Information System (INIS)

    Chou, Tom; Siegel, Michael

    2012-01-01

    We present a model of the mechanical and fluid forces associated with exudative retinal detachments where the retinal photoreceptor cells separate, typically from the underlying retinal pigment epithelium (RPE). By computing the total fluid volume flow arising from transretinal, vascular and RPE pump currents, we determine the conditions under which the subretinal fluid pressure exceeds the maximum yield stress holding the retina and RPE together, giving rise to an irreversible, extended retinal delamination. We also investigate localized, blister-like retinal detachments by balancing mechanical tension in the retina with both the retina–RPE adhesion energy and the hydraulic pressure jump across the retina. For detachments induced by traction forces, we find a critical radius beyond which the blister is unstable to growth. Growth of a detached blister can also be driven by inflamed lesions in which the tissue has a higher choroidal hydraulic conductivity, has insufficient RPE pump activity, or has defective adhesion bonds. We determine the parameter regimes in which the blister either becomes unstable to growth, remains stable and finite-sized, or shrinks, allowing possible healing. The corresponding stable blister radius and shape are calculated. Our analysis provides a quantitative description of the physical mechanisms involved in exudative retinal detachments and can help guide the development of retinal reattachment protocols or preventative procedures. (paper)

  4. A mechanical model of retinal detachment

    Science.gov (United States)

    Chou, Tom; Siegel, Michael

    2012-08-01

    We present a model of the mechanical and fluid forces associated with exudative retinal detachments where the retinal photoreceptor cells separate, typically from the underlying retinal pigment epithelium (RPE). By computing the total fluid volume flow arising from transretinal, vascular and RPE pump currents, we determine the conditions under which the subretinal fluid pressure exceeds the maximum yield stress holding the retina and RPE together, giving rise to an irreversible, extended retinal delamination. We also investigate localized, blister-like retinal detachments by balancing mechanical tension in the retina with both the retina-RPE adhesion energy and the hydraulic pressure jump across the retina. For detachments induced by traction forces, we find a critical radius beyond which the blister is unstable to growth. Growth of a detached blister can also be driven by inflamed lesions in which the tissue has a higher choroidal hydraulic conductivity, has insufficient RPE pump activity, or has defective adhesion bonds. We determine the parameter regimes in which the blister either becomes unstable to growth, remains stable and finite-sized, or shrinks, allowing possible healing. The corresponding stable blister radius and shape are calculated. Our analysis provides a quantitative description of the physical mechanisms involved in exudative retinal detachments and can help guide the development of retinal reattachment protocols or preventative procedures.

  5. Macroscopic erosion of divertor and first wall armour in future tokamaks

    Science.gov (United States)

    Würz, H.; Bazylev, B.; Landman, I.; Pestchanyi, S.; Safronov, V.

    2002-12-01

    Sputtering, evaporation and macroscopic erosion determine the lifetime of the 'in vessel' armour materials CFC, tungsten and beryllium presently under discussion for future tokamaks. For CFC armour macroscopic erosion means brittle destruction and dust formation whereas for metallic armour melt layer erosion by melt motion and droplet splashing. Available results on macroscopic erosion from hot plasma and e-beam simulation experiments and from tokamaks are critically evaluated and a comprehensive discussion of experimental and numerical macroscopic erosion and its extrapolation to future tokamaks is given. Shielding of divertor armour materials by their own vapor exists during plasma disruptions. The evolving plasma shield protects the armour from high heat loads, absorbs the incoming energy and reradiates it volumetrically thus reducing drastically the deposited energy. As a result, vertical target erosion by vaporization turns out to be of the order of a few microns per disruption event and macroscopic erosion becomes the dominant erosion source.

  6. The dynamic ergodic divertor in TEXTOR-A novel tool for studying magnetic perturbation field effects

    International Nuclear Information System (INIS)

    Neubauer, O.; Czymek, G.; Finken, K.H.; Giesen, B.; Huettemann, P.W.; Lambertz, H.T.; Schruff, J.

    2005-01-01

    Recently TEXTOR has been upgraded by the installation of the dynamic ergodic divertor (DED). The purpose of the DED is to influence transport parameters in plasma edge and core and to study the resulting effects on heat exhaust, edge cooling, impurity screening, plasma confinement and stability. Alternatively, the DED creates static or rotating multipolar helical magnetic perturbation fields of different mode patterns. A set of 16 helical coils has been installed on the inboard high-field side of the vacuum vessel. Rotating fields of up to 10 kHz can be generated. A novel coil design has been developed which fulfills the various mechanical, electrical, high frequency, thermal, and vacuum requirements. In addition to the various technical aspects of the DED design, implementation and commissioning, highlights of recent experiments will be presented. In particular the impact of the perturbation field on MHD stability and plasma rotation will be addressed

  7. Macroscopic erosion of divertor and first wall armour in future tokamaks

    International Nuclear Information System (INIS)

    Wuerz, H.; Bazylev, B.; Landman, I.; Pestchanyi, S.; Safronov, V.

    2002-01-01

    Sputtering, evaporation and macroscopic erosion determine the lifetime of the 'in vessel' armour materials CFC, tungsten and beryllium presently under discussion for future tokamaks. For CFC armour macroscopic erosion means brittle destruction and dust formation whereas for metallic armour melt layer erosion by melt motion and droplet splashing. Available results on macroscopic erosion from hot plasma and e-beam simulation experiments and from tokamaks are critically evaluated and a comprehensive discussion of experimental and numerical macroscopic erosion and its extrapolation to future tokamaks is given. Shielding of divertor armour materials by their own vapor exists during plasma disruptions. The evolving plasma shield protects the armour from high heat loads, absorbs the incoming energy and reradiates it volumetrically thus reducing drastically the deposited energy. As a result, vertical target erosion by vaporization turns out to be of the order of a few microns per disruption event and macroscopic erosion becomes the dominant erosion source

  8. Infrared thermography inspection methods applied to the target elements of W7-X divertor

    Energy Technology Data Exchange (ETDEWEB)

    Missirlian, M. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint Paul Lez Durance (France)], E-mail: marc.missirlian@cea.fr; Traxler, H. [PLANSEE SE, Technology Center, A-6600 Reutte (Austria); Boscary, J. [Max-Planck-Institut fuer Plasmaphysik, Euratom Association, Boltzmannstr. 2, D-85748 Garching (Germany); Durocher, A.; Escourbiac, F.; Schlosser, J. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint Paul Lez Durance (France); Schedler, B.; Schuler, P. [PLANSEE SE, Technology Center, A-6600 Reutte (Austria)

    2007-10-15

    The non-destructive examination (NDE) method is one of the key issues in developing highly loaded plasma-facing components (PFCs) for a next generation fusion devices such as W7-X and ITER. The most critical step is certainly the fabrication and the examination of the bond between the armour and the heat sink. Two inspection systems based on the infrared thermography methods, namely, the transient thermography (SATIR-CEA) and the pulsed thermography (ARGUS-PLANSEE), are being developed and have been applied to the pre-series of target elements of the W7-X divertor. Results obtained from qualification experiences performed on target elements with artificial calibrated defects allowed to demonstrate the capability of the two techniques and raised the efficiency of inspection to a level which is appropriate for industrial application.

  9. Infrared thermography inspection methods applied to the target elements of W7-X divertor

    International Nuclear Information System (INIS)

    Missirlian, M.; Traxler, H.; Boscary, J.; Durocher, A.; Escourbiac, F.; Schlosser, J.; Schedler, B.; Schuler, P.

    2007-01-01

    The non-destructive examination (NDE) method is one of the key issues in developing highly loaded plasma-facing components (PFCs) for a next generation fusion devices such as W7-X and ITER. The most critical step is certainly the fabrication and the examination of the bond between the armour and the heat sink. Two inspection systems based on the infrared thermography methods, namely, the transient thermography (SATIR-CEA) and the pulsed thermography (ARGUS-PLANSEE), are being developed and have been applied to the pre-series of target elements of the W7-X divertor. Results obtained from qualification experiences performed on target elements with artificial calibrated defects allowed to demonstrate the capability of the two techniques and raised the efficiency of inspection to a level which is appropriate for industrial application

  10. R and D and maintenance on graphite tile of divertor region at EAST

    International Nuclear Information System (INIS)

    Ji, X.; Song, Y.T.; Wu, S.T.; Hao, J.; Du, S.; Peng, Y.; Cao, L.; Wang, S.

    2012-01-01

    Highlights: ► Find out the reason of damage of graphite tile. ► Simulation the halo current. ► Stress analysis of graphite tile by ANSYS. ► Do the experiments to test the strength of graphite tile. ► Do the optimization and maintenance of graphite tile. - Abstract: EAST, with full superconducting magnetic coils, had been designed and constructed to address the scientific and engineering issues under steady state operation. The in-vessel components are full graphite tiles as first wall had been operated successfully. In the experiment campaign of 2010, the H mode operation and 1 MA operation have been gotten on EAST. However, in some case, some of the graphite tiles of divertor region are damaged with the plasma parameter enhanced. As most of the damaged graphite tiles are in the divertor region, they are probably damaged by the electro-magnetic force of the halo current when the VDEs occur. The force of the halo current is re-estimated. The structure analysis has been done by the ANSYS software. From the analysis result. It can be obtained that the stress is larger than the allowable stress when the halo current on the graphite tile is larger than 2.7 kA. The tensile testing of the graphite also has been done. As the result, the graphite tiles are damaged when the forces are up to 2400 N. To deal with the problem, two proposes are accepted. In the one hand, the new type graphite material is used, whose tensile strength is up to 45 MPa. In the other hand, the structure of the graphite tiles is optimized.

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

    NARCIS (Netherlands)

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

    2018-01-01

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  13. Spontaneous Detachment of Colloids from Primary Energy Minima by Brownian Diffusion.

    Directory of Open Access Journals (Sweden)

    Zhan Wang

    Full Text Available The Derjaguin-Landau-Verwey-Overbeek (DLVO interaction energy profile has been frequently used to interpret the mechanisms controlling colloid attachment/detachment and aggregation/disaggregation behavior. This study highlighted a type of energy profile that is characterized by a shallow primary energy well (i.e., comparable to the average kinetic energy of a colloid at a small separation distance and a monotonic decrease of interaction energy with separation distance beyond the primary energy well. This energy profile is present due to variations of height, curvature, and density of discrete physical heterogeneities on collector surfaces. The energy profile indicates that colloids can be spontaneously detached from the shallow primary energy well by Brownian diffusion. The spontaneous detachment from primary minima was unambiguously confirmed by conducting laboratory column transport experiments involving flow interruptions for two model colloids (polystyrene latex microspheres and engineered nanoparticles (fullerene C60 aggregates. Whereas the spontaneous detachment has been frequently attributed to attachment in secondary minima in the literature, our study indicates that the detached colloids could be initially attached at primary minima. Our study further suggests that the spontaneous disaggregation from primary minima is more significant than spontaneous detachment because the primary minimum depth between colloid themselves is lower than that between a colloid and a collector surface.

  14. Spontaneous Detachment of Colloids from Primary Energy Minima by Brownian Diffusion.

    Science.gov (United States)

    Wang, Zhan; Jin, Yan; Shen, Chongyang; Li, Tiantian; Huang, Yuanfang; Li, Baoguo

    2016-01-01

    The Derjaguin-Landau-Verwey-Overbeek (DLVO) interaction energy profile has been frequently used to interpret the mechanisms controlling colloid attachment/detachment and aggregation/disaggregation behavior. This study highlighted a type of energy profile that is characterized by a shallow primary energy well (i.e., comparable to the average kinetic energy of a colloid) at a small separation distance and a monotonic decrease of interaction energy with separation distance beyond the primary energy well. This energy profile is present due to variations of height, curvature, and density of discrete physical heterogeneities on collector surfaces. The energy profile indicates that colloids can be spontaneously detached from the shallow primary energy well by Brownian diffusion. The spontaneous detachment from primary minima was unambiguously confirmed by conducting laboratory column transport experiments involving flow interruptions for two model colloids (polystyrene latex microspheres) and engineered nanoparticles (fullerene C60 aggregates). Whereas the spontaneous detachment has been frequently attributed to attachment in secondary minima in the literature, our study indicates that the detached colloids could be initially attached at primary minima. Our study further suggests that the spontaneous disaggregation from primary minima is more significant than spontaneous detachment because the primary minimum depth between colloid themselves is lower than that between a colloid and a collector surface.

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    International Nuclear Information System (INIS)

    Cohen, S.A.

    1991-12-01

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

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

    Directory of Open Access Journals (Sweden)

    T. Lunt

    2017-08-01

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

  19. 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)

  20. Programmed subcellular release to study the dynamics of cell detachment

    Science.gov (United States)

    Wildt, Bridget

    Cell detachment is central to a broad range of physio-pathological changes however there are no quantitative methods to study this process. Here we report programmed subcellular release, a method for spatially and temporally controlled cellular detachment and present the first quantitative results of the detachment dynamics of 3T3 fibroblasts at the subcellular level. Programmed subcellular release is an in vitro technique designed to trigger the detachment of distinct parts of a single cell from a patterned substrate with both spatial and temporal control. Subcellular release is achieved by plating cells on an array of patterned gold electrodes created by standard microfabrication techniques. The electrodes are biochemically functionalized with an adhesion-promoting RGD peptide sequence that is attached to the gold electrode via a thiol linkage. Each electrode is electrically isolated so that a subcellular section of a single cell spanning multiple electrodes can be released independently. Upon application of a voltage pulse to a single electrode, RGD-thiol molecules on an individual electrode undergo rapid electrochemical desorption that leads to subsequent cell contraction. The dynamics of cell contraction are found to have characteristic induction and contraction times. This thesis presents the first molecular inhibition studies conducted using programmed subcellular release verifying that this technique can be used to study complex signaling pathways critical to cell motility. Molecular level dynamics of focal adhesion proteins and actin stress fibers provide some insight into the complexities associated with triggered cell detachment. In addition to subcellular release, the programmed release of alkanethiols provides a tool for to study the spatially and temporally controlled release of small molecules or particles from individually addressable gold electrodes. Here we report on experiments which determine the dynamics of programmed release using fluorophore

  1. Analytical method for determining rill detachment rate of purple soil as compared with that of loess soil

    Science.gov (United States)

    Chen, Xiao-yan; Huang, Yu-han; Zhao, Yu; Mo, Bin; Mi, Hong-xing; Huang, Chi-hua

    2017-06-01

    Rill detachment is an important process in rill erosion. The rill detachment rate is the fundamental basis for determination of the parameters of a rill erosion model. In this paper, an analytical method was proposed to estimate the rill detachment rate. The method is based on the exact analytical solution of rill erosion to the differential equation of rill detachment. The rill sediment concentration distribution as a function of rill length was identified through laboratory experiments under different slope gradients and flow rates. The sediment concentration processes from experiments on loess and purple rills were considered to estimate the rill detachment rates of both soils analytically. They were respectively used as a function of rill length and sediment concentration. The analytical detachment rates were compared with the numerically determined values to verify the analytical methods. The rill detachment rates of the two soils under different flow rates and slope gradients estimated by the analytical method were further compared on the basis of detachment-sediment function and detachment-rill length function. Results indicated that the analytically estimated values were very close to the numerically estimated values. Numerical and analytical methods were equally useful for rill detachment rate estimation. Therefore, the analytical method was verified to be rational and applicable for the rapid determination of the rill detachment rates based on either sediment concentration or rill length. The analytical detachment values of purple and loess soils suggested that the detachment rates of loess soil were significantly and considerably higher than those of purple soil. The erosion potentials of loess soil were also significant higher than those of purple soil. The differences in the erosion of the two soils decreased as the slope gradient and flow rate increased. These observations implied that the degree of loess soil erosion was greater than that of purple

  2. Surface erosion issues and analysis for dissipative divertors

    International Nuclear Information System (INIS)

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

    1994-05-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  4. Manufacture and installation of JET MKII divertor support structure

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  5. Infrared thermography inspection methods applied to the target elements of W7-X Divertor

    International Nuclear Information System (INIS)

    Missirlian, M.; Durocher, A.; Schlosser, J.; Farjon, J.-L.; Vignal, N.; Traxler, H.; Schedler, B.; Boscary, J.

    2006-01-01

    As heat exhaust capability and lifetime of plasma-facing component (PFC) during in-situ operation are linked to the manufacturing quality, a set of non-destructive testing must be operated during R-and-D and manufacturing phases. Within this framework, advanced non-destructive examination (NDE) methods are one of the key issues to achieve a high level of quality and reliability of joining techniques in the production of high heat flux components but also to develop and built successfully PFCs for a next generation of fusion devices. In this frame, two NDE infrared thermographic approaches, which have been recently applied to the qualification of CFC target elements of the W7-X divertor during the first series production will be discussed in this paper. The first one, developed by CEA (SATIR facility) and used with successfully to the control of the mass-produced actively cooled PFCs on Tore Supra, is based on the transient thermography where the testing protocol consists in inducing a thermal transient within the heat sink structure by an alternative hot/cold water flow. The second one, recently developed by PLANSEE (ARGUS facility), is based on the pulsed thermography where the component is heated externally by a single powerful flash of light. Results obtained on qualification experiences performed during the first series production of W7-X divertor components representing about thirty mock-ups with artificial and manufacturing defects, demonstrated the capabilities of these two methods and raised the efficiency of inspection to a level which is appropriate for industrial application. This comparative study, associated to a cross-checking analysis between the high heat flux performance tests and these inspection methods by infrared thermography, showed a good reproducibility and allowed to set a detectable limit specific at each method. Finally, the detectability of relevant defects showed excellent coincidence with thermal images obtained from high heat flux

  6. Overview of co-deposition and fuel inventory in castellated divertor structures at JET

    International Nuclear Information System (INIS)

    Rubel, M.J.; Coad, J.P.; Pitts, R.A.

    2007-01-01

    The main focus of this work is fuel retention in plasma components of the JET water-cooled Mk-I divertors operated with small tiles, first with carbon fibre composite (CFC) and then with castellated beryllium. Until recently these have been the only large-scale structures of this type used in fusion experiments. Three issues regarding fuel retention and material migration are addressed: (i) accumulation in gaps separating tiles and in the grooves of castellation; (ii) comparison of deposition on carbon and beryllium; (iii) in-depth migration of deuterium into the bulk of CFC. The essential results are summarised as follows: (i) co-deposition occurs up to a few cm deep in the gaps between the Mk-I tiles; (ii) fuel inventory in the CFC tile gaps exceeds that on plasma-facing surfaces by up to a factor of 2; (iii) in gaps between the beryllium tiles from the inner divertor corner the fuel content reaches 30% of that on plasma-facing surfaces, whereas in the grooves of castellation in Be the fuel content is less than 3.0% of that found on the top surface; (iv) fuel inventory on the Be tiles is strongly associated with the carbon co-deposition; (v) the D content measured in the bulk (1.5 mm below the surface) on cleaved CFC tiles exceeds 1 x 10 15 cm -2 . Implications of these results for a next-step device are addressed and the transport mechanism into the gaps is briefly discussed. The results presented here suggest that in a machine with non-carbon walls in the main chamber (as foreseen for ITER) the material transport and subsequent fuel inventory in the castellation would be reduced

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

    International Nuclear Information System (INIS)

    Federici, G; Loarte, A; Strohmayer, G

    2003-01-01

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

  8. [Retinal detachment with retinoschisis--case report].

    Science.gov (United States)

    Cristescu, R; Muşat, O; Toma, Oana; Coma, Corina; Gabej, Ioana; Burcea, M

    2013-01-01

    We present the case of a 43 year old patient diagnosed with rhegmatogenous retinal detachment and retinoschizis, a rare case of disease association. Surgery is recommended and we practice 23 gauge vitrectomy, laser retinopexy, criopexy in the periphery and internal heavy oil tamponade. Postoperatory evolution was favorable.

  9. 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

  10. Repair of Traumatic Rhegmatogenous Retinal Detachment Combined with Congenital Falciform Retinal Detachment

    Directory of Open Access Journals (Sweden)

    Fukutaro Mano

    2018-01-01

    Full Text Available Purpose: To report a case of surgical repair of traumatic rhegmatogenous retinal detachment combined with congenital falciform retinal detachment (FRD. Methods: A retrospective case report. Results: A 36-year-old man with traumatic rhegmatogenous retinal detachment complicating a previously known FRD was successfully treated despite residual FRD following pars plana lensectomy, vitrectomy, and encircling scleral buckling. His best corrected visual acuity improved from hand motion at 50 cm to 20/1,000. Conclusion: We concluded that the root of the FRD is susceptible to trauma because of the contraction of fibrovascular tissue. The early intervention of modern vitrectomy to traumatic rhegmatogenous retinal detachment complicating a previously known FRD is an important consideration for enhanced quality of care and optimal patient outcomes.

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    International Nuclear Information System (INIS)

    Brooks, J.N.

    1996-01-01

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

  14. Divertor and first wall design for TIBER-II

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  15. First demonstration of non-destructive tests on tungsten-coated JET divertor CFC tiles in the electron beam facility JUDITH-2

    International Nuclear Information System (INIS)

    Schmidt, A; Keusemann, S; Roedig, M; Pintsuk, G; Linke, J; Hirai, T; Maier, H; Riccardo, V; Matthews, G F; Hill, M; Altmann, H

    2009-01-01

    In the ITER-like wall project, the JET tokamak will employ tungsten-coated carbon fibre reinforced composite (CFCs) tiles for the outer and inner divertor tile rows. In order to assure the performance under thermal loads at the JET divertor, high heat flux (HHF) tests are performed for the qualification during the series production of the W-coated tiles, some of which might be carried out in the JUDITH-2 facility of FZJ. The electron beam (EB) facility JUDITH-2 consists of an EB-gun, a process chamber with a vacuum system (vacuum chamber extension) and different cooling circuits. This paper presents a summary of preparation steps and the first demonstration of cyclic EB loading experiments with the JUDITH-2 facility. One W-coated CFC tile was loaded at an absorbed average power density of 2.86 MW m -2 . The cyclic EB loading time was 10.2 s.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    International Nuclear Information System (INIS)

    Yoshino, Ryuji

    1988-11-01

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

  1. WIDE-FIELD INFRARED IMAGING: A Descriptive Review of Characteristics of Retinoschisis, Retinal Detachment, and Schisis Detachments.

    Science.gov (United States)

    Ho, Vincent Y; Wehmeier, Jarrod M; Shah, Gaurav K

    2016-08-01

    Retinoschisis and retinal detachments are primarily differentiated based on characteristic examination findings. In diagnostically challenging cases, noncontact wide-field infrared imaging can help diagnosis and visualize the extent/margins of retinoschisis, retinal detachment, or combined schisis detachments by comparing reflectivity patterns. This is a retrospective, observational, descriptive case series of 14 eyes of 14 nonconsecutive patients, ranging from 28 to 89 years old (mean 61), diagnosed with retinoschisis, retinal detachment, or schisis detachment from May 5, 2014 to March 4, 2015. Patients with secondary retinoschisis and/or retinal detachment from other causes were not included in the study. Heidelberg Wide-Field Module lens and Heidelberg Spectralis HRA+OCT machine (Heidelberg Engineering, Heidelberg, Germany) were used to obtain noncontact, wide-field infrared images on each study eye. Seven eyes with retinal detachments, four with retinoschises, and three with schisis detachments were imaged using this novel wide-field infrared technique. Retinoschisis appears light and translucent with prominent vasculature, retinal detachments appear dark and opaque, and combined retinoschisis/retinal detachment exhibit mixed reflectivity patterns. Wide-field infrared imaging provides a quick, noncontact, noninvasive method to accurately diagnose and to monitor for progression of retinoschisis, retinal detachment, or combined schisis detachments.

  2. Rhegmatogenous retinal detachment and conventional surgical treatment.

    Science.gov (United States)

    Golubovic, M

    2013-01-01

    The aim of the paper was to present the efficacy and indications for application of conventional surgical treatment of retinal detachment by using external implants, that is,application of encircling band and buckle. This study comprised patients from the University Eye Clinic in Skopje. A total of 33 patients were diagnosed and surgically treated in the period between May 2010 and August 2011. Conventional surgery was applied in smaller number of patients whose changes of the vitreous body were manifested by detachment of posterior hyaloid membrane, syneresis, with appearance of a small number of pigment cells in the vitreous body and synchysis, and the very retina was with fresh detachment without folds or epiretinal changes (that is, PVR A grade). There were a larger number of patients with more distinct proliferative changes of the vitreous body and of the retina, grades PVR B to C1-C2, and who also underwent the same surgical approach. Routine ophthalmologic examinations were performed, including: determination of visual acuity by Snellen's optotypes, determination of eye pressure with Schiotz's tonometer, examination of anterior segment on biomicroscopy, indirect biomicroscopy of posterior eye segment (vitreous body and retina) and examination on biomicroscopy with Goldmann prism, B scan echography of the eyes before and after surgical treatment. Conventional treatment was used by external application of buckle or application of buckle and encircling band. In case of one break, radial buckle was applied and in case of multiple breaks in one quadrant limbus parallel buckle was applied. Besides buckle, encircling band was applied in patients with total or subtotal retinal detachment with already present distinct changes in the vitreous body (PVR B or C1-C2) and degenerative changes in the vitreous body. Breaks were closed with cryopexy. The results obtained have shown that male gender was predominant and that the disease was manifested in younger male adults

  3. Identifying subsurface detachment defects by acoustic tracing

    Czech Academy of Sciences Publication Activity Database

    Sklodowski, R.; Drdácký, Miloš; Sklodowski, M.

    2013-01-01

    Roč. 56, June (2013), s. 56-64 ISSN 0963-8695 R&D Projects: GA ČR(CZ) GBP105/12/G059 Institutional support: RVO:68378297 Keywords : acoustic excitation * plaster detachment defects * frequency response * inspection systems * signal processing Subject RIV: JN - Civil Engineering Impact factor: 1.717, year: 2013 http://www.sciencedirect.com/science/article/pii/S0963869513000303

  4. Tonic shock induces detachment of Giardia lamblia.

    Directory of Open Access Journals (Sweden)

    Wendy R Hansen

    Full Text Available BACKGROUND: The parasite Giardia lamblia must remain attached to the host small intestine in order to proliferate and subsequently cause disease. However, little is known about the factors that may cause detachment in vivo, such as changes in the aqueous environment. Osmolality within the proximal small intestine can vary by nearly an order of magnitude between host fed and fasted states, while pH can vary by several orders of magnitude. Giardia cells are known to regulate their volume when exposed to changes in osmolality, but the short-timescale effects of osmolality and pH on parasite attachment are not known. METHODOLOGY AND PRINCIPAL FINDINGS: We used a closed flow chamber assay to test the effects of rapid changes in media osmolality, tonicity, and pH on Giardia attachment to both glass and C2(Bbe-1 intestinal cell monolayer surfaces. We found that Giardia detach from both surfaces in a tonicity-dependent manner, where tonicity is the effective osmolality experienced by the cell. Detachment occurs with a characteristic time constant of 25 seconds (SD = 10 sec, n = 17 in both hypo- and hypertonic media but is otherwise insensitive to physiologically relevant changes in media composition and pH. Interestingly, cells that remain attached are able to adapt to moderate changes in tonicity. By exposing cells to a timed pattern of tonicity variations and adjustment periods, we found that it is possible to maximize the tonicity change experienced by the cells, overcoming the adaptive response and resulting in extensive detachment. CONCLUSIONS AND SIGNIFICANCE: These results, conducted with human-infecting Giardia on human intestinal epithelial monolayers, highlight the ability of Giardia to adapt to the changing intestinal environment and suggest new possibilities for treatment of giardiasis by manipulation of tonicity in the intestinal lumen.

  5. The prognosis of retinal detachment due to lattice degeneration.

    Science.gov (United States)

    Benson, W E; Morse, P H

    1978-09-01

    In a series of 553 consecutive retinal detachments, 29% (120) were due to lattice degeneration. Forty-five percent of these were due to atrophic holes in the lattice degeneration and 55% were due to tears caused by traction posterior to or at the end of a patch of lattice. In phakic patients, retinal detachments due to atrophic holes were most common in young myopes. Detachments due to traction tears were seen in older, less myopic patients. The incidence of massive periretinal proliferation was less (5%) in detachments due to lattice degeneration than in detachments not due to lattice degeneration (6.5%).

  6. Fibrinogen and rhegmatogenous retinal detachment: a pilot prospective study.

    Science.gov (United States)

    Theocharis, Ip

    2010-02-18

    To examine the correlation, if any, between fibrinogen plasma levels (FPL) and the clinical features of rhegmatogenous retinal detachment (RRD). FPL were measured preoperatively in 33 patients with primary RRD. Patient characteristics and detachment features such as the numbers of breaks and the extent of the detachment were recorded; No statistically significant correlation was found between FPL and the number of breaks. A statistically significant correlation was found between FPL and the extent of the RRD, even if the influence of the number of breaks was excluded. FPL correlate with retinal detachment extent, which implicates an acute inflammatory response to detachment traumatic phenomenon or a role of the fibrinogen molecule in retinal adhesiveness.

  7. The seasonal cycle of Titan's detached haze

    Science.gov (United States)

    West, Robert A.; Seignovert, Benoît; Rannou, Pascal; Dumont, Philip; Turtle, Elizabeth P.; Perry, Jason; Roy, Mou; Ovanessian, Aida

    2018-04-01

    Titan's `detached' haze, seen in Voyager images in 1980 and 1981 and monitored by the Cassini Imaging Science Subsystem (ISS) during the period 2004-2017, provides a measure of seasonal activity in Titan's mesosphere with observations over almost half of Saturn's seasonal cycle. Here we report on retrieved haze extinction profiles that reveal a depleted layer (having a diminished aerosol content), visually manifested as a gap between the main haze and a thin, detached upper layer. Our measurements show the disappearance of the feature in 2012 and its reappearance in 2016, as well as details after the reappearance. These observations highlight the dynamical nature of the detached haze. The reappearance seems congruent with earlier descriptions by climate models but more complex than previously described. It occurs in two steps, first as haze reappearing at 450 ± 20 km and one year later at 510 ± 20 km. These observations provide additional tight and valuable constraints about the underlying mechanisms, especially for Titan's mesosphere, that control Titan's haze cycle.

  8. [Scleral buckling for inferior rhegmatogenous retinal detachments].

    Science.gov (United States)

    Abdellaoui, M; Chraibi, F; Benatiya Andaloussi, I; Tahri, H

    2014-10-01

    To investigate the epidemiological, clinical, therapeutic and prognostic factors in cases of inferior rhegmatogenous retinal detachments (RD) treated by scleral buckling surgery. A retrospective chart review was performed on 45 patients (45 eyes) with inferior RD with only inferior tears (4:00-8:00), who had been treated by scleral buckling surgery over a 6-year period from 2006 to 2011. The parameters studied included patient demographics, refractive status, time until consultation, clinical exam data, treatment modalities and functional and anatomic results. Forty-five cases were included in this study (45 eyes), with an average patient age of 44.5 years (14 to 75 years) and a slight male predominance (56%). Myopia was observed in 60%. Mean time until consultation was 3.5 months. Visual acuity on admission was less than 1/10 in 53.33%. Macular detachment was found in 80%. Causative lesions were holes in 26 eyes. Proliferative vitreoretinopathy was essentially stage B in 48.9%. Scleral buckling surgery was performed in all patients, with drainage of subretinal fluid in 37.8%. Retinal reattachment was obtained in 36 eyes (80%) with a final visual acuity greater than or equal to 1/10 in 71.11%. The mean follow-up in our study was 6.62 months. Inferior retinal detachment has a predilection for young myopes. The time until consultation is often long, and extraocular surgery, although difficult, exhibits documented efficacy. Copyright © 2014 Elsevier Masson SAS. All rights reserved.

  9. Numerical computational of fluid flow through a detached retina

    Science.gov (United States)

    Jiann, Lim Yeou; Ismail, Zuhaila; Shafie, Sharidan; Fitt, Alistair

    2015-02-01

    In this paper, a phenomenon of fluid flow through a detached retina is studied. Rhegmatogeneous retinal detachment happens when vitreous humour flow through a detached retina. The exact mechanism of Rhegmatogeneous retinal detachment is complex and remains incomplete. To understand the fluid flow, a paradigm mathematical model is developed and is approximated by the lubrication theory. The numerical results of the velocity profile and pressure distribution are computed by using Finite Element Method. The effects of fluid mechanical on the retinal detachment is discussed and analyzed. Based on the analysis, it is found that the retinal detachment deformation affects the pressure distribution. It is important to comprehend the development of the retinal detachment so that a new treatment method can be developed.

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  11. Stability and heating of a poloidal divertor tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-06-01

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

  12. Structural evaluation of a DTHR bundle divertor particle collector

    International Nuclear Information System (INIS)

    Prevenslik, T.V.

    1980-09-01

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

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

    International Nuclear Information System (INIS)

    Skinner, C.H.; Federici, G.

    2005-01-01

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

  14. Applicability of Different Hydraulic Parameters to Describe Soil Detachment in Eroding Rills

    NARCIS (Netherlands)

    Wirtz, S.; Seeger, K.M.; Zell, A.; Wagner, C.; Wagner, J.F.; Ries, J.B.

    2013-01-01

    This study presents the comparison of experimental results with assumptions used in numerical models. The aim of the field experiments is to test the linear relationship between different hydraulic parameters and soil detachment. For example correlations between shear stress, unit length shear

  15. Photo excitation and laser detachment of C60 − anions in a storage ring

    DEFF Research Database (Denmark)

    Støchkel, Kristian; Andersen, Jens Ulrik

    2013-01-01

    We have studied the photo physics of C60 − anions in the electrostatic storage ring ELISA with ions produced in a plasma source and cooled and bunched in a He filled ion trap. A previous study using delayed electron detachment as a signal of resonance-enhanced multiphoton electron detachment...... from a principal component analysis of these spectra. In good agreement with the earlier REMPED experiment, an origin band for transitions between the two lowest electronic levels of the anion, with t 1u and t 1g symmetry, is observed at 9380 cm−1, with strong sidebands from excitation of the two A g...... level, is much weaker in the new measurements and could be an H g vibrational sideband. Also earlier studies of direct laser detachment from C60 − in the storage ring ASTRID have been revisited, with ions cooled by liquid nitrogen in the ion trap. We confirm the previous measurement with a determination...

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-11-21

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

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

    International Nuclear Information System (INIS)

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

    2006-10-01

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    DEFF Research Database (Denmark)

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

    2012-01-01

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

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

    1997-06-01

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

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

    Science.gov (United States)

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

    2017-12-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-15

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  11. High-Z material erosion and its control in DIII-D carbon divertor

    Directory of Open Access Journals (Sweden)

    R. Ding

    2017-08-01

    Full Text Available As High-Z materials will likely be used as plasma-facing components (PFCs in future fusion devices, the erosion of high-Z materials is a key issue for high-power, long pulse operation. High-Z material erosion and redeposition have been studied using tungsten and molybdenum coated samples exposed in well-diagnosed DIII-D divertor plasma discharges. By coupling dedicated experiments and modelling using the 3D Monte Carlo code ERO, the roles of sheath potential and background carbon impurities in determining high-Z material erosion are identified. Different methods suggested by modelling have been investigated to control high-Z material erosion in DIII-D experiments. The erosion of Mo and W is found to be strongly suppressed by local injection of methane and deuterium gases. The 13C deposition resulting from local 13CH4 injection also provides information on radial transport due to E ×B drifts and cross field diffusion. Finally, D2 gas puffing is found to cause local plasma perturbation, suppressing W erosion because of the lower effective sputtering yield of W at lower plasma temperature and for higher carbon concentration in the mixed surface layer.

  12. Characterizing Low-Z erosion and deposition in the DIII-D divertor using aluminum

    Directory of Open Access Journals (Sweden)

    C.P. Chrobak

    2017-08-01

    Full Text Available We present measurements and modeling of aluminum erosion and redeposition experiments in separate helium and deuterium low power, low density L-mode plasmas at the outer divertor strike point of DIII-D to provide a low-Z material benchmark dataset for tokamak erosion-deposition modeling codes. Coatings of Al ∼100nm thick were applied to ideal (smooth and realistic (rough surfaces and exposed to repeat plasma discharges using the DiMES probe. Redeposition in all cases was primarily in the downstream toroidal field direction, evident from both in-situ spectroscopic and post-mortem non-spectroscopic measurements. The gross Al erosion yield was estimated from film thickness change measurements of small area samples, and was found to be ∼40–70% of the expected erosion yield based on theoretical physical sputtering yields after including sputtering by a 1–3% carbon impurity. The multi-step redeposition and re-erosion process, and hence the measured net erosion yield and material migration patterns, were found to be influenced by the surface roughness and/or porosity. A time-dependent model of material migration accounting for deposit accumulation in hidden areas was developed to reproduce the measurements in these experiments and determine a redeposition probability distribution function for sputtered atoms.

  13. Relationships between work-home segmentation and psychological detachment from work: the role of communication technology use at home.

    Science.gov (United States)

    Park, YoungAh; Fritz, Charlotte; Jex, Steve M

    2011-10-01

    Employees can have difficulty mentally distancing themselves from work during off-job time due to increasing use of communication technologies (e.g., e-mail, cell phone, etc.). However, psychological detachment from work during nonwork time is important for employee recovery and health. This study examined several antecedents of psychological detachment: work-home segmentation preference, perceived segmentation norm, and the use of communication technology at home. Results indicate that segmentation preference and segmentation norm were positively associated with psychological detachment. Further, technology use at home partially mediated these relationships. Findings indicate that segmenting work and nonwork roles can help employees detach and recover from work demands. In addition, findings show that the segmentation norm within a work group is associated with employee experiences outside of work. (PsycINFO Database Record (c) 2011 APA, all rights reserved).

  14. Simple relations between scrape-off layer parameters of high recycling divertors Part I: The relation between 'upstream' density and temperature

    International Nuclear Information System (INIS)

    Erents, S.K.; La Bombard, B.; Fundamenski, W.

    2000-01-01

    It has been found that values of n eu and T eu in the SOL of single null, high recycling divertor discharges in JET and Alcator C-Mod are related approximately as T eu ∝ n eu 1/2 , where (T eu , n eu ) are the values of plasma temperature and density on any given flux tube at the 'upstream' end, for example, halfway between the two targets. Using the standard two point model for the high recycling (conduction limited) divertor, this result is shown to be directly related to criteria based on plasma collisionality - which have to be satisfied if the flux tube is to be in the conduction limited (high recycling) regime rather than the sheath limited (low recycling) regime or the detached regime: the (T eu , n eu ) points for conduction limited flux tubes are predicted to fill a (somewhat narrow) band in the (T eu , n eu ) plane. Secondly, a tendency for the collisionality, as calculated for upstream conditions, to remain constant across the SOL is noted and an explanation is given. The latter then implies a tendency for T eu ∝ n eu 1/2 , to hold, not just at the separatrix but across the SOL. Thirdly, for the JET data a further re-enforcing factor tending to cause a rather strong T eu ∝ n eu 1/2 correlation is identified in the dependence of χ perpendicular SOL on the plasma parameters. The values of χ perpendicular SOL were extracted from the JET data using an onion skin method analysis. The first effect is expected to be general to tokamaks operated in the high recycling/conduction limited regime, although it only states that (T eu , n eu ) values will fall within a certain band, and a decreasing value of T eu with increasing neu is not ruled out. It is not known at this time how general the second two effects are. (author)

  15. FTIR measurements of OH in deformed quartz and feldspars of the South Tibetan Detachment, Greater Himalaya

    Science.gov (United States)

    Jezek, L.; Law, R. D.; Jessup, M. J.; Searle, M. P.; Kronenberg, A. K.

    2017-12-01

    OH absorption bands due to water in deformed quartz and feldspar grains of mylonites from the low-angle Lhotse Detachment (of the South Tibetan Detachment System, Rongbuk Valley north of Mount Everest) have been measured by Fourier Transform Infrared (FTIR) Spectroscopy. Previous microstructural studies have shown that these rocks deformed by dislocation creep at high temperature conditions in the middle crust (lower - middle amphibolite facies), and oxygen isotope studies suggest significant influx of meteoric water. OH absorption bands at 3400 cm-1 of quartz mylonites from the footwall of the Lhotse Detachment Fault are large, with the character of the molecular water band due to fluid inclusions in milky quartz. Mean water contents depend on structural position relative to the core of the Lhotse Detachment, from 1000 ppm (OH/106 Si) at 420 m below the fault to 11,350 (+/-1095) ppm near its center. The gradient in OH content shown by quartz grains implies influx of meteoric water along the Lhotse Detachment from the Tibetan Plateau ground surface to middle crustal depths, and significant fluid penetration into the extruding Himalayan slab by intergranular, permeable fluid flow processes. Feldspars of individual samples have comparable water contents to those of quartz and some are wetter. Large water contents of quartz and feldspar may have contributed to continued deformation and strain localization on the South Tibetan Detachment System. Dislocation creep in quartz is facilitated by water in laboratory experiments, and the water contents of the Lhotse fault rocks are similar to (and even larger than) water contents of quartz experimentally deformed during water weakening. Water contents of feldspars are comparable to those of plagioclase aggregates deformed experimentally by dislocation and diffusion creep under wet conditions.

  16. Kinematic evidence for downdip movement on the Mormon Peak detachment

    Science.gov (United States)

    Walker, Christopher D.; Anders, Mark H.; Christie-Blick, Nicholas

    2007-03-01

    The Mormon Peak detachment is considered to be one of the best examples of a rooted upper crustal detachment fault that propagated through the brittle crust at a low angle. The hanging wall of the detachment today consists of a number of isolated blocks that have been interpreted as remnants of a once-contiguous extensional allochthon. Here we present the results of a new study of directional indicators from the basal surfaces beneath these blocks. These measurements do not agree with the long-standing interpretation of a S75°W movement direction for the detachment hanging wall. Instead, the most recent movement on each section of the detachment took place approximately parallel to the present downdip direction. We conclude that the Mormon Peak detachment is best explained as the basal surfaces to a series of rootless gravity slides.

  17. Fibrinogen and rhegmatogenous retinal detachment: a pilot prospective study

    Directory of Open Access Journals (Sweden)

    IP Theocharis

    2010-02-01

    Full Text Available IP TheocharisOphthalmology Department, Iaso General Hospital, Athens, GreecePurpose: To examine the correlation, if any, between fibrinogen plasma levels (FPL and the clinical features of rhegmatogenous retinal detachment (RRD.Methods: FPL were measured preoperatively in 33 patients with primary RRD. Patient characteristics and detachment features such as the numbers of breaks and the extent of the detachment were recorded;Results: No statistically significant correlation was found between FPL and the number of breaks. A statistically significant correlation was found between FPL and the extent of the RRD, even if the influence of the number of breaks was excluded. Conclusions: FPL correlate with retinal detachment extent, which implicates an acute inflammatory response to detachment traumatic phenomenon or a role of the fibrinogen molecule in retinal adhesiveness.Keywords: fibrinogen, retinal detachment, pathogenesis

  18. Structural recovery of the detached macula after retinal detachment repair as assessed by optical coherence tomography.

    Science.gov (United States)

    Joe, Soo Geun; Kim, Yoon Jeon; Chae, Ju Byung; Yang, Sung Jae; Lee, Joo Yong; Kim, June-Gone; Yoon, Young Hee

    2013-06-01

    To investigate correlations between preoperative and postoperative foveal microstructures in patients with macula-off rhegmatogenous retinal detachment (RRD). We reviewed the records of 31 eyes from 31 patients with macula-off RRD who had undergone successful re-attachment surgery. We analyzed data obtained from complete ophthalmologic examinations and optical coherence tomography (OCT) before and 9 to 12 months after surgery. All postoperative OCT measurements were taken with spectral-domain OCT, but a subset of preoperative OCT measurements were taken with time-domain OCT. The mean duration of macular detachment was 15.5 ± 15.2 days, and mean preoperative best-corrected visual acuity (BCVA, logarithm of the minimum angle of resolution) was 1.03 ± 0.68. Preoperative visual acuity was correlated with retinal detachment height (p macula-off duration. The final BCVA was significantly correlated with integrity of the junction between the photoreceptor inner and outer segments (IS/OS) combined with the continuity of external limiting membrane (ELM) (p = 0.025). The presence of IRS and OLU on a detached macula were highly correlated with the final postoperative integrity of the IS/OS junction and the ELM (p = 0.017). Eyes preoperatively exhibiting IRS and OLU showed a higher incidence of disruption to the photoreceptor IS/OS junction and the ELM at final follow-up. Such a close correlation between preoperative and postoperative structural changes may explain why ultimate visual recovery in such eyes is poor.

  19. Detection of detachments and inhomogeneities in frescos by Compton scattering

    International Nuclear Information System (INIS)

    Castellano, A.; Cesareo, R.; Buccolieri, G.; Donativi, M.; Palama, F.; Quarta, S.; De Nunzio, G.; Brunetti, A.; Marabelli, M.; Santamaria, U.

    2005-01-01

    A mobile instrument has been developed for the detection and mapping of detachments in frescos by using Compton back scattered photons. The instrument is mainly composed of a high energy X-ray tube, an X-ray detection system and a translation table. The instrument was first applied to samples simulating various detachment situations, and then transferred to the Vatican Museum to detect detachments and inhomogeneities in the stanza di Eliodoro, one of the 'Raphael's stanze'

  20. Detection of detachments and inhomogeneities in frescos by Compton scattering

    Energy Technology Data Exchange (ETDEWEB)

    Castellano, A. [Dipartimento di Scienza dei Materiali, Universita di Lecce, 73100 Lecce (Italy); INFN, Sezione di Lecce, via per Arnesano, 73100 Lecce (Italy); Cesareo, R. [Istituto di Matematica e Fisica, Universita di Sassari, 07100 Sassari (Italy) and INFN, Sezione di Cagliari, Cittadella Universitaria di Monserrato, 09042 Cagliari (Italy)]. E-mail: cesareo@uniss.it; Buccolieri, G. [Dipartimento di Scienza dei Materiali, Universita di Lecce, 73100 Lecce (Italy); INFN, Sezione di Lecce, via per Arnesano, 73100 Lecce (Italy); Donativi, M. [Dipartimento di Scienza dei Materiali, Universita di Lecce, 73100 Lecce (Italy); Palama, F. [Dipartimento di Scienza dei Materiali, Universita di Lecce, 73100 Lecce (Italy); INFN, Sezione di Lecce, via per Arnesano, 73100 Lecce (Italy); Quarta, S. [Dipartimento di Scienza dei Materiali, Universita di Lecce, 73100 Lecce (Italy); INFN, Sezione di Lecce, via per Arnesano, 73100 Lecce (Italy); De Nunzio, G. [Dipartimento di Scienza dei Materiali, Universita di Lecce, 73100 Lecce (Italy); INFN, Sezione di Lecce, via per Arnesano, 73100 Lecce (Italy); Brunetti, A. [Istituto di Matematica e Fisica, Universita di Sassari, 07100 Sassari (Italy); INFN, Sezione di Cagliari, Cittadella Universitaria di Monserrato, 09042 Cagliari (Italy); Marabelli, M. [Istituto Centrale del Restauro, P.zza S. Francesco di Paola, 00184 Rome (Italy); Santamaria, U. [Laboratori dei Musei Vaticani, Citta del Vaticano, Rome (Italy)

    2005-07-01

    A mobile instrument has been developed for the detection and mapping of detachments in frescos by using Compton back scattered photons. The instrument is mainly composed of a high energy X-ray tube, an X-ray detection system and a translation table. The instrument was first applied to samples simulating various detachment situations, and then transferred to the Vatican Museum to detect detachments and inhomogeneities in the stanza di Eliodoro, one of the 'Raphael's stanze'.

  1. Meniscus Shapes in Detached Bridgman Growth

    Science.gov (United States)

    Volz, M. P.; Mazuruk, K

    2010-01-01

    In detached Bridgman crystal growth, most of the melt is in contact with the ampoule wall, but the crystal is separated from the wall by a small gap, typically 1-100 micrometers. A liquid free surface, or meniscus, bridges across this gap at the position of the melt-crystal interface. Meniscus shapes have been calculated for the case of detached Bridgman growth in cylindrical ampoules by solving the Young-Laplace equation. Key parameters affecting meniscus shapes are the growth angle, contact angle of the meniscus to the ampoule wall, the pressure differential across the meniscus, and the Bond number, a measure of the ratio of gravitational to capillary forces. In general, for specified values of growth and contact angles, solutions exist only over a finite range of pressure differentials. For intermediate values of the Bond number, there are multiple solutions to the Young-Laplace equations. There are also cases where, as a function of pressure differential, existence intervals alternate with intervals where no solutions exist. The implications of the meniscus shape calculations on meniscus stability are discussed.

  2. Extent of Detached Retina and Lens Status Influence Intravitreal Protein Expression in Rhegmatogenous Retinal Detachment.

    Science.gov (United States)

    Pollreisz, Andreas; Sacu, Stefan; Eibenberger, Katharina; Funk, Marion; Kivaranovic, Danijel; Zlabinger, Gerhard J; Georgopoulos, Michael; Schmidt-Erfurth, Ursula

    2015-08-01

    The aim of the study was to compare intravitreal cytokines and chemokines to clinical parameters in patients with rhegmatogenous retinal detachment (RRD). In this prospective study vitreous samples were taken undiluted from 60 patients with RRD and 20 age-matched controls with idiopathic epiretinal membranes at the beginning of primary vitrectomy. The following clinical parameters were assessed from RRD patients prior to surgery: number of quadrants detached, RD height, lens status, symptom duration, and refractive power. Concentrations of 40 different proteins in the vitreous of RRD eyes were measured by multiplex protein array, compared with controls and correlated to clinical parameters. Ten cytokines and chemokines were significantly upregulated in the vitreous of RRD eyes compared with controls (tissue inhibitors of metalloproteinases [TIMP]-1 and -2, macrophage inflammatory protein [MIP]-1α, monocyte chemoattractant protein [MCP]-1, IL-6, and -8, inducible protein (IP)-10, brain-derived neurotrophic factor [BDNF], TGFβ-3, and platelet-derived growth factor [PDGF]-AB/BB). Linear regression analysis revealed that IL-8 and TGFβ-3 increased with the number of retinal quadrants detached, while TIMP-1 rose in eyes with greater RD heights. Concentrations of IP-10 and myeloperoxidase (MPO) peaked in eyes with two or more quadrants detached, while TIMP-2 was highest expressed in the vitreous of eyes with great RD height. In pseudophakic eyes with higher detachment height levels of vascular cell adhesion molecule (VCAM)-1 were significantly increased, while neural cell adhesion molecule (NCAM) was decreased in pseudophakic patients with shallow RD height. Extent of RRD and lens status significantly influence intravitreal proinflammatory, profibrotic, and proapoptotic protein expression. These data contribute to the fundamental understanding of pathophysiological mechanisms in RRD and may serve as a basis for development of adjunct therapeutics to facilitate

  3. Risk of retinal detachment in patients with lattice degeneration.

    Science.gov (United States)

    Sasaki, K; Ideta, H; Yonemoto, J; Tanaka, S; Hirose, A; Oka, C

    1998-01-01

    To determine the risk of retinal detachment in patients with lattice degeneration of the retina, we statistically analyzed the incidence of retinal detachment in these patients. The data of hospital patients with retinal detachment associated with lattice degeneration in Kumamoto Prefecture, Japan, in 1990 were collected. The prevalence of lattice degeneration in Kumamoto was reported to be 9.5% in 1980. Based on population data from the 1990 census, the cumulative incidence of retinal detachment associated with lattice degeneration was calculated in this study. Among 1,840,000 residents in Kumamoto, there were 110 patients with retinal detachment associated with lattice degeneration; 72 with detachment resulting from tractional tears (tears), and 38 with detachment from atrophic holes. The cumulative incidence of retinal detachment from atrophic holes was 1.5% at the age of 40 years; from tears it was 3.6% at the age of 80 years. The cumulative incidence of detachment from both atrophic holes and tears was 5.3% at the age of 80 years. The results of this study are useful for clarifying the natural course of lattice degeneration.

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  5. Development, thermomechanical analysis, fabrication and testing of a macro-block-divertor specimens

    International Nuclear Information System (INIS)

    Scheerer, M.; Bolt, H.; Linke, J.; Lison, R.; Gervash, A.; Smid, I.

    1998-01-01

    For a divertor element with internal coolant flow reversal Carbon Fibre reinforced Carbon brazed on a Cu-base or refractory metal cooling tube was selected. In order to choose the optimum geometry and material combination, two criteria in the thermomechanical FEM analysis were used: the surface temperature under a thermal load of 10 MW/m 2 and the residual stress in the CFC-material after brazing. Based on the results of the FEM calculations test specimens in macro-block-geometry made of a 2D-CFC material brazed to an OFCu tube were manufactured using a CuTi braze. In order to study the influence of the brazing parameters, the size of the braze gap and the formation of the CuTi-phase were varied. To evaluate the thermal behaviour, heat load experiments up to 10 MW/m 2 for pulse durations of 5 s, 10 s and 15 s were applied in an electron beam facility. Before and after brazing all Cu-CFC interfaces were non-destructively tested using a high resolution ultrasonic inspection system. A very good agreement between thermography, ultrasonic inspection and metallography of the braze joints was observed. Due to these results new larger test specimens were produced using a smaller brazing gap and the vacuum plasma deposition of Ti for the formation of the braze. (author)

  6. Calculation of the radial electric field with RF sheath boundary conditions in divertor geometry

    Science.gov (United States)

    Gui, B.; Xia, T. Y.; Xu, X. Q.; Myra, J. R.; Xiao, X. T.

    2018-02-01

    The equilibrium electric field that results from an imposed DC bias potential, such as that driven by a radio frequency (RF) sheath, is calculated using a new minimal two-field model in the BOUT++ framework. Biasing, using an RF-modified sheath boundary condition, is applied to an axisymmetric limiter, and a thermal sheath boundary is applied to the divertor plates. The penetration of the bias potential into the plasma is studied with a minimal self-consistent model that includes the physics of vorticity (charge balance), ion polarization currents, force balance with E× B , ion diamagnetic flow (ion pressure gradient) and parallel electron charge loss to the thermal and biased sheaths. It is found that a positive radial electric field forms in the scrape-off layer and it smoothly connects across the separatrix to the force-balanced radial electric field in the closed flux surface region. The results are in qualitative agreement with the experiments. Plasma convection related to the E× B net flow in front of the limiter is also obtained from the calculation.

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  8. Local magnetic divertor for control of the plasma--limiter interaction in a tokamak

    International Nuclear Information System (INIS)

    Zweben, S.J.; Liewer, P.C.; Gould, R.W.

    1984-01-01

    An experiment is described in which plasma flow to a tokamak limiter is controlled through the use of a local toroidal divertor coil mounted inside the limiter itself. This coil produces a local perturbed field B/sub C/ approximately equal to the local unperturbed toroidal field B/sub T/approx. =3 kG, such that when B/sub C/ adds to B/sub T/ the field lines move into the limiter and the local plasma flow to it increases by a factor as great as 1.6, and when B/sub C/ subtracts from B/sub T/ the field lines move away from the limiter and the local plasma flow to it decreases by as much as a factor of 4. A simple theoretical model is used to interpret these results. Since these changes occur without significantly affecting global plasma confinement, such a control scheme may be useful for optimizing the performance of pumped limiters

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  10. Diagnostics for Evaluating Performance of NSTX Liquid Lihium Divertor

    Science.gov (United States)

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

    2009-11-01

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

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  12. Technologies for ITER divertor vertical target plasma facing components

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  13. DITE Mk2 bundle divertor power supplies and control system

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  14. Advances in optical thermometry for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-01-15

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  16. Effect of stationary high heat flux and transient ELMs-like heat loads on the divertor PFCs

    Energy Technology Data Exchange (ETDEWEB)

    Riccardi, B., E-mail: bruno.riccardi@f4e.europa.eu [Fusion for Energy, ITER Department, Josep Pla, 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Gavila, P. [Fusion for Energy, ITER Department, Josep Pla, 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Giniatulin, R. [Efremov Institute, 196641 St. Petersburg (Russian Federation); Kuznetsov, V. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, 142190 Troitsk, Moscow Region (Russian Federation); Rulev, R. [Efremov Institute, 196641 St. Petersburg (Russian Federation); Klimov, N.; Kovalenko, D.; Barsuk, V. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, 142190 Troitsk, Moscow Region (Russian Federation); Koidan, V.; Korshunov, S. [NRC “Kurchatov Institute”, Moscow (Russian Federation)

    2013-10-15

    The experimental evaluation of the divertor plasma facing components (PFCs) lifetime under transient events, such as edge localized modes (ELMs) and high heat flux (HHF) thermal fatigue expected during ITER normal operations and slow transient events is here presented. The experiments have been performed in the frame of an EU/RF collaboration. For carbon fiber composite material the erosion is caused by PAN fiber damage whilst the erosion of tungsten is determined by the melt layer movement and crack formation. The conclusion of this study is that, in addition to the structural change produced in the armor materials by ELMs-like loads, some mock ups showed also a degradation of the thermal fatigue performances.

  17. Hydrogen concentration of co-deposited carbon films produced in the vicinity of local island divertor in Large Helical Device

    International Nuclear Information System (INIS)

    Hino, T.; Hirata, T.; Ashikawa, N.

    2008-10-01

    It is quite important to evaluate hydrogen concentration of co-deposited carbon film/dust to estimate in-vessel tritium inventory in ITER. The co-deposited carbon films were prepared at the wall of pumping duct in Local Island Divertor experiments of LHD. The hydrogen concentration of the co-deposited carbon film at the wall not facing to the plasma with a low temperature was extremely high, 1.3 in the atomic ratio of H/C. This value is triple times higher than the previous value obtained by hydrogen ion irradiation to graphite. The crystal structure of the co-deposited carbon film observed by Raman spectroscopy showed very unique structure (polymeric a-C:H), which is well consistent with the high hydrogen concentration. The accumulation of in-vessel tritium inventory is also discussed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    Directory of Open Access Journals (Sweden)

    Jeong-Ha You

    2015-12-01

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  2. Effects of Crop Canopies on Rain Splash Detachment

    Science.gov (United States)

    Ma, Bo; Yu, Xiaoling; Ma, Fan; Li, Zhanbin; Wu, Faqi

    2014-01-01

    Crops are one of the main factors affecting soil erosion in sloping fields. To determine the characteristics of splash erosion under crop canopies, corn, soybean, millet, and winter wheat were collected, and the relationship among splash erosion, rainfall intensity, and throughfall intensity under different crop canopies was analyzed through artificial rainfall experiments. The results showed that, the mean splash detachment rate on the ground surface was 390.12 g/m2·h, which was lower by 67.81% than that on bare land. The inhibiting effects of crops on splash erosion increased as the crops grew, and the ability of the four crops to inhibit splash erosion was in the order of winter wheat>corn>soybeans>millet. An increase in rainfall intensity could significantly enhance the occurrence of splash erosion, but the ability of crops to inhibit splash erosion was 13% greater in cases of higher rainfall intensity. The throughfall intensity under crop canopies was positively related to the splash detachment rate, and this relationship was more significant when the rainfall intensity was 40 mm/h. Splash erosion tended to occur intensively in the central row of croplands as the crop grew, and the non-uniformity of splash erosion was substantial, with splash erosion occurring mainly between the rows and in the region directly under the leaf margin. This study has provided a theoretical basis for describing the erosion mechanisms of cropland and for assisting soil erosion prediction as well as irrigation and fertilizer management in cultivated fields. PMID:24992386

  3. Functional Detachment of Totalitarian Nazi Architecture

    Science.gov (United States)

    Antoszczyszyn, Marek

    2017-10-01

    The paper describes the systematization process of architectural styles in use during Nazi period in Germany between 1933-45. In the results of the research some regularity about strict concern between function & styling has been observed. Using comparison & case study as well as analytical methods there were pointed out characteristic features of more than 500 objects’ architectural appearance that helped to specify their styling & group them into architectural trends. Ultimately the paper proves that the found trends of architectural styling could be collected by functional detachment key. This observation explains easy to recognize even nowadays traceability – so characteristic to Nazi German architecture. Facing today pluralism in architecture, the findings could be a helpful key in the organization of spatial architectural identification process.

  4. Maritime Training Serbian Autonomous Vessel Protection Detachment

    Directory of Open Access Journals (Sweden)

    Šoškić Svetislav D.

    2014-06-01

    Full Text Available The crisis in Somalia has caused appearance of piracy at sea in the Gulf of Aden and the Western Indian Ocean. Somali pirates have become a threat to economic security of the world because almost 30 percent of world oil and 20 percent of global trade passes through the Gulf of Aden. Solving the problem of piracy in this part of the world have included international organizations, institutions, military alliances and the states, acting in accordance with international law and UN Security Council resolutions. The European Union will demonstrate the application of a comprehensive approach to solving the problem of piracy at sea and the crisis in Somalia conducting naval operation — EU NAVFOR Atalanta and operation EUTM under the Common Security and Defense Policy. The paper discusses approaches to solving the problem of piracy in the Gulf of Aden and the crisis in Somalia. Also, the paper points to the complexity of the crisis in Somalia and dilemmas correctness principles that are applied to solve the problem piracy at sea. One of goals is protections of vessels of the World Food Programme (WFP delivering food aid to displaced persons in Somalia. Republic of Serbia joined in this mission and trained and sent one a autonomous team in this military operation for protection WFP. This paper consist the problem of modern piracy, particularly in the area of the Horn of Africa became a real threat for the safety of maritime ships and educational process of Serbian Autonomous vessel protection detachment. Serbian Military Academy adopted and developed educational a training program against piracy applying all the provisions and recommendations of the IMO conventions and IMO model courses for Serbian Autonomous vessel protection detachment.

  5. Missed retinal breaks in rhegmatogenous retinal detachment

    Directory of Open Access Journals (Sweden)

    Brijesh Takkar

    2016-12-01

    Full Text Available AIM: To evaluate the causes and associations of missed retinal breaks (MRBs and posterior vitreous detachment (PVD in patients with rhegmatogenous retinal detachment (RRD. METHODS: Case sheets of patients undergoing vitreo retinal surgery for RRD at a tertiary eye care centre were evaluated retrospectively. Out of the 378 records screened, 253 were included for analysis of MRBs and 191 patients were included for analysis of PVD, depending on the inclusion criteria. Features of RRD and retinal breaks noted on examination were compared to the status of MRBs and PVD detected during surgery for possible associations. RESULTS: Overall, 27% patients had MRBs. Retinal holes were commonly missed in patients with lattice degeneration while missed retinal tears were associated with presence of complete PVD. Patients operated for cataract surgery were significantly associated with MRBs (P=0.033 with the odds of missing a retinal break being 1.91 as compared to patients with natural lens. Advanced proliferative vitreo retinopathy (PVR and retinal bullae were the most common reasons for missing a retinal break during examination. PVD was present in 52% of the cases and was wrongly assessed in 16%. Retinal bullae, pseudophakia/aphakia, myopia, and horse shoe retinal tears were strongly associated with presence of PVD. Traumatic RRDs were rarely associated with PVD. CONCLUSION: Pseudophakic patients, and patients with retinal bullae or advanced PVR should be carefully screened for MRBs. Though Weiss ring is a good indicator of PVD, it may still be over diagnosed in some cases. PVD is associated with retinal bullae and pseudophakia, and inversely with traumatic RRD.

  6. Direct observation of bacterial deposition on and detachment from nanocomposite membranes embedded with silver nanoparticles.

    Science.gov (United States)

    Liu, Yaolin; Rosenfield, Eric; Hu, Meng; Mi, Baoxia

    2013-06-01

    A microscope-equipped online monitoring system was used to investigate the bacterial deposition and detachment kinetics of a nanocomposite membrane that was synthesized by embedding silver nanoparticles in a polysulfone membrane. A pure polysulfone membrane was used as a control in the experiments. The deposition experiments with live bacteria showed that the bacterial deposition rates were the same for the nanocomposite and control polysulfone membranes. After the rinsing experiments, however, on average a high bacterial detachment ratio of 75% was observed for the nanocomposite membrane, compared with 18% for the control polysulfone membrane. These results indicate that the presence of silver nanoparticles as an antibacterial agent enhances the antiadhesive property of the nanocomposite membrane by decreasing the capability of bacteria in permanently attaching to the membrane surface. A quartz crystal microbalance with dissipation was used to study silver leaching. It was found that silver leaching significantly decreased within the first few hours. Deposition and rinsing experiments with dead bacterial cells revealed that the dead cell deposition rates were similar for both membranes, and so were the detachment ratios. Since the nanocomposite membrane did not show any antiadhesive action against dead cells, its antiadhesive property was most likely attributed to its ability to inhibit biological activities. Results of the antibacterial experiments confirmed that the nanocomposite membrane was highly effective in inhibiting bacterial growth with an antibacterial efficiency of over 98%, which did not decrease even after the membrane was soaked in DI water for seven days. Copyright © 2013 Elsevier Ltd. All rights reserved.