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Sample records for helium-cooled divertor finger

  1. Evaluation of helium cooling for fusion divertors

    SciTech Connect

    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{sup 2} at an average heat flux of 2 MW/m{sup 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{sup 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{sup 2}. The pumping power required was less than 1% of the power removed. These results verified the design prediction.

  2. Helium-cooled high temperature reactors

    SciTech Connect

    Trauger, D.B.

    1985-01-01

    Experience with several helium cooled reactors has been favorable, and two commercial plants are now operating. Both of these units are of the High Temperature Graphite Gas Cooled concept, one in the United States and the other in the Federal Republic of Germany. The initial helium charge for a reactor of the 1000 MW(e) size is modest, approx.15,000 kg.

  3. Helium cooling systems for large superconducting physics detector magnets

    NASA Astrophysics Data System (ADS)

    Green, M. A.

    The large superconducting detector magnets used for high energy physics experiments are virtually all indirectly cooled. In general, these detector magnets are not cryogenically stabilized. Therefore, there are a number of choices for cooling large indirectly cooled detector magnets. These choices include; 1) forced two-phase helium cooling driven by the helium refrigerator J-T circuit, 2) forced two-phase helium cooling driven by a helium pump, and 3) a peculation gravity feed cooling system which uses liquid helium from a large storage dewar. The choices for the cooling of a large detector magnet are illustrated by applying these concepts to a 4.2 meter diameter 0.5 tesla thin superconducting solenoid for an experiment at the Relativistic Heavy Ion Collider (RHIC).

  4. Helium-Cooled Refractory Alloys First Wall and Blanket Evaluation

    SciTech Connect

    Wong, C.P.C.; Nygren, R.E.; Baxi, C.B.; Fogarty, P.; Ghoniem, N.; Khater, H.; McCarthy, K.; Merrill, B.; Nelson, B.; Reis, E.E.; Sharafat, S.; Schleicher, R.; Sze, D.K.; Ulrickson, M.; Willms, S.; Youssef, M.; Zinkel, S.

    1999-08-01

    Under the APEX program the He-cooled system design task is to evaluate and recommend high power density refractory alloy first wall and blanket designs and to recommend and initiate tests to address critical issues. We completed the preliminary design of a helium-cooled, W-5Re alloy, lithium breeder design and the results are reported in this paper. Many areas of the design were assessed, including material selection, helium impurity control, and mechanical, nuclear and thermal hydraulics design, and waste disposal, tritium and safety design. System study results show that at a closed cycle gas turbine (CCGT) gross thermal efficiency of 57.5%, a superconducting coil tokamak reactor, with an aspect ratio of 4, and an output power of 2 GWe, can be projected to have a cost of electricity at 54.6 mill/kWh. Critical issues were identified and we plan to continue the design on some of the critical issues during the next phase of the APEX design study.

  5. Design of a helium-cooled molten salt fusion breeder

    SciTech Connect

    Moir, R.W.; Lee, J.D.; Fulton, F.J.; Huegel, F.; Neef, W.S. Jr.; Sherwood, A.E.; Berwald, D.H.; Whitley, R.H.; Wong, C.P.C.; DeVan, J.H.

    1985-02-01

    A new conceptual blanket design for a fusion reactor produces fissile material for fission power plants. Fission is suppressed by using beryllium, rather than uranium, to multiply neutrons and also by minimizing the fissile inventory. The molten-salt breeding media (LiF + BeF/sub 2/ + TghF/sub 4/) is circulated through the blanket and on to the online processing system where /sup 233/U and tritium are continuously removed. Helium cools the blanket including the steel pipes containing the molten salt. Austenitic steel was chosen because of its ease of fabrication, adequate radiation-damage lifetime, and low corrosion rate by molten salt. We estimate the breeder, having 3000 MW of fusion power, produces 6400 kg of /sup 233/U per year, which is enough to provide make up for 20 GWe of LWR per year (or 14 LWR plants of 4440 MWt) or twice that many HTGRs or CANDUs. Safety is enhanced because the afterheat is low and the blanket materials do not react with air or water. The fusion breeder based on a pre-MARS tandem mirror is estimated to cost $4.9B or 2.35 times an LWR of the same power. The estimated present value cost of the /sup 2/anumber/sup 3/U produced is $40/g if utility financed or $16/g if government financed.

  6. Development of a feed monitor system for a helium-cooled Michelson intererometer for the Spacelab

    NASA Technical Reports Server (NTRS)

    Essenwanger, P.

    1980-01-01

    A Michelson interferometer feed monitor system developed for Spacelab is described. The device is helium cooled and is to be used to measure far infrared radiation sources in space. Performance data and development sequence are presented.

  7. Helium-cooled molten-salt fusion breeder

    SciTech Connect

    Moir, R.W.; Lee, J.D.; Fulton, F.J.; Huegel, F.; Neef, W.S. Jr.; Sherwood, A.E.; Berwald, D.H.; Whitley, R.H.; Wong, C.P.C.; Devan, J.H.

    1984-12-01

    We present a new conceptual design for a fusion reactor blanket that is intended to produce fissile material for fission power plants. Fast fission is suppressed by using beryllium instead of uranium to multiply neutrons. Thermal fission is suppressed by minimizing the fissile inventory. The molten-salt breeding medium (LiF + BeF/sub 2/ + ThF/sub 4/) is circulated through the blanket and to the on-line processing system where /sup 233/U and tritium are continuously removed. Helium cools the blanket and the austenitic steel tubes that contain the molten salt. Austenitic steel was chosen because of its ease of fabrication, adequate radiation-damage lifetime, and low corrosion by molten salt. We estimate that a breeder having 3000 MW of fusion power will produce 6500 kg of /sup 233/U per year. This amount is enough to provide makeup for 20 GWe of light-water reactors per year or twice that many high-temperature gas-cooled reactors or Canadian heavy-water reactors. Safety is enhanced because the afterheat is low and blanket materials do not react with air or water. The fusion breeder based on a pre-MARS tandem mirror is estimated to cost $4.9B or 2.35 times a light-water reactor of the same power. The estimated cost of the /sup 233/U produced is $40/g for fusion plants costing 2.35 times that of a light-water reactor if utility owned or $16/g if government owned.

  8. Helium-Cooled Black Shroud for Subscale Cryogenic Testing

    NASA Technical Reports Server (NTRS)

    Tuttle, James; Jackson, Michael; DiPirro, Michael; Francis, John

    2011-01-01

    This shroud provides a deep-space simulating environment for testing scaled-down models of passively cooling systems for spaceflight optics and instruments. It is used inside a liquid-nitrogen- cooled vacuum chamber, and it is cooled by liquid helium to 5 K. It has an inside geometry of approximately 1.6 m diameter by 0.45 m tall. The inside surfaces of its top and sidewalls have a thermal absorptivity greater than 0.96. The bottom wall has a large central opening that is easily customized to allow a specific test item to extend through it. This enables testing of scale models of realistic passive cooling configurations that feature a very large temperature drop between the deepspace-facing cooled side and the Sun/Earth-facing warm side. This shroud has an innovative thermal closeout of the bottom wall, so that a test sample can have a hot (room temperature) side outside of the shroud, and a cold side inside the shroud. The combination of this closeout and the very black walls keeps radiated heat from the sample s warm end from entering the shroud, reflecting off the walls and heating the sample s cold end. The shroud includes 12 vertical rectangular sheet-copper side panels that are oriented in a circular pattern. Using tabs bent off from their edges, these side panels are bolted to each other and to a steel support ring on which they rest. The removable shroud top is a large copper sheet that rests on, and is bolted to, the support ring when the shroud is closed. The support ring stands on four fiberglass tube legs, which isolate it thermally from the vacuum chamber bottom. The insides of the cooper top and side panels are completely covered with 25- mm-thick aluminum honeycomb panels. This honeycomb is painted black before it is epoxied to the copper surfaces. A spiral-shaped copper tube, clamped at many different locations to the outside of the top copper plate, serves as part of the liquid helium cooling loop. Another copper tube, plumbed in a series to the

  9. Divertor detachment

    NASA Astrophysics Data System (ADS)

    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.

  10. Preparation of the helium-cooled German Infrared Laboratory (GIRL) interferometer for scientific use on Spacelab

    NASA Astrophysics Data System (ADS)

    Essenwanger, P.

    1985-07-01

    Preparation of a helium-cooled Michelson interferometer for high resolution IR spectroscopy of astronomical sources on the Spacelab telescope GIRL is summarized. Molecular parameters are provided as a basis for astrophysical observations. Critical problems in the adjustment and calibration of the sensitive interferometer were solved in the laboratory: a retroreflector was adjusted and tested, and a gas cell was built and tested for the calibration.

  11. Helium-cooled balloon-borne infrared experiment for measurements of stratospheric trace gas emissions

    NASA Astrophysics Data System (ADS)

    Rippel, H.; Kampf, D.; Hilbert, L.; Jarisch, M.; Offermann, D.

    1987-08-01

    A helium-cooled IR spectrometer with a diffraction-limited telescope was launched on Sept. 23, 1983, from Aire-sur-l'Adour (France) as part of the MAP/Globus 1983 campaign. The float altitude of the balloon was 38 km. Limb scan measurements of atmospheric emissions were taken in the 5.5-19 micron wavelength region. The measurements were performed at about 1 h before sunrise. From several spectral features volume mixing ratios of NO2, H2O, CH4, HNO3, O3, and N2O were derived.

  12. Deep Burn Develpment of Transuranic Fuel for High-Temperature Helium-Cooled Reactors - July 2010

    SciTech Connect

    Snead, Lance Lewis; Besmann, Theodore M; Collins, Emory D; Bell, Gary L

    2010-08-01

    The DB Program Quarterly Progress Report for April - June 2010, ORNL/TM/2010/140, was distributed to program participants on August 4. This report discusses the following: (1) TRU (transuranic elements) HTR (high temperature helium-cooled reactor) Fuel Modeling - (a) Thermochemical Modeling, (b) 5.3 Radiation Damage and Properties; (2) TRU HTR Fuel Qualification - (a) TRU Kernel Development, (b) Coating Development, (c) ZrC Properties and Handbook; and (3) HTR Fuel Recycle - (a) Recycle Processes, (b) Graphite Recycle, (c) Pyrochemical Reprocessing - METROX (metal recovery from oxide fuel) Process Development.

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

    SciTech Connect

    Youchison, Dennis L.

    2015-07-30

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

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

    DOE PAGES

    Youchison, Dennis L.

    2015-07-30

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

  15. Transport and deposition of activation products in a helium cooled fusion power plant

    SciTech Connect

    Bickford, W.E.

    1980-09-01

    The transport and deposition of neutron activation products in a helium cooled tokamak fusion power plant are investigated. Stainless steel is used as coolant channel material for a helium/steam system. The important gamma emitting nuclides /sup 56/Mn, /sup 54/Mn, /sup 57/Co, /sup 58/Co, /sup 60/Co, /sup 51/Cr, and /sup 99/Mo are considered. The dominant release mechanism identified is direct daughter recoil emission from (n,x) type reactions. Corrosion and evaporation are discussed. The radionuclide inventory released by these mechanisms is predicted to exceed 1 x 10/sup 4/ Ci for a reference reactor design after only several days of operation, and approach 3.5 x 10/sup 4/ Ci in equilibrium. A mass transport model is then used to predict the deposition pattern of this inventory in the reactor cooling system.

  16. The Helium Cooling System and Cold Mass Support System for theMICE Coupling Solenoid

    SciTech Connect

    Wang, L.; Wu, H.; Li, L.K.; Green, M.A.; Liu, C.S.; Li, L.Y.; Jia, L.X.; Virostek, S.P.

    2007-08-27

    The MICE cooling channel consists of alternating threeabsorber focus coil module (AFC) and two RF coupling coil module (RFCC)where the process of muon cooling and reacceleration occurs. The RFCCmodule comprises a superconducting coupling solenoid mounted around fourconventional conducting 201.25 MHz closed RF cavities and producing up to2.2T magnetic field on the centerline. The coupling coil magnetic fieldis to produce a low muon beam beta function in order to keep the beamwithin the RF cavities. The magnet is to be built using commercialniobium titanium MRI conductors and cooled by pulse tube coolers thatproduce 1.5 W of cooling capacity at 4.2 K each. A self-centering supportsystem is applied for the coupling magnet cold mass support, which isdesigned to carry a longitudinal force up to 500 kN. This report willdescribe the updated design for the MICE coupling magnet. The cold masssupport system and helium cooling system are discussed indetail.

  17. A Liquid-Helium-Cooled Absolute Reference Cold Load forLong-Wavelength Radiometric Calibration

    SciTech Connect

    Bensadoun, M.; Witebsky, C.; Smoot, George F.; De Amici,Giovanni; Kogut, A.; Levin, S.

    1990-05-01

    We describe a large (78-cm) diameter liquid-helium-cooled black-body absolute reference cold load for the calibration of microwave radiometers. The load provides an absolute calibration near the liquid helium (LHe) boiling point, accurate to better than 30 mK for wavelengths from 2.5 to 25 cm (12-1.2 GHz). The emission (from non-LHe temperature parts of the cold load) and reflection are small and well determined. Total corrections to the LHe boiling point temperature are {le} 50 mK over the operating range. This cold load has been used at several wavelengths at the South Pole and at the White Mountain Research Station. In operation, the average LHe loss rate was {le} 4.4 l/hr. Design considerations, radiometric and thermal performance and operational aspects are discussed. A comparison with other LHe-cooled reference loads including the predecessor of this cold load is given.

  18. A liquid-helium-cooled absolute reference cold load for long-wavelength radiometric calibration

    NASA Technical Reports Server (NTRS)

    Bensadoun, Marc; Witebsky, Chris; Smoot, George; De Amici, Giovanni; Kogut, AL; Levin, Steve

    1992-01-01

    Design, radiometric and thermal performance, and operation of a large diameter (78 cm) liquid-helium-cooled blackbody absolute reference cold load (CL) for the calibration of microwave radiometers is described. CL provides an absolute calibration near the liquid-helium (LHe) boiling point, with total uncertainty in the radiometric temperature of less than 30 mK over the 2.5-23 cm wavelength operating range. CL was used at several wavelengths at the South Pole, Antarctica and the White Mountain Research Center, California. Results show that, for the instruments operated at 20-, 12-, 7.9-, and 4.0 cm wavelength at the South Pole, the total corrections to the LHe boiling-point temperature (about 3.8 K) were 48 +/-23, 18 +/-10, 10 +/-18, and 15 +/-mK.

  19. Coupling a Supercritical Carbon Dioxide Brayton Cycle to a Helium-Cooled Reactor.

    SciTech Connect

    Middleton, Bobby; Pasch, James Jay; Kruizenga, Alan Michael; Walker, Matthew

    2016-01-01

    This report outlines the thermodynamics of a supercritical carbon dioxide (sCO2) recompression closed Brayton cycle (RCBC) coupled to a Helium-cooled nuclear reactor. The baseline reactor design for the study is the AREVA High Temperature Gas-Cooled Reactor (HTGR). Using the AREVA HTGR nominal operating parameters, an initial thermodynamic study was performed using Sandia's deterministic RCBC analysis program. Utilizing the output of the RCBC thermodynamic analysis, preliminary values of reactor power and of Helium flow rate through the reactor were calculated in Sandia's HelCO2 code. Some research regarding materials requirements was then conducted to determine aspects of corrosion related to both Helium and to sCO2 , as well as some mechanical considerations for pressures and temperatures that will be seen by the piping and other components. This analysis resulted in a list of materials-related research items that need to be conducted in the future. A short assessment of dry heat rejection advantages of sCO2> Brayton cycles was also included. This assessment lists some items that should be investigated in the future to better understand how sCO2 Brayton cycles and nuclear can maximally contribute to optimizing the water efficiency of carbon free power generation

  20. Forced Two-Phase Helium Cooling Scheme for the Mu2e Transport Solenoid

    SciTech Connect

    Tatkowski, G.; Cheban, S.; Dhanaraj, N.; Evbota, D.; Lopes, M.; Nicol, T.; Sanders, R.; Schmitt, R.; Voirin, E.

    2015-01-01

    The Mu2e Transport Solenoid (TS) is an S-shaped magnet formed by two separate but similar magnets, TS-u and TS-d. Each magnet is quarter-toroid shaped with a centerline radius of approximately 3 m utilizing a helium cooling loop consisting of 25 to 27 horizontal-axis rings connected in series. This cooling loop configuration has been deemed adequate for cooling via forced single phase liquid helium; however it presents major challenges to forced two-phase flow such as “garden hose” pressure drop, concerns of flow separation from tube walls, difficulty of calculation, etc. Even with these disadvantages, forced two-phase flow has certain inherent advantages which make it a more attractive option than forced single phase flow. It is for this reason that the use of forced two-phase flow was studied for the TS magnets. This paper will describe the analysis using helium-specific pressure drop correlations, conservative engineering approach, helium properties calculated and updated at over fifty points, and how the results compared with those in literature. Based on the findings, the use of forced-two phase helium is determined to be feasible for steady-state cooling of the TS solenoids

  1. THE VALUE OF HELIUM-COOLED REACTOR TECHNOLOGIES OF NUCLEAR WASTE

    SciTech Connect

    C. RODRIGUEZ; A. BAXTER

    2001-03-01

    Helium-cooled reactor technologies offer significant advantages in accomplishing the waste transmutation process. They are ideally suited for use with thermal, epithermal, or fast neutron energy spectra. They can provide a relatively hard thermal neutron spectrum for transmutation of fissionable materials such as Pu-239 using ceramic-coated transmutation fuel particles, a graphite moderator, and a non-fertile burnable poison. These features (1) allow deep levels of transmutation with minimal or no intermediate reprocessing, (2) enhance passive decay heat removal via heat conduction and radiation, (3) allow operation at relatively high temperatures for a highly efficient generation of electricity, and (4) discharge the transmuted waste in a form that is highly resistant to corrosion for long times. They also offer the possibility for the use of epithermal neutrons that can interact with transmutable materials more effectively because of the large atomic cross sections in this energy domain. A fast spectrum may be useful for deep burnup of certain minor actinides. For this application, helium is essentially transparent to neutrons, does not degrade neutron energies, and offers the hardest possible neutron energy environment. In this paper, we report results from recent work on materials transmutation balances, safety, value to a geological repository, and economic considerations.

  2. Helium-cooled, FLiBe-breeder, beryllium-multiplier blanket for MINIMARS

    SciTech Connect

    Moir, R.W.; Lee, J.D.

    1986-06-01

    We adapted the helium-cooled, FLiBe-breeder blanket to the commercial tandem-mirror fusion-reactor design, MINIMARS. Vanadium was used to achieve high performance from the high-energy-release neutron-capture reactions and from the high-temperature operation permitted by the refractory property of the material, which increases the conversion efficiency and decreases the helium-pumping power. Although this blanket had the highest performance among the MINIMARS blankets designs, measured by Mn/sub th/ (blanket energy multiplication times thermal conversion efficiency), it had a cost of electricity (COE) 18% higher than the University of Wisconsin (UW) blanket design (42.5 vs 35.9 mills/kW.h). This increased cost was due to using higher-cost blanket materials (beryllium and vanadium) and a thicker blanket, which resulted in higher-cost central-cell magnets and the need for more blanket materials. Apparently, the high efficiency does not substantially affect the COE. Therefore, in the future, we recommend lowering the helium temperature so that ferritic steel can be used. This will result in a lower-cost blanket, which may compensate for the lower performance resulting from lower efficiency.

  3. A methodology for accident analysis of fusion breeder blankets and its application to helium-cooled lead–lithium blanket

    DOE PAGES

    Panayotov, Dobromir; Poitevin, Yves; Grief, Andrew; ...

    2016-09-23

    'Fusion for Energy' (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials,more » and phenomena while remaining consistent with the approach already applied to ITER accident analyses. Furthermore, the methodology phases are illustrated in the paper by its application to the EU HCLL TBS using both MELCOR and RELAP5 codes.« less

  4. A methodology for accident analysis of fusion breeder blankets and its application to helium-cooled lead–lithium blanket

    SciTech Connect

    Panayotov, Dobromir; Poitevin, Yves; Grief, Andrew; Trow, Martin; Dillistone, Michael; Murgatroyd, Julian T.; Owen, Simon; Peers, Karen; Lyons, Alex; Heaton, Adam; Scott, Richard; Merrill, Brad J.; Humrickhouse, Paul

    2016-09-23

    'Fusion for Energy' (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. Furthermore, the methodology phases are illustrated in the paper by its application to the EU HCLL TBS using both MELCOR and RELAP5 codes.

  5. Design of the helium cooled lithium lead breeding blanket in CEA: from TBM to DEMO

    NASA Astrophysics Data System (ADS)

    Aiello, G.; Aubert, J.; Forest, L.; Jaboulay, J.-C.; Li Puma, A.; Boccaccini, L. V.

    2017-04-01

    The helium cooled lithium lead (HCLL) blanket concept was originally developed in CEA at the beginning of 2000: it is one of the two European blanket concepts to be tested in ITER in the form of a test blanket module (TBM) and one of the four blanket concepts currently being considered for the DEMOnstration reactor that will follow ITER. The TBM is a highly optimized component for the ITER environment that will provide crucial information for the development of the DEMO blanket, but its design needs to be adapted to the DEMO reactor. With respect to the TBM design, reduction of the steel content in the breeding zone (BZ) is sought in order to maximize tritium breeding reactions. Different options are being studied, with the potential of reaching tritium breeding ratio (TBR) values up to 1.21. At the same time, the design of the back supporting structure (BSS), which is a DEMO specific component that has to support the blanket modules inside the vacuum vessel (VV), is ongoing with the aim of maximizing the shielding power and minimizing pumping power. This implies a re-engineering of the modules’ attachment system. Design changes however, will have an impact on the manufacturing and assembly sequences that are being developed for the HCLL-TBM. Due to the differences in joint configurations, thicknesses to be welded, heat dissipation and the various technical constraints related to the accessibility of the welding tools and implementation of non-destructive examination (NDE), the manufacturing procedure should be adapted and optimized for DEMO design. Laser welding instead of TIG could be an option to reduce distortions. The time-of-flight diffraction (TOFD) technique is being investigated for NDE. Finally, essential information expected from the HCLL-TBM program that will be needed to finalize the DEMO design is discussed.

  6. Gas propagation in a liquid helium cooled vacuum tube following a sudden vacuum loss

    NASA Astrophysics Data System (ADS)

    Dhuley, Ram C.

    This dissertation describes the propagation of near atmospheric nitrogen gas that rushes into a liquid helium cooled vacuum tube after the tube suddenly loses vacuum. The loss-of-vacuum scenario resembles accidental venting of atmospheric air to the beam-line of a superconducting radio frequency particle accelerator and is investigated to understand how in the presence of condensation, the in-flowing air will propagate in such geometry. In a series of controlled experiments, room temperature nitrogen gas (a substitute for air) at a variety of mass flow rates was vented to a high vacuum tube immersed in a bath of liquid helium. Pressure probes and thermometers installed on the tube along its length measured respectively the tube pressure and tube wall temperature rise due to gas flooding and condensation. At high mass in-flow rates a gas front propagated down the vacuum tube but with a continuously decreasing speed. Regression analysis of the measured front arrival times indicates that the speed decreases nearly exponentially with the travel length. At low enough mass in-flow rates, no front propagated in the vacuum tube. Instead, the in-flowing gas steadily condensed over a short section of the tube near its entrance and the front appeared to `freeze-out'. An analytical expression is derived for gas front propagation speed in a vacuum tube in the presence of condensation. The analytical model qualitatively explains the front deceleration and flow freeze-out. The model is then simplified and supplemented with condensation heat/mass transfer data to again find the front to decelerate exponentially while going away from the tube entrance. Within the experimental and procedural uncertainty, the exponential decay length-scales obtained from the front arrival time regression and from the simplified model agree.

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

    SciTech Connect

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

    1999-11-01

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

  8. TPX divertor modeling studies

    SciTech Connect

    Rensink, M.E.; Braams, B.J.; Brooks, J.N.

    1995-06-20

    The Tokamak Physics Experiment (TPX) is designed to demonstrate features of an economically attractive steady state tokamak reactor. In this paper we present recent results from numerical studies of the proposed TPX divertor design (1), focusing on particle control and on radiative divertor scenarios for reducing the peak divertor heat flux. The configuration is an up/down symmetric double-null with a deep re-entrant slot geometry for the outer divertor legs.

  9. Deep Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors- Monthly Highlights October 2010

    SciTech Connect

    Snead, Lance Lewis; Besmann, Theodore M; Collins, Emory D; Bell, Gary L

    2010-11-01

    The DB Program monthly highlights report for September 2010, ORNL/TM-2010/252, was distributed to program participants by email on October 26. This report discusses: (1) Core and Fuel Analysis; (2) Spent Fuel Management; (3) Fuel Cycle Integration of the HTR (high temperature helium-cooled reactor); (4) TRU (transuranic elements) HTR Fuel Qualification; (5) HTR Spent Fuel Recycle - (a) TRU Kernel Development (ORNL), (b) Coating Development (ORNL), (c) Characterization Development and Support, (d) ZrC Properties and Handbook; and (6) HTR Fuel Recycle.

  10. Deep Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors- Monthly Highlights September 2010

    SciTech Connect

    Snead, Lance Lewis; Besmann, Theodore M; Collins, Emory D; Bell, Gary L

    2010-10-01

    The DB Program monthly highlights report for August 2010, ORNL/TM-2010/184, was distributed to program participants by email on September 17. This report discusses: (1) Core and Fuel Analysis - (a) Core Design Optimization in the HTR (high temperature helium-cooled reactor) Prismatic Design (Logos), (b) Core Design Optimization in the HTR Pebble Bed Design (INL), (c) Microfuel analysis for the DB HTR (INL, GA, Logos); (2) Spent Fuel Management - (a) TRISO (tri-structural isotropic) repository behavior (UNLV), (b) Repository performance of TRISO fuel (UCB); (3) Fuel Cycle Integration of the HTR (high temperature helium-cooled reactor) - Synergy with other reactor fuel cycles (GA, Logos); (4) TRU (transuranic elements) HTR Fuel Qualification - (a) Thermochemical Modeling, (b) Actinide and Fission Product Transport, (c) Radiation Damage and Properties; (5) HTR Spent Fuel Recycle - (a) TRU Kernel Development (ORNL), (b) Coating Development (ORNL), (c) Characterization Development and Support, (d) ZrC Properties and Handbook; and (6) HTR Fuel Recycle - (a) Graphite Recycle (ORNL), (b) Aqueous Reprocessing, (c) Pyrochemical Reprocessing METROX (metal recovery from oxide fuel) Process Development (ANL).

  11. Transmutation and activation analysis for divertor materials in a HCLL-type fusion power reactor

    NASA Astrophysics Data System (ADS)

    Fischer, U.; Pereslavtsev, P.; Möslang, A.; Rieth, M.

    2009-04-01

    The activation and transmutation of tungsten and tantalum as plasma facing materials was assessed for a helium cooled divertor irradiated in a typical fusion power reactor based on the use of Helium-cooled Lithium Lead (HCLL) blankets. 3D activation calculations were performed by applying a programme system linking the Monte Carlo transport code MCNP and the fusion inventory code FISPACT through an appropriate interface. Special attention was given to the proper treatment of the resonance shielding of tungsten and tantalum by using reaction rates provided directly by MCNP on the basis of continuous energy activation cross-section data. It was shown that the long-term activation behaviour is dominated by activation products of the assumed tramp material while the short-term behaviour is due to the activation of the stable Ta and W isotopes. The recycling limit for remote handling of 100 mSv/h can be achieved after decay times of 10 and 50 years for Ta and W, respectively. The elemental transmutation rates of Ta and W were shown to be on a moderate level for the HCLL-type fusion power reactor.

  12. Spectroscopy of divertor plasmas

    SciTech Connect

    Isler, R.C.

    1995-12-31

    The requirements for divertor spectroscopy are treated with respect to instrumentation and observations on present machines. Emphasis is placed on quantitative measurements.of impurity concentrations from the interpretation of spectral line intensities. The possible influence of non-Maxwellian electron distributions on spectral line excitation in the divertor is discussed. Finally the use of spectroscopy for determining plasma temperature, density, and flows is examined.

  13. The snowflake divertor

    SciTech Connect

    Ryutov, D. D.; Soukhanovskii, V. A.

    2015-11-17

    The snowflake magnetic configuration is characterized by the presence of two closely spaced poloidal field nulls that create a characteristic hexagonal (reminiscent of a snowflake) separatrix structure. The magnetic field properties and the plasma behaviour in the snowflake are determined by the simultaneous action of both nulls, this generating a lot of interesting physics, as well as providing a chance for improving divertor performance. One of the most interesting effects of the snowflake geometry is the heat flux sharing between multiple divertor channels. The authors summarise experimental results obtained with the snowflake configuration on several tokamaks. Wherever possible, relation to the existing theoretical models is described. Divertor concepts utilizing the properties of a snowflake configuration are briefly discussed.

  14. The snowflake divertor

    DOE PAGES

    Ryutov, D. D.; Soukhanovskii, V. A.

    2015-11-17

    The snowflake magnetic configuration is characterized by the presence of two closely spaced poloidal field nulls that create a characteristic hexagonal (reminiscent of a snowflake) separatrix structure. The magnetic field properties and the plasma behaviour in the snowflake are determined by the simultaneous action of both nulls, this generating a lot of interesting physics, as well as providing a chance for improving divertor performance. One of the most interesting effects of the snowflake geometry is the heat flux sharing between multiple divertor channels. The authors summarise experimental results obtained with the snowflake configuration on several tokamaks. Wherever possible, relation tomore » the existing theoretical models is described. Divertor concepts utilizing the properties of a snowflake configuration are briefly discussed.« less

  15. Initial assessment of environmental effects on SiC/SiC composites in helium-cooled nuclear systems

    SciTech Connect

    Contescu, Cristian I

    2013-09-01

    This report summarized the information available in the literature on the chemical reactivity of SiC/SiC composites and of their components in contact with the helium coolant used in HTGR, VHTR and GFR designs. In normal operation conditions, ultra-high purity helium will have chemically controlled impurities (water, oxygen, carbon dioxide, carbon monoxide, methane, hydrogen) that will create a slightly oxidizing gas environment. Little is known from direct experiments on the reactivity of third generation (nuclear grade) SiC/SiC composites in contact with low concentrations of water or oxygen in inert gas, at high temperature. However, there is ample information about the oxidation in dry and moist air of SiC/SiC composites at high temperatures. This information is reviewed first in the next chapters. The emphasis is places on the improvement in material oxidation, thermal, and mechanical properties during three stages of development of SiC fibers and at least two stages of development of the fiber/matrix interphase. The chemical stability of SiC/SiC composites in contact with oxygen or steam at temperatures that may develop in off-normal reactor conditions supports the conclusion that most advanced composites (also known as nuclear grade SiC/SiC composites) have the chemical resistance that would allow them maintain mechanical properties at temperatures up to 1200 1300 oC in the extreme conditions of an air or water ingress accident scenario. Further research is needed to assess the long-term stability of advanced SiC/SiC composites in inert gas (helium) in presence of very low concentrations (traces) of water and oxygen at the temperatures of normal operation of helium-cooled reactors. Another aspect that needs to be investigated is the effect of fast neutron irradiation on the oxidation stability of advanced SiC/SiC composites in normal operation conditions.

  16. Characterisation and reduction of the EEG artefact caused by the helium cooling pump in the MR environment: validation in epilepsy patient data.

    PubMed

    Rothlübbers, Sven; Relvas, Vânia; Leal, Alberto; Murta, Teresa; Lemieux, Louis; Figueiredo, Patrícia

    2015-03-01

    The EEG acquired simultaneously with fMRI is distorted by a number of artefacts related to the presence of strong magnetic fields, which must be reduced in order to allow for a useful interpretation and quantification of the EEG data. For the two most prominent artefacts, associated with magnetic field gradient switching and the heart beat, reduction methods have been developed and applied successfully. However, a number of artefacts related to the MR-environment can be found to distort the EEG data acquired even without ongoing fMRI acquisition. In this paper, we investigate the most prominent of those artefacts, caused by the Helium cooling pump, and propose a method for its reduction and respective validation in data collected from epilepsy patients. Since the Helium cooling pump artefact was found to be repetitive, an average template subtraction method was developed for its reduction with appropriate adjustments for minimizing the degradation of the physiological part of the signal. The new methodology was validated in a group of 15 EEG-fMRI datasets collected from six consecutive epilepsy patients, where it successfully reduced the amplitude of the artefact spectral peaks by 95 ± 2 % while the background spectral amplitude within those peaks was reduced by only -5 ± 4 %. Although the Helium cooling pump should ideally be switched off during simultaneous EEG-fMRI acquisitions, we have shown here that in cases where this is not possible the associated artefact can be effectively reduced in post processing.

  17. Far-Infrared Photometry with an 0.4-Meter Liquid Helium Cooled Balloon-Borne Telescope. Ph.D. Thesis

    NASA Technical Reports Server (NTRS)

    Jacobson, M. R.

    1977-01-01

    A 0.4-meter aperture, liquid helium cooled multichannel far-infrared balloon-borne telescope was constructed to survey the galactic plane. Nine new sources, above a 3-sigma confidence level of 1300 Jy, were identified. Although two-thirds of the scanned area was more than 10 degrees from the galactic plane, no sources were detected in that region; all nine fell within 10 degrees and eight of those within 4 degrees of the galactic equator. Correlations with visible, compact H lines associated with radio continuum and with sources displaying spectra steeply rising between 11 and 20 microns were noted, while stellar objects were not detected.

  18. The snowflake divertor

    SciTech Connect

    Ryutov, D. D.; Soukhanovskii, V. A.

    2015-11-15

    The snowflake magnetic configuration is characterized by the presence of two closely spaced poloidal field nulls that create a characteristic hexagonal (reminiscent of a snowflake) separatrix structure. The magnetic field properties and the plasma behaviour in the snowflake are determined by the simultaneous action of both nulls, this generating a lot of interesting physics, as well as providing a chance for improving divertor performance. Among potential beneficial effects of this geometry are: increased volume of a low poloidal field around the null, increased connection length, and the heat flux sharing between multiple divertor channels. The authors summarise experimental results obtained with the snowflake configuration on several tokamaks. Wherever possible, relation to the existing theoretical models is described.

  19. Divertor plasma detachment

    SciTech Connect

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

    2016-05-15

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

  20. Divertor plasma detachment

    NASA Astrophysics Data System (ADS)

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

    2016-05-01

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

  1. Design of the advanced divertor pump cryogenic system for DIII-D

    SciTech Connect

    Schaubel, K.M.; Baxi, C.B.; Campbell, G.L.; Gootgeld, A.M.; Langhorn, A.R.; Laughon, G.J.; Smith, J.P.; Anderson, P.M. ); Menon, M.M. )

    1991-11-01

    The design of the cryogenic system for the D3-D advanced divertor cryocondensation pump is presented. The advanced divertor incorporates a baffle chamber and bias ring located near the bottom of the D3-D vacuum vessel. A 50,000 l/s cryocondensation pump will be installed underneath the baffle for plasma particle exhaust. The pump consists of a liquid helium cooled tube operating at 4.3{degrees}K and a liquid nitrogen cooled radiation shield. Liquid helium is fed by forced flow through the cryopump. Compressed helium gas flowing through the high pressure side of a heat exchanger is regeneratively cooled by the two-phase helium leaving the pump. The cooled high pressure gaseous helium is than liquefied by a Joule-Thomson expansion valve. The liquid is returned to a storage dewar. The liquid nitrogen for the radiation shield is supplied by forced flow from a bulk storage system. Control of the cryogenic system is accomplished by a programmable logic controller.

  2. A cryocondensation pump for the DIII-D Advanced Divertor Program

    SciTech Connect

    Smith, J.P.; Baxi, C.; Reis, E.; Sevier, L.

    1992-03-01

    A cryocondensation pump was designed for the baffle chamber of General Atomics DIII-D tokamak and will be installed in the fall of 1992. The purpose of the pump is to study plasma density control by pumping the divertor. The pump is toroidally continuous, approximately 10 m long and located in the lower outer corner of the vacuum chamber of the machine. It consists of a 1 m{sup 2} liquid helium-cooled surface surrounded by a liquid nitrogen-cooled shield to limit the heat load on the helium-cooled surface. The liquid nitrogen-cooled surface is surrounded by a radiation/particle shield to prevent energetic particles from impacting and releasing condensed water molecules. A thermal enhancement coating was applied to the nitrogen shell to lower the maximum temperature of the shell. The coating is non-continuous to keep the toroidal electrical resistance high. The whole pump is supported off the water-cooled vacuum vessel wall. Supports for the pump were designed to accommodate the thermal differences between the 4 K helium surface, the 77 K nitrogen shells, and the 300 K vacuum vessel supporting the pump and to provide a low heat leak structural support. Disruption loading on the pump was analyzed and a finite element structural analysis of the pump was completed. A testing program was completed to evaluate coating techniques to enhance heat transfer and emissivity of the various surfaces. Fabrication tests were performed to determine the best method of attaching the liquid nitrogen flow tubes to their shield surfaces and to determine the best alternative to fabricating the different shells of the pump. A prototype sector of the pump was built to verify fabrication and assembly techniques.

  3. Asymmetric divertor biasing in MAST

    NASA Astrophysics Data System (ADS)

    Helander, P.; Cohen, R.; Counsell, G. C.; Ryutov, D. D.

    2002-11-01

    Experiments are being carried out on the Mega-Ampere Spherical Tokamak (MAST) where the divertor tiles are electrically biased in a toroidally alternating way. The aim is to induce convective cells in the divertor plasma, broaden the SOL and reduce the divertor heat load. This paper describes the underlying theory and experimental results. Criteria are presented for achieving strong broadening and exciting shear-flow turbulence in the SOL, and properties of the expected turbulence are derived. It is also shown that magnetic shear near the X-point is likely to confine the potential perturbations to the divertor region, leaving the part of the SOL that is in direct contact with the core plasma intact. Preliminary comparison of the theory with MAST data is encouraging: the distortion of the heat deposition pattern, its broadening, and the incremental heat load are qualitatively in agreement; quantitative comparisons are underway.

  4. Electric field divertor plasma pump

    DOEpatents

    Schaffer, M.J.

    1994-10-04

    An electric field plasma pump includes a toroidal ring bias electrode positioned near the divertor strike point of a poloidal divertor of a tokamak, or similar plasma-confining apparatus. For optimum plasma pumping, the separatrix of the poloidal divertor contacts the ring electrode, which then also acts as a divertor plate. A plenum or other duct near the electrode includes an entrance aperture open to receive electrically-driven plasma. The electrode is insulated laterally with insulators, one of which is positioned opposite the electrode at the entrance aperture. An electric field E is established between the ring electrode and a vacuum vessel wall, with the polarity of the bias applied to the electrode being relative to the vessel wall selected such that the resultant electric field E interacts with the magnetic field B already existing in the tokamak to create an E [times] B/B[sup 2] drift velocity that drives plasma into the entrance aperture. The pumped plasma flow into the entrance aperture is insensitive to variations, intentional or otherwise, of the pump and divertor geometry. Pressure buildups in the plenum or duct connected to the entrance aperture in excess of 10 mtorr are achievable. 11 figs.

  5. Electric field divertor plasma pump

    DOEpatents

    Schaffer, Michael J.

    1994-01-01

    An electric field plasma pump includes a toroidal ring bias electrode (56) positioned near the divertor strike point of a poloidal divertor of a tokamak (20), or similar plasma-confining apparatus. For optimum plasma pumping, the separatrix (40) of the poloidal divertor contacts the ring electrode (56), which then also acts as a divertor plate. A plenum (54) or other duct near the electrode (56) includes an entrance aperture open to receive electrically-driven plasma. The electrode (56) is insulated laterally with insulators (63,64), one of which (64) is positioned opposite the electrode at the entrance aperture. An electric field E is established between the ring electrode (56) and a vacuum vessel wall (22), with the polarity of the bias applied to the electrode being relative to the vessel wall selected such that the resultant electric field E interacts with the magnetic field B already existing in the tokamak to create an E.times.B/B.sup.2 drift velocity that drives plasma into the entrance aperture. The pumped plasma flow into the entrance aperture is insensitive to variations, intentional or otherwise, of the pump and divertor geometry. Pressure buildups in the plenum or duct connected to the entrance aperture in excess of 10 mtorr are achievable.

  6. Actively convected liquid metal divertor

    NASA Astrophysics Data System (ADS)

    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.

  7. VUV Spectroscopy in DIII-D Divertor

    SciTech Connect

    Alkesh Punjabi; Nelson Jalufka

    2004-11-04

    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.

  8. Tokamak Physics Experiment divertor design

    SciTech Connect

    Anderson, P.M.

    1995-12-31

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

  9. Controlling marginally detached divertor plasmas

    DOE PAGES

    Eldon, David; Kolemen, Egemen; Barton, Joseph L.; ...

    2017-05-04

    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 Te = 5 eV near the divertor target plate), the resulting Te 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 and it handles ELMing plasmas by using real time Dα measurements to remove during-ELM slices from real time Te measurements derived from divertor Thomson scattering. The difference between measured and requested inter-ELM Te is passed to a PID (proportionalintegral-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. However, the observed bifurcation in plasma conditions at the outer strike point with the ion B ×more » $$\

  10. Trigger finger

    MedlinePlus

    ... Redness in your cut or hand Swelling or warmth in your cut or hand Yellow or green drainage from the cut Hand pain or discomfort Fever If your trigger finger returns, call your surgeon. You may need another surgery.

  11. Three-dimensional neutronics optimization of helium-cooled blanket for multi-functional experimental fusion-fission hybrid reactor (FDS-MFX)

    SciTech Connect

    Jiang, J.; Yuan, B.; Jin, M.; Wang, M.; Long, P.; Hu, L.

    2012-07-01

    Three-dimensional neutronics optimization calculations were performed to analyse the parameters of Tritium Breeding Ratio (TBR) and maximum average Power Density (PDmax) in a helium-cooled multi-functional experimental fusion-fission hybrid reactor named FDS (Fusion-Driven hybrid System)-MFX (Multi-Functional experimental) blanket. Three-stage tests will be carried out successively, in which the tritium breeding blanket, uranium-fueled blanket and spent-fuel-fueled blanket will be utilized respectively. In this contribution, the most significant and main goal of the FDS-MFX blanket is to achieve the PDmax of about 100 MW/m3 with self-sustaining tritium (TBR {>=} 1.05) based on the second-stage test with uranium-fueled blanket to check and validate the demonstrator reactor blanket relevant technologies based on the viable fusion and fission technologies. Four different enriched uranium materials were taken into account to evaluate PDmax in subcritical blanket: (i) natural uranium, (ii) 3.2% enriched uranium, (iii) 19.75% enriched uranium, and (iv) 64.4% enriched uranium carbide. These calculations and analyses were performed using a home-developed code VisualBUS and Hybrid Evaluated Nuclear Data Library (HENDL). The results showed that the performance of the blanket loaded with 64.4% enriched uranium was the most attractive and it could be promising to effectively obtain tritium self-sufficiency (TBR-1.05) and a high maximum average power density ({approx}100 MW/m{sup 3}) when the blanket was loaded with the mass of {sup 235}U about 1 ton. (authors)

  12. Sensitivity and uncertainty analysis for the tritium breeding ratio of a DEMO fusion reactor with a helium cooled pebble bed blanket

    NASA Astrophysics Data System (ADS)

    Nunnenmann, Elena; Fischer, Ulrich; Stieglitz, Robert

    2017-09-01

    An uncertainty analysis was performed for the tritium breeding ratio (TBR) of a fusion power plant of the European DEMO type using the MCSEN patch to the MCNP Monte Carlo code. The breeding blanket was of the type Helium Cooled Pebble Bed (HCPB), currently under development in the European Power Plant Physics and Technology (PPPT) programme for a fusion power demonstration reactor (DEMO). A suitable 3D model of the DEMO reactor with HCPB blanket modules, as routinely used for blanket design calculations, was employed. The nuclear cross-section data were taken from the JEFF-3.2 data library. For the uncertainty analysis, the isotopes H-1, Li-6, Li-7, Be-9, O-16, Si-28, Si-29, Si-30, Cr-52, Fe-54, Fe-56, Ni-58, W-182, W-183, W-184 and W-186 were considered. The covariance data were taken from JEFF-3.2 where available. Otherwise a combination of FENDL-2.1 for Li-7, EFF-3 for Be-9 and JENDL-3.2 for O-16 were compared with data from TENDL-2014. Another comparison was performed with covariance data from JEFF-3.3T1. The analyses show an overall uncertainty of ± 3.2% for the TBR when using JEFF-3.2 covariance data with the mentioned additions. When using TENDL-2014 covariance data as replacement, the uncertainty increases to ± 8.6%. For JEFF-3.3T1 the uncertainty result is ± 5.6%. The uncertainty is dominated by O-16, Li-6 and Li-7 cross-sections.

  13. Advanced fuels modeling: Evaluating the steady-state performance of carbide fuel in helium-cooled reactors using FRAPCON 3.4

    NASA Astrophysics Data System (ADS)

    Hallman, Luther, Jr.

    Uranium carbide (UC) has long been considered a potential alternative to uranium dioxide (UO2) fuel, especially in the context of Gen IV gas-cooled reactors. It has shown promise because of its high uranium density, good irradiation stability, and especially high thermal conductivity. Despite its many benefits, UC is known to swell at a rate twice that of UO2. However, the swelling phenomenon is not well understood, and we are limited to a weak empirical understanding of the swelling mechanism. One suggested cladding for UC is silicon carbide (SiC), a ceramic that demonstrates a number of desirable properties. Among them are an increased corrosion resistance, high mechanical strength, and irradiation stability. However, with increased temperatures, SiC exhibits an extremely brittle nature. The brittle behavior of SiC is not fully understood and thus it is unknown how SiC would respond to the added stress of a swelling UC fuel. To better understand the interaction between these advanced materials, each has been implemented into FRAPCON, the preferred fuel performance code of the Nuclear Regulatory Commission (NRC); additionally, the material properties for a helium coolant have been incorporated. The implementation of UC within FRAPCON required the development of material models that described not only the thermophysical properties of UC, such as thermal conductivity and thermal expansion, but also models for the swelling, densification, and fission gas release associated with the fuel's irradiation behavior. This research is intended to supplement ongoing analysis of the performance and behavior of uranium carbide and silicon carbide in a helium-cooled reactor.

  14. Divertor Plasma Parameters During Radiative Divertor Operation on DIII--D

    NASA Astrophysics Data System (ADS)

    Allen, S. L.; Fenstermacher, M. E.; Hill, D. N.; Lasnier, C. J.; Meyer, W. H.; Porter, G. D.; Wood, R. D.; Leonard, A. W.; Mahdavi, M. A.; Petrie, T. W.; West, W. P.; Maingi, R.; Wade, M. R.; Whyte, D. G.

    1996-11-01

    A large array of divertor diagnostics has been used to characterize the DIII--D divertor conditions during radiative divertor operation. We have used both D2 and impurities to reduce the divertor heat flux. Several discharge conditions have been obtained, including attached and detached ELMing H-modes. The multi-chord Divertor Thomson Scattering (DTS) system has been used with divertor sweeping to obtain 2-D measurements of ne and Te in the divertor. The Te drops to <= 2 eV with D2 puffing, ne increases, and the electron pressure Pe decreases. The radiation zone, measured by multi-chord bolometry, moves from the inside leg of the divertor to the outside. Comparisons of the 2-D distribution of ne and Te and the radiation distribution will be presented.

  15. Kinetic Modeling of Divertor Plasma

    NASA Astrophysics Data System (ADS)

    Ishiguro, Seiji; Hasegawa, Hiroki; Pianpanit, Theerasarn

    2015-11-01

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

  16. Controlling marginally detached divertor plasmas

    NASA Astrophysics Data System (ADS)

    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   ×  \

  17. The lithium vapor box divertor

    NASA Astrophysics Data System (ADS)

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

    2016-02-01

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

  18. Bolometry for divertor characterization and control

    SciTech Connect

    Leonard, A.W.; Goetz, J.; Fuchs, C.; Marashek, M.; Mast, F.; Reichle, R.

    1995-10-01

    Operation of the divertor will provide one of the greatest challenges for ITER. Up to 400 MW of power is expected to be produced in the core plasma which must then be handled by plasma facing components. Power flowing across the separatrix and into the scrape-off-layer (SOL) can lead to a heat flux in the divertor of 30 MW/m{sup 2} if nothing is done to dissipate the power. This peak heat flux must be reduced to 5 MW/m{sup 2} for an acceptable engineering design. The current plan is to use impurity radiation and other atomic processes from intrinsic or injected impurities to spread out the power onto the first wall and divertor chamber walls. It is estimated that 300 MW of radiation in the divertor and SOL will be necessary to achieve this solution. Measurement of the magnitude and distribution of this radiated power with bolometry will be important for understanding and controlling the nER divertor. Present experiments have shown intense regions of radiation both in the divertor near the separatrix and in the X-point region. The task of a divertor bolometer system will be to measure the distribution and magnitude of this radiation. First, radiation measurements can be used for machine protection. Intense divertor radiation will heat plasma facing surfaces that are not in direct view of temperature monitors. Measurement of the radiation distribution will provide information about the power flux to these components. Secondly, a bolometer diagnostic is a basic tool for divertor characterization and understanding. Radiation measurements are important for power accounting, as a cross check for other power diagnostics, and gross characterisation of the plasma behavior. A divertor bolometer system can provide a 2-D measurement of the radiation profile for comparison with theory and modeling. Finally a bolometer system can provide realtime signals for control of the divertor operation.

  19. Finger Foods for Babies

    MedlinePlus

    ... Kids to Be Smart About Social Media Finger Foods for Babies KidsHealth > For Parents > Finger Foods for ... will accept a new food. previous continue Finger Foods to Avoid Finger feeding is fun and rewarding ...

  20. Homoclinic finger-rings in RN

    NASA Astrophysics Data System (ADS)

    Zhu, Changrong; Zhang, Weinian

    2017-09-01

    In this paper we investigate bifurcations of a degenerate homoclinic loop in RN. We prove that a homoclinic finger-ring, an invariant manifold of a definite dimension textured with homoclinic orbits, arises from the degenerate homoclinic orbit. The size of the homoclinic finger-ring is decided by not only its dimension of manifold but also its width. For the rise of homoclinic finger-rings of different dimensions we give conditions, which are proved to form bifurcation manifolds in the parameter space. We further estimate the width for the homoclinic finger-ring and give a method to compute the bifurcation manifolds approximately.

  1. Moving Divertor Plates in a Tokamak

    SciTech Connect

    S.J. Zweben, H. Zhang

    2009-02-12

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

  2. Rapidly Moving Divertor Plates In A Tokamak

    SciTech Connect

    S. Zweben

    2011-05-16

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

  3. The lithium vapor box divertor

    SciTech Connect

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

    2016-01-13

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

  4. The lithium vapor box divertor

    DOE PAGES

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

    2016-01-13

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

  5. DIII-D divertor reflectometer system

    SciTech Connect

    Rhodes, T.L.; Doyle, E.J.; Nguyen, X.V.; Kim, K.W.; Peebles, W.A.; Doane, J.L.

    1997-01-01

    Divertor density profiles, asymmetries, turbulence, and MARFE diagnosis are extremely important and affect the divertor design process for ITER and other future devices. In addition, a functioning divertor density profile system will be essential for the operation of these machines. It is thus critical to prototype and demonstrate diagnostics capable of operating in a divertor environment. To meet these needs a divertor reflectometer system has been designed and installed on DIII-D. The design stresses flexibility, modularity, and simplicity. It consists of a circular, smoothwall, overmoded waveguide followed by a TE{sub 11}{R_arrow}HE{sub 11} mode converter (the HE{sub 11} mode is a low loss Gaussian mode with a very symmetric radiation pattern, optimal for this use) thus allowing use of an arbitrary polarization (f{sub pe},f{sub LH},f{sub RH}). The design provides for testing of a variety of antennas/probing directions including: upward to probe the X-point region, including MARFEs, sideways to probe outboard/inboard divertor legs, and oppositely directed to probe both divertor legs simultaneously. System design, operational considerations, and experimental data are presented. {copyright} {ital 1997 American Institute of Physics.}

  6. Advanced Divertor Developments at DIII-D

    NASA Astrophysics Data System (ADS)

    Kolemen, E.; Allen, S. L.; Makowski, M. A.; Soukhanovskii, V. A.; Bray, B. D.; Eldon, D.; Humphreys, D. A.; Johnson, R.; Leonard, A. W.; Liu, C.; Penaflor, B. G.; Petrie, T. W.; McLean, A. G.; Unterberg, E. A.

    2013-10-01

    Novel divertor configurations and control schemes have been implemented at DIII-D to test and optimize heat and particle handling capabilities for advanced tokamaks. The snowflake configuration is stabilized by first calculating the position of the two null-points using real-time equilibrium reconstruction and then regulating the shaping coil currents. Experiments in which the snowflake divertor is stabilized for many confinement times show that it is compatible with high-performance operation and results in greatly reduced divertor heat flux. An advanced divertor control system regulates the gas injection to achieve partial or full detachment by using the divertor temperature measurements from real-time Thomson diagnostics and a line ratio measurement, and adjusts the core and divertor radiation via measurement of the real-time bolometer diagnostics. Prospects of achieving acceptable divertor target heat fluxes for future fusion reactors are analyzed and challenges are presented. Work supported by the US DOE under DE-AC02-09CH11466, DE-AC52-07NA27344, DE-FC02-04ER54698 and DE-AC05-00OR22725.

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

    SciTech Connect

    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.

  8. Dust divertor for a tokamak fusion reactor

    SciTech Connect

    Tang, X Z; Delzanno, G L

    2009-01-01

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

  9. Design of divertor plate and measurements of double-null open divertor plasma in the JFT-2M tokamak

    NASA Astrophysics Data System (ADS)

    Yanagisawa, Ichiro; Shoji, Teruaki; Mori, Masahiro; Odajima, Kazuo; Ohtsuka, Hideo; Suzuki, Norio; Hasegawa, Mitsuru; Ohta, Kanji; Sugihara, Masayoshi; Uesugi, Yoshihiko

    1987-10-01

    The Design of the divertor plate, the results of the computational simulation and the experimental results on the compact diverter of the JFT-2 tokamak are described. Graphite divertor plates have showed a good performance as divertor target materials through divertor discharges. The H-mode plasma and low temperature, high density divertor plasma are obtained. From computational results, this is in the intermediate region between low and high recycling region.

  10. Robotic hand and fingers

    DOEpatents

    Salisbury, Curt Michael; Dullea, Kevin J.

    2017-06-06

    Technologies pertaining to a robotic hand are described herein. The robotic hand includes one or more fingers releasably attached to a robotic hand frame. The fingers can abduct and adduct as well as flex and tense. The fingers are releasably attached to the frame by magnets that allow for the fingers to detach from the frame when excess force is applied to the fingers.

  11. Divertor interferometer diagnostic for ITER

    SciTech Connect

    Brower, D. L.; Deng, B. H.; Ding, W. X.

    2006-10-15

    In the harsh environment of the divertor region in ITER, plasmas spanning a huge density range from 10{sup 19} to 10{sup 22} m{sup -3} are anticipated making measurement of the electron density particularly challenging. For any reasonable wavelength choice, the total phase measured by a conventional two-color interferometer system is always >>2{pi} and therefore subject to fringe counting errors. This problem can be remedied by adding a polarimeter capability whereby the Cotton-Mouton effect is measured or by employing differential interferometry. Using either approach, the total phase is always <<2{pi}. The conceptual design of an interferometer system along with possible wavelength choices will be explored.

  12. Effect of Divertors in NCSX

    NASA Astrophysics Data System (ADS)

    Kaiser, Thomas B.; Hill, David N.

    2004-11-01

    We have used magnetic field data generated by the PIES 3D MHD equilibrium code (M50 coil set) and a new vacuum field code [1] together with the latest numerical model of the first wall [2] to compute wall heat-loading in the National Compact Stellarator Experiment (NCSX). Heat flow is traced by following field lines, with field-line diffusion used to mimic the effect of particle scattering, and the local heat flux estimated from the strike-point density of escaping field lines. This extends our earlier work [3] by including the effect of divertors, whose size, location and configuration are varied to minimize estimated wall damage. Error scaling of the field-line integrator is also presented. 1. Michael Drevlak, MPIPP, Greifswald, Germany, private communication 2. Art Brooks, PPPL, private communication. 3. T. B. Kaiser, et al, BAPPS 48, 304 (2003).

  13. High flux expansion divertor studies in NSTX

    SciTech Connect

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

    2009-06-29

    Projections for high-performance H-mode scenarios in spherical torus (ST)-based devices assume low electron collisionality for increased efficiency of the neutral beam current drive. At lower collisionality (lower density), the mitigation techniques based on induced divertor volumetric power and momentum losses may not be capable of reducing heat and material erosion to acceptable levels in a compact ST divertor. Divertor geometry can also be used to reduce high peak heat and particle fluxes by flaring a scrape-off layer (SOL) flux tube at the divertor plate, and by optimizing the angle at which the flux tube intersects the divertor plate, or reduce heat flow to the divertor by increasing the length of the flux tube. The recently proposed advanced divertor concepts [1, 2] take advantage of these geometry effects. In a high triangularity ST plasma configuration, the magnetic flux expansion at the divertor strike point (SP) is inherently high, leading to a reduction of heat and particle fluxes and a facilitated access to the outer SP detachment, as has been demonstrated recently in NSTX [3]. The natural synergy of the highly-shaped high-performance ST plasmas with beneficial divertor properties motivated a further systematic study of the high flux expansion divertor. The National Spherical Torus Experiment (NSTX) is a mid-sized device with the aspect ratio A = 1.3-1.5 [4]. In NSTX, the graphite tile divertor has an open horizontal plate geometry. The divertor magnetic configuration geometry was systematically changed in an experiment by either (1) changing the distance between the lower divertor X-point and the divertor plate (X-point height h{sub X}), or by (2) keeping the X-point height constant and increasing the outer SP radius. An initial analysis of the former experiment is presented below. Since in the divertor the poloidal field B{sub {theta}} strength is proportional to h{sub X}, the X-point height variation changed the divertor plasma wetted area due to

  14. ARIES-III divertor engineering design

    SciTech Connect

    Wong, C.P.C.; Schultz, K.R.; Cheng, E.T.; Grotz, S.; Hasan, M.A.; Najmabadi, F.; Sharafat, S.; Brooks, J.N.; Ehst, D.A.; Sze, D.K.; 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{sup 2}, a conventional divertor design can be used. The divertor surface is contoured to have a constant heat flux of 5.4 MW/m{sup 2}. The net erosion of the W-surface was found to be negligible at about 0.1 mm/year. After 3 years of operation, the W-3Re alloy ARIES-III divertor can be disposed of as Class A waste. In order to control the prompt dose release at site boundary to less than 200 Rem, isotopic tailoring of the W-alloy will be needed.

  15. Divertor experiment in large helical device

    NASA Astrophysics Data System (ADS)

    Motojima, O.; Ohyabu, N.; Komori, A.; Noda, N.; Yamazaki, K.; Yamada, H.; Sagara, A.; Kubota, Y.; Suzuki, H.; Inoue, N.; Morisaki, T.; Masuzaki, S.; Sakamoto, R.; Matsuoka, K.; Fujiwara, M.; Iiyoshi, A.

    1996-12-01

    This paper describes the major objectives of the LHD divertor experiment which is proposed to produce currentless-steady-state plasmas with high performance and without any current disruption. Since further improvement in confinement is a common and general requirement for fusion research including the LHD project, it is also necessary to develop the edge plasma control techniques and to understand the physical behaviour in the LHD divertor, i.e. the newly developed continuous helical divertor and a local island divertor (LID) concepts. In order to achieve these objectives, there were several key issues in physics and technology, which had to be resolved through careful investigation before the LHD experiment could start. In this paper, we summarize the recent progress of the physics understanding of divertor functions, divertor plasma operation scenarios, and properties of the LHD magnetic field structure in addition to the experimental planning. We also describe the recent result of an LID experiment in the CHS device, which demonstrated the possibility of edge particle and heat control by the LID.

  16. Recent DIII-D divertor research

    SciTech Connect

    Allen, S.L.; Bozek, A.S.; Brooks, N.H.

    1995-07-01

    DIII-D currently operates with a single- or double-null open divertor and graphite walls. Active particle control with a divertor cryopump has demonstrated density control, efficient helium exhaust, and reduction of the inventory of particles in the wall. Gas puffing of D{sub 2} and impurities has demonstrated reduction of the peak divertor beat flux by factors of 3--5 by radiation. A combination of active cryopumping and feedback-controlled D{sub 2} gas puffing has produced similar divertor heat flux reduction with density control. Experiments with neon puffing have shown that the radiation is equally-divided between a localized zone near the X-point and a mantle around the plasma core. The density in these experiments has also been controlled with cryopumping. These experimental results combined with modeling were used to develop the new Radiative Divertor for DIII-D. This is a double-null slot divertor with four cryopumps to provide particle control and neutral shielding for high-triangularity advanced tokamak discharges. UEDGE and DEGAS simulations, benchmarked to experimental data, have been used to optimize the design.

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

    SciTech Connect

    O'NEIL, RC; STAMBAUGH, RD

    2002-06-01

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

  18. Predictive modelling for EAST divertor operation

    NASA Astrophysics Data System (ADS)

    Chen, YiPing

    2011-06-01

    The predictive modelling study of the divertor operation in EAST tokamak [B. Wan et al., Nucl. Fusion 49, 104011 (2009)] with double null (DN) configuration is carried out by using the two-dimensional edge plasma code B2.5-SOLPS5.0 [D. P. Coster, X. Bonnin et al., J. Nucl. Mater. 337-339, 366 (2005)]. The modelling study includes the particle and power balance in the scrape-off-layer (SOL), the operation parameters of plasma density, temperature and plasma heat fluxes at the separatrix, the target plates and the wall, and the effect of the gas puffing, drifts, and vertical target plate on the divertor operation. The fluid model for the edge plasma is applied using the real magnetohydrodynamic (MHD) equilibrium from the MHD equilibrium code EFIT [L. L. Lao et al., Nucl. Fusion 25, 1611 (1985)] and the real divertor geometry in the device. Before EAST tokamak starts its experimental programme of divertor operation, the modelling plays an important role in the design of its experimental programme and the optimization of the divertor operation parameters. Based on the modelling results, EAST divertor can operate over a large wide of plasma parameters with different regimes. For a heating power of 8 MW and an edge density at core-SOL interface Nedge = 0.8 × 10191/m3 and Nedge = 1.3 × 10191/m3, the EAST divertor begins access to the high recycling operation regime at the outer and inner target plates, respectively, where the plasma temperature and the heat fluxes at the target plates decrease. The gas puffing can increase the plasma density at the separatrix and trigger the transition from the high recycling operation into detachment at the target plates. When E × B and B × ▿B drifts are taken into account, the asymmetry of plasma parameters and heat fluxes between up-down SOLs can be found. The vertical target plate in EAST divertor can reduce the peak values of heat fluxes at the target plate and enables detachment at lower plasma density. The divertor with the

  19. Predictive modelling for EAST divertor operation

    SciTech Connect

    Chen Yiping

    2011-06-15

    The predictive modelling study of the divertor operation in EAST tokamak [B. Wan et al., Nucl. Fusion 49, 104011 (2009)] with double null (DN) configuration is carried out by using the two-dimensional edge plasma code B2.5-SOLPS5.0 [D. P. Coster, X. Bonnin et al., J. Nucl. Mater. 337-339, 366 (2005)]. The modelling study includes the particle and power balance in the scrape-off-layer (SOL), the operation parameters of plasma density, temperature and plasma heat fluxes at the separatrix, the target plates and the wall, and the effect of the gas puffing, drifts, and vertical target plate on the divertor operation. The fluid model for the edge plasma is applied using the real magnetohydrodynamic (MHD) equilibrium from the MHD equilibrium code EFIT [L. L. Lao et al., Nucl. Fusion 25, 1611 (1985)] and the real divertor geometry in the device. Before EAST tokamak starts its experimental programme of divertor operation, the modelling plays an important role in the design of its experimental programme and the optimization of the divertor operation parameters. Based on the modelling results, EAST divertor can operate over a large wide of plasma parameters with different regimes. For a heating power of 8 MW and an edge density at core-SOL interface N{sub edge} = 0.8 x 10{sup 19}1/m{sup 3} and N{sub edge} = 1.3 x 10{sup 19}1/m{sup 3}, the EAST divertor begins access to the high recycling operation regime at the outer and inner target plates, respectively, where the plasma temperature and the heat fluxes at the target plates decrease. The gas puffing can increase the plasma density at the separatrix and trigger the transition from the high recycling operation into detachment at the target plates. When E x B and B x {nabla}B drifts are taken into account, the asymmetry of plasma parameters and heat fluxes between up-down SOLs can be found. The vertical target plate in EAST divertor can reduce the peak values of heat fluxes at the target plate and enables detachment at lower

  20. Snowflake divertor configuration studies in National Spherical Torus Experiment

    SciTech Connect

    Soukhanovskii, V. A.; McLean, A. G.; Rognlien, T. D.; Ryutov, D. D.; Umansky, M. V.; Bell, R. E.; Diallo, A.; Gerhardt, S.; Kaye, S.; Kolemen, E.; LeBlanc, B. P.; Menard, J. E.; Paul, S. F.; Podesta, M.; Roquemore, A. L.; Scotti, F.; Battaglia, D.; Bell, M. G.; Gates, D. A.; Kaita, R.; and others

    2012-08-15

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

  1. Snowflake divertor configuration studies for NSTX-Upgrade

    SciTech Connect

    Soukhanovskii, V A

    2011-11-12

    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.

  2. ELM heat flux in the ITER divertor

    SciTech Connect

    Leonard, A.W.; Osborne, T.H.; Hermann, A.; Suttrop, W.; Itami, K.; Lingertat, J.; Loarte, A.

    1998-07-01

    Edge-Localized-Modes (ELMs) have the potential to produce unacceptable levels of erosion of the ITER divertor. Ablation of the carbon divertor target will occur if the surface temperature rises above about 2,500 C. Because a large number of ELMs, {ge}1000, are expected in each discharge it is important that the surface temperature rise due to an individual ELM remain below this threshold. Calculations that have been carried out for the ITER carbon divertor target indicate ablation will occur for ELM energy {ge}0.5MJ/m{sup 2} if it is deposited in 0.1 ms, or 1.2 MJ/m{sup 2} if the deposition time is 1.0 ms. Since {Delta}T{proportional_to}Q{Delta}t{sup {minus}1/2}, an ablation threshold can be estimated at Q{Delta}t{sup {minus}1/2}{approx}45 MJm{sup {minus}2} s{sup {minus}1/2} where Q is the divertor ELM energy density in J-m{sup {minus}2} and {Delta}t is the time in seconds for that deposition. If a significant fraction of ELMs exceed this threshold then an unacceptable level of erosion may take place. The ablation parameter in ITER can be determined by scaling four factors from present experiments: the ELM energy loss from the core plasma, the fraction of ELM energy deposited on the divertor target, the area of the ELM profile onto the target, and the time for the ELM deposition. ELM data from JET, ASDEX-Upgrade, JT-60U, DIII-D and Compass-D have been assembled by the ITER Divertor Modeling and Database expert group into a database for the purpose of predicting these factors for ELMs in the ITER divertor.

  3. A "Snowflake" Divertor and its Properties

    SciTech Connect

    Ryutov, D

    2007-06-21

    Handling the power and particle exhaust in fusion reactors based on tokamaks is a challenging problem [1,2]. To bring the energy flux to the divertor plates to an acceptable level (< 10 MW/m2), it is desirable to significantly increase poloidal flux expansion in the divertor area. Some recent ideas include that of a so-called X divertor [3] and a 'snowflake' divertor [4]. We use an acronym SF to designate the latter. In this paper we concentrate on the SF divertor. The general idea behind this configuration is that, by a proper selection of divertor (poloidal field) coils, one can make the null point of the second, not of the first order as in the standard divertor. The separatrix in the vicinity of the X point then acquires a characteristic hexapole structure (Fig. 1), reminiscent of a snowflake, whence the name. The fact that the field has a second-order null, leads to a significant increase of the flux expansion. It was noted in Ref. [4] that the SF configuration is topologically unstable: if the current in the divertor coils is somewhat higher than the one that provides the SF configuration, it becomes a single-null X-point configuration. Conversely, if the coil current becomes somewhat lower, there appear two separate X-points. To solve this problem, one can operate the divertor at the current by roughly 5% higher than the value needed to create the second-order null. Then, configuration becomes robust enough and the shape of the separatrix does not change significantly if the coil current varies by 2-3%. At the same time, the flux expansion still remained by a factor of {approx}3 larger compared to a 'canonical' divertor. Following Ref. [4], we call this configuration a 'SF-plus' configuration. Specific examples in Ref. [4] were given for simple magnetic geometries The aim of this paper is to demonstrate that the SF concept will also work for a strongly shaped plasma. The other set of issues considered in the present paper relates to the possible presence of

  4. Impurity-induced divertor plasma oscillations

    SciTech Connect

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

    2016-01-07

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

  5. Impurity-induced divertor plasma oscillations

    SciTech Connect

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

    2016-01-15

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

  6. Plasma power recycling at the divertor surface

    SciTech Connect

    Tang, Xian -Zhu; Guo, Zehua

    2016-12-03

    With a divertor made of solid materials like carbon and tungsten, plasma ions are expected to be recycled at the divertor surface with a time-averaged particle recycling coefficient very close to unity in steady-state operation. This means that almost every plasma ion (hydrogen and helium) will be returned to the plasma, mostly as neutrals. The power flux deposited by the plasma on the divertor surface, on the other hand, can have varying recycling characteristics depending on the material choice of the divertor; the run-time atomic composition of the surface, which can be modified by material mix due to impurity migration in the chamber; and the surface morphology change over time. In general, a high-Z–material (such as tungsten) surface tends to reflect light ions and produce stronger power recycling, while a low-Z–material (such as carbon) surface tends to have a larger sticking coefficient for light ions and hence lower power recycling. Here, an explicit constraint on target plasma density and temperature is derived from the truncated bi-Maxwellian sheath model, in relation to the absorbed power load and power recycling coefficient at the divertor surface. Lastly, it is shown that because of the surface recombination energy flux, the attached plasma has a sharper response to power recycling in comparison to a detached plasma.

  7. Impurity-induced divertor plasma oscillations

    DOE PAGES

    Smirnov, R. D.; Kukushkin, A. S.; Krasheninnikov, S. I.; ...

    2016-01-07

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

  8. Plasma power recycling at the divertor surface

    DOE PAGES

    Tang, Xian -Zhu; Guo, Zehua

    2016-12-03

    With a divertor made of solid materials like carbon and tungsten, plasma ions are expected to be recycled at the divertor surface with a time-averaged particle recycling coefficient very close to unity in steady-state operation. This means that almost every plasma ion (hydrogen and helium) will be returned to the plasma, mostly as neutrals. The power flux deposited by the plasma on the divertor surface, on the other hand, can have varying recycling characteristics depending on the material choice of the divertor; the run-time atomic composition of the surface, which can be modified by material mix due to impurity migrationmore » in the chamber; and the surface morphology change over time. In general, a high-Z–material (such as tungsten) surface tends to reflect light ions and produce stronger power recycling, while a low-Z–material (such as carbon) surface tends to have a larger sticking coefficient for light ions and hence lower power recycling. Here, an explicit constraint on target plasma density and temperature is derived from the truncated bi-Maxwellian sheath model, in relation to the absorbed power load and power recycling coefficient at the divertor surface. Lastly, it is shown that because of the surface recombination energy flux, the attached plasma has a sharper response to power recycling in comparison to a detached plasma.« less

  9. Impurity-induced divertor plasma oscillations

    NASA Astrophysics Data System (ADS)

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

    2016-01-01

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

  10. Divertor design for the tokamak physics experiment

    NASA Astrophysics Data System (ADS)

    Hill, D. N.; Braams, B.; Brooks, J. N.; Ruzic, D. N.; Ulrickson, M.; Werley, K. A.; Campbell, R.; Goldston, R.; Kaiser, T.; Neilson, G. H.; Mioduszewski, P.; Rensink, M. E.; Rognlien, T. D.

    1995-04-01

    In this paper we discuss the divertor design for the planned TPX tokamak, which will explore the physics and technology of steady state (1000 s pulses) heat and particle removal in high confinement (up to 4 × L-mode), high beta (up to βN = 5) divertor plasmas sustained by non-inductive current drive. TPX will operate in the double-null divertor configuration, with actively cooled graphite targets forming a deep (0.57 m) slot at the outer strike point. The peak heat flux on the highly tilted (74° from normal) re-entrant divertor plate (tilted to recycle ions back toward the separatrix) will be in the range of 4-6 MW/m 2 with 17.5 MW of auxiliary heating power. The combination of pumping and gas puffing (D 2 plus impurities), along with higher heating power (45 MW maximum) will allow testing of radiative divertor concepts at ITER-like power densities.

  11. A design method of divertor in tokamak reactors

    NASA Astrophysics Data System (ADS)

    Ueda, N.; Itoh, S.-I.; Tanaka, M.; Itoh, K.

    1990-08-01

    Computational method to design the efficient divertor configuration in tokamak reactor is presented. The two dimensional code was developed to analyze the distributions of the plasma and neutral particles for realistic configurations. Using this code, a method to design the efficient divertor configuration is developed. An example of new divertor, which consists of the baffle and fin plates, is analyzed.

  12. Designing divertor targets for uniform power load

    NASA Astrophysics Data System (ADS)

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

    2015-08-01

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

  13. Liquid metal cooled divertor for ARIES

    SciTech Connect

    Muraviev, E.

    1995-01-01

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

  14. Liquid Surface Divertor Designs for Fusion Reactors

    SciTech Connect

    Nygren, R; Rognlien, T; Rensink, M

    2003-11-11

    As part of work in the US on free flowing liquid surfaces facing the plasma, we are studying issues of integrating a liquid surface divertor into a configuration based upon an advanced tokamak (ARIES-RS). The simplest form of such a divertor is to extend the flow of the liquid first wall and avoid introducing any separate fluid streams. A design and some of the issues in design integration are presented for a divertor (and first wall) with the molten salt Flinabe, a mixture of lithium and sodium fluorides. Thermal performance and the interactions with the plasma edge are treated. Sn and Sn-Li have also been considered, although the complicated 3-D MHD flows cannot yet be fully modeled.

  15. Design Integration of Liquid Surface Divertors

    SciTech Connect

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

    2003-11-13

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

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

    NASA Astrophysics Data System (ADS)

    Smith, J. P.; Anderson, P. M.; Baxi, C. B.; Chin, E.; Hollerbach, M. A.; Hyatt, A. W.; Junge, R.; Mahdavi, M. A.; Redler, K.; Reis, E. E.

    1994-10-01

    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(sub 2)) or the core Z(sub 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 (greater than 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.

  17. Current concepts: mallet finger.

    PubMed

    Alla, Sreenivasa R; Deal, Nicole D; Dempsey, Ian J

    2014-06-01

    Loss of the extensor mechanism at the distal interphalangeal (DIP) joint leads to mallet finger also known as baseball finger or drop finger. This can be secondary to tendon substance disruption or to a bony avulsion. Soft tissue mallet finger is the result of a rupture of the extensor tendon in Zone 1, and a bony mallet finger is the result of an avulsion of the extensor tendon from the distal phalanx with a small fragment of bone attached to the avulsed tendon. Mallet finger leads to an imbalance in the distribution of the extensor force between the proximal interphalangeal (PIP) and DIP joints. If left untreated, mallet finger leads to a swan neck deformity from PIP joint hyper extension and DIP joint flexion. Most mallet finger injuries can be managed non-surgically, but occasionally surgery is recommended for either an acute or a chronic mallet finger or for salvage of failed prior treatment.

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

    NASA Astrophysics Data System (ADS)

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

    2017-04-01

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

  19. Divertor-leg instability for finite beta and radially-tilted divertor plate

    NASA Astrophysics Data System (ADS)

    Cohen, R. H.; Ryutov, D. D.

    2004-11-01

    Plasma in the divertor leg may experience a fast instability caused by sheath boundary conditions (BC). Perturbations cannot penetrate beyond the X point because of very strong shearing in its vicinity. Accordingly, this instability could increase cross-field transport in the divertor leg, and thereby reduce the heat load on the divertor plate, without having any appreciable negative effect on core plasma confinement. A way of describing the role of shearing in terms of the surface resistivity attributed to a ``control plane'' below the X point has recently been suggested (Contr. Plasma Phys., v. 44, p. 168, 2004). We use this BC, plus sheath BC at the divertor plate. We include effects of finite beta and of the radial tilt of the divertor plate. We optimize the radial tilt in order to maximize radial transport in divertor legs. We discuss experimental signatures of the instability: i) phase velocity and wave-numbers of the most unstable modes; ii) correlations between fluctuations of various parameters; and iii) the differences between fluctuations in the common and private flux regions.

  20. The tungsten divertor experiment at ASDEX Upgrade

    NASA Astrophysics Data System (ADS)

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

    1996-12-01

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

  1. Divertor design for the Tokamak Physics Experiment

    SciTech Connect

    Hill, D.N.; Braams, B.; Brooks, J.N.

    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--4{times} L-mode), high beta ({beta}{sub N} {ge} 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{sup 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.

  2. Divertor design for the Tokamak Physics Experiment

    NASA Astrophysics Data System (ADS)

    Hill, D. N.; Braams, B.; Brooks, J. N.; Ruzic, D. N.; Ulrickson, M.; Werley, K. A.; Campbell, R.; Goldston, R.; Kaiser, T.; Nellson, G. H.

    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 (2-4 x L-mode), high beta (beta(sub N) greater than or equal to 3) divertor plasmas sustained by non-induct ive 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 deg) from normal) re-entrant (to recycle ions back toward the separatrix) will be in the range of 4-6 MW/sq m 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.

  3. Heat Load on Divertors in NCSX

    NASA Astrophysics Data System (ADS)

    Kaiser, T. B.; Hill, D. N.; Maingi, R.; Monticello, D.; Zarnstorff, M.; Grossman, A.

    2006-10-01

    We have continued our study[1-3] of the effect of divertors in NCSX, using magnetic field data generated by both the PIES and VMEC/MFBE equilibrium codes. Results for comparable equilibria from the two codes agree to within statistical uncertainty. We follow field lines from a surface just outside and conformal with the LCMS until they strike a divertor plate or the first wall, or exceed 1000m in length, with effects of particle scattering mimicked by field-line diffusion. Current candidate divertor designs efficiently collect field lines, allowing fewer than 0.1% to reach the wall. The sensitivity of localized power deposition, assumed to be proportional to the density of field- line strike-points, to adjustments in the divertor configuration is under investigation.1. T.B. Kaiser, et al, Bull. Am. Phys. Soc., 48, paper RP1-20, 2003.2. T.B. Kaiser, et al, Bull. Am. Phys. Soc., 49, paper PP1-73, 2004.3. R. Maingi, et al, EPS Conf. Rome, Italy, paper P5.116, 2006.

  4. Divertor target for magnetic containment device

    DOEpatents

    Luzzi, Jr., Theodore E.

    1982-01-01

    In a plasma containment device of a type having superconducting field coils for magnetically shaping the plasma into approximately the form of a torus, an improved divertor target for removing impurities from a "scrape off" region of the plasma comprises an array of water cooled swirl tubes onto which the scrape off flux is impinged. Impurities reflected from the divertor target are removed from the target region by a conventional vacuum getter system. The swirl tubes are oriented and spaced apart within the divertor region relative to the incident angle of the scrape off flux to cause only one side of each tube to be exposed to the flux to increase the burnout rating of the target. The divertor target plane is oriented relative to the plane of the path of the scrape off flux such that the maximum heat flux onto a swirl tube is less than the tube design flux. The containment device is used to contain the plasma of a tokamak fusion reactor and is applicable to other long pulse plasma containment systems.

  5. Divertor erosion in DIII-D

    SciTech Connect

    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{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.

  6. Divertor Materials Evaluation System (DiMES)

    SciTech Connect

    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 {sup 3}He(d,p) {sup 4}He nuclear reaction. For carbon, these measurements showed peak deuterium areal density of about 8 {times} 10 {sup 18} D/cm{sup 2} in a co-deposited layer about 6 {micro}m 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 {approx} 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.

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

    SciTech Connect

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

    2004-07-15

    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

  8. Flute mode fluctuations in the divertor mirror cell

    SciTech Connect

    Katanuma, I.; Yagi, K.; Nakashima, Y.; Ichimura, M.; Imai, T.

    2010-03-15

    The computer code by reduced magnetohydrodynamic equations were made which can simulate the flute interchange modes (similar to the Rayleigh-Taylor instability) and the instability associated with the presence of nonuniform plasma flows (similar to the Kelvin-Helmholtz instability). This code is applied to a model divertor and the GAMMA10 [M. Inutake et al., Phys. Rev. Lett. 55, 939 (1985)] with divertor in order to investigate the flute modes in these divertor cells. The linear growth rate of the flute instability determined by the nonlocal linear analysis agrees with that in the linear phase of the simulations. There is a stable nonlinear steady state in both divertor cells, but the nonlinear steady state is different between the model divertor and the GAMMA10 with divertor.

  9. Divertor research on the DIII-D tokamak

    SciTech Connect

    Hill, D.N.; Allen, S.L.; Brooks, N.H.

    1994-10-01

    In this paper the authors summarize recent progress on DIII-D in developing techniques for divertor power and particle control relevant to next generation tokamaks such as the proposed ITER and TPX devices. Density control and helium removal by divertor pumping have been demonstrated for the first time in high confinement ELMing H-mode discharges ({tau} {approximately} 2 {times} {tau}{sub ITER-89P}) following installation of a divertor cryopumping system. The peak divertor heat flux in similar H-mode discharges has been reduced through production of a radiating mantle with neon or argon puffing (reductions of 3--5). A number of diagnostics have been added to improve the understanding of the physical processes involved. They are now designing modified double-null divertor structures for DIII-D that will provide improved particle control for high-triangularity VH-mode plasmas while at the same time allowing for gas puffing to reduce the divertor heat flux.

  10. Modeling detachment physics in the NSTX snowflake divertor

    NASA Astrophysics Data System (ADS)

    Meier, E. T.; Soukhanovskii, V. A.; Bell, R. E.; Diallo, A.; Kaita, R.; LeBlanc, B. P.; McLean, A. G.; Podestà, M.; Rognlien, T. D.; Scotti, F.

    2015-08-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2016-10-01

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

  12. Neutral recirculation—the key to control of divertor operation

    NASA Astrophysics Data System (ADS)

    Kukushkin, A. S.; Pacher, H. D.

    2016-12-01

    Interaction of the plasma with neutral gas in the divertor affects virtually all aspects of divertor functionality (power loading of the targets, pumping and fuelling, sustaining the operational conditions of the core plasma). In the course of ITER design development, this interaction has been the subject of intense modelling analysis, supported by experiments on various tokamaks. Neutral gas puffing is found to be the most effective means of divertor control. The results of those studies are summarized and assessed in the paper.

  13. NSTX Plasma Response to Lithium Coated Divertor

    SciTech Connect

    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.

  14. Two-point model for divertor transport

    SciTech Connect

    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.

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

    SciTech Connect

    Boedo, J.A.; Schaffer, M.J.; Maingi, M.; Lasnier, C.J.; Watkins, J.G.

    1999-07-01

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

  16. Divertor bypass in the Alcator C-Mod tokamak

    SciTech Connect

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

    2001-01-01

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

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

    SciTech Connect

    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.

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

    NASA Astrophysics Data System (ADS)

    Bufferand, H.; Ciraolo, G.; Dif-Pradalier, G.; Ghendrih, P.; Tamain, Ph; Marandet, Y.; Serre, E.

    2014-12-01

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

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

    NASA Astrophysics Data System (ADS)

    Schmitz, O.; Becoulet, M.; Cahyna, P.; Evans, T. E.; Feng, Y.; Frerichs, H.; Loarte, A.; Pitts, R. A.; Reiser, D.; Fenstermacher, M. E.; Harting, D.; Kirschner, A.; Kukushkin, A.; Lunt, T.; Saibene, G.; Reiter, D.; Samm, U.; Wiesen, S.

    2016-06-01

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

  20. ADX - Advanced Divertor and RF Tokamak Experiment

    NASA Astrophysics Data System (ADS)

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

    2015-11-01

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

  1. THERMAL HYDRAULIC ANALYSIS OF FIRE DIVERTOR

    SciTech Connect

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

    2000-10-01

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

  2. Island divertor studies on W7-AS

    NASA Astrophysics Data System (ADS)

    Sardei, F.; Feng, Y.; Grigull, P.; Herre, G.; Hildebrandt, D.; Hofmann, J. V.; Kisslinger, J.; Brakel, R.; Das, J.; Geiger, J.; Heinrich, O.; Kühner, G.; Niedermeyer, H.; Reiter, D.; Richter-Glötzl, M.; Runov, A.; Schneider, R.; Stroth, U.; Verbeek, H.; Wagner, F.; Wolf, R.; W7-AS Team; NBI Group

    1997-02-01

    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 overlinene ≥ 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.

  3. A new scaling for divertor detachment

    DOE PAGES

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

    2017-03-29

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

  4. Constrained ripple optimization of Tokamak bundle divertors

    SciTech Connect

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

    1983-02-01

    Magnetic field ripple from a tokamak bundle divertor is localized to a small toroidal sector and must be treated differently from the usual (distributed) toroidal field (TF) coil ripple. Generally, in a tokamak with an unoptimized divertor design, all of the banana-trapped fast ions are quickly lost due to banana drift diffusion or to trapping between the 1/R variation in absolute value vector B ..xi.. B and local field maxima due to the divertor. A computer code has been written to optimize automatically on-axis ripple subject to these constraints, while varying up to nine design parameters. Optimum configurations have low on-axis ripple (<0.2%) so that, now, most banana-trapped fast ions are confined. Only those ions with banana tips near the outside region (absolute value theta < or equal to 45/sup 0/) are lost. However, because finite-sized TF coils have not been used in this study, the flux bundle is not expanded.

  5. Divertor Configurations which Optimize Helium Pumping

    NASA Astrophysics Data System (ADS)

    Strachan, James

    2008-11-01

    Helium accumulation in DT plasmas is often presumed to be one limitation to the fusion power production. The core helium density has an unavoidable central source and a confinement time which tends to be long as is consistent with the required energy confinement times. Any pumping of the helium can only act to reduce the helium recycling. Within that constraint, however, it is still valuable to efficiently pump helium. Helium pumping can be aided by optimal placement of the helium pump in the divertor. The pump should be on the SOL side of the separatrix displaced into the region where the current of impurity particles enters into the divertor and initially strike the target. A numerical example will be given of helium pumping by the ITER divertor. A factor-of-two reduction in core helium densities is possible by optimal pump placement. One difficulty is the need for low temperatures along the targets to prevent their erosion. On ITER, recycled DT near the strike points is hoped to cool this region. The angle between the separatrix and the target is such that recycled neutrals cause ionization, excitation, and dissociation power losses along the target. The ITER solution constrains the choice of pump locations. Alternatively, the strike point cooling can be achieved by local DT (or low Z impurity) injection at the strike point.

  6. Divertor materials evaluation system (DiMES)

    SciTech Connect

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

    1997-12-31

    The mission of the Divertor Materials Evaluation System (DiMES) in DIII-D is to establish an integrated data base from measurements in the divertor of a tokamak in order to address some of the ITER and fusion power reactor plasma material interaction issues. Carbon and metal coatings of Be, W, V, and Mo were exposed to the steady-state outer strike point on DIII-D for 4--18 s. These short exposure times ensure controlled exposure conditions, and the extensive arrays of DIII-D divertor diagnostics provide a well-characterized plasma for modeling efforts. Post-exposure analysis provides a direct measure of surface material erosion rates and the amount of retained deuterium. For carbon, these results match closely with the results of accumulated carbon deposition and erosion, and the corresponding deuterium retention of long term exposure tiles in DIII-D. Under the carbon-contaminated background plasma of DIII-D, metal coatings of Be, V, Mo, and W were exposed to the steady-state outer strike point under ELMing and ELM-free H-mode discharges. The rate of material erosion and deuterium retention were measured. As expected, W shows the lowest erosion rate at 0.1 mm/s and the lowest deuterium uptake of 2 {times} 10{sup 20}/m{sup 2}.

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

    NASA Astrophysics Data System (ADS)

    Kolemen, E.; Allen, S. L.; Bray, B. D.; Fenstermacher, M. E.; Humphreys, D. A.; Hyatt, A. W.; Lasnier, C. J.; Leonard, A. W.; Makowski, M. A.; McLean, A. G.; Maingi, R.; Nazikian, R.; Petrie, T. W.; Soukhanovskii, V. A.; Unterberg, E. A.

    2015-08-01

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

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

    SciTech Connect

    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.

  9. Plasma flow in the DIII-D divertor

    SciTech Connect

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

    1998-07-01

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

  10. Divertor impurity monitor for the International Thermonuclear Experimental Reactor

    NASA Astrophysics Data System (ADS)

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

    1999-01-01

    The divertor impurity monitoring system of the International Thermonuclear Experimental Reactor has been designed. The main functions of this system are to identify impurity species and to measure the two-dimensional distributions of the particle influxes in the divertor plasmas. The wavelength range is 200-1000 nm. The viewing fans are realized by molybdenum mirrors located in the divertor cassette. With additional viewing fans seeing through the gap between the divertor cassettes, the region approximately from the divertor leg to the x point will be observed. The light from the divertor region passes through the quartz windows on the divertor port plug and the cryostat, and goes through the dog-leg optics in the biological shield. Three different type of spectrometers: (i) survey spectrometers for impurity species monitoring, (ii) filter spectrometers for the particle influx measurement with the spatial resolution of 10 mm and the time resolution of 1 ms, and (iii) high dispersion spectrometers for high resolution wavelength measurements are designed. These spectrometers are installed just behind the biological shield (for 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.

  11. Rolling friction robot fingers

    NASA Technical Reports Server (NTRS)

    Vranish, John M. (Inventor)

    1992-01-01

    A low friction, object guidance, and gripping finger device for a robotic end effector on a robotic arm is disclosed, having a pair of robotic fingers each having a finger shaft slideably located on a gripper housing attached to the end effector. Each of the robotic fingers has a roller housing attached to the finger shaft. The roller housing has a ball bearing mounted centering roller located at the center, and a pair of ball bearing mounted clamping rollers located on either side of the centering roller. The object has a recess to engage the centering roller and a number of seating ramps for engaging the clamping rollers. The centering roller acts to position and hold the object symmetrically about the centering roller with respect to the X axis and the clamping rollers act to position and hold the object with respect to the Y and Z axis.

  12. OEDGE Modeling of Divertor Fueling at DIII-D

    NASA Astrophysics Data System (ADS)

    Bray, B. D.; Leonard, A. W.; Elder, J. D.; Stangeby, P. C.

    2015-11-01

    Onion-skin-modeling (OSM) is used to assess the affect of divertor closure on pedestal fueling sources. The OSM includes information from a wide range of diagnostic measurements at DIII-D to constrain the model background plasma for better simulation of neutrals and impurity ions and spectroscopy to compare to the results of the simulation. DIII-D has open lower divertor and closed upper divertor configurations which can be run with similar discharges. Progress toward modeling the pedestal fueling in low density plasmas for these cases will be presented as well as initial comparisons of recent lower single null discharges with the outer leg on the divertor shelf (fully open) and divertor floor (partially open). Work supported by the US DOE under DE-FC02-04ER54698 and DE-AC52-07NA27344.

  13. Super-X divertors and high power density fusion devices

    SciTech Connect

    Valanju, P. M.; Kotschenreuther, M.; Mahajan, S. M.; Canik, J.

    2009-05-15

    The Super-X Divertor (SXD), a robust axisymmetric redesign of the divertor magnetic geometry that can allow a fivefold increase in the core power density of toroidal fusion devices, is presented. With small changes in poloidal coils and currents for standard divertors, the SXD allows the largest divertor plate radius inside toroidal field coils. This increases the plasma-wetted area by 2-3 times over all flux-expansion-only methods (e.g., plate near main X point, plate tilting, X divertor, and snowflake), decreases parallel heat flux and hence plasma temperature at plate, and increases connection length by 2-5 times. Examples of high-power-density fusion devices enabled by SXD are discussed; the most promising near-term device is a 100 MW modular compact fusion neutron source 'battery' small enough to fit inside a conventional fission blanket.

  14. RELAP5 MODEL OF THE DIVERTOR PRIMARY HEAT TRANSFER SYSTEM

    SciTech Connect

    Popov, Emilian L; Yoder Jr, Graydon L; Kim, Seokho H

    2010-08-01

    This report describes the RELAP5 model that has been developed for the divertor primary heat transfer system (PHTS). The model is intended to be used to examine the transient performance of the divertor PHTS and evaluate control schemes necessary to maintain parameters within acceptable limits during transients. Some preliminary results are presented to show the maturity of the model and examine general divertor PHTS transient behavior. The model can be used as a starting point for developing transient modeling capability, including control system modeling, safety evaluations, etc., and is not intended to represent the final divertor PHTS design. Preliminary calculations using the models indicate that during normal pulsed operation, present pressurizer controls may not be sufficient to keep system pressures within their desired range. Additional divertor PHTS and control system design efforts may be required to ensure system pressure fluctuation during normal operation remains within specified limits.

  15. Comparison of ELM heat loads in snowflake and standard divertors

    SciTech Connect

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

    2012-05-08

    An analysis is given of the impact of the tokamak divertor magnetic structure on the temporal and spatial divertor heat flux from edge localized modes (ELMs). Two configurations are studied: the standard divertor where the poloidal magnetic field (B{sub p}) varies linearly with distance (r) from the magnetic null and the snowflake where B{sub p} varies quadratrically with r. Both one and two-dimensional models are used to analyze the effect of the longer magnetic field length between the midplane and the divertor plate for the snowflake that causes a temporal dilation of the ELM divertor heat flux. A second effect discussed is the appearance of a broad region near the null point where the poloidal plasma beta can substantially exceed unity, especially for the snowflake configuration during the ELM; such a condition is likely to drive additional radial ELM transport.

  16. High heat flux experiments of saddle type divertor module

    NASA Astrophysics Data System (ADS)

    Suzuki, Satoshi; Akiba, Masato; Araki, Masanori; Satoh, Kazuyoshi; Yokoyama, Kenji; Dairaku, Masayuki

    1994-09-01

    JAERI has been extensively developing plasma facing components for next tokomak devices. The authors have developed a saddle type divertor module which consists of saddle-shaped armor tiles brazed on metal heat sink. This paper presents the experimental and analytical results of thermal cycling experiments of the saddle type divertor module. The divertor module has unidirectional CFC armor tiles brazed on OFHC copper heat sink. A twisted tape was inserted in the cooling tube to enhance the heat transfer. In the experiments, thermal response of the divertor module was monitored by an infrared camera and thermocouples. The maximum incident heat flux was 24.5 MW/m 2 for a duration of 30 s. No degradation of thermal response was observed during the experiment. As a result, the saddle type divertor module successfully endured at an incident heat flux of over 20 MW/m 2 under steady state conditions for 1000 cycles.

  17. Alternative divertor target concepts for next step fusion devices

    NASA Astrophysics Data System (ADS)

    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.

  18. Magnetic configuration flexibility of snowflake divertor for HL-2M [Analysis of snowflake divertor configurations for HL-2M

    SciTech Connect

    Zheng, G. Y.; Xu, X. Q.; Ryutov, D. D.; Pan, Y. D.; Xia, T. Y.

    2014-07-09

    HL-2M (Li, 2013 [1]) is a tokamak device that is under construction. Based on the magnetic coils design of HL-2M, four kinds of divertor configurations are calculated by CORSICA code (Pearlstein et al., 2001 [2]) with the same main plasma parameters, which are standard divertor, exact snowflake divertor, snowflake-plus divertor and snowflake-minus divertor configurations. The potential properties of these divertors are analyzed and presented in this paper: low poloidal field area around X-point, connection length from outside mid-plane to the primary X-point, target plate design and magnetic field shear. The results show that the snowflake configurations not only can reduce the heat load at divertor target plates, but also may improve the magneto-hydrodynamic stability by stronger magnetic shear at the edge. Furthermore, a new divertor configuration, named “tripod divertor”, is designed by adjusting the positions of the two X-points according to plasma parameters and magnetic coils current of HL-2M.

  19. Magnetic configuration flexibility of snowflake divertor for HL-2M [Analysis of snowflake divertor configurations for HL-2M

    DOE PAGES

    Zheng, G. Y.; Xu, X. Q.; Ryutov, D. D.; ...

    2014-07-09

    HL-2M (Li, 2013 [1]) is a tokamak device that is under construction. Based on the magnetic coils design of HL-2M, four kinds of divertor configurations are calculated by CORSICA code (Pearlstein et al., 2001 [2]) with the same main plasma parameters, which are standard divertor, exact snowflake divertor, snowflake-plus divertor and snowflake-minus divertor configurations. The potential properties of these divertors are analyzed and presented in this paper: low poloidal field area around X-point, connection length from outside mid-plane to the primary X-point, target plate design and magnetic field shear. The results show that the snowflake configurations not only can reducemore » the heat load at divertor target plates, but also may improve the magneto-hydrodynamic stability by stronger magnetic shear at the edge. Furthermore, a new divertor configuration, named “tripod divertor”, is designed by adjusting the positions of the two X-points according to plasma parameters and magnetic coils current of HL-2M.« less

  20. Divertor for use in fusion reactors

    DOEpatents

    Christensen, Uffe R.

    1979-01-01

    A poloidal divertor for a toroidal plasma column ring having a set of poloidal coils co-axial with the plasma ring for providing a space for a thick shielding blanket close to the plasma along the entire length of the plasma ring cross section and all the way around the axis of rotation of the plasma ring. The poloidal coils of this invention also provide a stagnation point on the inside of the toroidal plasma column ring, gently curving field lines for vertical stability, an initial plasma current, and the shaping of the field lines of a separatrix up and around the shielding blanket.

  1. Liquid-helium-cooled Michelson interferometer

    NASA Technical Reports Server (NTRS)

    Augason, G. C.; Young, N.

    1972-01-01

    Interferometer serves as a rocket-flight spectrometer for examination of the far infrared emission spectra of astronomical objects. The double beam interferometer is readily adapted to make spectral scans and for use as a detector of discrete line emissions.

  2. Multiple Fingers - One Gestalt.

    PubMed

    Lezkan, Alexandra; Manuel, Steven G; Colgate, J Edward; Klatzky, Roberta L; Peshkin, Michael A; Drewing, Knut

    2016-01-01

    The Gestalt theory of perception offered principles by which distributed visual sensations are combined into a structured experience ("Gestalt"). We demonstrate conditions whereby haptic sensations at two fingertips are integrated in the perception of a single object. When virtual bumps were presented simultaneously to the right hand's thumb and index finger during lateral arm movements, participants reported perceiving a single bump. A discrimination task measured the bump's perceived location and perceptual reliability (assessed by differential thresholds) for four finger configurations, which varied in their adherence to the Gestalt principles of proximity (small versus large finger separation) and synchrony (virtual spring to link movements of the two fingers versus no spring). According to models of integration, reliability should increase with the degree to which multi-finger cues integrate into a unified percept. Differential thresholds were smaller in the virtual-spring condition (synchrony) than when fingers were unlinked. Additionally, in the condition with reduced synchrony, greater proximity led to lower differential thresholds. Thus, with greater adherence to Gestalt principles, thresholds approached values predicted for optimal integration. We conclude that the Gestalt principles of synchrony and proximity apply to haptic perception of surface properties and that these principles can interact to promote multi-finger integration.

  3. Osseointegrated finger prostheses.

    PubMed

    Doppen, P; Solomons, M; Kritzinger, S

    2009-02-01

    Amputation of a digit can lead to functional and psychological problems and patients can benefit from digital prostheses. Unfortunately, standard prostheses are often unstable, particularly when fitted over short amputation stumps. Prosthesis fixation by osseointegration is widely used in oral and extraoral applications and may help avoid the problem of instability. This paper reports the results of four patients with five finger amputations who were treated with osseointegrated implants to attach finger prostheses. One implant failed to osseointegrate and the procedure was abandoned. Three patients were successfully treated to completion of three finger prostheses and are extremely satisfied with their outcomes, both cosmetically and functionally, with osseoperception reported by all three patients.

  4. Extinguishing ELMs in detached radiative divertor plasmas

    NASA Astrophysics Data System (ADS)

    Pigarov, Alexander; Krasheninnikov, Sergei; Rognlien, Thomas

    2016-10-01

    In order to avoid deleterious effects of ELMs on PFCs in next-step fusion devices it has been suggested to operate with small-sized ELMs naturally extinguishing in the divertor. Our modeling effort is focusing at extinguishing type-I ELMs: conditions for expelled plasma dissipation; efficiency of ELM power handling by detached radiative divertors; and the ELM impact on detachment state. Here time-dependent modeling of a sequence of many ELMs was performed with 2-D edge plasma transport code UEDGE-MB-W which incorporates the Macro-Blob (MB) approach to simulate non-diffusive filamentary transport and various ``Wall'' (W) models for time-dependent hydrogen wall inventory and recycling. Three cases were modeled, in which extinguishing ELMs are achieved due to: (i) intrinsic impurities via graphite sputtering, (ii) extrinsic impurity gas puff (Ne), and (iii) =(i) +(ii). For each case, we performed a series of UEDGE-MB-W runs scanning the deuterium and impurity inventories, pedestal losses and ELM frequency. Temporal variations of the degree of detachment, ionization front shape, recombination sink strength, radiated fraction, peak power loads, OSP, impurity charge states, and in/out asymmetries were analyzed. We discuss the onset of extinguishing ELMs, conditions for not burning through and enhanced plasma recombination as functions of scanned parameters. Efficiencies of intrinsic and extrinsic impurities in ELM extinguishing are compared.

  5. Guidance of the divertor channel outside the main coil system for heliotron/torsatron devices

    NASA Astrophysics Data System (ADS)

    Takase, H.; Ohyabu, N.

    1995-02-01

    A divertor magnetic configuration is proposed that significantly reduces heat load on the divertor plates in heliotron/torsatron devices. The proposed configuration utilizes an octupole field for guiding the divertor channels to a remote area outside the main coil system, where the magnetic field is weak. This allows a significant reduction of the heat load due to expansion of the divertor channels as well as substantially easier access to the divertor plates for maintenance, the key requirements for toroidal fusion reactor designs

  6. Snowflake divertor experiments in the DIII-D, NSTX, and NSTX-U tokamaks aimed at the development of the divertor power exhaust solution

    DOE PAGES

    Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; ...

    2016-11-16

    Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in the future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment was performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large Edge Localized Modes (ELMs). However, a stable partial detachment ofmore » the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3-5, and peak ELM heat flux was reduced by up to 80% (see standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower ne, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D2-seeded SF divertor at PSOL = 3 - 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multifluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected PSOL ≃ 9 MW case. Furthermore, the radiative SF divertor with carbon impurity provides a wider ne operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar qpeak reduction factors (see standard divertor).« less

  7. Snowflake divertor experiments in the DIII-D, NSTX, and NSTX-U tokamaks aimed at the development of the divertor power exhaust solution

    SciTech Connect

    Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; Lasnier, C. J.; Makowski, M. A.; McLean, A. G.; Meier, E. T.; Meyer, W. H.; Rognlien, T. D.; Ryutov, D. D.; Scotti, F.; Kolemen, E.; Bell, R. E.; Diallo, A.; Gerhardt, S.; Kaita, R.; Kaye, S.; LeBlanc, B. P.; Maingi, R.; Menard, J. E.; Podesta, M.; Roquemore, A. L.; Groebner, R. J.; Hyatt, A. W.; Leonard, A. W.; Osborne, T. H.; Petrie, T. W.; Ahn, J. -W.; Raman, R.; Watkins, J. G.

    2016-11-16

    Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in the future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment was performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large Edge Localized Modes (ELMs). However, a stable partial detachment of the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3-5, and peak ELM heat flux was reduced by up to 80% (see standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower ne, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D2-seeded SF divertor at PSOL = 3 - 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multifluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected PSOL ≃ 9 MW case. Furthermore, the radiative SF divertor with carbon impurity provides a wider ne operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar qpeak reduction factors (see standard divertor).

  8. A super-cusp divertor configuration for tokamaks

    SciTech Connect

    Ryutov, D. D.

    2015-08-26

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

  9. A super-cusp divertor configuration for tokamaks

    DOE PAGES

    Ryutov, D. D.

    2015-08-26

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

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

    NASA Astrophysics Data System (ADS)

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

    2015-11-01

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

  11. Modeling of extinguishing ELMs in detached divertor plasmas

    NASA Astrophysics Data System (ADS)

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

    2015-11-01

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

  12. Divertor Optimization via Control at DIII-D

    NASA Astrophysics Data System (ADS)

    Kolemen, E.; Allen, S. L.; Makowski, M. A.; Soukhanovskii, V. A.; Bray, B. D.; Humphreys, D. A.; Johnson, R.; Leonard, A. W.; Liu, C.; Penaflor, B. G.; Petrie, T. W.; Eldon, D.; McLean, A. G.; Unterberg, E. A.

    2014-10-01

    DIII-D divertor performance and heat-handling capabilities are optimized using advanced control techniques. The world's first real-time snowflake divertor detection and control system was implemented on DIII-D in order to stabilize and optimize this configuration. A new control system was implemented to regulate and study detachment and radiation, since future fusion reactors will require detached or partially detached plasmas to achieve acceptable divertor target heat fluxes. The algorithm regulates the D2 and impurity gas injection level by using the divertor temperature measurements from real-time Thomson diagnostics to compute the detachment level, and the real-time bolometer diagnostics to determine core and divertor radiation. This control allows the optimization of the detachment and radiation from the core and the divertor to achieve high core performance compatible with reduced heat-flux to the divertor. Work supported by the US DOE under DE-AC02-09CH11466, DE-AC52-07NA27344, DE-FC02-04ER54698 and DE-AC05-00OR22725.

  13. A super-cusp divertor configuration for tokamaks

    NASA Astrophysics Data System (ADS)

    Ryutov, D. D.

    2015-10-01

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

  14. Finger Foods for Babies

    MedlinePlus

    ... textures. No longer are baby purees and mushy cereals the only things on the menu. By the ... ll still be helping out by spoon-feeding cereal and other important dietary elements. Encouraging finger feeding ...

  15. Nickel transfer by fingers.

    PubMed

    Isnardo, D; Vidal, J; Panyella, D; Vilaplana, J

    2015-06-01

    We investigated fingers as a potential source of nickel transfer to the face in patients with allergic contact dermatitis to nickel and a history of facial dermatitis. Samples were collected from the fingers and cheeks of volunteers using the stripping method with standard adhesive tape, and nickel levels were quantified using mass spectrometry. Fingers and cheeks of individuals who had handled coins were both positive for nickel, with levels ranging from 14.67 to 58.64 ppm and 1.28 to 8.52 ppm, respectively. The levels in a control group were considerably and significantly lower. Transfer of nickel from a person's fingers to their face after handling a nickel-containing object could explain the presence of facial dermatitis in patients with nickel hypersensitivity.

  16. Hand and Finger Exercises

    MedlinePlus

    ... each fingertip. Repeat ____ times for ____ seconds.  Bend the end joint of your finger, keeping the base and middle joints straight. Hold this position. Relax and then straighten the end joint. Hold this position. Repeat ____ times for ____ seconds.  ...

  17. Simulations of NSTX with a Liquid Lithium Divertor Module

    SciTech Connect

    D. P. Stotler, R. Maingi, H.W. Kugel, A. Yu. Pigarov, T.D. Rognlien, V.A. Soukhanovskii

    2008-07-08

    The UEDGE edge plasma transport code is used to model the effect of the reduced recycling provided by the Liquid Lithium Divertor (LLD) module that will be installed in NSTX. UEDGE's transport coefficients are calibrated against an existing NSTX shot using midplane and divertor diagnostic data. The LLD is then incorporated into the simulations as a reduction in the recycling coefficient over the outer divertor. Heat transfer calculations performed using the resulting heat flux profiles indicate that lithium evaporation will be negligible for pulse lengths < 2 s at low (~ 2 MW) input power. At high input power (~ 7 MW), the pulse length may have to be restricted.

  18. Simulations of NSTX with a Liquid Lithium Divertor Module

    SciTech Connect

    Stotler, D. P.; Maingi, R.; Zakharov, L. E.; Kugel, H. W.; Pigarov, A. Yu.; Rognlien, T. D.; Soukhanovskii, V. A.

    2010-02-18

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

  19. Two-chamber model for divertors with plasma recycling

    SciTech Connect

    Langer, W.D.; Singer, C.E.

    1984-11-01

    To model particle and heat loss terms at the edge of a tokamak with a divertor or pumped limiter, a simple two-chamber formulation of the scrapeoff has been constructed by integrating the fluid equations, including sources, along open field lines. The model is then solved for a wide range of density and temperature conditions in the scrapeoff, using geometrical parameters typical of the PDX poloidal divertor. The solutions characterize four divertor operating conditions for beam-heated plasmas: plugged, unplugged, blowthrough, and blowback.

  20. Disruption characteristics in PDX with limiter and divertor discharges

    SciTech Connect

    Couture, P.; McGuire, K.

    1986-09-01

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

  1. Tendon Driven Finger Actuation System

    NASA Technical Reports Server (NTRS)

    Ihrke, Chris A. (Inventor); Reich, David M. (Inventor); Bridgwater, Lyndon (Inventor); Linn, Douglas Martin (Inventor); Askew, Scott R. (Inventor); Diftler, Myron A. (Inventor); Platt, Robert (Inventor); Hargrave, Brian (Inventor); Valvo, Michael C. (Inventor); Abdallah, Muhammad E. (Inventor); Permenter, Frank Noble (Inventor); Mehling, Joshua S. (Inventor)

    2013-01-01

    A humanoid robot includes a robotic hand having at least one finger. An actuation system for the robotic finger includes an actuator assembly which is supported by the robot and is spaced apart from the finger. A tendon extends from the actuator assembly to the at least one finger and ends in a tendon terminator. The actuator assembly is operable to actuate the tendon to move the tendon terminator and, thus, the finger.

  2. Finger and toenail onycholysis.

    PubMed

    Zaias, N; Escovar, S X; Zaiac, M N

    2015-05-01

    Onycholysis - the separation of the nail plate from the nail bed occurs in fingers and toenails. It is diagnosed by the whitish appearance of the separated nail plate from the nail bed. In fingers, the majority is caused by trauma, manicuring, occupational or self-induced behavior. The most common disease producing fingernail onycholysis is psoriasis and pustular psoriasis. Phototoxic dermatitis, due to drugs can also produce finger onycholysis. Once the separation occurs, the environmental flora sets up temporary colonization in the available space. Finger onycholysis is most common in women. Candida albicans is often recovered from the onycholytic space. Many reports, want to associate the yeast as cause and effect, but the data are lacking and the treatment of the candida does not improve finger onycholysis. A reasonable explanation for the frequent isolation of Candida and Pseudomonas in fingernail onycholysis in women, is the close proximity the fingers have to the vaginal and gastrointestinal tract. Fifty per cent of humans harbour C. albicans in the GI tract and it is frequently carried to the vagina during hygienic practices. Finger onycholysis is best treated by drying the nail 'lytic' area with a hair blower, since all colonizing biota are moisture loving and perish in a dry environment. Toenail onycholysis has a very different etiology. It is mechanical, the result of pressure on the toes from the closed shoes, while walking, because of the ubiquitous uneven flat feet producing an asymmetric gait with more pressure on the foot with the flatter sole. © 2014 European Academy of Dermatology and Venereology.

  3. Divertor IR thermography on Alcator C-Mod

    SciTech Connect

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

    2010-10-15

    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 deg. toroidal sector has been given a 2 deg. 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.

  4. Divertor IR thermography on Alcator C-Moda)

    NASA Astrophysics Data System (ADS)

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

    2010-10-01

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

  5. Divertor IR thermography on Alcator C-Mod.

    PubMed

    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.

  6. Beryllium accumulation at the inner divertor of JET

    NASA Astrophysics Data System (ADS)

    Likonen, J.; Vainonen-Ahlgren, E.; Coad, J. P.; Zilliacus, R.; Renvall, T.; Hole, D. E.; Rubel, M.; Arstila, K.; Matthews, G. F.; Stamp, M.; JET-EFDA Contributors

    2005-03-01

    MkIIGB divertor tiles exposed in JET for the 1998-2001 and 1999-2001 campaigns have been used to assess the amount of beryllium and carbon deposited at the inner divertor wall. Total amount of Be at the inner divertor tiles was determined and integrated toroidally. Results were compared with data obtained with optical spectroscopy and good agreement was obtained. The amount of deposited C was computed from the amount of deposited Be assuming that the Be/C ratio arriving in the divertor is the same as the Be/C ratio in the main chamber. On the basis of this analysis we would expect there to be ˜0.4 kg of C deposited. This gives an average C deposition rate lower than during the MkIIA phase.

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

    NASA Astrophysics Data System (ADS)

    Leonard, A. W.; Makowski, M. A.; McLean, A. G.; Osborne, T. H.; Snyder, P. B.

    2015-08-01

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

  8. Diagnostics for the DIII-D radiative divertor

    SciTech Connect

    Nilson, D.G.; Brooks, N.H.; Smith, J.P.; Snider, R.T.

    1995-10-01

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

  9. Beryllium flux distribution and layer deposition in the ITER divertor

    NASA Astrophysics Data System (ADS)

    Schmid, K.

    2008-10-01

    The deposition of Be eroded from the main chamber wall on the W surfaces in the ITER divertor could result in the formation of Be rich Be/W mixed layers with a low melting temperature compared with pure W. To predict whether or not these layers form the Be flux distribution in the ITER divertor is required. This paper presents the results of a combination of plasma transport with erosion/deposition simulations that allow one to calculate both the Be flux distribution and the Be layer deposition in the ITER divertor. This model includes the Be source due to Be erosion in the main chamber and the deposition, re-erosion and re-deposition of Be in the ITER divertor. The calculations show that the fraction of Be in the incident particle flux in the divertor ranges from ≈10-3 to ≈5% with a pronounced inner-outer divertor asymmetry. The flux fractions in the inner divertor are on average ten times higher than in the outer divertor. Thick Be layers only form at the inner strike point and the dome baffles. The highest Be layer growth rate is found to be 1.0 nm s-1. Despite the Be deposition the formation of Be rich Be/W mixed layers is not to be expected in ITER. The expected surface temperature at these locations during steady-state operation is too low as to result in Be diffusion into W and thus Be/W mixed layers cannot form. The paper also discusses the influence of off normal events such as ELMs or VDEs on the formation of Be/W mixed layers.

  10. Turbulent Simulations of Divertor Detachment Based On BOUT + + Framework

    NASA Astrophysics Data System (ADS)

    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.

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

    SciTech Connect

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

    2010-01-01

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

  12. Divertor conditions near double null in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Brunner, Dan; Labombard, Brian; Kuang, Adam; Terry, Jim; Mumgaard, Bob; Wolfe, Steve

    2016-10-01

    Many tokamak reactor designs utilize a double-null equilibrium for the boundary plasma because of the expected benefits of heat flux sharing between the two outer divertor leg as well as the attractiveness of the high-field side scrape-off layer plasma in double-null for RF actuators. However, there has been very little reported on boundary plasma conditions near double null, especially at the divertor plate. And, due to the narrow boundary plasma width, there is concern of the precision to which a double-null equilibrium must be controlled to maintain divertor heat flux sharing. To this end, a series of experiments were performed varying the magnetic balance around double null. The magnetic balance between the two nulls was scanned shot-to-shot in L-, I-, and H-mode plasmas. In addition, current and density scans were performed in L-mode plasmas. Results will be presented for relative balances of divertor particle and energy fluxes to the four divertors (inboard/outboard, upper/lower) as well as the sensitivity of changes in divertor conditions to the magnetic balance. Supported by USDoE Award DE-FC02-99ER54512.

  13. Divertor research on the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

    Hill, D. N.; Allen, S. L.; Brooks, N. H.; Buchenauer, D.; Cuthbertson, J. W.; Evans, T. E.; Fenstermacher, M. E.; Ghendrih, Ph.; Hillis, D. L.; Hogan, J. T.

    1994-10-01

    In this paper the authors summarize recent progress on DIII-D in developing techniques for divertor power and particle control relevant to next generation tokamaks such as the proposed ITER and TPX devices. Density control and helium removal by divertor pumping have been demonstrated for the first time in high confinement ELMing H-mode discharges (tau is approximately 2 times tau(sub ITER-89P)) following installation of a divertor cryopumping system. The peak divertor heat flux in similar H-mode discharges has been reduced through production of a radiating mantle with neon or argon puffing (reductions of 3-5). A number of diagnostics have been added to improve the understanding of the physical processes involved. They are now designing modified double-null divertor structures for DIII-D that will provide improved particle control for high-triangularity VH-mode plasmas while at the same time allowing for gas puffing to reduce the divertor heat flux.

  14. The Magnetic Field Structure of a Snowflake Divertor

    SciTech Connect

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

    2008-05-30

    The snowflake divertor exploits a tokamak geometry in which the poloidal magnetic field null approaches second order; the name stems from the characteristic hexagonal, snowflake-like, shape of the separatrix for an exact second-order null. The proximity of the poloidal field structure to that of a second-order null substantially modifies edge magnetic properties compared to the standard X-point geometry; this, in turn, affects the edge plasma behavior. Modifications include: (1) The flux expansion near the null-point becomes 2-3 times larger; (2) The connection length between the equatorial plane and divertor plate significantly increases; (3) Magnetic shear just inside the separatrix becomes much larger; and (4) In the open-field-line region, the squeezing of the flux-tubes near the null-point increases, thereby causing stronger decoupling of the plasma turbulence in the divertor legs and in the main SOL. These effects can be used to reduce the power load on the divertor plates and/or to suppress the 'bursty' component of the heat flux. It is emphasized that the snowflake divertor can be created by a relatively simple set of poloidal field coils situated beyond the toroidal field coils. Analysis of the robustness of the proposed divertor configuration with respect to changes of the plasma current distribution is presented and it is concluded that, even if the null is close to the second order, the configuration is quite robust.

  15. Snowflake Divertor Configuration Studies in DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

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

    2014-10-01

    Recent DIII-D studies show that the snowflake (SF) divertor enables significant manipulation of divertor heat transport for power exhaust in attached and radiative divertor conditions, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Results include: 1) Increased scrape-off layer (SOL) width suggesting enhanced divertor heat transport; 2) Direct measurements of divertor null-region poloidal beta βp >> 1 in support of the theoretically proposed instability mechanism leading to fast convective plasma redistribution, especially efficient during ELMs, and contribution to 1); 3) Weak effect on pedestal profile and stability resulting in essentially unchanged ELM regime; 4) Reduction of Type-I ELM energy loss; 5) In radiative SF divertor regimes with D2 seeding, a significant reduction of peak heat fluxes between and during ELMs, as in standard H-modes. Work supported by the US Department of Energy under DE-AC52-07NA27344, DE-AC02-09CH11466, DE-FC02-04ER54698, and DE-AC04-94AL85000.

  16. Development of a radiative divertor for DIII-D

    SciTech Connect

    Allen, S.L.; Brooks, N.H.; Campbell, R.B.

    1994-07-01

    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 ({approximately}10 cm diameter) radiation zone which results in substantial reduction (3--10) in the divertor heat flux while {delta}{sub E} remains {approximately}2 times ITER-89P scaling. However, ne increases with D{sub 2} puffing, and Z{sub eff} increases with neon puffing. Divertor structures are required to minimize the effects on the core plasma. The UEDGE fluid code, benchmarked with DIII-D data, and the DEGAS neutrals transport code are used to estimate the effectiveness of divertor configurations; slots reduce the core ionization more than baffles. The overall divertor shape is set by confinement studies which indicate that high triangularity ({delta} {approximately}0.8) is important for high {tau}{sub E} VH-modes. Results from engineering feasibility studies, including diagnostic access, will be presented.

  17. Multi-fingered robotic hand

    NASA Technical Reports Server (NTRS)

    Ruoff, Carl F. (Inventor); Salisbury, Kenneth, Jr. (Inventor)

    1990-01-01

    A robotic hand is presented having a plurality of fingers, each having a plurality of joints pivotally connected one to the other. Actuators are connected at one end to an actuating and control mechanism mounted remotely from the hand and at the other end to the joints of the fingers for manipulating the fingers and passing externally of the robot manipulating arm in between the hand and the actuating and control mechanism. The fingers include pulleys to route the actuators within the fingers. Cable tension sensing structure mounted on a portion of the hand are disclosed, as is covering of the tip of each finger with a resilient and pliable friction enhancing surface.

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

    DOE PAGES

    Rognlien, Thomas D.; McLean, Adam G.; Fenstermacher, Max E.; ...

    2017-01-27

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

  19. Developing snowflake divertor physics basis in the DIII-D, NSTX and NSTX-U tokamaks aimed at the divertor power exhaust solution [Snowflake divertor experiments in the DIII-D, NSTX and NSTX-U tokamaks aimed at the development of the divertor power exhaust solution

    DOE PAGES

    Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; ...

    2016-06-02

    Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment were performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large ELMs. However, a stable partial detachment of the outer strike pointmore » was also achieved where inter-ELM peak heat flux was reduced by factors 3-5, and peak ELM heat flux was reduced by up to 80% (cf. standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower ne, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D2-seeded SF divertor at PSOL = 3 - 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multi-fluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected PSOL ≃9 MW case. In conclusion, the radiative SF divertor with carbon impurity provides a wider ne operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar qpeak reduction factors (cf. standard divertor).« less

  20. Three-Fingered Robot Hand

    NASA Technical Reports Server (NTRS)

    Ruoff, C. F.; Salisbury, J. K.

    1984-01-01

    Mechanical joints and tendons resemble human hand. Robot hand has three "human-like" fingers. "Thumb" at top. Rounded tips of fingers covered with resilient material provides high friction for griping. Hand potential as prosthesis for humans.

  1. Finger agnosia in Alzheimer disease.

    PubMed

    Shenal, Brian V; Jackson, Melissa D; Crucian, Gregory P; Heilman, Kenneth M

    2006-12-01

    The purpose of this study was to learn if a deficit of finger naming (finger agnosia or anomia) is a sensitive test for Alzheimer disease (AD) and the best means of testing for finger agnosia. The subjects were 38 patients with AD and 10 matched normal controls. All subjects were asked to name the thumb, index, and pinky fingers. No control subject had trouble naming any of these fingers, but 37% of the AD subjects did. When AD patients had difficulty with finger naming, they always had trouble naming the index finger. In the absence of stroke, the inability to name the index finger seems as an indicator of dementia. Although brief, this test is not extremely sensitive test for AD.

  2. Three-Fingered Robot Hand

    NASA Technical Reports Server (NTRS)

    Ruoff, C. F.; Salisbury, J. K.

    1984-01-01

    Mechanical joints and tendons resemble human hand. Robot hand has three "human-like" fingers. "Thumb" at top. Rounded tips of fingers covered with resilient material provides high friction for griping. Hand potential as prosthesis for humans.

  3. Spiral viscous fingering.

    NASA Astrophysics Data System (ADS)

    Nagatsu, Yuichiro; Hayashi, Atsushi; Kato, Yoshihito; Tada, Yutaka

    2006-11-01

    When a less-viscous fluid displaces a more-viscous fluid in a radial Hele-Shaw cell, viscous fingering pattern is believed to develop in a radial direction. We performed experiments on viscous fingering in a radial Hele-Shaw cell when a polymer solution, a sodium polyacrylate (SPA) solution is used as the more-viscous fluid and the trivalent iron (Fe^3+) solution is as the less-viscous fluid. The experiment was done by varying the concentration of Fe^3+, cFe3+. We have found that viscous fingering pattern develops spirally when cFe3+ is larger than a threshold value, while the pattern develops in a radial direction for small cFe3+. We confirmed from different experiments that an instantaneous chemical reaction takes place between SPA solution and Fe^3+ solution. The chemical reaction produces precipitation and significantly reduces the viscosity of the SPA solution. The quantity of the precipitation is increased with cFe3+. We will make a discussion on the relationship between the formation of spiral viscous fingering and the chemical reaction taking place between the two fluids.

  4. Finger Lakes LPG

    EPA Pesticide Factsheets

    Finger Lakes LPG Storage, LLC; Two Brush Creek Blvd, Suite 200; Kansas City; Missouri 64112 (Applicant) has applied to the U.S. Environmental Protection Agency (EPA) under the provisions of the Safe Drinking Water Act, 42 U.S.C. 300f et. seq (the Act), for

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

    NASA Astrophysics Data System (ADS)

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

    2016-02-01

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

  6. Model for particle balance in pumped divertors (pre-VORTEX)

    SciTech Connect

    Hogan, J.T.

    1990-08-01

    An internally consistent model for particle transport in an open divertor geometry has been developed. Embodied in a new code, pre-VORTEX, the model couples the particle balance in the plasma core, the scrape-off layer, the open divertor channels, and the vacuum'' regions. This mutual coupling is particularly important in determining the conditions required for high recycling in the divertor. The plasma core is considered to have a relatively quiescent core region and a less well confined edge-localized mode''(ELM) region. The scrape-off layer is modeled with one-dimensional parallel and perpendicular transport. A two-point divertor channel model is used; it is similar to previous models, but with the addition of new physical processes: hydrogen charge exchange, impurity thermal charge exchange, and flux-limited parallel transport. Wall recycling data are required to describe the differing recycling properties of the wall regions and the divertor plates. Given local plasma diffusivities and wall recycling properties, the model predicts the volume-averaged density and global particle confinement time. The input data are uncertain, and a major use for the model is to permit comparison with data. The final model, VORTEX, is intended for application to the analysis of divertor confinement experiments; it is coupled to a one-and-one-half--dimensional transport code and uses detailed geometric input from equilibrium fitting codes, experimentally measured core profiles, and such parameters as can be measured in the scrape-off layer. The pre-VORTEX model is compared as a stand-alone code with typical data from the DIII-D experiment and applied to the proposed DIII-D Advanced Divertor Project.

  7. Variation of Particle Control with Changes in Divertor Geometry

    SciTech Connect

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

    2004-10-18

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

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

    SciTech Connect

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

    1985-01-01

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

  9. X-Ray Exam: Finger

    MedlinePlus

    ... Habits for TV, Video Games, and the Internet X-Ray Exam: Finger KidsHealth > For Parents > X-Ray Exam: Finger Print A A A What's in ... You Have Questions What It Is A finger X-ray is a safe and painless test that uses ...

  10. Overview of the DIII-D Divertor Tungsten Rings Campaign

    NASA Astrophysics Data System (ADS)

    Unterberg, E. A.; Thomas, D. M.; Petrie, T. W.; Abrams, T.; Garofalo, A. M.; Stangeby, P. C.; Rudakov, D. L.; Schmitz, O.; Grierson, B. A.; Victor, B.

    2016-10-01

    Experiments have recently been carried out with toroidal arrays of W-coated metal inserts at two distinct locations in the lower divertor region. The purpose of the experiments is to determine the high-Z divertor erosion and migration, and its effect on core contamination in high performance, ELM-y H-mode, tokamak discharges in a mixed-material, i.e. C and W, environment. The experiments focused on characterizing the sputtering source from each location, the SOL transport of W, and the subsequent impact on core performance. A wide range of ELM-y conditions was studied, including ELM controlled and ELM-free regimes, to determine the importance of the divertor strike point position relative to W sources in these various regimes. The W penetration efficiency was characterized by using a far-SOL collector probe related to core W density. Correlations between source strength (as measured by W-I spectroscopy) relative to the distance of the strikepoint to each W array, the divertor target magnetic flux expansion, and ELM frequency was seen. These experiments aid in understanding the impact of high-Z divertor source location on core performance in future mixed-material fusion devices, e.g. ITER. Supported by US DOE under DE- AC05-00OR22725, DE-FC02-04ER54698, DE-FG02-07ER54917, DE-SC0013911, DE-AC02-09CH11466, DE-AC52-07NA27344.

  11. Radiative snowflake divertor studies in DIII-D

    NASA Astrophysics Data System (ADS)

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

    2015-08-01

    Recent DIII-D experiments assessed the snowflake divertor (SF) configuration in a radiative regime in H-mode discharges with D2 seeding. The SF configuration was maintained for many energy confinement times (2-3 s) in H-mode discharges (Ip = 1.2 MA, PNBI = 4- 5 MW, and B × ∇B down (favorable direction toward the divertor)), and found to be compatible with high performance operation (H98y2 ⩾ 1). The two studied SF configurations, the SF-plus and the SF-minus, have a small finite distance between the primary X-point and the secondary Bp null located in the private flux region or the common flux region, respectively. In H-mode discharges with the SF configurations (cf. H-mode discharges with the standard divertor with similar conditions) the stored energy lost per the edge localized mode (ELM) was reduced, and significant divertor heat flux reduction between and during ELMs was observed over a range of collisionalities, from lower density conditions toward a higher density H-modes with the radiative SF divertor.

  12. Compatibility of Detached Divertor Operation with Robust Edge Pedestal Performance

    NASA Astrophysics Data System (ADS)

    Leonard, A. W.; Osborne, T. H.; Snyder, P. B.; Makowski, M. A.; McLean, A. G.

    2014-10-01

    The compatibility of radiative detached divertor operation with the maintenance of a robust H-mode pedestal is examined in DIII-D. A density scan with deuterium injection into H-mode spanned a range of divertor conditions from fully attached, ~30 eV at the target, with little divertor radiation to a fully detached with Te < 5 eV throughout the divertor up to the X-point. Over this scan of pedestal density from n /nGW = 30% to 60% the pedestal Te was reduced from 800 eV to 350 eV, representing a ~20% reduction in pedestal pressure with a similar reduction in normalized energy confinement. The reduction in pedestal pressure at high density was found to be consistent with a reduced pedestal ELM MHD stability limit at high collisionality. The scaling of the pedestal top pressure with density was also consistent with the EPED model, which assumes an additional constraint on the local pressure gradient. The MHD stability limit at the highest collisionality depends on details of the ELM instability growth rate normalization. This result is encouraging for future burning plasmas where a low collisionality pedestal is expected to be maintained even for high density detached divertor operation. Work supported by the US DOE under DE-FC02-04ER54698 and DE-AC52-07NA27344.

  13. Upgraded divertor Thomson scattering system on DIII-D

    SciTech Connect

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

    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.

  14. Analysis of sweeping heat loads on divertor plate materials

    SciTech Connect

    Hassanein, A.

    1991-12-31

    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{sup 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.

  15. Analysis of sweeping heat loads on divertor plate materials

    SciTech Connect

    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{sup 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.

  16. Upgraded divertor Thomson scattering system on DIII-D

    NASA Astrophysics Data System (ADS)

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

    2016-11-01

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

  17. Solid tungsten Divertor-III for ASDEX Upgrade and contributions to ITER

    NASA Astrophysics Data System (ADS)

    Herrmann, A.; Greuner, H.; Jaksic, N.; Balden, M.; Kallenbach, A.; Krieger, K.; de Marné, P.; Rohde, V.; Scarabosio, A.; Schall, G.; the ASDEX Upgrade Team

    2015-06-01

    ASDEX Upgrade became a full tungsten experiment in 2007 by coating its graphite plasma facing components with tungsten. In 2013 a redesigned solid tungsten divertor, Div-III, was installed and came into operation in 2014. The redesign of the outer divertor geometry provided the opportunity to increase the pumping efficiency in the lower divertor by increasing the gap between divertor and vessel. In parallel, a by-pass was installed into the cryo-pump in the divertor region allowing adapting of the pumping speed to the required edge density. Safe divertor operation and heat removal becomes more and more significant for future fusion devices. This requires developing ‘tools’ for divertor heat load control and to optimize the divertor design. The new divertor manipulator, DIM-II, allows retracting a relevant part of the outer divertor into a target exchange box without venting ASDEX Upgrade. Different front-ends can be installed and exposed to the plasma. At present, front-ends for probe exposition, gas puffing, electrical probes and actively cooled prototype targets are under construction. The installation of solid tungsten, the control of the pumping speed and the flexibility for testing divertor modifications on a weekly base is a unique feature of ASDEX Upgrade and offers together with the extended set of diagnostics the possibility to investigate dedicated questions for a future divertor design.

  18. A review of radiative detachment studies in tokamak advanced magnetic divertor configurations

    NASA Astrophysics Data System (ADS)

    Soukhanovskii, V. A.

    2017-06-01

    The present vision for a plasma-material interface in the tokamak is an axisymmetric poloidal magnetic X-point divertor. Four tasks are accomplished by the standard poloidal X-point divertor: plasma power exhaust; particle control (D/T and He pumping); reduction of impurity production (source); and impurity screening by the divertor scrape-off layer. A low-temperature, low heat flux divertor operating regime called radiative detachment is viewed as the main option that addresses these tasks for present and future tokamaks. Advanced magnetic divertor configuration has the capability to modify divertor parallel and cross-field transport, radiative and dissipative losses, and detachment front stability. Advanced magnetic divertor configurations are divided into four categories based on their salient qualitative features: (1) multiple standard X-point divertors; (2) divertors with higher order nulls; (3) divertors with multiple X-points; and (4) long poloidal leg divertors (and also with multiple X-points). This paper reviews experiments and modeling in the area of radiative detachment in the advanced magnetic divertor configurations.

  19. Safe Finger Tourniquet--Ideas.

    PubMed

    Wei, Lin-Gwei; Chen, Chieh-Feng; Hwang, Chun-Yuan; Chang, Chiung-Wen; Chiu, Wen-Kuan; Li, Chun-Chang; Wang, Hsian-Jenn

    2016-03-01

    Tourniquets are often needed for optimized phalangeal surgeries. However, few surgeons forget to remove them and caused ischemic injuries. We have a modified method to create a safe finger tourniquet for short duration finger surgeries, which can avoid such tragedy. It is done by donning a glove, cutting the tip of the glove over the finger of interest, and rolling the glove finger to the base. From 2010 to 2013, approximately 54 patients underwent digital surgical procedures with our safe finger tourniquet. Because the glove cannot be forgotten to be removed, the tourniquet must be released and removed. This is a simple and efficient way to apply a safe finger tourniquet by using hand rubber glove for a short-term bloodless finger surgery and can achieve an excellent surgical result.

  20. Hemangioma of the fingers.

    PubMed

    Kodachi, K; Kojima, T; Shimbashi, T; Furusato, M

    1990-01-01

    Fingers often suffer trauma and the clinician is continuously faced with the difficult task of clarifying the distinction between a hemangioma and a traumatic lesion. This study was undertaken to examine ten cases in which a small skin mass located on a finger had been diagnosed preoperatively as hemangioma. Our results showed that seven masses were confirmed pathologically as hemangioma (five cavernous hemangiomas and two capillary hemangiomas), two as traumatic thrombosis and one varix. The clinical manifestations of the two cases of traumatic thrombosis were related to those of hemangioma. In the varix, endothelial proliferation was observed in the area of the thrombosis. This phenomenon is called "intravascular papillary endothelial hyperplasia", and can confuse the differential diagnosis between a vascular neoplasm and a traumatic thrombosis. Our findings demonstrate that since the traumatic lesions were firmer than the hemangiomas, hardness on physical examination may be a helpful indicator in the differential diagnosis of a hemangioma and a traumatic lesion.

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

    SciTech Connect

    Boozer, A. ); Braams, B.; Weitzner, H. . Courant Inst. of Mathematical Sciences); Cohen, R. ); Hazeltine, R. . Inst. for Fusion Studies); Hinton, F. ); Houlberg, W. (Oak

    1992-12-22

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

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

    SciTech Connect

    Boozer, A.; Braams, B.; Weitzner, H.; Cohen, R.; Hazeltine, R.; Hinton, F.; Houlberg, W.; Oktay, E.; Sadowski, W.; Post, D.; Sigmar, D.; Wootton, A.

    1992-12-22

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

  3. Innovative divertor concept development on DIII-D and EAST

    SciTech Connect

    Guo, H. Y.; Allen, S.; Canik, J.; Hill, D. N.; Leonard, T.; Sang, C. F.; Stangeby, P. C.; Thomas, D. M.; Unterberg, Z.; Luo, G. N.; Wang, L.; Wan, B. N.; Xu, G. S.

    2016-06-02

    A critical issue facing the design and operation of next-step high-power steady-state fusion devices is the control of heat fluxes and erosion at the plasma-facing components, in particular, the divertor target plates. A new initiative has been launched on DIII-D to develop and demonstrate innovative boundary plasma-materials interface solutions. The central purposes of this new initiative are to advance scientific understanding in this critical area and develop an advanced divertor concept for application to next-step fusion devices. Finally, DIII-D will leverage strong collaborative efforts on the EAST superconducting tokamak for extending integrated high performance advanced divertor solutions to true steady-state.

  4. Radiative divertor plasmas with convection in DIII-D

    SciTech Connect

    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.

  5. Plasma transport in a simulated magnetic-divertor configuration

    SciTech Connect

    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.

  6. Turbulence studies in Tokamak boundary plasmas with realistic divertor geometry

    SciTech Connect

    Xu, X.Q.

    1998-10-14

    Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT [1] and the linearized shooting code BAL[2] to study 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 finite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters.

  7. Deuterium and tritium separation in a tokamak reactor divertor layer

    NASA Astrophysics Data System (ADS)

    Tokar', M. Z.

    1989-04-01

    It's shown that the plasma isotope composition in a tokamak reactor divertor layer changes along the magnetic field and can notable differ from the gas composition in a pumping chamber. Heavier tritium must concentrate in the hot plasma far from the divertor plate due to thermal force stipulated by mutial collisions of deuterium and tritium ions. This circumstance is favourable from the point of view of tritium cycle optimization and must facilitate solution of the problem of tritium accumulation in the reactor construction elements.

  8. Taming the heat flux problem: Advanced divertors towards fusion power

    SciTech Connect

    Kotschenreuther, M.; Mahajan, S.; Valanju, P. M.; Covele, B.; Waelbroeck, F. L.; Canik, John M.; LaBombard, Brian

    2015-09-11

    The next generation fusion machines are likely to face enormous heat exhaust problems. In addition to summarizing major issues and physical processes connected with these problems, we discuss how advanced divertors, obtained by modifying the local geometry, may yield workable solutions. We also point out that: (1) the initial interpretation of recent experiments show that the advantages, predicted, for instance, for the X-divertor (in particular, being able to run a detached operation at high pedestal pressure) correlate very well with observations, and (2) the X-D geometry could be implemented on ITER (and DEMOS) respecting all the relevant constraints. As a result, a roadmap for future research efforts is proposed.

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

    SciTech Connect

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

    2011-02-01

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

  10. A review of radiative detachment studies in tokamak advanced magnetic divertor configurations

    DOE PAGES

    Soukhanovskii, V. A.

    2017-04-28

    The present vision for a plasma–material interface in the tokamak is an axisymmetric poloidal magnetic X-point divertor. Four tasks are accomplished by the standard poloidal X-point divertor: plasma power exhaust; particle control (D/T and He pumping); reduction of impurity production (source); and impurity screening by the divertor scrape-off layer. A low-temperature, low heat flux divertor operating regime called radiative detachment is viewed as the main option that addresses these tasks for present and future tokamaks. Advanced magnetic divertor configuration has the capability to modify divertor parallel and cross-field transport, radiative and dissipative losses, and detachment front stability. Advanced magnetic divertormore » configurations are divided into four categories based on their salient qualitative features: (1) multiple standard X-point divertors; (2) divertors with higher order nulls; (3) divertors with multiple X-points; and (4) long poloidal leg divertors (and also with multiple X-points). As a result, this paper reviews experiments and modeling in the area of radiative detachment in the advanced magnetic divertor configurations.« less

  11. Attainment of a stable, fully detached plasma state in innovative divertor configurations

    NASA Astrophysics Data System (ADS)

    Umansky, M. V.; LaBombard, B.; Brunner, D.; Rensink, M. E.; Rognlien, T. D.; Terry, J. L.; Whyte, D. G.

    2017-05-01

    A computational study of long-legged tokamak divertor configurations is performed with the edge transport code UEDGE. Several divertor configurations are considered, with radially or vertically extended, tightly baffled, outer divertor legs and with or without a secondary X-point in the divertor leg volume. For otherwise identical conditions, a scan of the input power from the core plasma is performed. As the power is reduced to a threshold value, the plasma in the outer leg transitions to a fully detached state, which defines the upper limit on the power for detached divertor operation. Reducing the power further results in the detachment front shifting upstream but remains stable. At low power, the detachment front eventually moves all the way to the primary X-point, which is usually associated with degradation of the core plasma, and this defines the lower limit on the power for the detached divertor operation. For the studied parameters, for long-legged divertors, the detached operation window is quite large, in particular, for the X-point target configuration using a secondary X-point in the divertor leg volume, allowing a factor of 5-10 variations in the input power. For the same parameters, for the standard divertor configuration, the detached operation window is very small or even non-existent. The present modeling results suggest the possibility of stable fully detached divertor operation for a tokamak with tightly baffled extended divertor legs.

  12. Progress in snowflake divertor research in DIII-D, NSTX and NSTX-U

    NASA Astrophysics Data System (ADS)

    Soukhanovskii, V. A.; Allen, S.; Fenstermacher, M.; Izacard, O.; Lasnier, C.; Makowski, M.; McLean, A.; Myer, W.; Ryutov, D.; Scotti, F.; Eldon, D.; Kolemen, E.; Vail, P.; Canal, G.; Groebner, R.; Hyatt, A.; Leonard, A.; Osborne, T.; Bell, R.; Diallo, A.; Gerhardt, S.; Kaye, S.; Leblanc, B.; Menard, J.; Podesta, M.

    2016-10-01

    Recent snowflake (SF) divertor DIII-D experiments focused on divertor heat transport under attached and radiative divertor conditions, incl 1-understanding of increased scrape-off layer width in SF-plus configuration at lower densities; 2-particle, heat and radiation distribution in the SF divertor with CD4 seeding. NSTX data was analyzed to understand the link between SF divertor and ELM (de)stabilization with and without CD4 seeding and lithium conditioning. Prep for SF divertor experiments in NSTX-U include 1-equilibria modeling with ISOLVER code using various sets of divertor coils and L- and H-mode plasma scenarios; 2-transport and impurity radiation modeling with UEDGE code; 3-new diagnostics (ie-a 100-200 kHz camera for null-region mode observations). Supported by DOE under DE-AC52-07NA27344, DE-AC02-09CH11466, DE-FC02-04ER54698.

  13. Robotic Finger Assembly

    NASA Technical Reports Server (NTRS)

    Ihrke, Chris A. (Inventor); Bridgwater, Lyndon (Inventor); Diftler, Myron A. (Inventor); Linn, Douglas M. (Inventor); Platt, Robert (Inventor); Hargrave, Brian (Inventor); Askew, Scott R. (Inventor); Valvo, Michael C. (Inventor)

    2013-01-01

    A robotic hand includes a finger with first, second, and third phalanges. A first joint rotatably connects the first phalange to a base structure. A second joint rotatably connects the first phalange to the second phalange. A third joint rotatably connects the third phalange to the second phalange. The second joint and the third joint are kinematically linked such that the position of the third phalange with respect to the second phalange is determined by the position of the second phalange with respect to the first phalange.

  14. Robotic Finger Assembly

    NASA Technical Reports Server (NTRS)

    Ihrke, Chris A. (Inventor); Bridgwater, Lyndon (Inventor); Diftler, Myron A. (Inventor); Linn, Douglas Martin (Inventor); Platt, Robert J., Jr. (Inventor); Hargrave, Brian (Inventor); Askew, Scott R. (Inventor); Valvo, Michael C. (Inventor)

    2014-01-01

    A robotic hand includes a finger with first, second, and third phalanges. A first joint rotatably connects the first phalange to a base structure. A second joint rotatably connects the first phalange to the second phalange. A third joint rotatably connects the third phalange to the second phalange. The second joint and the third joint are kinematically linked such that the position of the third phalange with respect to the second phalange is determined by the position of the second phalange with respect to the first phalange.

  15. Finger Forces in Clarinet Playing

    PubMed Central

    Hofmann, Alex; Goebl, Werner

    2016-01-01

    Clarinettists close and open multiple tone holes to alter the pitch of the tones. Their fingering technique must be fast, precise, and coordinated with the tongue articulation. In this empirical study, finger force profiles and tongue techniques of clarinet students (N = 17) and professional clarinettists (N = 6) were investigated under controlled performance conditions. First, in an expressive-performance task, eight selected excerpts from the first Weber Concerto were performed. These excerpts were chosen to fit in a 2 × 2 × 2 design (register: low–high; tempo: slow–fast, dynamics: soft–loud). There was an additional condition controlled by the experimenter, which determined the expression levels (low–high) of the performers. Second, a technical-exercise task, an isochronous 23-tone melody was designed that required different effectors to produce the sequence (finger-only, tongue-only, combined tongue-finger actions). The melody was performed in three tempo conditions (slow, medium, fast) in a synchronization-continuation paradigm. Participants played on a sensor-equipped Viennese clarinet, which tracked finger forces and reed oscillations simultaneously. From the data, average finger force (Fmean) and peak force (Fmax) were calculated. The overall finger forces were low (Fmean = 1.17 N, Fmax = 3.05 N) compared to those on other musical instruments (e.g., guitar). Participants applied the largest finger forces during the high expression level performance conditions (Fmean = 1.21 N). For the technical exercise task, timing and articulation information were extracted from the reed signal. Here, the timing precision of the fingers deteriorated the timing precision of the tongue for combined tongue-finger actions, especially for faster tempi. Although individual finger force profiles were overlapping, the group of professional players applied less finger force overall (Fmean = 0.54 N). Such sensor instruments provide useful insights into player

  16. The effects of the Snowflake Divertor on upstream SOL profiles

    NASA Astrophysics Data System (ADS)

    Tsui, C. K.; Boedo, J. A.; Coda, S.; Labit, B.; Maurizio, R.; Nespoli, F.; Reimerdes, H.; Theiler, C.; Spolaore, M.; Vianello, N.; Lunt, T.; Vijvers, W. A. J.; Walkden, N.; the EUROfusion MST1 Team Team; the TCV Team Team

    2016-10-01

    The Snowflake Divertor creates separated volumes within the SOL and divertor that feature strikingly different ne, Te profiles, and decay lengths, as measured with a scanning probe. Profiles were taken at the outer midplane of TCV plasmas with snowflake divertors as well as just above the X-points within the region of enhanced βpol. Density shoulders in the far SOL in single null plasmas are relaxed by secondary X-points, while effects are more complex in the near SOL. These changes were observed whether the secondary X-point was placed in the low field side SOL, or in the high field side SOL. Additionally, target profiles measured with IR camera and Langmiur probes that were taken in the divertor leg opposite the secondary X-point also show features on the flux surface corresponding to the secondary X-point. Fluctuation statistics from the reciprocating probe as well as comparisons made between upstream and downstream measurements are considered for their implications on SOL transport. Support from EUROfusion Grant 633053 and US DOE Grant DE-SC0010529 are gratefully acknowledged.

  17. Taming the plasma-material interface with the snowflake divertor.

    SciTech Connect

    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.

  18. Line Shapes and Opacity Studies in Divertor Plasmas

    SciTech Connect

    Rosato, J.

    2008-10-22

    Large or dense divertor plasmas of magnetic fusion devices can be optically thick to the resonance lines of the hydrogen isotopes. In this work we examine the sensitivity of the line radiation transport to the detailed structure of the spectral profiles.

  19. Visible spectroscopy in the DIII-D divertor

    SciTech Connect

    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-circle spectrograph into a rastered format on the rectangular sensor area of a two-dimensional CCD camera. Vertical grouping of pixels during CCD readout integrates the signal intensity over the height of each spectral segment in the rastered image, minimizing readout time. For the full visible spectrum, readout time is 50 ms. Faster response time (< 10 ms) may be obtained by selecting for readout just a small number of the twenty spectral segments in the image on the CCD. Simultaneous recording of low charge states of carbon, oxygen and injected impurities has yielded information about gas recycling and impurity behavior at the divertor strike points. Transport of lithium to the divertor region during lithium pellet injection has been studied, as well as cumulative deposition of lithium on the divertor targets from pellet injection over many successive discharges.

  20. Mechanical Design of the NSTX Liquid Lithium Divertor

    SciTech Connect

    R. Ellis, R. Kaita, H. Kugel, G. Paluzzi, M. Viola and R. Nygren

    2009-02-19

    The Liquid Lithium Divertor (LLD) on NSTX will be the first test of a fully-toroidal liquid lithium divertor in a high-power magnetic confinement device. It will replace part of the lower outboard divertor between a specified inside and outside radius, and ultimately provide a lithium surface exposed to the plasma with enough depth to absorb a significant particle flux. There are numerous technical challenges involved in the design. The lithium layer must be as thin as possible, and maintained at a temperature between 200 and 400 degrees Celsius to minimize lithium evaporation. This requirement leads to the use of a thick copper substrate, with a thin stainless steel layer bonded to the plasma-facing surface. A porous molybdenum layer is then plasma-sprayed onto the stainless steel, to provide a coating that facilitates full wetting of the surface by the liquid lithium. Other challenges include the design of a robust, vacuumcompatible heating and cooling system for the LLD. Replacement graphite tiles that provided the proper interface between the existing outer divertor and the LLD also had to be designed, as well as accommodation for special LLD diagnostics. This paper describes the mechanical design of the LLD, and presents analyses showing the performance limits of the LLD.

  1. Snowflake Divertor Configuration Studies in DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

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

    2013-10-01

    Experiments in DIII-D show the snowflake divertor (SFD) configuration is compatible with high performance operation (H98 y 2 >= 1) and results in greatly reduced divertor heat flux between and during edge localized modes (ELMs). The SFD was sustained for many energy confinement times using the standard poloidal field shaping coils in 3-5 MW neutral beam injection-heated discharges. Pedestal and divertor effects resulting from a large region of reduced poloidal magnetic field in the SFD are measured and studied using the 2D multi-fluid code UEDGE. The pedestal pressure appeared to be unchanged, while the energy loss per ELM was reduced by 50%. Partial detachment of the SFD was observed at higher ne, with an expanded divertor radiation zone and peak ELM heat flux reduced by up to 80%. Work supported by the US Department of Energy under DE-AC52-07NA27344, DE-AC02-09CH11466, DE-FC02-04ER54698, DE-FG02-07ER54917, and DE-AC04-94AL85000.

  2. Modeling results for a linear simulator of a divertor

    SciTech Connect

    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.

  3. Theoretical design of a compact energy recovering divertor

    NASA Astrophysics Data System (ADS)

    Baver, D. A.

    2015-11-01

    An energy recovering divertor (ERD) is a type of plasma direct converter (PDC) designed to fit in the divertor channel of a tokamak. Such a device reduces the heat load to the divertor plate by converting a portion of it into electrical energy. This recovered energy can then be used for auxiliary heating and current drive, fundamentally altering the relationship between scientific and engineering breakeven and reducing dependence on bootstrap current. Previous work on the ERD concept focused on amplification of Alfven waves in a manner similar to a free-electron laser. While conceptually straightforward, this concept was also bulky, thus limiting its applicability to existing tokamak experiments. A design is presented for an ERD based on sheath-localized waves. This makes possible a device sufficiently compact to fit in the divertor channel of many existing tokamak experiments, and moreover requires no new shaping coils to achieve the desired magnetic geometry or topology. In addition, incidental advantages of this concept will be discussed.

  4. Fibonacci-compliant finger design.

    PubMed

    El-Sheikh, Mogeeb A

    2016-11-11

    This work presents the mechanical design of 4 configurations of compliant fingers in order to address the need for commercially feasible prosthetic and robotic hands. The fingers consist of a single part and utilize a compliant mechanism to reduce the cost and control complexity. The geometric parameters of the compliant finger designs follow the Fibonacci series. The first and second compliant fingers have 2 joints and 2 degrees of freedom. The others have 3 joints and 3 degrees of freedom. The type of flexure hinges of the compliant finger are single and multiple nonsymmetrical circular hinges. The finite element method (FEM) was used to verify the range of motion of the joints in the compliant finger. In addition, the study defines the finger tip trajectory of these configurations. The multiple flexure hinges have minimum stress. This study presents affordable, single-element, compliant finger designs and their presumable hypothetical design variables are defined by the Fibonacci series. This method is faster and simpler than optimization. The study identifies the application of each finger design for either prosthetic or robotic purposes.

  5. Finger Injuries in Ball Sports.

    PubMed

    Netscher, David T; Pham, Dang T; Staines, Kimberly Goldie

    2017-02-01

    Finger injuries are common in athletes playing in professional ball sports. Understanding the intricate anatomy of the digit is necessary to properly diagnose and manage finger injuries. Unrecognized or poorly managed finger injuries can lead to chronic deformities that can affect an athlete's performance. Multiple factors and treatment options should be considered to provide the best functional outcome and rapid return to play for an athlete. This article discusses the mechanism of injury, diagnosis, treatment, and return-to-play recommendations for common finger injuries in ball sports.

  6. Repair of webbed fingers or toes

    MedlinePlus

    ... skin grafts Stiffness of the fingers or toes Injuries to the blood vessels, tendons, or bones in the fingers Call your provider if you notice the following: Fever Fingers that tingle, are numb, or have a bluish ... fingers or toes to protect the repaired area from injury. Small children who had webbed finger repair may ...

  7. Finger vein recognition based on finger crease location

    NASA Astrophysics Data System (ADS)

    Lu, Zhiying; Ding, Shumeng; Yin, Jing

    2016-07-01

    Finger vein recognition technology has significant advantages over other methods in terms of accuracy, uniqueness, and stability, and it has wide promising applications in the field of biometric recognition. We propose using finger creases to locate and extract an object region. Then we use linear fitting to overcome the problem of finger rotation in the plane. The method of modular adaptive histogram equalization (MAHE) is presented to enhance image contrast and reduce computational cost. To extract the finger vein features, we use a fusion method, which can obtain clear and distinguishable vein patterns under different conditions. We used the Hausdorff average distance algorithm to examine the recognition performance of the system. The experimental results demonstrate that MAHE can better balance the recognition accuracy and the expenditure of time compared with three other methods. Our resulting equal error rate throughout the total procedure was 3.268% in a database of 153 finger vein images.

  8. Differences in finger localisation performance of patients with finger agnosia.

    PubMed

    Anema, Helen A; Kessels, Roy P C; de Haan, Edward H F; Kappelle, L Jaap; Leijten, Frans S; van Zandvoort, Martine J E; Dijkerman, H Chris

    2008-09-17

    Several neuropsychological studies have suggested parallel processing of somatosensory input when localising a tactile stimulus on one's own by pointing towards it (body schema) and when localising this touched location by pointing to it on a map of a hand (body image). Usually these reports describe patients with impaired detection, but intact sensorimotor localisation. This study examined three patients with a lesion of the angular gyrus with intact somatosensory processing, but with selectively disturbed finger identification (finger agnosia). These patients performed normally when pointing towards the touched finger on their own hand but failed to indicate this finger on a drawing of a hand or to name it. Similar defects in the perception of other body parts were not observed. The findings provide converging evidence for the dissociation between body image and body schema and, more importantly, reveal for the first time that this distinction is also present in higher-order cognitive processes selectively for the fingers.

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

    SciTech Connect

    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).

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

    SciTech Connect

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

    2012-10-15

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

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

    PubMed

    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 (q(peak) ≤ 15 MW/m(2)), radiative divertor experiments have demonstrated a significant reduction of divertor peak heat flux simultaneously with good core H-mode confinement using pre-programmed D(2) or CD(4) gas injections. In this work diagnostic options for a new real-time feedback control system for active radiative divertor detachment control in NSTX-U, where steady-state peak divertor heat fluxes are projected to reach 20-30 MW/m(2), are discussed. Based on the NSTX divertor detachment measurements and analysis, the control diagnostic signals available for NSTX-U include divertor radiated power, neutral pressure, spectroscopic deuterium recombination signatures, infrared thermography of PFC surfaces, and thermoelectric scrape-off layer current. In addition, spectroscopic "security" monitoring of possible confinement or pedestal degradation is recommended. These signals would be implemented in a digital plasma control system to manage the divertor detachment process via an actuator (impurity gas seeding rate).

  12. Finger Movements in Transcription Typing

    DTIC Science & Technology

    1980-05-07

    learned motor movements.- DD , , 1473 00f n,.. ofo I. OV.o 95c ’ ~ 15 O,/BSOL."ETE, .[ -. 4 a/ 11L 1.5U NTARF~ SO. ~ ~ ~ ~ ~ ~ ~ ~ ~ ECRT AUSTRAC ...fingers. Sensory information and proprioceptive feedback from the mus- cles controlling the fingers could play a role . Neural impulses take about 70

  13. Skilled Finger Movements in Typing.

    ERIC Educational Resources Information Center

    Gentner, Donald R.

    Six skilled typists were studied while they transcribed English text. The typists showed stable patterns of performance, but with significant individual differences among themselves. Inter-keypress latencies for two-finger digraphs (typed by two fingers on the same hand) were particularly variable among typists. Two typists showed large…

  14. Gert Finger Becomes Emeritus Physicist

    NASA Astrophysics Data System (ADS)

    de Zeeuw, T.; Lucuix, C.; Péron, M.

    2016-03-01

    Gert Finger has retired after almost 33 years service and he has been made the first Emeritus Physicist at ESO. An appreciation of some of his many achievements in the development of infrared instrumentation and detector controllers is given. A retirement party for Gert Finger was held in February 2016.

  15. Competition between anisotropic viscous fingers

    NASA Astrophysics Data System (ADS)

    Pecelerowicz, M.; Budek, A.; Szymczak, P.

    2014-09-01

    We consider viscous fingers created by injection of low viscosity fluid into the network of capillaries initially filled with a more viscous fluid (motor oil). Due to the anisotropy of the system and its geometry, such a setup promotes the formation of long-and-thin fingers which then grow and compete for the available flow, interacting through the pressure field. The interaction between the fingers is analyzed using the branched growth formalism of Halsey and Leibig (Phys. Rev. A 46, 7723, 1992) using a number of simple, analytically tractable models. It is shown that as soon as the fingers are allowed to capture the flow from one another, the fixed point appears in the phase space, corresponding to the asymptotic state in which the growth of one of the fingers in hindered by the other. The properties of phase space flows in such systems are shown to be remarkably insensitive to the details of the dynamics.

  16. Optimal three finger grasps

    NASA Technical Reports Server (NTRS)

    Demmel, J.; Lafferriere, G.

    1989-01-01

    Consideration is given to the problem of optimal force distribution among three point fingers holding a planar object. A scheme that reduces the nonlinear optimization problem to an easily solved generalized eigenvalue problem is proposed. This scheme generalizes and simplifies results of Ji and Roth (1988). The generalizations include all possible geometric arrangements and extensions to three dimensions and to the case of variable coefficients of friction. For the two-dimensional case with constant coefficients of friction, it is proved that, except for some special cases, the optimal grasping forces (in the sense of minimizing the dependence on friction) are those for which the angles with the corresponding normals are all equal (in absolute value).

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

    SciTech Connect

    Zhao, M. L.; Chen, Y. P.; Li, G. Q.; Luo, Z. P.; Guo, H. Y.; Ye, M. Y.; Tendler, M.

    2014-05-15

    The China Fusion Engineering Testing Reactor (CFETR) is currently under design. The SOLPS/B2-Eirene code package is utilized for the design and optimization of the divertor geometry for CFETR. Detailed modeling is carried out for an ITER-like divertor configuration and one with relatively open inner divertor structure, to assess, in particular, peak power loading on the divertor target, which is a key issue for the operation of a next-step fusion machine, such as ITER and CFETR. As expected, the divertor peak heat flux greatly exceeds the maximum steady-state heat load of 10 MW/m{sup 2}, which is a limit dictated by engineering, for both divertor configurations with a wide range of edge plasma conditions. Ar puffing is effective at reducing divertor peak heat fluxes below 10 MW/m{sup 2} even at relatively low densities for both cases, favoring the divertor configuration with more open inner divertor structure.

  18. Developing snowflake divertor physics basis in the DIII-D, NSTX and NSTX-U tokamaks aimed at the divertor power exhaust solution [Snowflake divertor experiments in the DIII-D, NSTX and NSTX-U tokamaks aimed at the development of the divertor power exhaust solution

    SciTech Connect

    Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; Lasnier, C. J.; Makowski, M. A.; McLean, A. G.; Meier, E. T.; Meyer, W. H.; Rognlien, T. D.; Ryutov, D. D.; Scotti, F.; Kolemen, E.; Bell, R. E.; Diallo, A.; Gerhardt, S.; Kaita, R.; Kaye, S.; LeBlanc, B. P.; Maingi, R.; Menard, J. E.; Podesta, M.; Roquemore, A. L.; Groebner, R. J.; Hyatt, A. W.; Leonard, A. W.; Osborne, T. H.; Petrie, T. W.; Ahn, J. -W.; Raman, R.; Watkins, J. G.

    2016-06-02

    Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment were performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large ELMs. However, a stable partial detachment of the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3-5, and peak ELM heat flux was reduced by up to 80% (cf. standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower ne, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D2-seeded SF divertor at PSOL = 3 - 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multi-fluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected PSOL ≃9 MW case. In conclusion, the radiative SF divertor with carbon impurity provides a wider ne operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar qpeak reduction factors (cf. standard divertor).

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

    SciTech Connect

    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.

  20. Preliminary study of divertor particle exhaust in the EAST superconducting tokamak

    NASA Astrophysics Data System (ADS)

    Liu, Huan; Wang, Liang; Xu, Guosheng; Ding, Fang; Liu, Jianbin; Xu, Jichan; Feng, Wei; Deng, Guozhong; Zheng, Xingwei; Yu, Yaowei; Si, Hang; Liu, Haiqing; Yang, Qingquan; Sun, Zhen; Guo, Houyang

    2017-09-01

    The particle exhaust of the upper tungsten and lower carbon divertors in EAST has been preliminarily studied during the 2016 experimental campaign. The density decay time during terminating gas puffing has been employed as a key parameter to evaluate the divertor particle exhaust performance. Comparative plasma discharges have been carried out on the particle exhaust performance between two toroidal field directions in the upper single null and lower single null divertor configurations. This work has enhanced the understanding of the effects of the in-out asymmetry and divertor geometry on the efficiency of the divertor particle exhaust. In addition, the sensitivity of the particle exhaust capability on different strike point locations has been analyzed. The experimental results are expected to provide important information on the future upgrade of EAST bottom divertor and facilitate the realization of longer pulse operation.

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

    NASA Astrophysics Data System (ADS)

    Himabindu, M.; Tyagi, Anil K.; Sharma, Deepti; Sharma, Devendra; Srinivasan, R.

    2017-04-01

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

  2. Kinetic effects in edge plasma: kinetic modeling for edge plasma and detached divertor

    NASA Astrophysics Data System (ADS)

    Takizuka, T.

    2017-03-01

    Detached divertor is considered a solution for the heat control in magnetic-confinement fusion reactors. Numerical simulations using the comprehensive divertor codes based on the plasma fluid modeling are indispensable for the design of the detached divertor in future reactors. Since the agreement in the results between detached-divertor experiments and simulations has been rather fair but not satisfactory, further improvement of the modeling is required. The kinetic effect is one of key issues for improving the modeling. Complete kinetic behaviors are able to be simulated by the kinetic modeling. In this paper at first, major kinetic effects in edge plasma and detached divertor are listed. One of the most powerful kinetic models, particle-in-cell (PIC) model, is described in detail. Several results of PIC simulations of edge-plasma kinetic natures are presented. Future works on PIC modeling and simulation for the deeper understanding of edge plasma and detached divertor are discussed.

  3. Tokamak power exhaust with the snowflake divertor: Present results and outstanding issues

    SciTech Connect

    Soukhanovskii, V. A.; Xu, X.

    2015-09-15

    Here, a snowflake divertor magnetic configuration (Ryutov in Phys Plasmas 14(6):064502, 2007) with the second-order poloidal field null offers a number of possible advantages for tokamak plasma heat and particle exhaust in comparison with the standard poloidal divertor with the first-order null. Results from snowflake divertor experiments are briefly reviewed and future directions for research in this area are outlined.

  4. An automated approach to magnetic divertor configuration design

    NASA Astrophysics Data System (ADS)

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

    2015-01-01

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

  5. Multiplexing thermography for International Thermonuclear Experimental Reactor divertor targets

    SciTech Connect

    Itami, K.; Sugie, T.; Vayakis, G.; Walker, C.

    2004-10-01

    The concept of multiplexing thermography is applied to the design of the divertor thermography system for International Thermonuclear Experimental Reactor (ITER). The combination of the front mirror with multiellipticity and a Czerney-Turner spectrometer with a 0.2 mm pitched multichannel detector enables a spatial resolution of 3 mm and a time resolution of 20 {mu}s above a target temperature of 300 deg. C to be achieved. This should be sufficient to measure ELM heat fluxes to the targets in ITER. To satisfy the measurement requirement, it is very important to keep an accurate alignment around the optical axis against movement of the vessel during the plasma discharges. Several key engineering problems, such as the survivability of components against mirror coating by redeposited divertor material, remain to be solved. Potential solutions have been identified.

  6. Modelling of Divertor Plasma Transport in Stochastic Magnetic Boundary

    SciTech Connect

    Kobayashi, Masahiro

    2010-05-20

    Impacts of stochastic magnetic field structure on divertor functions are discussed based on analyses with the three dimensional (3D) edge transport code package EMC3-EIRENE with Braginskii type fluid equations, in the Large Helical Device (LHD), in comparison with the experimental data. It is shown that the three dimensional field line topology introduced by the stochasticity provides controllability of the edge plasma transport such as divertor regime, impurity transport. The observations in other devices with stochastic magnetic boundary regarding these issues are discussed as well. Also presented are the traditional formulation of the magnetic field and the transport in the stochastic layer based on diffusive picture, which are contrasted with the 3D treatment of the flux tube topology and of the transport.

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

    NASA Astrophysics Data System (ADS)

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

    2016-09-01

    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.

  8. Modelling of Divertor Plasma Transport in Stochastic Magnetic Boundary

    NASA Astrophysics Data System (ADS)

    Kobayashi, Masahiro

    2010-05-01

    Impacts of stochastic magnetic field structure on divertor functions are discussed based on analyses with the three dimensional (3D) edge transport code package EMC3-EIRENE with Braginskii type fluid equations, in the Large Helical Device (LHD), in comparison with the experimental data. It is shown that the three dimensional field line topology introduced by the stochasticity provides controllability of the edge plasma transport such as divertor regime, impurity transport. The observations in other devices with stochastic magnetic boundary regarding these issues are discussed as well. Also presented are the traditional formulation of the magnetic field and the transport in the stochastic layer based on diffusive picture, which are contrasted with the 3D treatment of the flux tube topology and of the transport.

  9. Edge exposure of poloidal divertor target plate tiles

    SciTech Connect

    Mohanti, R.B.; Gilligan, J.G.; Bourham, M.A.

    1996-12-01

    Exposure to near normal surfaces of poloidal divertor target plate tiles is a limiting feature of the power handling capability of the tiles. The problems associated with the design of poloidal divertor tiles, with beryllium chosen as the tile material, and possible methods of solving the problem are discussed. Thermal two- and three-dimensional analyses are carried out for the assessment of relative merits in performance due to modifications to the surface. The power handling capability (time to reach melting temperature of beryllium) of the target plate tiles is presented for unswept and swept plasma cases. Results have shown that sweeping the plasma improves the power handling capability by a factor of up to 10. 20 refs., 7 figs., 3 tabs.

  10. Characterizing the Outer Divertor Leg Transition to Full Detachment

    NASA Astrophysics Data System (ADS)

    McLean, A. G.; Allen, S. L.; Fenstermacher, M. E.; Lasnier, C. J.; Meyer, W. H.; Porter, G. D.; Soukhanovskii, V. A.; Bray, B. D.; Carlstrom, T. N.; Leonard, A. W.; Liu, C.; Eldon, D.; Groth, M.; Stangeby, P. C.; Tsui, C. K.

    2013-10-01

    Experiments at DIII-D have explored the transition from an attached to fully detached divertor condition in L- and H-mode with an unprecedented level of detail. Improved divertor Thomson scattering capturing Te <= 1 eV, coupled with high resolution spectroscopic studies of molecular and neutral emissions, and Stark broadening of the deuterium Paschen series provide essential data for modeling the transition to detachment. 2D Te and ne profiles of the outer leg reveal movement of the ionization front away from the plate not replicated in modeling. Measured Paschen and molecular emissions suggest the onset of recombination occurs prior to, and to a greater extent than modeled. These data help guide and expose any missing physics in predictions for detached operation in future devices. This work supported in part by the US Department of Energy under DE-AC52-07NA27344 and DE-FC02-04ER54698.

  11. Preparation of the liquid lithium divertor plates for NSTX

    NASA Astrophysics Data System (ADS)

    Nygren, R. E.; McKee, G. R.; Fordham, J. A.; Lewis, S. A.; Kugel, H.; Ellis, R. A.; Viola, M. E.; O'Dell, J. S.

    2011-10-01

    Each of the four toroidal panels of the liquid lithium divertor being installed in NSTX for operation in the 2010 campaign is a conical section inclined at 22° like the previous graphite divertor tiles. Each panel is a copper plate clad with stainless steel and a surface layer of porous plasma sprayed molybdenum (Mo) that will host lithium deposited from an evaporator. This paper describes the processes in fabrication; these include cutting to rough shape, die pressing into conical sections, machining to near final shape with holes for electrical heaters, thermocouples and a groove for a cooling tube, brazing of the 0.25-mm cladding and vacuum plasma spraying of the Mo coating.

  12. Finger-Circumference-Measuring Device

    NASA Technical Reports Server (NTRS)

    Le, Suy

    1995-01-01

    Easy-to-use device quickly measures circumference of finger (including thumb) on human hand. Includes polytetrafluoroethylene band 1/8 in. wide, bent into loop and attached to tab that slides on scale graduated in millimeters. Sliding tab preloaded with constant-force tension spring, which pulls tab toward closure of loop. Designed to facilitate measurements at various points along fingers to obtain data for studies of volumetric changes of fingers in microgravity. Also used in normal Earth gravity studies of growth and in assessment of diseases like arthritis.

  13. [Multiple finger geodes in children].

    PubMed

    Hoeffel, J C; Oprisescu, B; Bresson, A; Ploier, R; Vidailhet, M

    1993-06-01

    Three pediatric patients with multiple geodes in the fingers are reported. This condition occurs mainly between one and three years and at seven years of age and is more common in winter. Affected fingers are swollen. Roentgenograms disclose several small lucent defects which are usually located in the middle phalanx. Several fingers are usually involved. The erythrocyte sedimentation rate is increased in virtually every case. Resolution occurs spontaneously within a few weeks or months. There is no tendency towards recurrence. Although the condition is inflammatory, exposure to cold is probably a precipitating factor.

  14. Finger-Circumference-Measuring Device

    NASA Technical Reports Server (NTRS)

    Le, Suy

    1995-01-01

    Easy-to-use device quickly measures circumference of finger (including thumb) on human hand. Includes polytetrafluoroethylene band 1/8 in. wide, bent into loop and attached to tab that slides on scale graduated in millimeters. Sliding tab preloaded with constant-force tension spring, which pulls tab toward closure of loop. Designed to facilitate measurements at various points along fingers to obtain data for studies of volumetric changes of fingers in microgravity. Also used in normal Earth gravity studies of growth and in assessment of diseases like arthritis.

  15. Thermal and structural analysis of the TPX divertor

    SciTech Connect

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

    1995-12-31

    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{sup 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.

  16. Experimental investigation of the natural divertor configuration in Heliotron-E

    SciTech Connect

    Hillis, D.L.; Mioduszewski, P.K.; Fowler, R.H.; Rome, J.A.; Motojima, O.; Mizuuchi, T.; Noda, N.; Mutoh, T.; Zushi, H.; Takahashi, R.; Obiki, T.; Iiyoshi, A.; Uo, K.

    1988-01-01

    Particle control with pump limiters and divertors has been successfully demonstrated in a number of present-day tokamaks. In a heliotron/stellarator configuration, plasma flows to the wall in distinct flux bundles, often called ''divertor stripes''. This complicated three-dimensional characteristic of the plasma edge presents a new challenge for active particle control systems such as pump limiters and divertors. The experiment described here has obtained data with an instrumented pump particle collector that is located in the ''natural'' magnetic divertor stripe of Heliotron-E. The particle collector consists of a moveable graphite assembly with single-sided particle collection and active pumping. By scanning the particle collector assembly through the plasma edge of Heliotron-E, the divertor stripe is observed to be about 2-3 cm (FWHM) in width, and pressure rises of 0.01-0.01 mTorr are observed in the particle collector pumping chamber. These measurements have demonstrated that particles leaving the bulk plasma via the divertor stripes can be collected and provide a basis for developing a divertor scheme for particle control in helical systems. Modelling of the Heliotron-E magnetic configuration at the plasma edge is used to determine the collection efficiency of the particle collector in the divertor stripes. The modeling is further extended to describe a helical divertor concept. 18 refs., 6 figs.

  17. First annual report of the Divertor Task Force: Progress and plans

    SciTech Connect

    1995-10-01

    This report describes the work of the Divertor Task Force of the Massachusetts Institute of Technology Plasma Fusion Center, particularly the Task Force`s founding meeting, original research and development needs, organization, and achievements of its first year. The Task Force`s goal is to obtain an increasingly complete physics understanding of existing divertor plasmas, to build analytical and numerical models of the scrape-off-layer divertor plasmas, and to extrapolate them to find design solutions for the high power divertors of ignited tokamak plasmas such as those of ITER and other high performance future tokamaks. 67 refs., 2 figs.

  18. Results from recent detachment experiments in alternative divertor configurations on TCV

    NASA Astrophysics Data System (ADS)

    Theiler, C.; Lipschultz, B.; Harrison, J.; Labit, B.; Reimerdes, H.; Tsui, C.; Vijvers, W. A. J.; Boedo, J. A.; Duval, B. P.; Elmore, S.; Innocente, P.; Kruezi, U.; Lunt, T.; Maurizio, R.; Nespoli, F.; Sheikh, U.; Thornton, A. J.; van Limpt, S. H. M.; Verhaegh, K.; Vianello, N.; the TCV Team; the EUROfusion MST1 Team

    2017-07-01

    Divertor detachment is explored on the TCV tokamak in alternative magnetic geometries. Starting from typical TCV single-null shapes, the poloidal flux expansion at the outer strikepoint is varied by a factor of 10 to investigate the X-divertor characteristics, and the total flux expansion is varied by 70 % to study the properties of the super-X divertor. The effect of an additional X-point near the target is investigated in X-point target divertors. Detachment of the outer target is studied in these plasmas during Ohmic density ramps and with the ion \

  19. Modelling of radiative divertor operation towards detachment in experimental advanced superconducting tokamak

    NASA Astrophysics Data System (ADS)

    Chen, YiPing; Wang, F. Q.; Zha, X. J.; Hu, L. Q.; Guo, H. Y.; Wu, Z. W.; Zhang, X. D.; Wan, B. N.; Li, J. G.

    2013-02-01

    In order to actively control power load on the divertor target plates and study the effect of radiative divertor on plasma parameters in divertor plasmas and heat fluxes to the targets, dedicated experiments with Ar impurity seeding have been performed on experimental advanced superconducting tokamak in typical L-mode discharge with single null divertor configuration, ohmic heating power of 0.5 MW, and lower hybrid wave heating power of 1.0 MW. Ar is puffed into the divertor plasma at the outer target plate near the separatrix strike point with the puffing rate 1.26×1020 s-1. The radiative divertor is formed during the Ar puffing. The SOL/divertor plasma in the L-mode discharge with radiative divertor has been modelled by using SOLPS5.2 code package [V. Rozhansky et al., Nucl. Fusion 49, 025007 (2009)]. The modelling shows the cooling of the divertor plasma due to Ar seeding and is compared with the experimental measurement. The changes of peak electron temperature and heat fluxes at the targets with the shot time from the modelling results are similar to the experimental measurement before and during the Ar impurity seeding, but there is a major difference in time scales when Ar affects the plasma in between experiment and modelling.

  20. Measuring the effect of divertor closure on detachment in DIII-D

    NASA Astrophysics Data System (ADS)

    Moser, Auna; Leonard, A. W.; Petrie, T. W.; Sang, C. F.; Allen, S. L.; McLean, A. G.; Fenstermacher, M. E.; Joseph, I.; Lasnier, C. J.; Makowski, M. A.; Watkins, J. G.; Briesemeister, A. R.

    2015-11-01

    Recent experiments compared the open lower divertor and semi-closed upper divertor in DIII-D to measure the effect of divertor closure on detachment onset and heat flux control, extending past work showing reduced core fueling with the more-closed upper DIII-D divertor. Experiments were performed to determine the extent to which closure may facilitate detachment at collisionalities more relevant to future devices. This work builds on previous experiments that quantified effects of divertor magnetic geometry, including connection length, ∇B-drift direction, incidence angle, and flux expansion; efforts were made to match these parameters while comparing single null configurations in the upper and lower divertor in order to isolate the effects of closure. Experimental measurements coupled with simulation results will help weigh the benefits of a more-closed divertor in facilitating detachment and reducing heat flux against the constraints imposed on the magnetic geometry by a more-closed divertor tile structure, aiding in the design of a future advanced divertor for DIII-D. Supported by the US Department of Energy under DE-FC02-04ER54698, DE-AC52-07NA27344, DE-AC04-94AL85000, and DE-AC05-00OR22725.

  1. Effect of divertor closure and impurities on detachment onset in DIII-D

    NASA Astrophysics Data System (ADS)

    Moser, A. L.; Leonard, A. W.; Groebner, R. J.; Petrie, T. W.; Sang, C. F.; Wang, H.; Allen, S. L.; McLean, A. G.; Fenstermacher, M. E.; Lasnier, C. J.; Makowski, M.; Watkins, J. G.; Briesemeister, A. R.

    2016-10-01

    Heat flux control in future devices requires a detached divertor with upstream parameters compatible with core performance, e.g., at a lower upstream density than presently achievable. Comparison between matched H-mode discharges in the upper and lower divertors of DIII-D demonstrates onset of detachment at a reduced pedestal density for the more-closed geometry of the upper divertor. The upper divertor also produces a lower pedestal density with a less-steep profile than the lower divertor for matched discharges with no additional fueling, presumably due to a reduction in ionization source for the upper divertor. Recent experiments further compare the upper and lower divertors with the addition of impurities injected into the private flux region. These experiments measure the interplay between increased closure and radiating impurities and the effect on divertor detachment, as well as the ability of the more-closed divertor geometry to prevent the accumulation of impurities in the core. Work supported by US DOE under DE-FC02-04ER54698, DE-AC52-07NA27344, DE-AC04-94AL85000, DE-AC05-00OR22725.

  2. Assessment of issues for the MAST divertor biasing experiment

    NASA Astrophysics Data System (ADS)

    Helander, P.; Cohen, R. H.; Fielding, S.; Ryutov, D.

    2001-10-01

    A biasing experiment is being undertaken in the MAST scrape-off layer; the goal is to induce intense convection by a toroidally alternating biasing of divertor tiles. This would lead to a thickening of the SOL and a reduction of the heat load on the divertor plates. In addition, by studying the reaction of a plasma to a varying bias, one can collect new information regarding pre-existing SOL turbulence. We consider the following issues: 1. The bias amplitude required to produce significant SOL broadening; 2. Excitation of shear-flow turbulence in convective cells; 3. The role of magnetic shear; 4. Effects of electrostatic sheaths at the divertor plates; 5. Redistribution of heat fluxes during biasing. We show that a significant effect of the biasing on the SOL structure can be reached at relatively small bias voltages 30 V. We also show that the potential perturbations will be limited to a zone between the X-point and the biased tiles, and will be essentially decoupled from the main SOL plasma. Preliminary experimental results may be shown.

  3. Modeling of Divertor Plates in the Compact Toroidal Hybrid

    NASA Astrophysics Data System (ADS)

    Hartwell, G. J.; Small, C. M.; Ennis, D. A.; Hanson, J. D.; Knowlton, S. F.; Maurer, D. A.

    2014-10-01

    In long pulse length stellarator experiments, edge island divertors can be used as a method of plasma particle and heat exhaust. Knowledge of the detailed power loading on these structures and its relationship to the long connection length scrape off layer physics is a new Compact Toroidal Hybrid research thrust. We report the results of connection length studies for divertor plates to be installed in the Compact Toroidal Hybrid (CTH), a five field period torsatron with R0 = 0 . 75 m, ap ~ 0 . 2 m, and B <= 0 . 7 T. For these studies, CTH will be operated as a pure stellarator with no ohmically generated plasma current. The CTH edge rotational transform can be varied from tvac (a) = 0.02-0.35 by adjusting the ratio of currents in the helical and toroidal field coils. A poloidal field coil is used to adjust the shear of the rotational transform profile, and hence the size of edge islands, while the phase of the island is rotated with a set of five error coils producing an n = 1 perturbation. For the studies conducted, a magnetic configuration with a large n = 1, m = 3 magnetic island at the edge is generated. Results from multiple possible divertor plate locations relative to the island structure will be presented. This work is supported by U.S. Department of Energy Grant No. DE-FG02-00ER54610.

  4. Fast reciprocating Langmuir probe for the DIII-D divertor

    NASA Astrophysics Data System (ADS)

    Watkins, J. G.; Hunter, J.; Tafoya, B.; Ulrickson, M.; Watson, R. D.; Moyer, R. A.; Cuthbertson, J. W.; Gunner, G.; Lehmer, R.; Luong, P.; Hill, D. N.; Mascaro, M.; Robinson, J. I.; Snider, R.; Stambaugh, R.

    1997-01-01

    A new reciprocating Langmuir probe was used to measure density and temperature profiles, ion flow, and potential fluctuation levels from the lower divertor floor up to the X point on the DIII-D Tokamak. This probe is designed to make fast (2 kHz swept, 20 kHz Mach, 500 kHz Vfloat) measurements with 2 mm spatial resolution in the region where the largest gradients on the plasma open flux tubes are found and therefore provide the best benchmarks for scrap-off layer and divertor numerical models. Profiles are constructed using the 300 ms time history of the probe measurements during the 25 cm reciprocating stroke. Both single and double null plasmas can be measured and compared with a 20 Hz divertor Thomson scattering system. The probe head is constructed of four different kinds of graphite to optimize the electrical and thermal characteristics. Electrically insulated pyrolytic graphite rings act as a heat shield to absorb the plasma heat flux on the probe shaft and are mounted on a carbon/carbon composite core for mechanical strength. The Langmuir probe sampling tips are made of a linear carbon fiber composite. The mechanical, electrical, data acquisition, and power supply systems will be described. Initial measurements will also be presented.

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

    NASA Astrophysics Data System (ADS)

    Groth, M.; Brezinsek, S.; Belo, P.; Brix, M.; Calabro, G.; Chankin, A.; Clever, M.; Coenen, J. W.; Corrigan, G.; Drewelow, P.; Guillemaut, C.; Harting, D.; Huber, A.; Jachmich, S.; Järvinen, A.; Kruezi, U.; Lawson, K. D.; Lehnen, M.; Maggi, C. F.; Marchetto, C.; Marsen, S.; Maviglia, F.; Meigs, A. G.; Moulton, D.; Silva, C.; Stamp, M. F.; Wiesen, S.

    2015-08-01

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

  6. Neural correlates of finger gnosis.

    PubMed

    Rusconi, Elena; Tamè, Luigi; Furlan, Michele; Haggard, Patrick; Demarchi, Gianpaolo; Adriani, Michela; Ferrari, Paolo; Braun, Christoph; Schwarzbach, Jens

    2014-07-02

    Neuropsychological studies have described patients with a selective impairment of finger identification in association with posterior parietal lesions. However, evidence of the role of these areas in finger gnosis from studies of the healthy human brain is still scarce. Here we used functional magnetic resonance imaging to identify the brain network engaged in a novel finger gnosis task, the intermanual in-between task (IIBT), in healthy participants. Several brain regions exhibited a stronger blood oxygenation level-dependent (BOLD) response in IIBT than in a control task that did not explicitly rely on finger gnosis but used identical stimuli and motor responses as the IIBT. The IIBT involved stronger signal in the left inferior parietal lobule (IPL), bilateral precuneus (PCN), bilateral premotor cortex, and left inferior frontal gyrus. In all regions, stimulation of nonhomologous fingers of the two hands elicited higher BOLD signal than stimulation of homologous fingers. Only in the left anteromedial IPL (a-mIPL) and left PCN did signal strength decrease parametrically from nonhomology, through partial homology, to total homology with stimulation delivered synchronously to the two hands. With asynchronous stimulation, the signal was stronger in the left a-mIPL than in any other region, possibly indicating retention of task-relevant information. We suggest that the left PCN may contribute a supporting visuospatial representation via its functional connection to the right PCN. The a-mIPL may instead provide the core substrate of an explicit bilateral body structure representation for the fingers that when disrupted can produce the typical symptoms of finger agnosia.

  7. Multimodal biometric authentication based on the fusion of finger vein and finger geometry

    NASA Astrophysics Data System (ADS)

    Kang, Byung Jun; Park, Kang Ryoung

    2009-09-01

    We propose a new multimodal biometric recognition based on the fusion of finger vein and finger geometry. This research shows three novelties compared to previous works. First, this is the first approach to combine the finger vein and finger geometry information at the same time. Second, the proposed method includes a new finger geometry recognition based on the sequential deviation values of finger thickness extracted from a single finger. Third, we integrate finger vein and finger geometry by a score-level fusion method based on a support vector machine. Results show that recognition accuracy is significantly enhanced using the proposed method.

  8. Investigations on the heat flux and impurity for the HL-2M divertor

    NASA Astrophysics Data System (ADS)

    Zheng, G. Y.; Cai, L. Z.; Duan, X. R.; Xu, X. Q.; Ryutov, D. D.; Cai, L. J.; Liu, X.; Li, J. X.; Pan, Y. D.

    2016-12-01

    The controllability of the heat load and impurity in the divertor is very important, which could be one of the critical problems to be solved in order to ensure the success for a steady state tokamak. HL-2M has the advantage of the poloidal field (PF) coils placed inside the demountable toroidal field (TF) coils and close to the main plasma. As a result, it is possible to make highly accurate configuration control of the advanced divertor for HL-2M. The divertor target geometry of HL-2M has been designed to be compatible with different divertor configurations to study the divertor physics and support the high performance plasma operations. In this paper, the heat loads and impurities with different divertor configurations, including the standard X-point divertor, the snowflake-minus divertor and two tripod divertor configurations for HL-2M, are investigated by numerical simulations with the SOLPS5.0 code under the current design of the HL-2M divertor geometry. The plasmas with different conditions, such as the low discharge parameters with {{I}\\text{p}}   =  0.5 MA at the first stage of HL-2M and the high parameters with {{I}\\text{p}}   =  2.0 MA during the normal operations, are simulated. The heat load profiles and the impurity distributions are obtained, and the control of the peak heat load and the effect of impurity on the core plasma are discussed. The compatibility of different divertor configurations for HL-2M is also evaluated. It is seen that the excellent compatibility of different divertor configurations with the current divertor geometry has been verified. The results show that the snowflake-minus divertor and the tripod divertor with {{d}x}=30 \\text{cm} present good performance in terms of the heat load profiles and the impurity distributions under different conditions, which may not have a big effect on the core plasma. In addition, it is possible to optimize the distance between the two X-points, {{d}x} , to achieve a better

  9. Near-infrared spectroscopy for divertor plasma diagnosis and control in DIII-D tokamaka)

    NASA Astrophysics Data System (ADS)

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

    2014-11-01

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

  10. Design, Engineering, and Testing for the Alcator C-Mod Outer Divertor Upgrade

    NASA Astrophysics Data System (ADS)

    Harrison, S.; Vieira, R.; Lipschultz, B.; Ellis, R.; Karnes, D.; Doody, J.; Zhou, L.; Titus, P.; Zhang, H.; Beck, W.; Granetz, R.

    2012-10-01

    Alcator C-mod's major outer divertor upgrade will enable significant advances in our understanding of reactor relevant physics and operations. Two primary features of the new outer divertor are its toroidally continuous design (electrical and mechanical), and ability to be operated up to or independently heated to 600 C. Full control of the divertor PFC temperature from ambient vessel temperature to 600 C, will enable new and important tokamak research into the temperature dependence of fuel retention, PFC deposition and erosion, and divertor recycling. Significant design, analysis, and testing is underway to complete this important and challenging upgrade, which will provide valuable information for ITER and future reactors. Among other aspects of the innovative approach, the divertor plate supports, halo current shunts, and thermal shield assemblies will be discussed. The divertor supports enable pure radial motion of the divertor ring as it expands thermally and robustness to massive disruption induced electro-mechanical loads. Halo current shunts conduct 400kA in an 8T magnetic field and allow for divertor displacement relative to the vessel. Thermal shielding significantly reduces radiation and conduction to surrounding vessel structures.

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

    SciTech Connect

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

    2014-11-15

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

  12. Gyrokinetic projection of the divertor heat-flux width from present tokamaks to ITER

    DOE PAGES

    Chang, Choong Seock; Ku, Seung -Hoe; Loarte, Alberto; ...

    2017-07-11

    Here, the XGC1 edge gyrokinetic code is used to study the width of the heat-flux to divertor plates in attached plasma condition. The flux-driven simulation is performed until an approximate power balance is achieved between the heat-flux across the steep pedestal pressure gradient and the heat-flux on the divertor plates.

  13. HL-2M Divertor Geometry Exploration with SOLPS5.0

    NASA Astrophysics Data System (ADS)

    Cui, Xuewu; Pan, Yudong; Cui, Zhengying; Li, Jiaxian; Zhang, Jinhua; Mao, Rui

    2013-12-01

    One of the critical issues to be solved for HL-2M is the power and particle exhaust. Divertor target plate geometry strongly influences the plasma profiles by controlling the neutral recycling pattern, which has in turn a strong effect on the symmetry and stability of the divertor plasma and finally on the whole edge region. The numerical simulation software SOLPS5.0 Package is used to design and explore the divertor target plates for HL-2M. We choose two divertor geometries, and assess the heat flux on the target plates and first wall, then further discuss the divertor plasma parameters, and how private flux baffling affects both neutral recirculation pattern and pumping efficiency.

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

    SciTech Connect

    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.

  15. The dynamics of coherent scrape-off layer structures in a snowflake divertor

    NASA Astrophysics Data System (ADS)

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

    2008-11-01

    A characteristic feature of a snowflake divertor is the quadratic dependence of the poloidal magnetic field strength vs the distance from the field null. Compared to a standard X-point divertor, where the magnetic field dependence over distance is linear, this leads to significant changes in the geometry of flux tubes passing in the vicinity of the null. In particular, squeezing of flux tubes by the magnetic shear becomes stronger; the field line mapping from the midplane to the divertor plate indicates much higher poloidal velocities of plasma filaments near the divertor plates. Thus, significant changes are expected in the dynamics of coherent structures (sometimes called ``blobs'') in the scrape-off layer. An analysis of the dynamical effects associated with curvature drive, divertor boundary conditions, and strong magnetic shearing is presented. Regimes of enhanced blob transport are identified. Prepared by LLNL under Contract DE-AC52-07NA27344.

  16. Reconstruction of Detached Divertor Plasma Conditions in DIII-D Using Spectroscopic and Probe Data

    SciTech Connect

    Stangeby, P; Fenstermacher, M

    2004-12-03

    For some divertor aspects, such as detached plasmas or the private flux zone, it is not clear that the controlling physics has been fully identified. This is a particular concern when the details of the plasma are likely to be important in modeling the problem--for example, modeling co-deposition in detached inner divertors. An empirical method of ''reconstructing'' the plasma based on direct experimental measurements may be useful in such situations. It is shown that a detached plasma in the outer divertor leg of DIII-D can be reconstructed reasonably well using spectroscopic and probe data as input to a simple onion-skin model and the Monte Carlo hydrogenic code, EIRENE. The calculated 2D distributions of n{sub e} and T{sub e} in the detached divertor were compared with direct measurements from the divertor Thomson scattering system, a diagnostic capability unique to DIII-D.

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

    NASA Astrophysics Data System (ADS)

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

    2016-02-01

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

  18. Calorimeter probe for the DIII-D divertor

    NASA Astrophysics Data System (ADS)

    Watkins, J. G.; Lasnier, C. J.; Whyte, D. G.; Stangeby, P. C.; Ulrickson, M. A.

    2003-03-01

    Heat flux measurements of the DIII-D divertor plate have been obtained with 6 mm spatial resolution using a calorimeter probe. These measurements complement the infrared camera system normally used for heat flux measurements on DIII-D but at higher-spatial resolution. The calorimeter probe is inserted into the tokamak from below to a position which is flush with the lower divertor plate tiles using the divertor materials experimental station (DiMES). The DiMES mechanism allows for retraction of the probe behind a gate valve and removal from the tokamak for modification or calibration. A 6 mm diameter insulated graphite cylinder for collecting energy is mounted within a standard DiMES sample. A 0.8 mm diameter thermocouple, installed 4 mm below the surface, provides a measurement of the temperature during and after the plasma exposure. The 80 ms time constant for the measurement is fast enough to determine heat flux changes during the 5 s plasma discharge and heat flux profiles have been obtained using both fixed strike points and slow strike point sweeps across the calorimeter. Special electronics and isolation is necessary as the sample is in direct electrical contact with the plasma. The calorimeter observes approximately 100 °C temperature rise over one tokamak discharge. The thermocouple signals are typically less than 1 mV and must be amplified near the vacuum feedthrough, passed through a low-pass filter to eliminate magnetic pickup, isolated, and sent to the data acquisition system approximately 8 m away. Initial measurements are included.

  19. The Effect of Magnetic Balance and Particle Drifts on Radiating Divertor Behavior in DIII-D

    SciTech Connect

    Petrie, T; Porter, G; Brooks, N; Fenstermacher, M; Ferron, J; Groth, M; Hyatt, A; La Haye, R; Lasnier, C; Leonard, A; Politzer, P; Rensink, M; Schaffer, M; Wade, M; Watkins, J; West, W

    2008-10-14

    Success of the puff-and-pump radiating divertor approach depends sensitively on both the divertor magnetic geometry and the ion B x {del}B drift direction. In the puff-and-pump scenario used in this study, argon impurities were injected into the private flux region, while plasma flows into both the inner and outer divertors were enhanced by a combination of particle pumping near both divertor targets and deuterium gas puffing upstream of the divertor targets. For single-null (SN) configurations, argon accumulation was 2-3 times lower in the main plasma when the ion B x {del}B drift was directed away from the divertor. The puff-and-pump approach was much less effective in screening argon from the main plasma of double-null (DN) discharges than of SN discharges, such that argon impurities accumulated in the main plasma of DNs at a rate {approx}2-3 times higher than in corresponding SNs. Regardless of which divertor in DN had argon injection, argon accumulated in the divertor that was opposite the B x {del}B drift direction. The argon density in the main plasma during puff-and-pump operation fell by a factor of three for dRsep {ge} +0.4 cm when the ion B x {del}B drift was directed away from the dominant divertor, and this represents the transition from DN to SN behavior during puff-and-pump application. Comparison of identically-prepared SN H-mode plasmas showed that core density control of deuterium and the argon was far more sensitive to the ion B x {del}B drift direction than to divertor closure in DIII-D.

  20. The role of atomic and molecular physics for dissipative divertor operation in helium and deuterium plasmas

    NASA Astrophysics Data System (ADS)

    Canik, J. M.

    2016-10-01

    Recent experiments in DIII-D helium plasmas are examined to resolve the role of atomic and molecular physics in major discrepancies between experiment and modeling of dissipative divertor operation. Helium operation removes the complicated molecular processes of deuterium plasmas that are a prime candidate for the inability of standard fluid models (SOLPS, UEDGE) to reproduce dissipative divertor operation, primarily the consistent under-prediction of radiated power. With helium fueling, a high-recycling divertor was established with divertor densities increasing to ne,div >= 3 ×1020m-3 and temperatures decreasing to Te,div <= 2 eV as measured by divertor Thomson scattering (DTS). The electron pressure, pe , div decreased gradually with increasing density to less than 30% of the low density value. However, the ion flux to the divertor target did not decrease until the highest densities and lowest temperatures, Te,div <= 2 eV. In contrast, with deuterium operation, increasing density leads to a rapid transition from Te,div >= 10 eV to Te,div <= 3 eV, though both pe , div and ion flux do not decrease until Te,div <= 2 eV. These differences indicate an important role for molecular and atomic physics in the dynamics of divertor dissipation. Initial SOLPS modeling has reproduced ne and Te profiles at the midplane and divertor target, as well as the spatial structure of radiation patterns measured in moderate density helium plasmas. However, the modeled divertor radiation is less than measured, similar to deuterium simulations, suggesting processes more universal than species-specific atomic or molecular physics may be the source of radiation deficit. Detailed assessments of ne, Te profiles in the divertor volume, uniquely determined at DIII-D using DTS, are made along with analysis of measured and modeled line radiation to shed more light on these intriguing findings. Supported by the US DOE under DE-AC05-00OR22725.

  1. Ballooning Modes in the Systems Stabilized by Divertors

    SciTech Connect

    Arsenin, V.V.; Skovoroda, A.A.; Zvonkov, A.V.

    2005-01-15

    MHD stability of a plasma in systems with closed magnetic field lines and open systems containing the nonparaxial stabilizing cells with large field lines curvature, in particular, divertors is analyzed. It is shown that population of particles trapped in such cells has a stabilizing effect not only on flute modes, but also on ballooning modes that determine the {beta} limit. At kinetic description that accounts for different effect of trapped and passing particles on perturbations, {beta} limit permitted by stability may be much greater then it follows from MHD model.

  2. Performance characteristics of the DIII-D advanced divertor cryopump

    SciTech Connect

    Menon, M.M.; Maingi, R.; Wade, M.R.; Baxi, C.B.; Campbell, G.L.; Holtrop, K.L.; Hyatt, A.W.; Laughon, G.J.; Makariou, C.C.; Mahdavi, M.A.; Reis, E.E.; Schaffer, M.J.; Schaubel, K.M.; Scoville, J.T.; Smith, J.P.; Stambaugh, R.D.

    1993-10-01

    A cryocondensation pump, cooled by forced flow of two-phase helium, has been installed for particle exhaust from the divertor region of the DIII-D tokamak. The Inconel pumping surface is of coaxial geometry, 25.4 mm in outer diameter and 11.65 m in length. Because of the tokamak environment, the pump is designed to perform under relatively high pulsed heat loads (300 Wm{sup {minus}2}). Results of measurements made on the pumping characteristics for D{sub 2}, H{sub 2}, and Ar are discussed.

  3. Crossed-field divertor for a plasma device

    DOEpatents

    Kerst, Donald W.; Strait, Edward J.

    1981-01-01

    A divertor for removal of unwanted materials from the interior of a magnetic plasma confinement device includes the division of the wall of the device into segments insulated from each other in order to apply an electric field having a component perpendicular to the confining magnetic field. The resulting crossed-field drift causes electrically charged particles to be removed from the outer part of the confinement chamber to a pumping chamber. This method moves the particles quickly past the saddle point in the poloidal magnetic field where they would otherwise tend to stall, and provides external control over the rate of removal by controlling the magnitude of the electric field.

  4. Surface erosion issues and analysis for dissipative divertors

    SciTech Connect

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

    1994-08-01

    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.

  5. Volume Recombination in Alcator C-Mod Divertor Plasmas

    NASA Astrophysics Data System (ADS)

    Terry, J. L.

    1997-11-01

    Volume recombination has been predicted(See, for example, A. Loarte, Proc. 12th PSI Conf, J. Nucl. Mater (1996) I9, in press.) to be a significant sink for plasma ions under the detached divertor conditions achieved on many tokamaks. This volume recombination sink was observed initially in Alcator C-Mod and shown to be a major fraction of the ion loss. Signatures of recombination have now been observed on DIII-D(R.C. Isler, et al., paper submitted for publication), Asdex-UG (B. Napiontek, et al. 24th EPS Conf., Berchtesgaden, Germany, 1997, P4.007, in press.), and JET(R.D. Monk, et al. 24th EPS Conf., Berchtesgaden, Germany, 1997, P1.030, in press.). It is important primarily because the recombined atoms are not accelerated through the sheath - thus reducing divertor plate sputtering, and because most of the potential energy of recombination (13.6 eV) is released as radiation before the ion strikes the plate. The Alcator C-Mod measurements show that the recombination occurs in low Te ( ~1 eV), high ne ( ~1× 10^21 m-3) regions, and is significantly larger in detached regions. At the inboard, detached divertor plate the measured volume recombination rate is typically greater than the rate of ion collection at that plate and is about an order of magnitude higher than on the attached, outer plate. These spatially resolved measurements also show that the recombination rate is peaked near the strike point and imply that the recombination is occurring close to the plate surface. The C-Mod observations about the magnitude and spatial distribution of the recombination are consistent with the modelling of similar discharges(F. Wising et al., Contrib. Plasma Phys. 36, p 136 (1996).). The experimental evidence for recombination is found in the deuterium emission spectra from the divertor, in particular in the Balmer- and/or Lyman-series. The spectra show that the dominant recombination mechanism is 3-body recombination into excited states of deuterium and that the populations

  6. Review on mallet finger treatment.

    PubMed

    Cheung, Jason Pui Yin; Fung, Boris; Ip, Wing Yuk

    2012-01-01

    Mallet finger is a common injury involving either an extensor tendon rupture at its insertion or an avulsion fracture involving the insertion of the terminal extensor tendon. It is usually caused by a forceful blow to the tip of the finger causing sudden flexion or a hyperextension injury. Fracture at the dorsal aspect of the base of the distal phalanx is commonly associated with palmar subluxation of the distal phalanx. Most mallet finger injuries are recommended to be treated with immobilisation of the distal interphalangeal joint in extension by splints. There is no consensus on the type of splint and the duration of use. Most studies have shown comparable results with different splints. Surgical fixation is still indicated in certain conditions such as open injuries, avulsion fracture involving at least one third of the articular surface with or without palmar subluxation of the distal phalanx and also failed splinting treatment.

  7. Mesofluidic controlled robotic or prosthetic finger

    DOEpatents

    Lind, Randall F; Jansen, John F; Love, Lonnie J

    2013-11-19

    A mesofluidic powered robotic and/or prosthetic finger joint includes a first finger section having at least one mesofluidic actuator in fluid communication with a first actuator, a second mesofluidic actuator in fluid communication with a second actuator and a second prosthetic finger section pivotally connected to the first finger section by a joint pivot, wherein the first actuator pivotally cooperates with the second finger to provide a first mechanical advantage relative to the joint point and wherein the second actuator pivotally cooperates with the second finger section to provide a second mechanical advantage relative to the joint point.

  8. Distribution of Hydrogen Isotopes, Carbon and Beryllium on In-Vessel Surfaces in the Various JET Divertors

    SciTech Connect

    Coad, J.P.; Rubel, M.; Bekris, N.; Brennan, D.; Hole, D.; Likonen, J.; Vainonen-Ahlgren, E

    2005-07-15

    JET has operated with divertors of differing geometries since 1994. Impurities accumulated in the inner leg of all the divertors, and operation of the first (Mk I) divertor with beryllium tiles demonstrated that most are eroded from the main chamber walls and swept along the scrape-off layer to the inner divertor. Carbon deposited at the inner divertor is then locally transported to shadowed regions such as the inner louvres, where, for example, most of the tritium was trapped during the deuterium-tritium experiment (DTE1). Factors affecting these transport processes (e.g. temperature) are important for ITER, but are not well understood.

  9. Fingering inside the coffee ring

    NASA Astrophysics Data System (ADS)

    Weon, Byung Mook; Je, Jung Ho

    2013-01-01

    Colloidal droplets including micro- and nanoparticles generally leave a ringlike stain, called the “coffee ring,” after evaporation. We show that fingering emerges during evaporation inside the coffee ring, resulting from a bidispersed colloidal mixture of micro- and nanoparticles. Microscopic observations suggest that finger formation is driven by competition between the coffee-ring and Marangoni effects, especially when the inward Marangoni flow is overwhelmed by the outward coffee-ring flow. This finding could help to understand the variety of the final deposition patterns of colloidal droplets.

  10. SOLPS modeling of the effect on plasma detachment of closing the lower divertor in DIII-D

    NASA Astrophysics Data System (ADS)

    Sang, C. F.; Stangeby, P. C.; Guo, H. Y.; Leonard, A. W.; Covele, B.; Lao, L. L.; Moser, A. L.; Thomas, D. M.

    2017-02-01

    Scrape-off layer plasma simulation modeling has been carried out to assess the effect of tightly closing the lower divertor in DIII-D, which at present is almost fully open, on the achievement of cold dissipative/detached divertor conditions. To isolate the impact of other factors on the divertor plasma solution and to make direct comparisons, most of the parameters including the meshes were kept as similar as possible. Only the neutral baffling was modified to compare a fully open divertor with a tightly closed one. The modeling shows that the tightly closed divertor greatly improves trapping of recycling neutrals, thereby increasing radiative and charge exchange losses in the divertor and reducing the electron temperature T et and deposited power density q dep at the target plate. Furthermore, the closed structure enables the divertor plasma to enter into highly dissipative and detached divertor conditions at a significantly lower upstream density. The effects of divertor closure on the neutral density and pressure, and their correlation with the divertor plasma conditions are also demonstrated. The effect of molecular D2-ion D+ elastic collisions and neutral-neutral collisions on the divertor plasma solution are assessed.

  11. SOLPS modeling of the effect on plasma detachment of closing the lower divertor in DIII-D

    DOE PAGES

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

    2016-12-15

    SOLPS modeling has been carried out to assess the effect of tightly closing the lower divertor in DIII-D, which at present is almost fully open, on the achievement of cold dissipative/detached divertor conditions. To isolate the impact of other factors on the divertor plasma solution and to make direct comparisons, most of the parameters including the meshes were kept as similar as possible. Only the neutral baffling was modified to compare a fully open divertor with a tightly closed one. The modeling shows that the tightly closed divertor greatly improves trapping of recycling neutrals, thereby increasing radiative and charge exchangemore » losses in the divertor and reducing the electron temperature Tet and deposited power density qdep at the target plate. Furthermore, the closed structure enables the divertor plasma to enter into highly dissipative and detached divertor conditions at a significantly lower upstream density. The effects of divertor closure on the neutral density and pressure, and their correlation with the divertor plasma conditions are also demonstrated. The effect of molecular D2- ion D+ elastic collisions and neutral-neutral collisions on the divertor plasma solution are assessed.« less

  12. SOLPS modeling of the effect on plasma detachment of closing the lower divertor in DIII-D

    SciTech Connect

    Sang, C. F.; Stangeby, P. C.; Guo, H. Y.; Leonard, A. W.; Covele, B.; Lao, L. L.; Moser, A. L.; Thomas, D. M.

    2016-12-15

    SOLPS modeling has been carried out to assess the effect of tightly closing the lower divertor in DIII-D, which at present is almost fully open, on the achievement of cold dissipative/detached divertor conditions. To isolate the impact of other factors on the divertor plasma solution and to make direct comparisons, most of the parameters including the meshes were kept as similar as possible. Only the neutral baffling was modified to compare a fully open divertor with a tightly closed one. The modeling shows that the tightly closed divertor greatly improves trapping of recycling neutrals, thereby increasing radiative and charge exchange losses in the divertor and reducing the electron temperature Tet and deposited power density qdep at the target plate. Furthermore, the closed structure enables the divertor plasma to enter into highly dissipative and detached divertor conditions at a significantly lower upstream density. The effects of divertor closure on the neutral density and pressure, and their correlation with the divertor plasma conditions are also demonstrated. The effect of molecular D2- ion D+ elastic collisions and neutral-neutral collisions on the divertor plasma solution are assessed.

  13. Experiments and computational modeling focused on divertor and SOL optimization for advanced tokamak operation on DIII-D

    NASA Astrophysics Data System (ADS)

    Allen, S. L.; Boedo, J. A.; Bozek, A. S.; Brooks, N. H.; Carlstrom, T. N.; Casper, T. A.; Colchin, R. J.; Evans, T. E.; Fenstermacher, M. E.; Friend, M. E.; Isler, R. C.; Jayakumar, R.; Lasnier, C. J.; Leonard, A. W.; Mahdavi, M. A.; Maingi, R.; McKee, G. R.; Moyer, R. A.; Murakami, M.; Osborne, T. H.; O'Neill, R. C.; Petrie, T. W.; Porter, G. D.; Ramsey, A. T.; Schaffer, M. J.; Stangeby, P. C.; Stambaugh, R. D.; Wade, M. R.; Watking, J. G.; West, W. P.; Whyte, D. G.; Wolf, N. S.

    2001-03-01

    We present the results from DIII-D experiments and modeling focused on the divertor issues of an `Advanced Tokamak' (AT). Operation at high plasma pressure β with good energy confinement H requires core and divertor plasma shaping and current profile J( r) control with ECH current drive. Transport modeling indicates that the available DIII-D ECH power determines a density and temperature regime for sustained DIII-D AT experiments. We demonstrate that a high-δ, unbalanced double null divertor with cryopumping (D-2000) is a flexible AT divertor. Impurity levels in AT experiments have been reduced by careful alignment of the divertor tiles; this, in turn has changed the time evolution of the core J( r) profiles. New physics has been observed near the X-point and private flux regions, including flow reversal and recombination, that is important in understanding and controlling the flows and thereby the radiation in the divertor region, which reduces the divertor heat flux.

  14. Evaluation of a monoblock divertor design for the ITER tokamak

    SciTech Connect

    Lee, Y.T.; Hoffman, M.A.; Hafez, M.

    1996-12-31

    A subcooled nucleate boiling computer code (with 3D heat conduction in solid and 1D forced convection in fluid) that incorporates a good estimation of the single-phase and two-phase pressure drop was developed to evaluate a monoblock design of the divertor with smooth tubes as well as a wide variety of cooling designs. Using one of the monoblock divertor designs proposed by the European International Thermonuclear Experimental Reactor (ITER) team as of March 1995, it was found that under a normal steady state operating condition with a peak heat flux of about 5 MW/m{sup 2}, the water flow remained in the single phase liquid regime. Under an abnormal operating condition with a peak heat flux of about 20 MW/m{sup 2}, the partially developed boiling (PDB) regime occurred where the local critical heat flux safety factor (SF{sub CHF}=V@CHF(z)/q{sub ({theta}}=0{degree})), was estimated to be about 1.4 using the Tong-75 CHF correlation. This indicates that further increases in the magnitude of the heat flux beyond 20 MW/m{sup 2} may raise safety concerns for the design. By increasing the mass flux, decreasing the inlet water temperature, or increasing the inlet water pressure, the CHF safety margin of the design can be increased without inserting twisted tapes inside cooling tubes. 8 refs., 6 figs.

  15. On the W7-X divertor performance under detached conditions

    NASA Astrophysics Data System (ADS)

    Feng, Y.; Beidler, C. D.; Geiger, J.; Helander, P.; Hölbe, H.; Maassberg, H.; Turkin, Y.; Reiter, D.; W7-X Team

    2016-12-01

    We present a theoretical/numerical predictive analysis of the performance of the W7-X island divertor under conditions of detachment characterized by intensive radiation. The analysis is based on EMC3-Eirene simulations and the earlier W7-AS experimental and numerical experience. Carbon is employed as a representative radiator. The associated drawbacks, i.e. core contamination and recycling degradation (reduced recycling flux), are evaluated by determining the carbon density at the last closed flux surface (LCFS) and the neutral pressure in the divertor chamber. Optimum conditions are explored in both configuration and plasma parameter space. This study aims to identify the key geometric/magnetic and plasma parameters that affect the performance of detached plasmas in W7-X. Emphasis is placed on what occurs when the islands are enlarged far beyond the maximum size available in W7-AS and whether an island size limit for optimal detachment operation exists, and why. Further issues addressed are the power removal ability of the W7-X edge islands, potentially limiting factors, compatibility between particle and power exhaust, and particle refueling capability of the recycling neutrals.

  16. Near infrared spectroscopy of the DIII-D divertor

    NASA Astrophysics Data System (ADS)

    McLean, A. G.; Soukhanovskii, V. A.; Brooks, N. H.; Bray, B. D.; Carlstrom, T. N.

    2012-10-01

    A high speed, high resolution near infrared (NIR) spectrometer has been installed at DIII-D to make first-of-its-kind observations of the 0.8-2.2 μm region in a tokamak divertor. The goals of this diagnostic are (1) to study Paschen spectra for line-averaged measurement of low temperature plasma parameters, (2) to benchmark the chemical and physically sputtered sources of neutral carbon using the lineshape of the CI, 910 nm multiplet, and (3) to quantify contamination of the 0.75-1.1 μm region where Thomson-shifted laser light is measured by the Thomson scattering diagnostic. Diagnostic capabilities include a 300 mm, f/3.9 design, 300-2400 Gr/mm gratings providing optical resolution of ˜0.65-0.04 nm, and readout at up to 900 frames/second. Data are presented in L-mode plasmas, and in H-mode between ELMs and during the ELM peak. Results acquired by this diagnostic will be applied to design of a proposed divertor Thomson diagnostic for NSTX-U and aid validation of the Thomson system on ITER.

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

    SciTech Connect

    Farmer, W. A.; Ryutov, D. D.

    2013-09-15

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

  18. Surface heat loads on the ITER divertor vertical targets

    NASA Astrophysics Data System (ADS)

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

    2017-04-01

    The heating of tungsten monoblocks at the ITER divertor vertical targets is calculated using the heat flux predicted by three-dimensional ion orbit modelling. The monoblocks are beveled to a depth of 0.5 mm in the toroidal direction to provide magnetic shadowing of the poloidal leading edges within the range of specified assembly tolerances, but this increases the magnetic field incidence angle resulting in a reduction of toroidal wetted fraction and concentration of the local heat flux to the unshadowed surfaces. This shaping solution successfully protects the leading edges from inter-ELM heat loads, but at the expense of (1) temperatures on the main loaded surface that could exceed the tungsten recrystallization temperature in the nominal partially detached regime, and (2) melting and loss of margin against critical heat flux during transient loss of detachment control. During ELMs, the risk of monoblock edge melting is found to be greater than the risk of full surface melting on the plasma-wetted zone. Full surface and edge melting will be triggered by uncontrolled ELMs in the burning plasma phase of ITER operation if current models of the likely ELM ion impact energies at the divertor targets are correct. During uncontrolled ELMs in pre-nuclear deuterium or helium plasmas at half the nominal plasma current and magnetic field, full surface melting should be avoided, but edge melting is predicted.

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

    SciTech Connect

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

    1999-09-15

    The Advanced Limiter-divertor Plasma-facing Systems (ALPS) program was initiated in order to evaluate the potential for improved performance and lifetime for plasma-facing systems. The main goal of the program is to demonstrate the advantages of advanced limiter/divertor systems over conventional systems in terms of power density capability, component lifetime, and power conversion efficiency, while providing for safe operation and minimizing impurity concerns for the plasma. Most of the work to date has been applied to free surface liquids. A multi-disciplinary team from several institutions has been organized to address the key issues associated with these systems. The main performance goals for advanced limiters and diverters are a peak heat flux of >50 MW/m{sup 2},elimination of a lifetime limit for erosion, and the ability to extract useful heat at high power conversion efficiency ({approximately}40%). The evaluation of various options is being conducted through a combination of laboratory experiments, modeling of key processes, and conceptual design studies. The current emphasis for the work is on the effects of free surface liquids on plasma edge performance.

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

    NASA Astrophysics Data System (ADS)

    Covele, B.; Valanju, P.; Kotschenreuther, M.; Mahajan, S.

    2014-07-01

    It is found that the X-divertor (XD) configuration (Kotschenreuther et al 2004 Proc. 20th Int. Conf. on Fusion Energy (Vilamoura, Portugal, 2004) (Vienna: IAEA) CD-ROM file [IC/P6-43] www-naweb.iaea.org/napc/physics/fec/fec2004/datasets/index.html, Kotschenreuther et al 2006 Proc. 21st Int. Conf. on Fusion Energy 2006 (Chengdu, China, 2006) (Vienna: IAEA), CD-ROM file [IC/P7-12] www-naweb.iaea.org/napc/physics/FEC/FEC2006/html/index.htm, Kotschenreuther et al 2007 Phys. Plasmas 14 072502) can be made with the conventional poloidal field (PF) coil set on ITER (Tomabechi et al and Team 1991 Nucl. Fusion 31 1135), where all PF coils are outside the TF coils. Starting from the standard divertor, a sequence of desirable XD configurations are possible where the PF currents are below the present maximum design limits on ITER, and where the baseline divertor cassette is used. This opens the possibility that the XD could be tested and used to assist in high-power operation on ITER, but some further issues need examination. Note that the increased major radius of the super-X-divertor (Kotschenreuther et al 2007 Bull. Am. Phys. Soc. 53 11, Valanju et al 2009 Phys. Plasmas 16 5, Kotschenreuther et al 2010 Nucl. Fusion 50 035003, Valanju et al 2010 Fusion Eng. Des. 85 46) is not a feature of the XD geometry. In addition, we present an XD configuration for K-DEMO (Kim et al 2013 Fusion Eng. Des. 88 123) to demonstrate that it is also possible to attain the XD configuration in advanced tokamak reactors with all PF coils outside the TF coils. The results given here for the XD are far more encouraging than recent calculations by Lackner and Zohm (2012 Fusion Sci. Technol. 63 43) for the Snowflake (Ryutov 2007 Phys. Plasmas 14 064502, Ryutov et al 2008 Phys. Plasmas 15 092501), where the required high PF currents represent a major technological challenge. The magnetic field structure in the outboard divertor SOL (Kotschenreuther 2013 Phys. Plasmas 20 102507) in the recently created

  1. 'Frozen finger' in anal fissures.

    PubMed

    Chintamani; Tandon, Megha; Khandelwal, Rohan

    2009-10-01

    Acute anal fissures are usually managed by various invasive and non-invasive modalities ranging from simple lifestyle changes to chemical and surgical sphincterotomies. Frozen finger, prepared using a water-filled ordinary rubber glove, was successfully used in one hundred patients, thus providing a cost-effective and simple solution to the problem.

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

    NASA Astrophysics Data System (ADS)

    Chen, L.; Xu, G. S.; Gao, W.; Zhang, L.; Nielsen, A. H.; Luo, Z. P.; Si, H.; Wang, Y. M.; Qu, H.; Sun, Z.; Duan, Y. M.; Liu, H. Q.; Wang, S. X.; Li, M. H.; Zhang, X. J.; Wu, B.; Chen, R.; Wang, L.; Wang, H. Q.; Ding, S. Y.; Yan, N.; Liu, S. C.; Shao, L. M.; Zhang, W.; Hu, G. H.; Li, J.; Li, Y. L.; Wu, X. Q.; Zhao, N.; Jia, M. N.

    2016-05-01

    The power threshold for low (L) to high (H) confinement mode transition achieved by radio-frequency heating and molybdenum first wall with lithium coating has been experimentally investigated on the EAST tokamak for two sets of divertor geometries and materials: tungsten/carbon divertor and full 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 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 plays an important part in the observed lower threshold. In addition, the H-mode in the Quasi-Snowflake divertor configuration has been obtained on EAST, exhibiting higher L-H power threshold compared to the lower single null configuration with similar IP/BT pairs.

  3. Motivation and goals of the new heated outer divertor for Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Lipschultz, B.; Doody, J.; Ellis, R.; Granetz, R.; Harrison, S.; Labombard, B.; Vieira, R.; Zhang, H.; Zhou, L.

    2012-10-01

    A precision-aligned, high-temperature outer divertor is being developed for Alcator C-Mod to enhance heatflux handling and to advance our knowledge and experience with high-Z Plasma Facing Components (PFCs) in a reactor-level power density environment. Several departures from the design of the current divertor will be implemented: Instead of 10 toroidal divertor segments that expand toroidally as they heat up, the divertor plate will be toroidally continuous, with no openings or leading edges in the high-heat flux region. It will expand in the radial direction when heated while maintaining good alignment with shallow field line angles (˜ 2 degrees), a requirement for future divertors. Those characteristics will reduce both impurity sources and disruption forces. A second design goal is to be able to control the divertor temperature up to 600^oC by installing heaters in the structure. Given the Arrhenius relation between hydrogen diffusivity and temperature in tungsten (and molybdenum) this will open up a new area of study for tokamaks - exploration of the effect of PFC temperature on fuel retention. Temperature control may also open up a new area of study into the effect of changes in divertor recycling on fueling and core confinement.

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

    SciTech Connect

    Sizyuk, V. Hassanein, A.

    2015-01-15

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

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

    SciTech Connect

    Fenstermacher, M.E.; Allen, S.L.; Hill, D.N.

    1996-02-01

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

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

    SciTech Connect

    Schaffer, M.J.; Wade, M.R.; Maingi, R.; Monier-Garbet, P.; West, W.P.; Whyte, D.G.; Wood, R.D.; Mahdavi, M.A.

    1996-06-01

    We report first direct measurements of divertor exhaust gas impurity enrichment, {eta}{sub exh}=(exhaust impurity concentration){divided_by}(core impurity concentration), for both unpumped and D{sub 2} puff-with-divertor-pump conditions. The experiment was performed with neutral beam heated, ELMing H-mode, single-null diverted deuterium plasmas with matched core and exhaust parameters in the DIII-D tokamak. Neon gas impurity was puffed into the divertor. Neon density was measured in the exhaust by a specially modified Penning gauge and in the core by absolute charge exchange recombination spectroscopy. Neon particle accounting indicates that much of the puffed neon entered a temporary unmeasured reservoir, inferred to be the graphite divertor target, which makes direct measurements necessary to calculate divertor enrichments. D{sub 2} puff into the SOL (scrape-off layer) with pumping increased {eta}{sub exh} threefold over either unpumped conditions or D{sub 2} puff directly into the divertor with pumping. These results show that SOL flow plays an important role in divertor exhaust impurity enrichment.

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

    NASA Astrophysics Data System (ADS)

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

    2012-08-01

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

  8. Role of cross-field drifts in the onset of divertor detachment

    NASA Astrophysics Data System (ADS)

    Groth, Mathias; Allen, S. L.; Fenstermacher, M. E.; Hill, D. H.; Makowski, M. A.; McLean, A. G.; Lasnier, C. J.; Porter, G. D.; Rognlien, T. D.; Briesemeister, A. R.; Unterberg, E. A.; Leonard, A. W.; Watkins, J. G.

    2015-11-01

    The impact of cross-field drifts in divertor configurations was investigated in DIII-D L and H-mode discharges. The studies show that the electron temperature at the outer divertor plate is reduced to below 2 eV at about 20 % lower pedestal density in configurations with the ion Bx ∇B direction toward the divertor X-point. When attached, these plasmas have significantly lower electron temperatures and and higher densities in the inner than in the outer divertor as directly measured with divertor Thomson scattering and inferred from line emission imaging using tangentially viewing cameras. Upon reversal of the toroidal field direction, the divertor conditions were observed in-out symmetric. Simulations with the edge fluid code UEDGE show that poloidal flows due to the radial electric field in the private flux region dominate the divertor asymmetries. Work supported by US DOE under DE-AC52-07NA27344, DE-FC02-04ER54698, DE-AC05-00OR22725, and DE-AC04-94AL85000.

  9. Impact of real-time magnetic axis sweeping on steady state divertor operation in LHD

    NASA Astrophysics Data System (ADS)

    Nakamura, Y.; Masuzaki, S.; Morisaki, T.; Ogawa, H.; Watanabe, T.; Kubota, Y.; Sakamoto, R.; Ashikawa, N.; Sato, K.; Chikaraishi, H.; Saito, K.; Seki, T.; Kumazawa, R.; Mutoh, T.; Kubo, S.; Takeiri, Y.; Peterson, B. J.; Komori, A.; Motojima, O.; LHD experimental Group

    2006-07-01

    Steady state divertor operation with high performance plasmas (ne ~ 0.7 × 1019 cm-3, Ti ~ 2 keV) was demonstrated for half an hour in the Large Helical Device (LHD), the superconducting helical device (R = 3.6-3.9 m, a = 0.6 m, B = 3 T, l/m = 2/10). The high performance plasmas have been sustained with an averaged heating power of 680 kW and achieved an injected energy of 1.3 GJ. This required both advanced technological integration of heating systems and divertor heat flux control. In particular, optimization of divertor heat flux distribution along the divertor leg trace on divertor plates and real-time magnetic axis sweeping (R = 3.67-3.7 m) have allowed LHD to access a steady state regime with a margin of safety for the actively cooled divertor plates. The distribution of divertor heat load along the traces was investigated with calorimetric measurements and it was found that there was a localized heat load connected with the loss of high-energy ions produced by ion cyclotron radio frequency near-fields. Orbit analysis shows that the behaviour of high-energy ions is qualitatively in good agreement with the experimental result. Long-pulse discharges were terminated by radiation collapse due to penetration of metallic flakes into the plasma.

  10. Attainment of a stable, fully detached plasma state in innovative divertor configurations

    NASA Astrophysics Data System (ADS)

    Umansky, Maxim

    2016-10-01

    The heat load on plasma facing components is a critical engineering constraint for future tokamaks, which has stimulated the community to consider innovative magnetic divertor geometries for future high power devices. Present-day advanced divertor scenarios generally rely on partially detached regimes, also planned for ITER; a fully detached state would usually lead to MARFE and degradation of core confinement. Modeling reveals that novel magnetic geometries can have a major impact on plasma detachment and power handling. Using the UEDGE tokamak edge transport model for configurations with tightly baffled long divertor legs, extended radially, or vertically, we find stable, fully detached divertor operation. Including a secondary X-point in the outer leg volume extends the attainment of a stable detached state to the highest power. As the input power is reduced to a threshold value, the outer leg transitions to a fully detached state with the detachment front localized at the secondary X-point or in the leg volume; reducing the power further results in the detachment front steady-state location shifting upstream. As the power is reduced, the detachment front eventually moves to the primary X-point, which sets the lower power limit for the range of stable operation. Still, for a long-legged divertor, a fully detached, stable divertor regime is maintained over an order-of-magnitude variation in exhaust power. In contrast, a standard divertor has a much smaller detachment operational window. These results suggest that stable fully detached divertor operation can be realized in tokamaks with extended divertor legs.

  11. The influence of Filaments in the Private Flux Region on Divertor Power and Particle Deposition

    NASA Astrophysics Data System (ADS)

    Harrison, James

    2014-10-01

    Recent advances in imaging of the MAST divertor have revealed, for the first time, evidence for filaments in the private flux region (PFR). Detailed analysis of the image data shows 3 distinct types of fluctuations occurring within the divertor volume: highly sheared filaments in the SOL originating from the outer midplane, high frequency (>50 kHz) filaments near the separatrix of the outer divertor leg and filaments in the private flux region originating from inner divertor leg. With the need to extrapolate divertor performance from existing machines to future devices, these observations can contribute to our quantitative understanding of transport in the PFR. In particular, they suggest that transport in the PFR is, at least in part, driven by turbulence, which may not be well captured by the Eich/Wagner description of the divertor footprint, expressed in terms of exponential decay in space above the X-point and Gaussian spreading below the X-point. The PFR filaments are observed to move largely parallel with the flux surfaces in a way equivalent to a toroidal angular velocity of order 2 ×104 rad/s in H-mode, and slower by a factor of order 2 in L-mode. During their transit parallel to the flux surfaces across the PFR, the filaments eject plasma in bursts, away from the separatrix, deeper into the private flux region. Correlation analysis suggests that they are generated by processes local to the inner divertor leg, as there is a weak correlation between fluctuations in the SOL and PFR above what is expected from line integration effects. Scaling of filament properties with machine operating parameters, such as plasma current, density and auxiliary heating power will be presented, together with a comparison with data from divertor Langmuir probes and IR thermography to estimate the role PFR filaments play in determining the width of the divertor footprint.

  12. Finger posture modulates structural body representations

    PubMed Central

    Tamè, Luigi; Dransfield, Elanah; Quettier, Thomas; Longo, Matthew R.

    2017-01-01

    Patients with lesions of the left posterior parietal cortex commonly fail in identifying their fingers, a condition known as finger agnosia, yet are relatively unimpaired in sensation and skilled action. Such dissociations have traditionally been interpreted as evidence that structural body representations (BSR), such as the body structural description, are distinct from sensorimotor representations, such as the body schema. We investigated whether performance on tasks commonly used to assess finger agnosia is modulated by changes in hand posture. We used the ‘in between’ test in which participants estimate the number of unstimulated fingers between two touched fingers or a localization task in which participants judge which two fingers were stimulated. Across blocks, the fingers were placed in three levels of splay. Judged finger numerosity was analysed, in Exp. 1 by direct report and in Exp. 2 as the actual number of fingers between the fingers named. In both experiments, judgments were greater when non-adjacent stimulated fingers were positioned far apart compared to when they were close together or touching, whereas judgements were unaltered when adjacent fingers were stimulated. This demonstrates that BSRs are not fixed, but are modulated by the real-time physical distances between body parts. PMID:28223685

  13. 27 CFR 9.34 - Finger Lakes.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 27 Alcohol, Tobacco Products and Firearms 1 2013-04-01 2013-04-01 false Finger Lakes. 9.34 Section... Lakes. (a) Name. The name of the viticultural area described in this section is “Finger Lakes.” (b) Approved maps. The appropriate maps for determining the boundaries of the Finger Lakes viticultural area...

  14. Compatibility of the Radiating Divertor with High Performance Plasmas in DIII-D

    SciTech Connect

    Petrie, T; Wade, M; Allen, S; Brooks, N; Fenstermacher, M; Ferron, J; Greenfield, C; Groth, M; Hyatt, A; Lasnier, C; Leonard, A; Luce, T; Mahdavi, M; Schaffer, M; Watkins, J; West, W

    2005-06-24

    Excessive thermal power loading on the divertor structures presents a design difficulty for future-generation, high powered tokamaks. This difficulty may be mitigated by ''seeding'' the divertor with impurities which radiate a significant fraction of the power upstream of the divertor targets. For this ''radiating divertor'' concept to be practical, however, the confinement and stability of the plasma cannot be compromised by excessive leakage of the seeded impurities into the core plasma. One proposed way of reducing impurity influx is to enhance the directed scrape-off layer (SOL) flow of deuterium ions toward the divertor [1-5]. We report here on the successful application of the radiating divertor scenario to high performance plasma operation in a DIII-D ''hybrid'' H-mode regime. The ''hybrid'' regime [6,7] has many features in common with conventional ELMing H-mode regimes, such as high confinement, e.g., H{sub ITER89P} > 2, where H{sub ITER89P} is the energy confinement normalized to the 1989 ITER L-mode scaling [8]. The main difference is the absence of sawtooth activity in the hybrid. Argon was selected as the seeded impurity for this experiment because argon radiates effectively at both the divertor and pedestal temperatures found in DIII-D hybrid H-mode operation and has a relatively short ionization mean free path. Carbon is also present as the dominant intrinsic impurity in DIII-D discharges. The geometry of this experiment is shown in Fig. 1. A double-null cross-sectional shape was biased upward (dRsep = +1.0 cm). To increase the deuterium ion flow toward the divertor at the top of the vessel, deuterium gas was introduced near the bottom. Argon was injected directly into the private flux region (PFR) of the upper divertor. In-vessel pumping of deuterium and argon was done by cryopumps located in the two upper divertor plenums, shown in cross-hatching [9]. The upper divertor, which we hereafter will simply refer to as the ''divertor'', is the region

  15. Enhanced visible and near-infrared capabilities of the JET mirror-linked divertor spectroscopy system

    SciTech Connect

    Lomanowski, B. A. Sharples, R. M.; Meigs, A. G.; Conway, N. J.; Zastrow, K.-D.; Heesterman, P.; Kinna, D. [EURATOM Collaboration: JET-EFDA Team

    2014-11-15

    The mirror-linked divertor spectroscopy diagnostic on JET has been upgraded with a new visible and near-infrared grating and filtered spectroscopy system. New capabilities include extended near-infrared coverage up to 1875 nm, capturing the hydrogen Paschen series, as well as a 2 kHz frame rate filtered imaging camera system for fast measurements of impurity (Be II) and deuterium Dα, Dβ, Dγ line emission in the outer divertor. The expanded system provides unique capabilities for studying spatially resolved divertor plasma dynamics at near-ELM resolved timescales as well as a test bed for feasibility assessment of near-infrared spectroscopy.

  16. Diagnostic tools for studying divertor detachment: bolometry, spectroscopy, and thermography for surface heat-flux

    NASA Astrophysics Data System (ADS)

    Terry, J. L.; Reinke, M. L.

    2017-04-01

    Some of the key aspects of divertor detachment that are addressed by bolometry, impurity spectroscopy, hydrogen spectroscopy, and measurements of divertor target heat-flux are reviewed. Measurement requirements for these diagnostic areas are defined, and brief descriptions of the techniques used for these diagnostics are given. Examples from the literature of measurements using these tools applied to detachment are presented. Feedback control of detachment using some of these diagnostics as the ‘sensors’ is reviewed. Challenges and some future directions for these diagnostics in the context of studying divertor detachment are described.

  17. Enhanced visible and near-infrared capabilities of the JET mirror-linked divertor spectroscopy systema)

    NASA Astrophysics Data System (ADS)

    Lomanowski, B. A.; Meigs, A. G.; Conway, N. J.; Zastrow, K.-D.; Sharples, R. M.; Heesterman, P.; Kinna, D.

    2014-11-01

    The mirror-linked divertor spectroscopy diagnostic on JET has been upgraded with a new visible and near-infrared grating and filtered spectroscopy system. New capabilities include extended near-infrared coverage up to 1875 nm, capturing the hydrogen Paschen series, as well as a 2 kHz frame rate filtered imaging camera system for fast measurements of impurity (Be II) and deuterium Dα, Dβ, Dγ line emission in the outer divertor. The expanded system provides unique capabilities for studying spatially resolved divertor plasma dynamics at near-ELM resolved timescales as well as a test bed for feasibility assessment of near-infrared spectroscopy.

  18. Development of a liquid-metal fusion reactor divertor with a capillary-pore system

    NASA Astrophysics Data System (ADS)

    Golubchikov, L. G.; Evtikhin, V. A.; Lyublinski, I. E.; Pistunovich, V. I.; Potapov, I. N.; Chumanov, A. N.

    1996-10-01

    The absence of a satisfactorily developed fusion reactor (FR) divertor approach (having no lost layers of sputtered plate materials and/or replaceable blocks) has become the reason for the development of the new concept of liquid-metal divertor (LMD) with a capillary-pore (CP) lithium protection system. Creative and novel design and material solutions, combined with unique natural thermophysical properties of Li working in a gas target evaporation—radiation mode, ensures the prolonged and steady performance of a FR divertor (D).

  19. Magnetic Field Structure near the Plasma Boundary in Helical Systems and Divertor Tokamaks

    NASA Astrophysics Data System (ADS)

    Nagasaki, Kazunobu; Itoh, Kimitaka

    1990-07-01

    The magnetic field structure of the scrape-off layer (SOL) region in both helical systems and divertor tokamaks is studied numerically by using model fields. The connection length of the field line to the wall, L, is calculated. In helical systems, L has logarithmic properties in the SOL region. The effect of axisymmetric fields on the field structure is determined. In divertor tokamaks, L also has logarithmic properties near the separatrix. Even when the perturbations which resonate to rational surfaces near the plasma boundary are added, the logarithmic properties are not changed. The connection length of torsatron/helical-heliotron systems is compared with that of divertor tokamaks.

  20. Viscous fingering of a draining suspension

    NASA Astrophysics Data System (ADS)

    Chen, Yun; Malambri, Frank; Lee, Sungyon

    2016-11-01

    The Saffman-Taylor viscous fingering arises when a viscous oil is withdrawn from a Hele-Shaw cell that is filled with a less viscous fluid. When particles are introduced into the draining fluid, new behaviors emerge, which are unobserved in the well-established pure oil case. We experimentally investigate the particle-modified inward fingering for varying particle concentrations. In particular, the fingering growth rate and number of fingers are experimentally quantified and are shown to be directly affected by the presence of particles. The physical mechanism of the particle-modified fingering is also discussed.

  1. Mechanical model of a single tendon finger

    NASA Astrophysics Data System (ADS)

    Rossi, Cesare; Savino, Sergio

    2013-10-01

    The mechanical model of a single tendon three phalanxes finger is presented. By means of the model both kinematic and dynamical behavior of the finger itself can be studied. This finger is a part of a more complex mechanical system that consists in a four finger grasping device for robots or in a five finger human hand prosthesis. A first prototype has been realized in our department in order to verify the real behavior of the model. Some results of both kinematic and dynamical behavior are presented.

  2. Acrylic Finger Prosthesis: A Case Report

    PubMed Central

    Bandela, Vinod; M, Bharathi; S V, Giridhar Reddy

    2014-01-01

    Hands basic function is to grasp, hold and manipulate items. Hand gesture is perhaps the most blatant example of non-verbal communication. Finger and partial finger amputations are most frequently encountered forms of partial hand loss. Common causes are traumatic injuries, congenital absence or malformations present great clinical challenges. In addition to immediate loss of grasp strength, finger absence may cause marked psychological trauma. Individuals who desire finger replacement usually have high expectation for the appearance of prosthesis. This clinical report portrays simple method to retain acrylic finger prosthesis. PMID:25302271

  3. Impact of Finger Type in Fingerprint Authentication

    NASA Astrophysics Data System (ADS)

    Gafurov, Davrondzhon; Bours, Patrick; Yang, Bian; Busch, Christoph

    Nowadays fingerprint verification system is the most widespread and accepted biometric technology that explores various features of the human fingers for this purpose. In general, every normal person has 10 fingers with different size. Although it is claimed that recognition performance with little fingers can be less accurate compared to other finger types, to our best knowledge, this has not been investigated yet. This paper presents our study on the topic of influence of the finger type into fingerprint recognition performance. For analysis we employ two fingerprint verification software packages (one public and one commercial). We conduct test on GUC100 multi sensor fingerprint database which contains fingerprint images of all 10 fingers from 100 subjects. Our analysis indeed confirms that performance with small fingers is less accurate than performance with the others fingers of the hand. It also appears that best performance is being obtained with thumb or index fingers. For example, performance deterioration from the best finger (i.e. index or thumb) to the worst fingers (i.e. small ones) can be in the range of 184%-1352%.

  4. Materials selection for the US INTOR divertor collector plate

    SciTech Connect

    Mattas, R.F.; Misra, B.; Smith, D.L.; Morgan, G.D.; Delaney, M.; Gold, R.

    1981-01-01

    The divertor collector plate in the INTOR reactor will be subjected to high heat, particle, and neutron fluxes, making it the most severely damaged torus component. The collector plate is composed of a protection plate, which is directly exposed to the particle flux, and a heat sink which provides support for the protection plate and carries the water coolant. The high-Z refractory metals have been considered for use as the protection plate material, and austenitic stainless steels and copper alloys have been considered as the heat sink material. Tungsten and Type 316 stainless steels have been selected for the protection plate and heat sink, respectively. The protection plate has a sputtering lifetime of 1.75 y at a 50% duty factor, while the heat sink is expected to last the lifetime of the reactor.

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

    NASA Astrophysics Data System (ADS)

    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

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

    SciTech Connect

    C.H. Skinner; G. Federici

    2005-05-13

    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.

  7. Prediction of Pressure Drop in the ITER Divertor Cooling Channels

    SciTech Connect

    Yin, S.T.; Chen, J.L.

    2005-04-15

    This study investigated the pressure drop in the divertor cooling channels of the International Thermonuclear Experimental Reactor (ITER). The water in the cooling channels will encounter the following flow and boiling regimes: 1) single-phase convection, 2) highly-subcooled boiling, 3) onset of nucleate boiling (ONB), and 4) fully-developed subcooled boiling. The upper operating boundary is limited by the departure from nucleate boiling (DNB) or burnout conditions. Twisted-tape insert will be used to enhance local heat transfer. Analytical models, validated with relevant databases, were proposed for the above-identified flow regimes. A user-friendly computer code was developed to calculate the overall pressure drop and the exit pressure of a specific local segment throughout the entire flow circuit. Although the operating parameters were based on the CDA phase input the results are found in general agreement when compared with the ITER EDA results.

  8. Structural evaluation of a DTHR bundle divertor particle collector

    SciTech Connect

    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.

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

    SciTech Connect

    Hua, T.Q.; Brooks, J.N.

    1994-05-01

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

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

    SciTech Connect

    Kajita, Shin; Hatae, Takaki; Tojo, Hiroshi; Hamano, Takashi; Shimizu, Katsuhiro; Kawashima, Hisato; Enokuchi, Akito

    2015-08-15

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

  11. Integration of tactile input across fingers in a patient with finger agnosia.

    PubMed

    Anema, Helen A; Overvliet, Krista E; Smeets, Jeroen B J; Brenner, Eli; Dijkerman, H Chris

    2011-01-01

    Finger agnosia has been described as an inability to explicitly individuate between the fingers, which is possibly due to fused neural representations of these fingers. Hence, are patients with finger agnosia unable to keep tactile information perceived over several fingers separate? Here, we tested a finger agnosic patient (GO) on two tasks that measured the ability to keep tactile information simultaneously perceived by individual fingers separate. In experiment 1 GO performed a haptic search task, in which a target (the absence of a protruded line) needed to be identified among distracters (protruded lines). The lines were presented simultaneously to the fingertips of both hands. Similarly to the controls, her reaction time decreased when her fingers were aligned as compared to when her fingers were stretched and in an unaligned position. This suggests that she can keep tactile input from different fingers separate. In experiment two, GO was required to judge the position of a target tactile stimulus to the index finger, relatively to a reference tactile stimulus to the middle finger, both in fingers uncrossed and crossed position. GO was able to indicate the relative position of the target stimulus as well as healthy controls, which indicates that she was able to keep tactile information perceived by two neighbouring fingers separate. Interestingly, GO performed better as compared to the healthy controls in the finger crossed condition. Together, these results suggest the GO is able to implicitly distinguish between tactile information perceived by multiple fingers. We therefore conclude that finger agnosia is not caused by minor disruptions of low-level somatosensory processing. These findings further underpin the idea of a selective impaired higher order body representation restricted to the fingers as underlying cause of finger agnosia.

  12. Erosion and deposition on JET divertor and limiter tiles during the experimental campaigns 2005-2009

    NASA Astrophysics Data System (ADS)

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

    2013-07-01

    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/cm3.

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

    NASA Astrophysics Data System (ADS)

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

    2013-07-01

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

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

    NASA Astrophysics Data System (ADS)

    Rimza, Sandeep; Khirwadkar, Samir; Velusamy, Karupanna

    2017-04-01

    Development of an efficient divertor concept is an important task to meet in the scenario of the future fusion power plant. The divertor, which is a vital part of the reactor has to discharge the considerable fraction of the total fusion thermal power (∼15%). Therefore, it has to survive very high thermal fluxes (∼10 MW/m2). In the present paper, an efficient divertor heat exchanger cooled by helium is proposed for the fusion tokamak. The Plasma facing surface of divertor made-up of several modules to overcome the stresses caused by high heat flux. The thermal hydraulic performance of one such module is numerically investigated in the present work. The result shows that the proposed design is capable of handling target heat flux values of 10 MW/m2. The computational model has been validated against high-heat flux experiments and a satisfactory agreement is noticed between the present simulation and the reported results.

  15. Divertor particle and power deposition profiles in JET ELMy H-mode discharges

    NASA Astrophysics Data System (ADS)

    Jet-Efda Contributors; Jachmich, S.; Eich, T.; Fundamenski, W.; Kallenbach, A.; Pitts, R. A.; JET-EFDA Contributors1

    2007-06-01

    The transient pulses of heat and particles arriving at the divertor target plates as a consequence of upstream ELM activity have been characterised at JET using an array of target embedded Langmuir probes in the MkIIGB-SRP divertor. High temporal and spatial resolution of the ELM time behaviour has been achieved by slow divertor strike point sweeps during ELMing H-mode discharges and subsequent coherent averaging of the data. One key result is the observation of target particle flux profile broadening with an e-folding length twice the inter-ELM during Type-I ELMs, presumably as a consequence of the enhanced radial transport. During the ELMs large divertor target currents have been observed, which change sign when the direction of the ion B × ∇B drift is reversed. First comparisons of IR and Langmuir probe derived power deposition profiles have shown a clear increase in the total sheath heat transmission coefficient during the ELMs.

  16. Flute instability and the associated radial transport in the tandem mirror with a divertor mirror cell

    SciTech Connect

    Katanuma, I.; Yagi, K.; Haraguchi, Y.; Ichioka, N.; Masaki, S.; Ichimura, M.; Imai, T.

    2010-11-15

    The flute instability and the associated radial transport are investigated in the tandem mirror with a divertor mirror cell (the GAMMA10 A-divertor) with help of computer simulation, where GAMMA10 is introduced [Inutake et al., Phys. Rev. Lett. 55, 939 (1985)]. The basic equations used in the simulation were derived on the assumption of an axisymmetric magnetic field. So the high plasma pressure in a nonaxisymmetric minimum-B anchor mirror cell, which is important for the flute mode stability, is taken into account by redefining the specific volume of a magnetic field line. It is found that the flute modes are stabilized by the minimum-B magnetic field even with a divertor mirror although its stabilizing effects are weaker than that without the divertor mirror. The flute instability enhances the radial transport by intermittently repeating the growing up and down of the Fourier amplitude of the flute instability in time.

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

    SciTech Connect

    McLean, A. G. Soukhanovskii, V. A.; Allen, S. L.; Carlstrom, T. N.; LeBlanc, B. P.; Ono, M.; Stratton, B. C.

    2014-11-15

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

  18. Initial Results from the C-Mod Divertor Thomson Scattering System

    NASA Astrophysics Data System (ADS)

    Grek, B.; Johnson, D.; Paladino, R.; Bartolick, J.; Dimock, D.; Lowrance, J.; Lipshultz, B.; Labombard, B.

    1996-11-01

    Thomson scattering system has been installed recently to diagnose the x-point and divertor plasma regions with a resolution of 2-3 mm over a 12 cm field. The light scattered from a 30 HZ Nd:YAG laser is viewed from below through a slot in the outer divertor plate with a reentrant, high throughput collection system. A compact laser dump is located inside the inner divertor plate. Laser alignment is maintained under feedback control to track vessel motion. A filter polychromator spectrally resolves the scattered light from 25 spatial positions onto four 25 element avalanche photodiode arrays. System performance is described in terms of both calibration results and initial measurements of divertor plasma parameters. Supported by U.S. DOE Contract No. DE-AC02-78ET51013, DE-AC02-76-CHO-3073 and SBIR Grant No. 20431-92-II.

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

    NASA Astrophysics Data System (ADS)

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

    2012-12-01

    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.

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

    SciTech Connect

    Klepper, C Christopher; Biewer, T. M.; Graves, Van B; Andrew, P.; Marcus, Chris; Shimada, M.; Hughes, S.; Boussier, B.; Johnson, D. W.; Gardner, W. L.; Hillis, D. L.; Vayakis, G.; Vayakis, G.; Walsh, M.

    2015-01-01

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

  1. Thermal-hydraulic analysis of the HL-2M divertor using an homogeneous equilibrium model

    NASA Astrophysics Data System (ADS)

    Lu, Yong; Cai, Lijun; Liu, Yuxiang; Liu, Jian; Yuan, Yinglong; Zheng, Guoyao; Liu, Dequan

    2017-09-01

    The heat flux of the HL-2M divertor would reach 10 MW m-2 or more at the local area when the device operates at high parameters. Subcooled boiling could occur at high thermal load, which would be simulated based on the homogeneous equilibrium model. The results show that the current design of the HL-2M divertor could withstand the local heat flux 10 MW m-2 at a plasma pulse duration of 5 s, inlet coolant pressure of 1.5 MPa and flow velocity of 4 m s-1. The pulse duration that the HL-2M divertor could withstand is closely related to the coolant velocity. In addition, at the time of 2 min after plasma discharge, the flow velocity decreased from 4 m s-1 to 1 m s-1, and the divertor could also be cooled to the initial temperature before the next plasma discharge commences.

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

    NASA Astrophysics Data System (ADS)

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

    2015-12-01

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

  3. Design and testing of a superfluid liquid helium cooling loop

    SciTech Connect

    Gavin, L.M.; Green, M.A.; Levin, S.M.; Smoot, G.F.; Witebsky, C.

    1989-07-01

    This paper describes the design and preliminary testing of a cryogenic cooling loop that uses a thermomechanical pump to circulate superfluid liquid helium. The cooling loop test apparatus is designed to prove forced liquid helium flow concepts that will be used on the Astromag superconducting magnet facility. 3 refs., 2 figs.

  4. The cryogenic helium cooling system for the Tokamak physics experiment

    SciTech Connect

    Felker, B.; Slack, D.S.; Wendland, C.R.

    1995-09-29

    The Tokamak Physics Experiment (TPX) will use supercritical helium to cool all the magnets and supply helium to the Vacuum cryopumping subsystem. The heat loads will come from the standard steady state conduction and thermal radiation sources and from the pulsed loads of the nuclear and eddy currents caused by the Central Solenoid Coils and the plasma positioning coils. The operations of the TPX will begin with pulses of up to 1000 seconds in duration every 75 minutes. The helium system utilizes a pulse load leveling scheme to buffer out the effects of the pulse load and maintain a constant cryogenic plant operation. The pulse load leveling scheme utilizes the thermal mass of liquid and gaseous helium stored in a remote dewar to absorb the pulses of the tokamak loads. The mass of the stored helium will buffer out the temperature pulses allowing 5 K helium to be delivered to the magnets throughout the length of the pulse. The temperature of the dewar will remain below 5 K with all the energy of the pulse absorbed. This paper will present the details of the heat load sources, of the pulse load leveling scheme operations, a partial helium schematic, dewar temperature as a function of time, the heat load sources as a function of time and the helium temperature as a function of length along the various components that will be cooled.

  5. Sensitive method for characterizing liquid helium cooled preamplifier feedback resistors

    NASA Technical Reports Server (NTRS)

    Smeins, L. G.; Arentz, R. F.

    1983-01-01

    It is pointed out that the simple and traditional method of measuring resistance using an electrometer is ineffective since it is limited to a narrow and nonrepresentative range of terminal voltages. The present investigation is concerned with a resistor measurement technique which was developed to select and calibrate the Transimpedance Mode Amplifier (TIA) load resistors on the Infrared Astronomical Satellite (IRAS) for the wide variety of time and voltage varying signals which will be processed during the flight. The developed method has great versatility and power, and makes it possible to measure the varied and complex responses of nonideal feedback resistors to IR photo-detector currents. When employed with a stable input coupling capacitor, and a narrow band RMS voltmeter, the five input waveforms thouroughly test and calibrate all the features of interest in a load resistor and its associated TIA circuitry.

  6. Overview of Stellarator Divertor Studies: Final Report of LDRD Project 01-ERD-069

    SciTech Connect

    Fenstermacher, M E; Rognlien, T D; Koniges, A; Unmansky, M; Hill, D N

    2003-01-21

    A summary is given of the work carried out under the LDRD project 01-ERD-069 entitled Stellarator Divertor Studies. This project has contributed to the development of a three-dimensional edge-plasma modeling and divertor diagnostic design capabilities at LLNL. Results are demonstrated by sample calculations and diagnostic possibilities for the edge plasma of the proposed U.S. National Compact Stellarator Experiment device. Details of the work are contained in accompanying LLNL reports that have been accepted for publication.

  7. Divertor impurity sources; effects of hot surfaces and thin films on impurity production

    NASA Astrophysics Data System (ADS)

    Stamp, M. F.; Andrew, P.; Brezinsek, S.; Huber, A.; JET EFDA Contributors

    2005-03-01

    Strong continuum emission has been observed from divertor tiles at visible wavelengths and identified as Planck radiation from surfaces with temperatures of typically ˜ 2600 K. Such hot spots (which are not tile edges) can persist for several seconds and are more common at the inner divertor, than the outer. Surprisingly, these hot spots do not usually produce significant impurity fluxes. In contrast, ELMs may produce a significant enhancement of impurity fluxes, depending on strike point location and ELM size.

  8. Survivability of dust in tokamaks: Dust transport in the divertor sheath

    SciTech Connect

    Delzanno, Gian Luca; Tang, Xianzhu

    2014-02-15

    The survivability of dust being transported in the magnetized sheath near the divertor plate of a tokamak and its impact on the desired balance of erosion and redeposition for a steady-state reactor are investigated. Two different divertor scenarios are considered. The first is characterized by an energy flux perpendicular to the plate q{sub 0}≃1 MW/m{sup 2} typical of current short-pulse tokamaks. The second has q{sub 0}≃10 MW/m{sup 2} and is relevant to long-pulse machines like ITER or Demonstration Power Plant. It is shown that micrometer dust particles can survive rather easily near the plates of a divertor plasma with q{sub 0}≃1 MW/m{sup 2} because thermal radiation provides adequate cooling for the dust particle. On the other hand, the survivability of micrometer dust particles near the divertor plates is drastically reduced when q{sub 0}≃10 MW/m{sup 2}. Micrometer dust particles redeposit their material non-locally, leading to a net poloidal mass migration across the divertor. Smaller particles (with radius ∼0.1 μm) cannot survive near the divertor and redeposit their material locally. Bigger particle (with radius ∼10 μm) can instead survive partially and move outside the divertor strike points, thus causing a net loss of divertor material to dust accumulation inside the chamber and some non-local redeposition. The implications of these results for ITER are discussed.

  9. Effect of 3D magnetic perturbations on divertor conditions and detachment in tokamak and stellarator

    NASA Astrophysics Data System (ADS)

    Ahn, J.-W.; Briesemester, A. R.; Kobayashi, M.; Lore, J. D.; Schmitz, O.; Diallo, A.; Gray, T. K.; Lasnier, C. J.; LeBlanc, B. P.; Maingi, R.; McLean, A. G.; Sabbagh, S. A.; Soukhanovskii, V. A.

    2017-08-01

    Enhanced perpendicular heat and momentum transport induces parallel pressure loss leading to divertor detachment, which can be produced by the increase of density in 2D tokamaks. However, in the 3D configurations such as tokamaks with 3D fields and stellarators, the fraction of perpendicular transport can be higher even in a lower density regime, which could lead to the early transition to detachment without passing through the high-recycling regime. 3D fields applied to the limiter tokamak plasmas produce edge stochastic layers close to the last closed flux surface (LCFS), which can allow for enhanced perpendicular transport and indeed the absence of high recycling regime and early detachment have been observed in TEXTOR and Tore Supra. However, in the X-point divertor tokamaks with the applied 3D fields, the parallel transport is still dominant and the detachment facilitation has not been observed yet. Rather, 3D fields affected detachment adversely under certain conditions, either by preventing detachment onset as seen in DIII-D or by re-attaching the existing detached plasma as shown in NSTX. The possible way for strong 3D effects to induce access to the early detachment in divertor tokamaks appears to be via significant perpendicular loss of parallel momentum by frictional force for the counter-streaming flows between neighboring flow channels in the divertor. In principle, the adjacent lobes in the 3D divertor tokamak may generate the counter-streaming flow channels. However, an EMC3-EIRENE simulation for ITER H-mode plasmas demonstrated that screened RMP leads to significantly reduced counter-flows near the divertor target, therefore the momentum loss effect leading to detachment facilitation is expected to be small. This is consistent with the observation in LHD, which showed screening (amplification) of RMP fields in the attachment (stable detachment) case. Work for optimal parameter window for best divertor operation scenario is needed particularly for

  10. A Fast Visible Camera Divertor-Imaging Diagnostic on DIII-D

    SciTech Connect

    Roquemore, A; Maingi, R; Lasnier, C; Nishino, N; Evans, T; Fenstermacher, M; Nagy, A

    2007-06-19

    In recent campaigns, the Photron Ultima SE fast framing camera has proven to be a powerful diagnostic when applied to imaging divertor phenomena on the National Spherical Torus Experiment (NSTX). Active areas of NSTX divertor research addressed with the fast camera include identification of types of EDGE Localized Modes (ELMs)[1], dust migration, impurity behavior and a number of phenomena related to turbulence. To compare such edge and divertor phenomena in low and high aspect ratio plasmas, a multi-institutional collaboration was developed for fast visible imaging on NSTX and DIII-D. More specifically, the collaboration was proposed to compare the NSTX small type V ELM regime [2] and the residual ELMs observed during Type I ELM suppression with external magnetic perturbations on DIII-D[3]. As part of the collaboration effort, the Photron camera was installed recently on DIII-D with a tangential view similar to the view implemented on NSTX, enabling a direct comparison between the two machines. The rapid implementation was facilitated by utilization of the existing optics that coupled the visible spectral output from the divertor vacuum ultraviolet UVTV system, which has a view similar to the view developed for the divertor tangential TV camera [4]. A remote controlled filter wheel was implemented, as was the radiation shield required for the DIII-D installation. The installation and initial operation of the camera are described in this paper, and the first images from the DIII-D divertor are presented.

  11. Compatibility of the Radiating Divertor with High Performance Plasmas in DIII-D

    SciTech Connect

    Petrie, T W; Wade, M R; Brooks, N H; Fenstermacher, M E; Groth, M; Hyatt, A W; Isler, R C; Lasnier, C J; Leonard, A W; Mahdavi, M A; Porter, G D; Schaffer, M J; Watkins, J G; West, W P

    2006-05-18

    A radiating divertor approach was successfully applied to high performance 'hybrid' plasmas [M.R. Wade, et al., Proc. 20th IAEA Fusion Energy Conf., Vilamoura, (2004)]. Our technique included: (1) injecting argon near the outer divertor target, (2) enhancing the plasma flow into the inner and outer divertors by a combination of particle pumping and deuterium gas puffing upstream of the divertor targets, and (3) isolating the inner divertor from the outer by a structure in the private flux region. Good hybrid conditions were maintained, as the peak heat flux at the outer divertor target was reduced by a factor of 2.5; the peak heat flux at the inner target decreased by 20%. This difference was caused by a higher concentration of argon at the outer target than at the inner target. Argon accumulation in the main plasma was modest (n{sub AR}/n{sub e} {le}0.004 on axis), although the argon profile was more peaked than the electron profile.

  12. Divertor Target Heat Load Reduction by Electrical Biasing, and Application to COMPASS-D

    SciTech Connect

    Fielding, S J; Cohen, R H; Helander, P; Ryutov, D D

    2001-03-07

    A toroidally-asymmetric potential structure in the scrape-off layer (SOL) plasma may be formed by toroidally distributed electrical biasing of the divertor target tiles. The resulting ExB convective motions should increase the plasma radial transport in the SOL and thereby reduce the heat load at the divertor [1]. In this paper we develop theoretical modeling and describe the implementation of this concept to the COMPASS-D divertor. We show that strong magnetic shear near the X-point should cause significant squeezing of the convective cells preventing convection from penetrating above the X-point. This should result in reduced heat load at the divertor target without increasing the radial transport in the portion of the SOL in direct contact with the core plasma, potentially avoiding any confinement degradation. implementation of divertor biasing is in hand on COMPASS-D involving insulation of, and modifications to, the present divertor tiles. Calculations based on measured edge parameters suggest that modest currents {approx} 8 A/tile are required, at up to 150V, to drive the convection. A technical test is preceeding full bias experiments.

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

    NASA Astrophysics Data System (ADS)

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

    2013-10-01

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

  14. Divertor target heat load reduction by electrical biasing, and application to COMPASS-D

    NASA Astrophysics Data System (ADS)

    Fielding, S. J.; Cohen, R. H.; Helander, P.; Ryutov, D. D.

    2001-03-01

    A toroidally asymmetric potential structure in the scrape-off layer (SOL) plasma may be formed by toroidally distributed electrical biasing of the divertor target tiles. The resulting E× B convective motions should increase the plasma radial transport in the SOL and thereby reduce the heat load at the divertor [R.H. Cohen, D.D. Ryutov, Nucl. Fus. 37 (1997) 621]. In this paper, we develop theoretical modelling and describe the implementation of this concept to the COMPASS-D divertor. We show that a strong magnetic shear near the X-point should cause significant squeezing of the convective cells preventing convection from penetrating above the X-point. This should result in reduced heat load at the divertor target without increasing the radial transport in the portion of the SOL in direct contact with the core plasma, potentially avoiding any confinement degradation. Implementation of divertor biasing is in hand on COMPASS-D involving insulation of, and modifications to, the present divertor tiles. Calculations based on measured edge parameters suggest that modest currents ˜8 A/tile are required, at up to 150 V, to drive the convection. A technical test is preceding full bias experiments.

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

    DOE Data Explorer

    Frerichs, H. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Waters, I. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Schmitz, O. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Canal, G. P. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Evans, T. E. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Feng, Y. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Soukhanovskii, V. A. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)

    2016-06-01

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

  16. Field reversal effects on divertor plasmas under radiative and detached conditions in JT-60U

    NASA Astrophysics Data System (ADS)

    Asakura, N.; Hosogane, H.; Tsuji-Iio, S.; Itami, K.; Shimizu, K.; Shimada, M.

    1996-06-01

    Reversal effects of the toroidal field Bt on the principal divertor plasma parameters were investigated under radiative and detached divertor conditions in L mode discharges. The ion flux to the inboard separatrix strike point decreased before a MARFE occurred, irrespective of the ion Del B drift direction. The local electron temperature, Te, div, decreased to around 10 eV. The maximum fraction of power radiated in the divertor was comparable between the two directions of Bt. With the power flowing into the two divertor fans being slightly larger on the outboard than on the inboard, a nearly symmetric in-out heat load was observed for the ion Del B drift away from the target. This was due to the outboard enhanced asymmetries in the particle flux and radiation loss distributions. From the viewpoint of in-out symmetry in the target heat load and Te, div, operation with the ion Del B drift away from the target plate is desirable as long as the attached divertor condition is maintained. On the contrary, during the MARFE, although deterioration of the energy confinement as well as the increase in the fuelling efficiency were comparable, for the ion Del B drift towards the target the plasma did not detach completely, and the strong in-out asymmetry in the particle recycling was relaxed to a relatively symmetric distribution. From the viewpoint of particle exhaust to the divertor, operation with the ion Del B drift towards the target is favourable

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

    NASA Astrophysics Data System (ADS)

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

    2015-11-01

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

  18. Erosion and deposition in the JET divertor during the first ILW campaign

    NASA Astrophysics Data System (ADS)

    Mayer, M.; Krat, S.; Van Renterghem, W.; Baron-Wiechec, A.; Brezinsek, S.; Bykov, I.; Coad, P.; Gasparyan, Yu; Heinola, K.; Likonen, J.; Pisarev, A.; Ruset, C.; de Saint-Aubin, G.; Widdowson, A.; Contributors, JET

    2016-02-01

    Erosion and deposition were studied in the JET divertor during the first JET ITER-like wall campaign 2011 to 2012 using marker tiles. An almost complete poloidal section consisting of tiles 0, 1, 3, 4, 6, 7, 8 was studied. The data from divertor tile surfaces were completed by the analysis of samples from remote divertor areas and from the inner wall cladding. The total mass of material deposited in the divertor decreased by a factor of 4-9 compared to the deposition of carbon during all-carbon JET operation before 2010. Deposits in 2011 to 2012 consist mainly of beryllium with 5-20 at.% of carbon and oxygen, respectively, and small amounts of Ni, Cr, Fe and W. This decrease of material deposition in the divertor is accompanied by a decrease of total deuterium retention inside the JET vessel by a factor of 10 to 20. The detailed erosion/deposition pattern in the divertor with the ITER-like wall configuration shows rigorous changes compared to the pattern with the all-carbon JET configuration.

  19. Viscous fingering in an elastic channel

    NASA Astrophysics Data System (ADS)

    Hazel, Andrew L.; Ducloué, Lucie; Juel, Anne

    2016-11-01

    We investigate experimentally the fingering instability of a flat, steadily propagating interface in a Hele-Shaw channel, where the top boundary has been replaced by an elastic membrane. In order to create a steadily propagating flat front, we exploit the reopening modes of fluid-filled elasto-rigid channels. The collapsed upper boundary reopens through the steady propagation of a wide finger, when air is injected from one end at a constant flow rate. For high levels of collapse and high finger speed, the tip of the finger becomes flat, creating a leading edge normal to the direction of propagation, which in turn is subject to a smaller scale viscous fingering instability. By modifying the cross-sectional geometry of the channel, we can actuate the finger shape to observe a variety of small-scale fingering phenomena including growth in a direction normal to the propagation and dendrite formation. The instability of the flat front exhibits constant-length fingers, very similar to the stubby fingers observed in radial compliant Hele-Shaw cells, and reminiscent of the printer's instability travel with the front. We investigate the geometry of those fingers in terms of the speed of the front, and the geometry of the reopening region. The financial support of the Leverhulme Trust is gratefully acknowledged.

  20. Comparison of radiating divertor behaviour in single-null and double-null plasmas in DIII-D

    SciTech Connect

    Petrie, T W; Brooks, N. H.; Fenstermacher, M. E.; Groth, M.; Hyatt, A. W.; Isler, R.C.; Lasnier, C. J.; Leonard, A. W.; Porter, G. D.; Schaffer, M. J.; Watkins, J. G.; Wade, M R; West, W.P.

    2008-01-01

    Puff-and-pump' radiating divertor scenarios, applied to both upper single-null (SN) and double-null (DN) H-mode plasmas, result in a 30-60% increase in radiated power with little or no decrease in tau(E). Argon was injected into the private flux region of the upper divertor, and plasma flow into the upper divertor was enhanced by a combination of deuterium gas puffing upstream of the divertor targets and particle pumping at the targets. For the same constant deuterium injection rate, argon penetrated the main plasma of SNs more rapidly and reached a higher steady-state concentration when the B x del B-ion drift direction was towards the divertor (V-del B up arrow) rather than away from the divertor (V-del B down arrow). We also found that the initial rate at which argon accumulated inside DN plasmas was more than twice that of comparable SN plasmas having the same B x del B-ion drift direction. In DNs, the radiated power was not shared equally between divertors during argon injection. Only when the B x del B ion drift direction was away from the divertor were both significant increases in divertor radiated power and an accumulation of argon in the divertor observed, based on spectroscopic measurements of Ar II. Our data suggest that an unbalanced DN shape where the B x del B-ion drift is directed away from the dominant divertor may provide the best chance of successfully coupling a radiating divertor approach with a higher performance H-mode plasma.

  1. Current status of ultrasonography of the finger

    PubMed Central

    2016-01-01

    The recent development of advanced high-resolution transducers has enabled the fast, easy, and dynamic ultrasonographic evaluation of small, superficial structures such as the finger. In order to best exploit these advances, it is important to understand the normal anatomy and the basic pathologies of the finger, as exemplified by the following conditions involving the dorsal, volar, and lateral sections of the finger: sagittal band injuries, mallet finger, and Boutonnière deformity (dorsal aspect); flexor tendon tears, trigger finger, and volar plate injuries (volar aspect); gamekeeper’s thumb (Stener lesions) and other collateral ligament tears (lateral aspect); and other lesions. This review provides a basis for understanding the ultrasonography of the finger and will therefore be useful for radiologists. PMID:26753604

  2. Finger-Jointed Wood Products.

    DTIC Science & Technology

    1981-04-01

    long enough to be useful (14, 36, 38, 59, 124). Nonstructural finger joints are primarily found in molding stock, trim, siding, fascia boards, door...all beams but two in series 7 and 8 to the grain. The average modulus of be slightly higher than that for apparently was related to the joints, rupture ...inch, and a tip thickness of combinations laminated by the plant . (a)A bolt hole on tensile strength 0.031 inch. The other was classified Straight-bevel

  3. Prosthetic Hand With Two Gripping Fingers

    NASA Technical Reports Server (NTRS)

    Norton, William E.; Belcher, Jewell B.; Vest, Thomas W.; Carden, James R.

    1993-01-01

    Prosthetic hand developed for amputee who retains significant portion of forearm. Outer end of device is end effector including two fingers, one moved by rotating remaining part of forearm about its longitudinal axis. Main body of end effector is end member supporting fingers, roller bearing assembly, and rack-and-pinion mechanism. Advantage of rack-and-pinion mechanism enables user to open or close gap between fingers with precision and force.

  4. Prosthetic Hand With Two Gripping Fingers

    NASA Technical Reports Server (NTRS)

    Norton, William E.; Belcher, Jewell B.; Vest, Thomas W.; Carden, James R.

    1993-01-01

    Prosthetic hand developed for amputee who retains significant portion of forearm. Outer end of device is end effector including two fingers, one moved by rotating remaining part of forearm about its longitudinal axis. Main body of end effector is end member supporting fingers, roller bearing assembly, and rack-and-pinion mechanism. Advantage of rack-and-pinion mechanism enables user to open or close gap between fingers with precision and force.

  5. Pediatric finger fractures: which ones turn ugly?

    PubMed

    Cornwall, Roger

    2012-06-01

    The majority of pediatric finger fractures can be treated by closed means with expected excellent outcomes. However, a subset of fractures can turn "ugly," with complications such as growth arrest, malunion, and joint dysfunction if not recognized and treated appropriately. The present paper discusses several fractures in a child's fingers that can cause substantial problems if not recognized promptly, highlighting important themes in the evaluation and treatment of a child's injured finger.

  6. The impact of ELMs on the ITER divertor

    SciTech Connect

    Leonard, A.W.; Osborne, T.H.; Suttrop, W.; Hermann, A.; Itami, K.; Lingertat, J.; Loarte, A.

    1998-07-01

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

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

    NASA Astrophysics Data System (ADS)

    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.

  8. Comprehensive 2D measurements of radiative divertor plasmas in DIII-D

    SciTech Connect

    Fenstermacher, M.E.; Wood, R.D.; Allen, S.L.; Hill, D.N.

    1997-07-01

    This paper presents a comparison of the total radiated power profile and impurity line emission distributions in the SOL and divertor of DIII-D. This is done for ELMing H-mode plasmas with heavy deuterium injection (Partially Detached Divertor operation, PDD) and those without deuterium puffing. Results are described from a series of dedicated experiments performed on DIII-D to systematically measure the 2-D (R,Z) structure of the divertor plasma. The discharges were designed to optimize measurements with new divertor diagnostics including a divertor Thomson scattering system. Discharge sequences were designed to produce optimized data sets against which SOL and divertor theories and simulation codes could be benchmarked. During PDD operation the regions of significant radiated power shift from the inner divertor leg and SOL to the outer leg and X-point regions. D{alpha} emission shifts from the inner strikepoint to the outer strikepoint. Carbon emissions (visible CII and CIII) shift from the inner SOL near the X-point to a distributed region from the X-point to partially down the outer leg during moderate D2 puffing. In heavy puffing discharges the carbon emission coalesces on the outer separatrix near the X-point and for very heavy puffing it appears inside the last closed flux surface above the X-point. Calibrated spectroscopic measurements indicate that hydrogenic and carbon radiation can account for all of the radiated power. L{alpha} and CIV radiation are comparable and when combined account for as much as 90% of the total radiated power along chords viewing the significant radiating regions of the outer leg.

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

    SciTech Connect

    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.

    2016-09-14

    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). In conclusion, this paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.

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

    DOE PAGES

    Guo, H. Y.; Hill, D. N.; Leonard, A. W.; ...

    2016-09-14

    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, whichmore » 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). In conclusion, this paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.« less

  11. Characterization of intermittent divertor filaments in L-mode discharges in NSTX and NSTX-U

    NASA Astrophysics Data System (ADS)

    Scotti, F.; Maqueda, R. J.; Soukhanovskii, V. A.; Zweben, S.

    2016-10-01

    Divertor filaments due to intermittent fluctuations are studied in L-mode discharges in NSTX and NSTX-U to understand transport due to edge blobs and their role in the divertor particle fluxes. In diverted Ohmic L-mode NSTX discharges, intermittent filaments on the divertor target were imaged via neutral lithium emission with frame rates up to 200 kHz and <= 1 cm resolution. Broadband fluctuations up to 20-50% in RMS/mean are observed between ΨN 1.02 and 1.3. Spiral-shaped divertor correlation regions are observed up to ΨN 1.02 and extend for over a toroidal turn. The spiral motion of the filaments at the target is consistent with a radial and poloidal downward motion upstream as previously observed in NSTX H-mode discharges. Divertor filaments are correlated with midplane blobs measured by the gas puff imaging diagnostic. The cross-correlation with midplane blobs is observed to peak at zero delay at every radius, with values up to 0.8 in the far SOL and decreasing to 0.4 at ΨN 1.05. In NSTX-U, a more sensitive camera with optimized throughput allowed divertor turbulence imaging using C III emission at up to f = 100 kHz, enabling the study of filament dynamics along the inner and outer divertor legs in NBI-heated L-mode discharges. Work supported by the US Department of Energy under DE-AC52-07NA27344 and DE-AC02-09CH11466.

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

    SciTech Connect

    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.

    2016-09-14

    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). In conclusion, this paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.

  13. Deposition of 13C tracer in the JET MkII-HD divertor

    NASA Astrophysics Data System (ADS)

    Likonen, J.; Airila, M.; Alves, E.; Barradas, N.; Brezinsek, S.; Coad, J. P.; Devaux, S.; Groth, M.; Grünhagen, S.; Hakola, A.; Jachmich, S.; Koivuranta, S.; Makkonen, T.; Rubel, M.; Strachan, J.; Stamp, M.; Widdowson, A.; EFDA contributors, JET-

    2011-12-01

    Migration of 13C has been investigated at JET by injecting 13C-labelled methane at the outer divertor base at the end of the 2009 campaign. The 13C deposition profiles on carbon fibre composite divertor tiles were measured by secondary ion mass spectrometry and Rutherford backscattering techniques. 13C was mainly deposited near the puffing location on the outer divertor base tiles. High amounts of 13C were also found at the outer vertical target: at the bottom of the lower and at the top of the upper plates. Thirty-three percent of puffed 13CH4 was instantly pumped out by the divertor cryopump, which is close to the pump duct entrance. Global 13C transport in the torus was modelled by the EDGE2D/EIRENE and DIVIMP codes, and local 13C migration in the vicinity of the injection location by the ERO code. The DIVIMP and EDGE2D simulations show strong prompt deposition of 13C directly adjacent to the injection point as well as in the far scrape-off layer (SOL) along both the inner and outer divertor targets. In addition, the measured 13C deposition along the outer divertor wall tiles is qualitatively reproduced. However, EDGE2D and DIVIMP do not predict any deposition along the divertor surfaces facing the private plasma on the inner floor tile and inboard of the outer strike point on tile 5. The ERO calculations also indicate that most of the deposition occurs close to the injection location on the vertical face of the LBSRP tile and the horizontal part of tile 6.

  14. Fokker-Planck Modelling of PISCES Linear Divertor Simulator

    NASA Astrophysics Data System (ADS)

    Batishchev, O. V.; Krasheninnikov, S. I.; Schmitz, L.

    1996-11-01

    The gas target operating regime in the PISCES [1] linear divertor simulator is characterized by a relatively high plasma density, 2.5 × 10^19 m-3, and low temperature, 8 eV, in the middle section of an ≈ 1 m long plasma column. Near the target, the plasma temperature and density as measured by Langmuir probes drop to 2 eV and 3.5 × 10^18 m-3, respectively, as a result of electron energy loss due to dissociation, ionization, and radiation. Such a sharp gradient in the plasma parameters can enhance non-local effects. To study these, we performed kinetic simulations of the relaxation of the electron energy distribution function on the experimentally measured background plasma using the adaptive finite-volumes code ALLA [2]. We discuss the effects of the observed incompletely equilibrated electron distribution function on key plasma parameter measurements and plasma - neutral particle interactions. cm [1] L.Schmitz et al., Physics of Plasmas 2 (1995) 3081. cm [2] A.A.Batishcheva et al., Physics of Plasmas 3 (1996) 1634. cm ^*Under U.S. DoE Contracts No.DE-FG02-91-ER-54109 at MIT, DE-FG02-88-ER-53263 at Lodestar, and DE-FG03-95ER54301 at UCSD.

  15. Tokamak edge Er studies by turbulence and divertor simulations

    NASA Astrophysics Data System (ADS)

    Nishimura, Y.; Coster, D.; Scott, B.

    2002-11-01

    Numerical coupling of the divertor code B2(B. J. Braams, Next European Torus Technical Report 68 (1987).) and the turbulence code DALF(B. D. Scott, Phys. Fluids B 4), 2468 (1992). is pursued. Within this model, space and time dependent transport coefficients (D and i) respond to the dynamics of drift wave turbulence. The Braginskii transport model of the B2 code incorporates guiding-center plasma drifts self-consistently and generate Er shear in the presence of steep pressure gradients. This Braginskii type Er can enter the turbulence model as a background E × B shear flow which suppresses the radial flux together with Reynolds stress induced electric fields. As an example of L-H transition, influx at the core boundary is controlled to produce steepening of the edge gradients. ( Y.Hamada et al.), in Proceedings of the 17th IAEA Fusion Energy Conference (IAEA-F1-CN-69/PD, 1998) reveals heat pulse induced L-H transitions after sawtooth events.

  16. ITER divertor performance in the low-activation phase

    NASA Astrophysics Data System (ADS)

    Kukushkin, A. S.; Pacher, H. D.; Kotov, V.; Pacher, G. W.; Pitts, R. A.; Reiter, D.

    2013-12-01

    The paper presents results of SOLPS modelling of the edge plasma performance during the low-activation phase of ITER operation. The calculations show that the peak power loading of the divertor targets can reach the reactor-relevant level of 3 to 5 MW m-2, even without the fusion reactions, rendering commissioning of the high heat flux components possible in this phase. Parametrization of the output of the SOLPS runs for the predominantly helium plasma concerned by the studies reported here is performed, thus providing the boundary conditions for modelling of the core and allowing efficient integration of the core and edge models. This approach, using the ASTRA code for core simulations, is applied to the analysis of hydrogen accumulation in helium plasmas due to H pellet injection. The latter is the only available option for early testing of ELM pace-making as an ELM control tool assuming H-mode in hydrogen will not be possible. Critical dilution with H down to 70% He in the core plasma can be reached in only 0.5 to 1 s or even shorter, depending on the assumptions made.

  17. Plasma flow interaction with ITER divertor related surfaces

    NASA Astrophysics Data System (ADS)

    Dojčinović, Ivan P.

    2010-11-01

    It has been found that the plasma flow generated by quasistationary plasma accelerators can be used for simulation of high energy plasma interaction with different materials of interest for fusion experiments. It is especially important for the studies of the processes such as ELMs (edge localized modes), plasma disruptions and VDEs (vertical displacement events), during which a significant part of the confined hot plasma is lost from the core to the SOL (scrape off layer) enveloping the core region. Experiments using plasma guns have been used to assess erosion from disruptions and ELMs. Namely, in this experiment modification of different targets, like tungsten, molybdenum, CFC and silicon single crystal surface by the action of hydrogen and nitrogen quasistationary compression plasma flow (CPF) generated by magnetoplasma compressor (MPC) has been studied. MPC plasma flow with standard parameters (1 MJ/m2 in 0.1 ms) can be used for simulation of transient peak thermal loads during Type I ELMs and disruptions. Analysis of the targets erosion, brittle destruction, melting processes, and dust formation has been performed. These surface phenomena are results of specific conditions during CPF interaction with target surface. The investigations are related to the fundamental aspects of high energy plasma flow interaction with different material of interest for fusion. One of the purposes is a study of competition between melting and cleavage of treated solid surface. The other is investigation of plasma interaction with first wall and divertor component materials related to the ITER experiment.

  18. EUV Spectroscopy During the DIII-D Tungsten Divertor Campaign

    NASA Astrophysics Data System (ADS)

    Allen, S. L.; Victor, B. S.; Beiersdorfer, P.; Magee, E.; Soukhanovskii, V.; Weller, M.; Loch, S.; Thomas, D.

    2016-10-01

    Two toroidal rings of tungsten-coated tile inserts were installed in the DIII-D lower divertor and a range of L- and H-mode plasma discharges were compared during a dedicated two week run campaign. A high resolution (1340 spectral channels) variable-ruling grating spectrometer viewing the core of the plasma was used to study the spectral region 10-70 Å a second spectrometer viewing 20 - 150 Å was also used. At DIII-D core plasma temperatures 2-3 keV, several emission lines from W38+ through W43+ were identified, including a quasi-continuum feature of W near 50 Å whose structure depends on core Te. Molybdenum (TZM substrate) emissions between 20-30 Å and near 70 Å were also observed. ADAS calculations are used to guide the identification of W emission lines for the measured core plasma Te and ne profiles. The behavior of W emissions during both ``benign'', pellet injection, and impurity accumulation conditions will be presented. Supported by US DOE under DE-AC52-07NA27344, and DE-FC02-04ER54698.

  19. Effects of ELMs on ITER divertor armour materials

    NASA Astrophysics Data System (ADS)

    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.

  20. Imaging divertor strike point splitting in RMP ELM suppression experiments

    NASA Astrophysics Data System (ADS)

    Moyer, R. A.; Bykov, I.; Orlov, D. M.; Lee, J. S.; Evans, T. E.; Nazikian, R.; Makowski, M.; Lasnier, C. S.; Wang, H.; Abrams, T.; Watkins, J. G.

    2016-10-01

    Fast visible imaging of the lower divertor has been implemented at DIII-D to study the structure and dynamics of lobes induced by 3D fields in RMP ELM suppression experiments. The sharpest imaging was obtained with spatially localized molecular D2 emission indicative of the D flux to the surface. Multiple D2 emission peaks are readily resolved during RMPs, in contrast to the heat flux profile (from IR), which often shows little structure. The brightest D2 lobe is often farthest from the primary inner strike point (ISP). Mitigated ELMs perturb the position and intensity of the ISP lobes and spread the outer strike point emission into the far SOL, where it may be caused by ELM filament propagation. RMP current ramps affect the lobe locations and separations. Implications of the lobe dynamics for plasma response is being studied. Work supported by U.S. DOE under Grants DE-FG02-07ER54917 and DE-FG02-05ER54809, and Contracts DE-FC02-04ER54698, DE-AC52-07NA27344, DE-AC04-94AL85000, DE-AC05-06OR23100 and DE-AC02-09CH11466.

  1. Current understanding of divertor detachment: experiments and modelling

    SciTech Connect

    Wischmeier, W; Groth, M; Kallenbach, A; Chankin, A; Coster, D; Dux, R; Herrmann, A; Muller, H; Pugno, R; Reiter, D; Scarabosio, A; Watkins, J; Team, T D; Team, A U

    2008-05-23

    A qualitative as well as quantitative evaluation of experimentally observed plasma parameters in the detached regime proves to be difficult for several tokamaks. A series of ohmic discharges have been performed in ASDEX Upgrade and DIII-D at similar as possible plasma parameters and at different line averaged densities, {bar n}{sub e}. The experimental data represent a set of well diagnosed discharges against which numerical simulations are compared. For the numerical modeling the fluid-code B2.5 coupled to the Monte Carlo neutrals transport code EIRENE is used. Only the combined enhancement of effects, such as geometry, drift terms, neutral conductance, increased radial transport and divertor target composition, explains a significant fraction of the experimentally observed asymmetries of the ion fluxes as a function of {bar n}{sub e} to the inner and outer target plates in ASDEX Upgrade. The relative importance of the mechanisms leading to detachment are different in DIII-D and ASDEX Upgrade.

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

    SciTech Connect

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

    2012-07-18

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

  3. Geometrical Effects in Plasma Stability and Dynamics of Coherent Structures in the Divertor

    SciTech Connect

    Ryutov, D D; Cohen, R H

    2007-05-16

    Plasma dynamics in the divertor region is strongly affected by a variety of phenomena associated with the magnetic field geometry and the shape of the divertor plates. One of the most universal effects is the squeezing of a normal cross-section of a thin magnetic flux-tube on its way from the divertor plate to the main SOL. It leads to decoupling of the most unstable perturbations in the divertor legs from those in the main SOL. For perturbations on either side of the X-point, this effect can be cast as a boundary condition at some 'control surface' situated near the X-point. We discuss several boundary conditions proposed thus far and assess the influence of the magnetic field geometry on them. Another set of geometrical effects is related to the transformation of a flux-tube that occurs when it is displaced in such a way that its central magnetic field line coincides with some other field line, and the magnetic field is not perturbed. These flute-like displacements are of a particular interest for the low-beta edge plasmas. It turns out that this transformation may also lead to a considerable deformation of a flux-tube cross-section; in addition, the distance between plasma particles occupying the flux-tube may change significantly even if there is no parallel plasma motion. We present expressions describing aforementioned transformations for the general tokamak geometry and simplify them for the divertor region (using the proximity of the X-point). We also discuss the effects associated with the shape of the plasma-limiting surfaces, both those designed to intercept the plasma (like divertor plates and limiters) and those that can be hit in some 'abnormal' events, e.g., in the course of a radial motion of an isolated plasma filament. The orientation of the limiting surface with respect to the magnetic field affects the plasma dynamics via the sheath boundary conditions. One can enhance or suppress plasma instabilities in the divertor legs by tilting the divertor

  4. Error compensation during finger force production after oneand four-finger voluntarily fatiguing exercise

    PubMed Central

    Kruger, Eric S.; Hoopes, Josh A.; Cordial, Rory J.; Li, Sheng

    2010-01-01

    The effect of muscle fatigue on error compensation strategies during multi-finger ramp force production tasks was investigated. Thirteen young, healthy subjects were instructed to produce a total force with four fingers of the right hand to accurately match a visually displayed template. The template consisted of a 3-s waiting period, a 3-s ramp force production (from 0 to 30% maximal voluntary contraction, MVC), and a 3-s constant force production. A series of twelve ramp trials was performed before and after fatigue. Fatigue was induced by a 60-s maximal isometric force production with either the index finger only or with all four fingers during two separate testing sessions. The average percent of drop was 38.2% in the MVC of the index finger after index-finger fatiguing exercise and 38.3% in the MVC of all fingers after four-finger fatiguing exercise. The ability of individual fingers to compensate for each other's errors in order for the total force to match the preset template was quantified as the error compensation index (ECI), i.e. the ratio of the sum of variances of individual finger forces and the variance of the total force. By comparing pre- and post-fatigue performance during four-finger ramp force production, we observed that the variance of the total force was not significantly changed after one- or four-finger fatiguing exercise. The ECI significantly decreased after four-finger fatiguing exercise, especially during the last second of the ramp; while the ECI remained unchanged after index finger single-finger fatiguing exercise. These results suggest that the central nervous system is able to utilize the abundant degrees of freedom to compensate for partial impairment of the motor apparatus induced by muscle fatigue to maintain the desired performance. However, this ability is significantly decreased when all elements of the motor apparatus are impaired. PMID:17443316

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

    NASA Astrophysics Data System (ADS)

    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.

  6. Engineering design of cryocondensation pumps for the DIII-D Radiative Divertor Program

    SciTech Connect

    Bozek, A.S.; Baxi, C.B.; Del Bene, J.V.; Laughon, G.J.; Reis, E.E.; Shatoff, H.D.; Smith, J.P.

    1995-10-01

    A new double-null, slotted divertor configuration will be installed for the DIII-D Radiative Divertor Program at General Atomics in late 1996. Four cryocondensation pumps, three new and one existing, will be part of this new divertor. The purpose of the pumps is to provide plasma density control and to limit the impurities entering the plasma core by providing pumping at each divertor strike point. The three new pumps are based on the design of the existing pump, installed in 1992 as part of the Advanced Divertor Program. The new pumps require geometry modifications to the original design. Therefore, extensive modal and dynamic analyses were performed to determine the behavior of these pumps and their helium and nitrogen feed lines during disruption events. Thermal and fluid analyses were also performed to characterize the helium two-phase flow regime in the pumps and their feedlines. A flow testing program was completed to test the change in geometry of the pump feed lines with respect to helium flow stability. The results were compared to the helium thermal and fluid analyses to verify predicted flow regimes and flow stability.

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

    NASA Astrophysics Data System (ADS)

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

    2015-11-01

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

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

    NASA Astrophysics Data System (ADS)

    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.

  9. Divertor ExB and Parallel Flows on the DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

    Boedo, J.; Rudakov, D.

    2016-10-01

    E ×B convection is an important particle transport mechanism responsible for up to 50 % of the total particle flux into the divertor, changing direction with B, and playing a role in divertor asymmetries. The gradient of the plasma potential, Vp =Vf + 2.5Te , reaches 5 kV/m across the SOL-private boundary, causing a poloidal particle flux, calculated as, Γθ = 2 πRne (Vp 1 -Vp 2) /BT , (along flux surfaces) of about 1022 s-1 , comparable to the target flow of 2 ×1022 s-1 , and consistent with previous work. Floating potential Vf, temperature Te, density Ne, and D+ flow were measured in the DIII-D divertor. The data will be compared to simulations by SOLPS and UEDGE. The D+ parallel flow velocity, V ∥ , calculated by multiplying the Mach number by the local sound speed cs =(γ ZkTe /mi) 1 / 2 show increasing velocity towards the plate in attached conditions and bulk sonic flows over the whole detached region in detached conditions. We compare measurements in the divertor to similar measurements made at the midplane to show how divertor conditions reflect upstream. Supported under USDOE Grant DE-FC02-04ER54698.

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

    SciTech Connect

    R.J. Maqueda, D.P. Stotler and the NSTX Team.

    2010-05-19

    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 Dα; 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).

  11. Numerical analyses of JT-60SA tokamak with tungsten divertor by COREDIV code

    NASA Astrophysics Data System (ADS)

    Gałązka, K.; Ivanova-Stanik, I.; Stępniewski, W.; Zagórski, R.; Neu, R.; Romanelli, M.; Nakano, T.

    2017-04-01

    An analysis of radiative power exhaust for the JT-60SA tokamak with a tungsten divertor is performed with the help of the self-consistent, core-edge integrated COREDIV code. Two scenarios of operation (low and high density) were investigated in the scope of different parameters (electron density at the separatrix and the perpendicular transport in the scrape-off layer) with impurity seeding (Ne and Kr). The calculations show that in the case of the tungsten divertor the power load to the divertor plate is mitigated and the central plasma dilution is smaller compared to the carbon divertor. In the most cases the energy flux through the separatrix is above the L-H transition threshold. For the high density case with neon seeding operation in full detachment mode is observed. Changing the diffusion coefficient in the SOL has a strong influence on the result of the calculations as increased radial transport causes stronger screening effect. Also by changing the electron density on the separatrix the influx of heavy impurities (W, Kr) into the core region can be reduced. The results demonstrate that it is easier to achieve sustainable conditions in the divertor region for the high density scenario, whereas for the low density one reducing the auxiliary heating power seems unavoidable to prevent damaging of the target plate, even for strong seeding gas influx.

  12. Reduction in resonant magnetic field induced heat flux splitting caused by detachment of the divertor

    NASA Astrophysics Data System (ADS)

    Briesemeister, A. R.; Ahn, J.-W.; Hillis, D. L.; Lore, J. D.; Shafer, M. W.; Unterberg, E. A.; Wingen, A.; Schmitz, O.; Frerichs, H.; Makowski, M. A.; McLean, A. G.; Ferraro, N. M.

    2015-11-01

    Measurements in DIII-D show that in high-density detached divertor conditions, the inter-ELM non-axisymmetric heat flux striations generated by resonant magnetic perturbations (RMPs) are eliminated. Non-axisymmetric heat loads caused by the RMP fields used to mitigate ELMs could reduce the lifetime of divertor components in ITER and future devices. It is shown that for RMPs with an n=3 toroidal mode number low levels of gas puffing can cause an increase in the heat flux splitting, but at high densities where the divertor becomes detached this splitting is eliminated. VUV imaging and 2D divertor Thomson scattering are used to measure RMP induced perturbations to the plasma conditions above the target plates. Modeling performed with the 3D fluid transport code EMC3-EIRENE both with and without the plasma response calculated by M3D-C1 is compared to the measured divertor conditions. Work supported by the US DOE under DE-AC05-00OR22725, DE-FC02-04ER54698, DE-AC52-07NA27344 & DE-FG02-92ER54139.

  13. Influence of helium puff on divertor asymmetry in Experimental Advanced Superconducting Tokamak

    SciTech Connect

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

    2014-02-15

    Divertor asymmetries with helium puffing are investigated in various divertor configurations on Experimental Advanced Superconducting Tokamak (EAST). The outer divertor electron temperature decreases significantly during the gas injection at the outer midplane. As soon as the gas is injected into the edge plasma, the power deposition drops sharply at the lower outer target while increases gradually at the lower inner target in LSN configuration; the power deposition increases quickly at the upper outer target while remains unchanged at the upper inner target in upper single null configuration; the power deposition increases slightly at the outer targets while shows no obvious variation at the inner targets in double null configuration. The radiated power measured by the extreme ultraviolet arrays increases significantly due to helium gas injection, especially in the outer divertor. The edge parameters are measured by reciprocating probes at the outer midplane, showing that the electron temperature and density increase but the parallel Mach number decreases significantly due to the gas injection. Effects of poloidal E × B drifts and parallel SOL flows on the divertor asymmetry observed in EAST are also discussed.

  14. Influence of helium puff on divertor asymmetry in Experimental Advanced Superconducting Tokamak

    NASA Astrophysics Data System (ADS)

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

    2014-02-01

    Divertor asymmetries with helium puffing are investigated in various divertor configurations on Experimental Advanced Superconducting Tokamak (EAST). The outer divertor electron temperature decreases significantly during the gas injection at the outer midplane. As soon as the gas is injected into the edge plasma, the power deposition drops sharply at the lower outer target while increases gradually at the lower inner target in LSN configuration; the power deposition increases quickly at the upper outer target while remains unchanged at the upper inner target in upper single null configuration; the power deposition increases slightly at the outer targets while shows no obvious variation at the inner targets in double null configuration. The radiated power measured by the extreme ultraviolet arrays increases significantly due to helium gas injection, especially in the outer divertor. The edge parameters are measured by reciprocating probes at the outer midplane, showing that the electron temperature and density increase but the parallel Mach number decreases significantly due to the gas injection. Effects of poloidal E × B drifts and parallel SOL flows on the divertor asymmetry observed in EAST are also discussed.

  15. Sputtering and Reflection Data for Mixed Tungsten/Beryllium Layers Under Typical FIRE Divertor Fluxes

    NASA Astrophysics Data System (ADS)

    Ruzic, D. N.; Nieto, M.; Alman, D. A.; Brooks, J. N.

    2001-10-01

    Computer modeling has been done as part of the Fusion Ignition Research Experiment (FIRE) design study. The current focus is on beryllium/tungsten mixed-material erosion. The FIRE design calls for a beryllium first wall and tungsten divertors. Beryllium can be sputtered from the first wall and transported to the divertor, forming a Be/W mixture on the divertor. The beryllium sputtering from the first wall is obtained from fluxes calculated by the DEGAS2 neutral transport code. Subsequent transport to the divertor is calculated by the REDEP code. VFTRIM-3D, a variant of the TRIM-SP binary-collision code, is used to investigate the sputtering properties of the Be/W divertor. Finally, WBC can compute beryllium and tungsten erosion and core plasma contamination using the sputtering and reflection coefficients obtained with VFTRIM-3D. In the present work, the VFTRIM-3D code was run on a W/Be surface with the Be content varied from 0 to 100 atomic percent. Deuterium and tritium (ions and neutrals), oxygen, beryllium from the first wall, and tungsten being redeposited are all incident on this mixed W/Be layer. Data on reflection and sputtering coefficients as a function of beryllium content in the bombarded surface will be presented.

  16. Achieving temporary divertor plasma detachment with MARFE events by pellet injection in the EAST superconducting tokamak

    NASA Astrophysics Data System (ADS)

    Guozhong, Deng; Liang, Wang; Xiaoju, Liu; Yanmin, Duan; Jiansheng, Hu; Changzheng, Li; Ling, Zhang; Shaocheng, Liu; Huiqian, Wang; Liang, Chen; Jichan, Xu; Wei, Feng; Jianbin, Liu; Huan, Liu; Guosheng, Xu; Houyang, Guo; Xiang, Gao; the EAST Team

    2017-01-01

    A new pellet injection system has been equipped on the experimental advanced superconducting tokamak (EAST) in the 2012 campaign, with a pellet size of ϕ 2 mm × 2 mm, a frequency of 1 Hz-10 Hz and velocity of 150 m s-1-300 m s-1. The deuterium pellet is well-known for plasma fuelling as well as for triggering the edge localized mode (ELM). In the 2012 campaign, pellet injection experiments were successfully carried out on EAST. Temporary plasma detachment achieved by deuterium pellets has been observed in a double null (DN) divertor configuration, with multi-pellet injections at a repetition frequency of 2 Hz. The partial detachment of the outer divertors and complete detachment of the inner divertors was achieved after 35 ms of each pellet injection, which have a duration of 30-60 ms with the maximum degree of detachment (DOD) reaching 3.5 and 37, respectively. Meanwhile, the multifaceted asymmetric radiation from the edge (MARFE) phenomena was also observed at the high field side (HFS) near both the lower and upper X-points with radiation loss suddenly increased to about 15%-70%, which may be the main cause of divertor plasma detachment. The temporary detachment induced by pellet injection may act as a new way to study divertor detachment behaviors.

  17. Deuterium and tritium fuelding in an ETF/INTOR plasma with divertor

    SciTech Connect

    Houlberg, W.A.; Howe, H.C.; Attenberger, S.E.

    1980-01-01

    Fueling by pellets and neutral gas in the presence of a divertor is examined with a one-dimensional multispecies transport code. Deuterium, tritium, and alpha particles are treated as independent thermal species. With an efficiently operating divertor, it becomes impossible to maintain high plasma density (anti n approx. 10/sup 14/ cm/sup -3/) with neutral gas fueling alone because of the high probability of the gas being ionized in the scrapeoff layer. Pellet fueling significantly reduces the particle load on the divertor and, with feedback control, can maintain the plasma density at the desired level. A low level of deuterium gas fueling can then be used to maintain the density of the scrapeoff layer and increase shielding against sputtered impurities. Even with an effective shielding divertor, the energy and particle fluxes to the first wall from both charged and neutral particles may be significant. The fluctuations at the plasma edge and in the scrapeoff layer introduced by the pellets cause the particle and energy fluxes to the divertor and first wall to fluctuate. This makes simulation more difficult but may be used to experimentally determine radial and parallel transport properties in the scrapeoff layer. Recommendations for further study are made.

  18. Initial Results of Local Island Divertor Experiments in the Large Helical Device

    SciTech Connect

    Komori, Akio; Morisaki, Tomohiro; Masuzaki, Suguru

    2004-07-15

    A local island divertor (LID) experiment has begun in the Large Helical Device (LHD) to demonstrate improved plasma confinement, and fundamental LID functions were demonstrated in the sixth experimental campaign in 2002-2003. It was clearly shown that when an m/n = 1/1 island is generated by adding a resonant perturbation field to the LHD magnetic configuration, the particle flow is guided along the island separatrix to the backside of the island, where carbon plates are located on a divertor head. The particles recycled there are pumped out efficiently so that the line-averaged core plasma density is reduced by a factor of {approx}2 at the same gas puff rate, compared with non-LID discharges. Obvious improvement of the global plasma confinement was, however, not observed yet, because the discharge could not be optimized, due to a large amount of outgas from the divertor head to the core plasma. The size of the divertor head was found to be larger than the optimum one; hence, the core plasma impacted slightly on the core plasma-facing portion of the divertor head with which the core plasma was not expected to collide.

  19. Characterizing the DIII-D divertor conditions during the tungsten ring experiment

    NASA Astrophysics Data System (ADS)

    Barton, J. L.; Watkins, J. G.; Wang, H. Q.; Nygren, R. E.; McLean, A.; Makowski, M.; Unterberg, E.; Thomas, D. M.; Guo, H. Y.; Guterl, J.; Buchenauer, B.

    2016-10-01

    Tungsten (W) is the leading divertor material in tokamaks, but the core W impurity fraction must be kept below 5 ×10-5 in a reactor. The DIII-D tokamak, having all graphite PFCs, has done a series of experiments with two W-coated molybdenum rings in the lower divertor to track W migration after plasma exposure. We characterize the divertor plasma conditions at the DIII-D target plate in L- and ELMing H-mode, and ELM suppressed plasmas. We will present data from an array of Langmuir probes in the divertor and divertor Thomson-scattering. We also compare the heat flux from fast thermocouples (7.5 mm below the surface of the metal tile inserts) and IRTV heat flux profiles from graphite tiles. The plasma conditions will be used to benchmark ERO modeling to aid in understanding the migration of sputtered W onto other plasma facing surfaces and will be compared to post exposure W distribution measured on the graphite tiles. Supported by US DOE under DE-AC04-94AL85000, DE-FC02-04ER54698, DE-AC05-000R22725, and DE-AC52-07NA27344.

  20. Creating Number Semantics through Finger Movement Perception

    ERIC Educational Resources Information Center

    Badets, Arnaud; Pesenti, Mauro

    2010-01-01

    Communication, language and conceptual knowledge related to concrete objects may rely on the sensory-motor systems from which they emerge. How abstract concepts can emerge from these systems is however still unknown. Here we report a functional interaction between a specific meaningful finger movement, such as a finger grip closing, and a concept…

  1. Correcting Finger Counting to Snellen Acuity.

    PubMed

    Karanjia, Rustum; Hwang, Tiffany Jean; Chen, Alexander Francis; Pouw, Andrew; Tian, Jack J; Chu, Edward R; Wang, Michelle Y; Tran, Jeffrey Show; Sadun, Alfredo A

    2016-10-01

    In this paper, the authors describe an online tool with which to convert and thus quantify count finger measurements of visual acuity into Snellen equivalents. It is hoped that this tool allows for the re-interpretation of retrospectively collected data that provide visual acuity in terms of qualitative count finger measurements.

  2. Correcting Finger Counting to Snellen Acuity

    PubMed Central

    Karanjia, Rustum; Hwang, Tiffany Jean; Chen, Alexander Francis; Pouw, Andrew; Tian, Jack J.; Chu, Edward R.; Wang, Michelle Y.; Tran, Jeffrey Show; Sadun, Alfredo A.

    2016-01-01

    ABSTRACT In this paper, the authors describe an online tool with which to convert and thus quantify count finger measurements of visual acuity into Snellen equivalents. It is hoped that this tool allows for the re-interpretation of retrospectively collected data that provide visual acuity in terms of qualitative count finger measurements. PMID:27928408

  3. Finger Mathematics: A Method for All Children.

    ERIC Educational Resources Information Center

    Ogletree, Earl J.; Chavez, Maria

    The instruction of finger counting and finger calculation, also known as Chisanbop, is promoted as a natural method of introducing and teaching the basic processes of addition, subtraction, multiplication and division to children, particularly to those who are mentally and physically handicapped. The sequential process for teaching finger…

  4. Creating Number Semantics through Finger Movement Perception

    ERIC Educational Resources Information Center

    Badets, Arnaud; Pesenti, Mauro

    2010-01-01

    Communication, language and conceptual knowledge related to concrete objects may rely on the sensory-motor systems from which they emerge. How abstract concepts can emerge from these systems is however still unknown. Here we report a functional interaction between a specific meaningful finger movement, such as a finger grip closing, and a concept…

  5. Generic Automated Multi-function Finger Design

    NASA Astrophysics Data System (ADS)

    Honarpardaz, M.; Tarkian, M.; Sirkett, D.; Ölvander, J.; Feng, X.; Elf, J.; Sjögren, R.

    2016-11-01

    Multi-function fingers that are able to handle multiple workpieces are crucial in improvement of a robot workcell. Design automation of multi-function fingers is highly demanded by robot industries to overcome the current iterative, time consuming and complex manual design process. However, the existing approaches for the multi-function finger design automation are unable to entirely meet the robot industries’ need. This paper proposes a generic approach for design automation of multi-function fingers. The proposed approach completely automates the design process and requires no expert skill. In addition, this approach executes the design process much faster than the current manual process. To validate the approach, multi-function fingers are successfully designed for two case studies. Further, the results are discussed and benchmarked with existing approaches.

  6. Laplacian trees - fingered growth in channel geometry

    NASA Astrophysics Data System (ADS)

    Szymczak, P.; Gubiec, T.

    2009-04-01

    A variety of natural growth processes, including viscous fingering, electrodeposition, or solidification can be modeled in terms of Laplacian growth. Laplacian growth patterns are formed when the boundary of a domain moves with a velocity proportional to the gradient of a field Ψ, which satisfies the Laplace equation, ‡2Ψ = 0, outside the domain. A simple model of Laplacian growth is considered, in which the growth takes place only at the tips of long, thin fingers [1]. The evolution of the fingers is studied by conformal mapping techniques. Analytical and numerical solutions are obtained for different domains and boundary conditions. In particular, a screening process is analyzed, when longer fingers suppress growth of the shorter ones. Possible geophysical applications of the model are discussed, including formation and evolution of the channels in a dissolving rock fracture. [1] T. Gubiec, P. Szymczak, Fingered growth in channel geometry: A Loewner equation approach , Phys. Rev. E, 77 , 041602, 2008

  7. Elastic fingering patterns in confined lifting flows.

    PubMed

    Fontana, João V; Miranda, José A

    2016-09-01

    The elastic fingering phenomenon occurs when two confined fluids are brought into contact, and due to a chemical reaction, the interface separating them becomes elastic. We study elastic fingering pattern formation in Newtonian fluids flowing in a lifting (time-dependent gap) Hele-Shaw cell. Using a mode-coupling approach, nonlinear effects induced by the interplay between viscous and elastic forces are investigated and the weakly nonlinear behavior of the fluid-fluid interfacial patterns is analyzed. Our results indicate that the existence of the elastic interface allows the development of unexpected morphological behaviors in such Newtonian fluid flow systems. More specifically, we show that depending on the values of the governing physical parameters, the observed elastic fingering structures are characterized by the occurrence of either finger tip splitting or side branching. The impact of the elastic interface on finger-competition events is also discussed.

  8. Use of twin dorsal middle phalangeal finger flaps for thumb or index finger reconstruction.

    PubMed

    Qi, W; Chen, K J

    2013-05-01

    Amputation or degloving injuries of the thumb or index finger are highly disabling. We describe the use of twin dorsal middle finger flaps harvested from the dorsal aspects of the middle and ring fingers, and based on one palmar proper digital artery, its venae comitantes, and the dorsal branches of the palmar digital nerves of the middle and ring fingers, respectively. These flaps offer advantages when large soft tissue defects of the thumb or index finger are present. In this study, twin dorsal middle finger flaps were used in nine patients (six thumbs, three index fingers). All flaps completely survived. At the mean follow-up of 20 months, the appearance of the reconstructed thumbs or index fingers was acceptable, the length was maintained, and the mean static 2-point discrimination values were 10 mm in the palmar flap and 13 mm in the dorsal flap of the reconstructed digit. All patients were satisfied with the appearance and mobility of the donor fingers. All but one donor finger showed normal finger pulp sensibility, with a static 2-point discrimination between 3 and 6 mm.

  9. Fingering in Stochastic Growth Models

    PubMed Central

    Aristotelous, Andreas C.; Durrett, Richard

    2015-01-01

    Motivated by the widespread use of hybrid-discrete cellular automata in modeling cancer, two simple growth models are studied on the two dimensional lattice that incorporate a nutrient, assumed to be oxygen. In the first model the oxygen concentration u(x, t) is computed based on the geometry of the growing blob, while in the second one u(x, t) satisfies a reaction-diffusion equation. A threshold θ value exists such that cells give birth at rate β(u(x, t) − θ)+ and die at rate δ(θ − u(x, t)+. In the first model, a phase transition was found between growth as a solid blob and “fingering” at a threshold θc = 0.5, while in the second case fingering always occurs, i.e., θc = 0. PMID:26430353

  10. Elastic Suppression of Viscous Fingering

    NASA Astrophysics Data System (ADS)

    Peng, Gunnar; Lister, John

    2016-11-01

    Consider peeling an elastic tape or beam away from a rigid base to which it is stuck by a film of viscous liquid. The peeling motion requires air to invade the viscous liquid and is thus susceptible to the Saffman-Taylor fingering instability. We analyse the fundamental travelling-wave solution and show that the advancing air-liquid interface remains linearly stable at higher capillary numbers than in a standard Hele-Shaw cell. A short-wavelength expansion yields an analytical expression for the growth rate which is valid for all unstable modes throughout the parameter space, allowing us to identify and quantify four distinct physical mechanisms that each help suppress the instability. Applying our method to the experiments by Pihler-Puzovic et al. (2012) reveals that the radial geometry and time-variation stabilize the system further.

  11. Fingering instability of Bingham fluids

    NASA Astrophysics Data System (ADS)

    Ghadge, Shilpa; Myers, Tim

    2005-11-01

    Contact line instabilities have been extensively studied and many useful results obtained for industrial applications. Our research in this area is to explore these instabilities for non-Newtonian fluids which has wide scope in geological, biological as well as industrial areas. In this talk, we will present an analysis of fingering instability near a contact line of the thin sheet of fluid flowing down on a moderately inclined plane. This instability has been well studied for Newtonian fluids. We explore the effect of a yield strength of the fluid on this instability. We have conveniently assumed the presence of the precussor film of small thickness ahead of the fluid film to avoid some mathematical singularities. Using a lubrication-type approximation, we perform a linear stability analysis of a straight contact line. We will show comparison with some experimental results using suspensions of kaolin in silicone oil as a yield strength fluid.

  12. Finger tremor in Parkinson's disease.

    PubMed

    Lakie, M; Mutch, W J

    1989-03-01

    Finger tremor was investigated in 20 patients (age range 54-88 yr) diagnosed as suffering from idiopathic Parkinson's disease and six controls of a similar age and no known neurological abnormality. In nine of the patients tremor was not clinically obvious. When the tremor of these patients was recorded immediately after voluntary movement and subjected to instrumental analysis there were consistently observable differences from the controls. Such analysis may have diagnostic potential when there is clinical uncertainty. Surface EMG recordings were obtained from four patients. One patient had a large resting tremor with obvious reciprocating activity in flexors and extensors; in the others who had no symptomatic tremor there was reciprocating activity only after movement, and this died away in a few seconds as the induced tremor disappeared.

  13. An effective preprocessing method for finger vein recognition

    NASA Astrophysics Data System (ADS)

    Peng, JiaLiang; Li, Qiong; Wang, Ning; Abd El-Latif, Ahmed A.; Niu, Xiamu

    2013-07-01

    The image preprocessing plays an important role in finger vein recognition system. However, previous preprocessing schemes remind weakness to be resolved for the high finger vein recongtion performance. In this paper, we propose a new finger vein preprocessing that includes finger region localization, alignment, finger vein ROI segmentation and enhancement. The experimental results show that the proposed scheme is capable of enhancing the quality of finger vein image effectively and reliably.

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

    SciTech Connect

    Petrie, T; Brooks, N; Fenstermacher, M; Groth, M; Hyatt, A; Isler, R; Lasnier, C; Leonard, A; Porter, G; Schaffer, M; Watkins, J; Wade, M; West, W

    2007-06-27

    'Puff and pump' radiating divertor scenarios [1,2] were applied to upper SN and DN H-mode plasmas. Under similar operating conditions, argon (Ar) accumulated in the main plasma of single-null (SN) plasmas more rapidly and reached a higher steady-state concentration when the B x {del}B ion drift direction was toward the divertor than when the B x {del}B ion drift direction was out of the divertor. The initial rate that Ar accumulated inside double-null (DN) plasmas was more than twice that of comparably-prepared SNs with the same B x {del}B direction. One way to reduce power loading at the divertor targets is to 'seed' the divertor plasma with impurities that radiatively reduce the conducted power. Studies have shown that the concentration of impurities in the divertor are increased by raising the flow of deuterium ions (D{sup +}) into the divertor by a combination of upstream deuterium gas puffing and active particle exhaust at the divertor targets, i.e., puff-and-pump. An enhanced D{sup +} particle flow toward the divertor targets exerts a frictional drag on impurities, and inhibits their escape from the divertor. A puff-and-pump approach using Ar as the impurity was successfully applied in recent DIII-D experiments to SN plasmas [3] while maintaining good H-mode performance. Studies on DIII-D and other tokamaks have shown that both the direction of the toroidal magnetic field B{sub T} and the degree of magnetic balance between divertors [i.e., the degree to which the plasma shape is considered SN or DN] are important factors in determining recycling and particle pumping [4,5]. It is unclear whether the favorable results of Ref. [3] can be extended to cases with different magnetic balance and/or B{sub T} direction. We show in this paper that reversing the direction of B{sub T} or altering the divertor magnetic balance does have an impact on how plasmas behave under puff-and-pump conditions. Our study takes advantage of DIII-D's capabilities to actively pump SN and

  15. Differing Dynamics of Intrapersonal and Interpersonal Coordination: Two-finger and Four-Finger Tapping Experiments

    PubMed Central

    Kodama, Kentaro; Furuyama, Nobuhiro; Inamura, Tetsunari

    2015-01-01

    Finger-tapping experiments were conducted to examine whether the dynamics of intrapersonal and interpersonal coordination systems can be described equally by the Haken—Kelso—Bunz model, which describes inter-limb coordination dynamics. This article reports the results of finger-tapping experiments conducted in both systems. Two within-subject factors were investigated: the phase mode and the number of fingers. In the intrapersonal experiment (Experiment 1), the participants were asked to tap, paced by a gradually hastening auditory metronome, looking at their fingers moving, using the index finger in the two finger condition, or the index and middle finger in the four-finger condition. In the interpersonal experiment (Experiment 2), pairs of participants performed the task while each participant used the outside hand, tapping with the index finger in the two finger condition, or the index and middle finger in the four-finger condition. Some results did not agree with the HKB model predictions. First, from Experiment 1, no significant difference was observed in the movement stability between the in-phase and anti-phase modes in the two finger condition. Second, from Experiment 2, no significant difference was found in the movement stability between the in-phase and anti-phase mode in the four-finger condition. From these findings, different coordination dynamics were inferred between intrapersonal and interpersonal coordination systems against prediction from the previous studies. Results were discussed according to differences between intrapersonal and interpersonal coordination systems in the availability of perceptual information and the complexity in the interaction between limbs derived from a nested structure. PMID:26070119

  16. Enhancement of cross-field transport into the private region of detached-divertor in Large Helical Device

    NASA Astrophysics Data System (ADS)

    Tanaka, H.; Ohno, N.; Tsuji, Y.; Kajita, S.; Masuzaki, S.; Kobayashi, M.; Morisaki, T.; Tsuchiya, H.; Komori, A.; LHD Experimental Group

    2010-10-01

    The fluctuation of ion saturation currents in the attached- and detached-divertor plasmas of the Large Helical Device [Fujiwara et al., Nucl. Fusion 41, 1355 (2001)] has been measured using a Langmuir probe array embedded in a divertor plate. Analytical results indicate that these fluctuation properties differ considerably from each other; for instance, the mean value distribution expands to and positive spikes propagate toward a private region from the divertor leg in the detached-divertor. We investigated the magnetic field lines traced from probe electrodes by using the KMAG code [Nakamura et al., J. Plasma Fusion Res. 69, 41 (1993)], and it is then confirmed that the propagation direction of positive spikes corresponds to that predicted by the theory of blobby plasma transport. This phenomenon is expected to lead to the broadening of plasma particle and heat fluxes to the divertor plate.

  17. Enhancement of cross-field transport into the private region of detached-divertor in Large Helical Device

    SciTech Connect

    Tanaka, H.; Ohno, N.; Tsuji, Y.; Kajita, S.; Masuzaki, S.; Kobayashi, M.; Morisaki, T.; Tsuchiya, H.; Komori, A.

    2010-10-15

    The fluctuation of ion saturation currents in the attached- and detached-divertor plasmas of the Large Helical Device [Fujiwara et al., Nucl. Fusion 41, 1355 (2001)] has been measured using a Langmuir probe array embedded in a divertor plate. Analytical results indicate that these fluctuation properties differ considerably from each other; for instance, the mean value distribution expands to and positive spikes propagate toward a private region from the divertor leg in the detached-divertor. We investigated the magnetic field lines traced from probe electrodes by using the KMAG code [Nakamura et al., J. Plasma Fusion Res. 69, 41 (1993)], and it is then confirmed that the propagation direction of positive spikes corresponds to that predicted by the theory of blobby plasma transport. This phenomenon is expected to lead to the broadening of plasma particle and heat fluxes to the divertor plate.

  18. Increased heat dissipation with the X-divertor geometry facilitating detachment onset at lower density in DIII-D

    NASA Astrophysics Data System (ADS)

    Covele, B.; Kotschenreuther, M.; Mahajan, S.; Valanju, P.; Leonard, A.; Watkins, J.; Makowski, M.; Fenstermacher, M.; Si, H.

    2017-08-01

    The X-divertor geometry on DIII-D has demonstrated reduced particle and heat fluxes to the target, facilitating detachment onset at 10-20% lower upstream density and higher H-mode pedestal pressure than a standard divertor. SOLPS modeling suggests that this effect cannot be explained by an increase in total connection length alone, but rather by the addition of connection length specifically in the power-dissipating volume near the target, via poloidal flux expansion and flaring. However, poloidal flaring must work synergistically with divertor closure to most effectively reduce the detachment density threshold. The model also points to carbon radiation as the primary driver of power dissipation in divertors on the DIII-D floor, which is consistent with experimental observations. Sustainable divertor detachment at lower density has beneficial consequences for energy confinement and current drive efficiency for core operation, while simultaneously satisfying the exhaust requirements of the plasma-facing components.

  19. Flute instability in the tandem mirror with the divertor/dipole regions

    SciTech Connect

    Katanuma, I.; Masaki, S.; Sato, S.; Sekiya, K.; Ichimura, M.; Imai, T.

    2011-11-15

    The numerical simulation is performed in GAMMA10 A-divertor magnetic configuration, which is a candidate of remodeled device of the GAMMA10 tandem mirror [M. Inutake et al., Phys. Rev. Lett. 55, 939 (1985)]. Both divertor and dipole regions are included in the numerical calculation, which is a new point. The electron short circuit effect along x-point, therefore, is not assumed so that it is not used the boundary condition of the electrostatic perturbations being zero at the separatrix on which the magnetic field lines pass through x-point. The simulation results reveal that the dipole field plays a role of a good magnetic field line curvature to the GAMMA10 A-divertor, and so the flute modes are stabilized without help of electron short circuit effects.

  20. Carbon Ion Flow Measurements in DIII-D Divertors Using Coherence Imaging Spectroscopy

    NASA Astrophysics Data System (ADS)

    Allen, S. L.; Meyer, W. H.; Porter, G. D.; Howard, J.

    2013-10-01

    New, single-crystal imaging interferometers along with improved relay optics have been installed in the upper and lower DIII-D divertors. These provide improved images of the Doppler shift and thereby flow of CIII (465 nm) ions. An improved in-situ calibration technique has been implemented, providing zero velocity reference images and measured spectrometer phase vs wavelength. The temperature resolution of the system has been greatly improved, resulting in a stable wavelength calibration. Image intensified cameras have made possible measurements of flow during ELMS and in the non-active divertor. Streamlined data analysis has been used to look for flow trends. In general, we see flow in opposite directions on the inner and outer scrape-off layers in the divertor. Supported in part by the US Department of Energy under DE-AC52-07NA27344 and DE-FC02-04ER54698.

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

    NASA Astrophysics Data System (ADS)

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

    2015-08-01

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

  2. Using Divertor Strike Point Splitting to Study Plasma Response and Its Sensitivity to Equilibrium Uncertainties

    NASA Astrophysics Data System (ADS)

    Lee, J. S.; Orlov, D. M.; Moyer, R. A.; Bykov, I.; Evans, T. E.; Wu, W.; Lyons, B. C.; Sugiyama, L. E.

    2016-10-01

    Magnetic field perturbations (RMPs) split the strike points in divertor tokamaks. This splitting is measured using fast imaging of filtered visible light from the divertor. We compare the observed splitting during n=3 RMP experiments to vacuum and plasma response modeling to determine if the measured splitting provides a sensitive diagnostic for the plasma response to the RMP. We also investigate the sensitivity of the computed plasma response to uncertainties in the initial 2D equilibrium. Strike point splitting was also observed in ELMing H-mode without the RMP, possibly due to n=1 error- and error-field correction fields. We compare the measured splitting during ELMs to linear plasma response modeling of the divertor footprints, and to nonlinear M3D ELM simulations. Work supported by U.S. DOE under Grant Numbers DE-FG02-07ER54917, DE-FG02-05ER54809.

  3. Calculation of Divertor Thermal Response as a Function of Material Composition for NSTX

    NASA Astrophysics Data System (ADS)

    Chaffin, Michael; Maingi, Rajesh

    2007-11-01

    Present tokamak designs use a magnetic divertor to deposit heat from the edge plasma onto Plasma Facing Components (PFCs) designed to remove the heat. Studying how this heat is distributed under various discharge conditions gives insight into how heat deposition can be optimized, and how different materials respond to plasma heating. In the National Spherical Torus eXperiment (NSTX), infrared cameras are used to measure divertor surface temperature, from which heat flux is computed using a 1D semi-infinite slab model with constant thermal conductivity. Here, a 1D simulation of the PFCs incorporating temperature-dependent thermal properties is used to compute heat flux profiles resolved across time and tile thickness. The PFC response to a given heat flux is also computed, and comparisons of resulting temperature profiles are made for a variety of materials including ATJ graphite (presently in the NSTX divertor), pyrolytic graphite, molybdenum, and tungsten.

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

    NASA Astrophysics Data System (ADS)

    Petersson, P.; Rubel, M.; Esser, H. G.; Likonen, J.; Koivuranta, S.; Widdowson, A.

    2015-08-01

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

  5. Structural studies of deposited layers on JET MkII-SRP inner divertor tiles

    NASA Astrophysics Data System (ADS)

    Likonen, J.; Coad, J. P.; Vainonen-Ahlgren, E.; Renvall, T.; Hole, D. E.; Rubel, M.; Widdowson, A.; JET-EFDA Contributors

    2007-06-01

    Deposited layers formed on JET inner divertor tiles during 1998-2004 and 2001-2004 campaigns have been investigated using secondary ion mass spectrometry (SIMS), Rutherford Backscattering (RBS) and optical microscopy. The thickness of the deposit decreases from the top of vertical tile 1 to the bottom and then increases on vertical tile 3 reaching ∼60 μm. There are even thicker deposits on the small sloping section of the floor tile 4 that can be accessed by the plasma at the inner divertor legs. Deposited films on divertor inner wall tiles are enriched in Be indicating chemical erosion of C and a multi-step transport of C to the shadowed area on floor tile 4. The films have generally a layered and globular structure in the areas with plasma contact.

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

    SciTech Connect

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

    2008-09-17

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

  7. Simulation of turbulence in the divertor region of tokamak edge plasma

    NASA Astrophysics Data System (ADS)

    Umansky, M. V.; Rognlien, T. D.; Xu, X. Q.

    2005-03-01

    Results are presented for turbulence simulations with the fluid edge turbulence code BOUT [X.Q. Xu, R.H. Cohen, Contr. Plas. Phys. 36 (1998) 158]. The present study is focussed on turbulence in the divertor leg region and on the role of the X-point in the structure of turbulence. Results of the present calculations indicate that the ballooning effects are important for the divertor fluctuations. The X-point shear leads to weak correlation of turbulence across the X-point regions, in particular for large toroidal wavenumber. For the saturated amplitudes of the divertor region turbulence it is found that amplitudes of density fluctuations are roughly proportional to the local density of the background plasma. The amplitudes of electron temperature and electric potential fluctuations are roughly proportional to the local electron temperature of the background plasma.

  8. SOLPS modeling of an innovative small-angle slot divertor concept for low-density detachment

    NASA Astrophysics Data System (ADS)

    Covele, B.; Sang, C.; Guo, H.; Lao, L.; Stangeby, P.; Thomas, D.

    2016-10-01

    SOLPS modeling offers insight into how a new Small-Angle Slot (SAS) divertor concept exploits the role of neutral trapping to exhaust power and particles at lower core densities than even highly slanted divertors. The special SAS baffling structure enhances volumetric power and momentum losses across the entire target profile, flattening temperatures even in the far SOL. SOLPS characterizes SAS heat and temperature handling for a spectrum of plasma and neutral source conditions, varying ne,sep, PSOL, heat flux width, gas puffing rates and locations, and pumping rates. Certain aspects of the baffling structure were also systematically varied to observe the effect on the neutral dynamics, particularly pressure gradients in D2 near the target. Radial transport coefficients were controlled to match midplane profiles to experimental H-mode profiles. The SAS divertor is an excellent testbed for probing the interplay between plasma and neutrals at the onset of detachment. The SAS concept is developed under General Atomics corporate funding.

  9. The super X divertor (SXD) and a compact fusion neutron source (CFNS)

    SciTech Connect

    Kotschenreuther, M.; Valanju, P.; Mahajan, S.; Zheng, L. J.; Pearlstein, L. D.; Bulmer, R. H.; Canik, John; Maingi, R.

    2010-01-01

    A new magnetic geometry, the super X divertor (SXD), is invented to solve severe heat exhaust problems in high power density fusion plasmas. SXD divertor plates are moved to the largest major radii inside the TF coils, increasing the wetted area by 2-3 and the line length by 2-5. Two-dimensional fluid simulations with SOLPS (Schneider et al 2006 SOLPS 2-D edge calculation code Contrib. Plasma Phys. 46) show a several-fold decrease in divertor heat flux and plasma temperature at the plate. A small high power density tokamak using SXD is proposed, for either (1) useful fusion applications using conservative physics, such as a component test facility (CTF) or fission fusion hybrid, or (2) to develop more advanced physics modes for a pure fusion reactor in an integrated fusion environment.

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

    SciTech Connect

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

    2016-01-22

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

  11. Local deposition of 13C tracer in the JET MKII-HD divertor

    NASA Astrophysics Data System (ADS)

    Likonen, Jari; Airila, M. I.; Coad, J. P.; Hakola, A.; Koivuranta, S.; Ahonen, E.; Alves, E.; Barradas, N.; Widdowson, A.; Rubel, M.; Brezinsek, S.; Groth, M.; JET-EFDA Contributors

    2013-07-01

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

  12. Finger wear detection for production line battery tester

    DOEpatents

    Depiante, E.V.

    1997-11-18

    A method is described for detecting wear in a battery tester probe. The method includes providing a battery tester unit having at least one tester finger, generating a tester signal using the tester fingers and battery tester unit with the signal characteristic of the electrochemical condition of the battery and the tester finger, applying wavelet transformation to the tester signal including computing a mother wavelet to produce finger wear indicator signals, analyzing the signals to create a finger wear index, comparing the wear index for the tester finger with the index for a new tester finger and generating a tester finger signal change signal to indicate achieving a threshold wear change. 9 figs.

  13. Finger wear detection for production line battery tester

    DOEpatents

    Depiante, Eduardo V.

    1997-01-01

    A method for detecting wear in a battery tester probe. The method includes providing a battery tester unit having at least one tester finger, generating a tester signal using the tester fingers and battery tester unit with the signal characteristic of the electrochemical condition of the battery and the tester finger, applying wavelet transformation to the tester signal including computing a mother wavelet to produce finger wear indicator signals, analyzing the signals to create a finger wear index, comparing the wear index for the tester finger with the index for a new tester finger and generating a tester finger signal change signal to indicate achieving a threshold wear change.

  14. ELM PARTICLE AND ENERGY TRANSPORT IN THE SOL AND DIVERTOR OF DIII-D

    SciTech Connect

    FENSTERMACHER,ME; LEONARD,AW; SNYDER,PB; BOEDO,JA; COLCHIN,RJ; GROEBNER,RJ; GRAY,DS; GROTH,M; HOLLMANN,E; LASNIER,CJ; OSBORNE,TH; PETRIE,TW; RUDAKOV,DL; TAKAHASHI,H; WATKINS,JG; ZENG,L

    2003-04-01

    A271 ELM PARTICLE AND ENERGY TRANSPORT IN THE SOL AND DIVERTOR OF DIII-D. Results from a series of dedicated experiments measuring the effect of particle and energy pulses from Type-I Edge Localized Modes (ELMs) in the DIII-D scrape-off layer (SOL) and divertor are compared with a simple model of ELM propagation in the boundary plasma. The simple model asserts that the propagation of ELM particle and energy perturbations is dominated by ion parallel convection along SOL fields lines and the recovery from the ELM perturbation is determined by recycling physics. Time scales associated with the initial changes of boundary plasma parameters are expected to be on the order of the ion transit time from the outer midplane, where the ELM instability is initiated, to the divertor targets. To test the model, the ion convection velocity is changed in the experiment by varying the plasma density. At moderate to high density, n{sub e}/n{sub Gr} = 0.5-0.8, the delays in the response of the boundary plasma to the midplane ELM pulses, the density dependence of those delays and other observations are consistent with the model. However, at the lowest densities, n{sub e}/n{sub Gr} {approx} 0.35, small delays between the response sin the two divertors, and changes in the response of the pedestal thermal energy to ELM events, indicate that additional factors including electron conduction in the SOL, the pre-ELM condition of the divertor plasma, and the ratio of ELM instability duration to SOL transit time, may be playing a role. The results show that understanding the response of the SOL and divertor plasmas to ELMs, for various pre-ELM conditions, is just as important to predicting the effect of ELM pulses on the target surfaces of future devices as is predicting the characteristics of the ELM perturbation of the core plasma.

  15. A multichannel visible spectroscopy system for the ITER-like W divertor on EAST

    NASA Astrophysics Data System (ADS)

    Mao, Hongmin; Ding, Fang; Luo, Guang-Nan; Hu, Zhenhua; Chen, Xiahua; Xu, Feng; Yang, Zhongshi; Chen, Jingbo; Wang, Liang; Ding, Rui; Zhang, Ling; Gao, Wei; Xu, Jichan; Wu, Chengrui

    2017-04-01

    To facilitate long-pulse high power operation, an ITER-like actively cooled tungsten (W) divertor was installed in Experimental Advanced Superconducting Tokamak (EAST) to replace the original upper graphite divertor in 2014. A dedicated multichannel visible spectroscopic diagnostic system has been accordingly developed for the characterization of the plasma and impurities in the W divertor. An array of 22 lines-of-sight (LOSs) provides a profile measurement of the light emitted from the plasma along upper outer divertor, and the other 17 vertical LOSs view the upper inner divertor, achieving a 13 mm poloidal resolution in both regions. The light emitted from the plasma is collected by a specially designed optical lens assembly and then transferred to a Czerny-Turner spectrometer via 40 m quartz fibers. At the end, the spectra dispersed by the spectrometer are recorded with an Electron-Multiplying Charge Coupled Device (EMCCD). The optical throughput and quantum efficiency of the system are optimized in the wavelength range 350-700 nm. The spectral resolution/coverage can be adjusted from 0.01 nm/3 nm to 0.41 nm/140 nm by switching the grating with suitable groove density. The frame rate depends on the setting of LOS number in EMCCD and can reach nearly 2 kHz for single LOS detection. The light collected by the front optical lens can also be divided and partly transferred to a photomultiplier tube array with specified bandpass filter, which can provide faster sampling rates by up to 200 kHz. The spectroscopic diagnostic is routinely operated in EAST discharges with absolute optical calibrations applied before and after each campaign, monitoring photon fluxes from impurities and H recycling in the upper divertor. This paper presents the technical details of the diagnostic and typical measurements during EAST discharges.

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

    NASA Astrophysics Data System (ADS)

    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. Simulating Divertor Detachment of Ohmic Discharges in ASDEX Upgrade Using SOLPS: the Role of Carbon

    SciTech Connect

    Wischmeier, M; Coster, D; Chankin, A; Fuchs, C; Groth, M; Harhausen, J; Kallenbach, A; Muller, H; Tsalas, M; Wolfrum, E

    2007-06-27

    With divertor detachment being a prerequisite for burning plasma operation in ITER, numerical codes such as SOLPS [1] have been developed for predicting and interpreting the divertor performance at all operational regimes in current tokamaks and ITER. In ITER complete detachment from the outer divertor target is not permitted as this might result in an X-point MARFE, imposing an upper limit for the upstream separatrix density, n{sub e}{sup sep}. Despite the knowledge of the basic mechanisms required for achieving detachment, such as radiative power exhaust, volumetric momentum and charge removal [1], a quantitative evaluation of experimentally observed detached regimes proves to be particularly difficult for several tokamaks. In particular the strong asymmetry of the ion flux density between the inner, {Lambda}{sub it}, and the outer target {Lambda}{sub ot} with increasing line averaged density, {bar n}{sub e}, and in particular ''vanishing'' of the ion flux, defined as full/complete detachment, at the inner target cannot be reproduced. It is unclear how this is related to divertor target plates or other plasma facing components containing carbon. As part of a combined effort at various experimental devices this paper contributes to the validation of the SOLPS code against experimental data from ASDEX Upgrade, AUG, at the onset of divertor detachment. In the framework established under the International Tokamak Physics Activity (ITPA) Divertor and SOL working group a series of ohmic discharges have been performed in AUG, which had as similar as possible plasma parameters as companion discharges undertaken in DIII-D [2]. The effect of activating drift terms, the influence of the chemical sputtering yield at the inner target and in addition to [3] the role of impurity influx from the inner heat shield are analyzed.

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

    SciTech Connect

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

    1997-11-01

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

  19. A Comparison of Radiating Divertor Behavior in Single- and Double-Null Plasmas in DIII-D

    SciTech Connect

    Petrie, T W; Brooks, N H; Fenstermacher, M E; Groth, M; Hyatt, A W; Isler, R C; Lasnier, C J; Leonard, A W; Porter, G D; Schaffer, M J; Watkins, J G; Wade, M R; West, W P

    2008-03-25

    'Puff and pump' radiating divertor scenarios, applied to both upper single-null (SN) and double-null (DN) H-mode plasmas, result in a 30-60% increase in radiated power with little or no decrease in {tau}{sub E}. Argon was injected into the private flux region of the upper divertor, and plasma flow into the upper divertor was enhanced by a combination of deuterium gas puffing upstream of the divertor targets and particle pumping at the targets. For the same constant deuterium injection rate, argon penetrated the main plasma of SNs more rapidly and reached a higher steady-state concentration when the Bx{del}B-ion drift direction was toward the divertor (V{sub {del}B{up_arrow}}) rather than away from the divertor (V{sub {del}B{down_arrow}}). We also found that the initial rate at which argon accumulated inside DN plasmas was more than twice that of comparable SN plasmas having the same Bx{del}B-ion drift direction. In DNs, the radiated power was not shared equally between divertors during argon injection. Only in the divertor opposite Bx{del}B ion drift direction were both significant increases in divertor radiated power and an accumulation of argon, based on spectroscopic measurements of ArII, observed. Our data suggests that a DN shape that is biased in the direction away from the Bx{del}B-ion drift direction may provide the best prospect of successfully coupling a radiating divertor approach with a higher performance H-mode plasma.

  20. Plasma Flow Interaction With Iter Divertor Related Surfaces

    NASA Astrophysics Data System (ADS)

    Dojcinovic, I. P.

    2010-07-01

    It has been found that the plasma flow generated by quasistationary plasma accelerators can be used for simulation of high energy plasma interaction with different materials of interest for fusion experiments (Arkhipov et al. 2000, Federici et al. 2005). It is especially important for the studies of the processes such as ELMs (edge localized modes), plasma disruptions and VDEs (vertical displacement events), during which a significant part of the confined hot plasma is lost from the core to the SOL (scrape off layer) enveloping the core region. Experiments using plasma guns have been used to assess erosion from disruptions and ELMs. Namely, in this experiment modification of different targets, like molybdenum, CFC and silicon single crystal surface by the action of hydrogen and nitrogen quasistationary compression plasma flow (CPF) generated by magnetoplasma compressor (MPC) has been studied. MPC plasma flow with standard parameters (1 MJ/m^2 in 0.1 ms) can be used for simulation of transient peak thermal loads during Type I ELMs and disruptions (Dojcinovic et al. 2007). Analysis of the targets erosion, brittle destruction, melting processes, and dust formation has been performed (Dojcinovic et al. 2006). These surface phenomena are results of specific conditions during CPF interaction with target surface. The investigations are related to the fundamental aspects of high energy plasma flow interaction with different material of interest for fusion. One of the purposes is a study of competition between melting and cleavage of treated solid surface. The other is investigation of plasma interaction with first wall and divertor component materials related to the ITER experiment.

  1. Theory of probe measurements at the divertor plate

    SciTech Connect

    Porter, G.D.; Ryutov, D.D.

    1996-12-31

    Probe measurements represent a technically simple and inexpensive approach to the characterization of plasma parameters in the divertor region. On the other hand, the interpretation of the probe signals is sometimes not straightforward, with discrepancies between the results of probe measurements and Thomson scattering measurements often arising. The difficulty of interpretation of probe measurements stems predominantly from the unknown influence of a strong magnetic field on the probe current-voltage characteristics (CVC). There have been many studies of this issue, among the most recent ones papers. In our paper, we present analysis of the physics issues which determine the performance of so called {open_quotes}flush mounted probes{close_quotes}. We note that, in case of an infinitely strong magnetic field, the flux-tube whose footprint coincides with the probe surface, is completely isolated from the rest of the plasma, and the probe CVC becomes more similar to the CVC of a double probe. As a next step in the analysis, we consider classical cross-field transport processes and conclude that, for the flux tubes with dimensions of a few ion gyroradii (as is the case for the probes of the type used in DMD) the cross-field currents are dominated by the ion viscosity. We derive the probe CVC for this case and find that it has a remarkable similarity with the aforementioned characteristics of the double probe. We consider possible effects of plasma turbulence on the cross field transport through a thin flux tube leaning on the probe. We conclude that the character of this effect strongly depends on the spatial and temporal scale of the plasma fluctuations: the influence of fast short-wavelength fluctuations can be described in terms of enhanced diffusion across the flux tube, while the influence of the slow ones can not. As the SOL turbulence is usually slow, we consider in more detail the effect of slow fluctuations.

  2. The effect of enslaving on perception of finger forces.

    PubMed

    Li, Sheng; Leonard, Charles T

    2006-07-01

    The primary purpose was to examine the effect of enslaving on finger force perception during isometric finger force production using an ipsilateral force-matching paradigm. Fourteen subjects were instructed to produce varying levels of reference forces [10, 20, 30, and 40% maximal voluntary contraction (MVC)] force using one finger (index, I or little, L) and to reproduce these forces using the same finger (homo-finger tasks, I/I and L/L) or a different finger (hetero-finger tasks, I/L and L/I). Forces of all fingers were recorded. During homo-finger tasks, no differences were found in force magnitude or relative level of force (expressed as a proportion of MVC). The index finger matching force magnitudes were greater than the little finger reference force magnitudes, with significantly lower levels of relative force during L/I tasks; while the little finger matching forces underestimated the index finger reference forces with significantly higher levels of relative force during I/L tasks. The difference in the matching and reference forces by the instructed finger(s), i.e., matching error, was larger in hetero-finger tasks than in homo-finger tasks, particularly at high reference force levels (30, 40% MVC). When forces of all fingers were considered, enslaving (uninstructed finger forces) significantly minimized matching errors of the total force during both I/L and L/I hetero-finger tasks, especially at high reference force levels. Our results show that there is a tendency to match the absolute magnitude of the total force during ipsilateral finger force-matching tasks. This tendency is likely related to enslaving effects. Our results provide evidence that all (instructed and uninstructed) finger forces are sensed, thus resulting in perception of the absolute magnitude of total finger force.

  3. Analysis of pumping requirement for exhausting duct in close vicinity of divertor in Tokamak Reactor

    SciTech Connect

    Saito, S.; Abe, T.; Fujisawa, N.; Sugihara, M.; Veda, K.

    1983-11-01

    An improved method for Monte Carlo simulation is described to calculate the neutral-particle transport in a divertor throat and to evaluate the helium removal efficiency from a burning plasma. The required pumping speed for the helium removal is discussed with special emphasis placed on the effects of long exhausting duct and of scrape-off plasma variables. The analysis for International Tokamak Reactor (INTOR) single null divertor suggests a possibility that the pumping requirement for INTOR could be drastically eased--e.g., <10/sup 4/ l/s, for the high scrape-off plasma density of the order of 10/sup 13/ cm/sup -3/.

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

    SciTech Connect

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

    2009-02-26

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

  5. Achievements and challenges in automated parameter, shape and topology optimization for divertor design

    NASA Astrophysics Data System (ADS)

    Baelmans, M.; Blommaert, M.; Dekeyser, W.; Van Oevelen, T.

    2017-03-01

    Plasma edge transport codes play a key role in the design of future divertor concepts. Their long simulation times in combination with a large number of control parameters turn the design into a challenging task. In aerodynamics and structural mechanics, adjoint-based optimization techniques have proven successful to tackle similar design challenges. This paper provides an overview of achievements and remaining challenges with these techniques for complex divertor design. It is shown how these developments pave the way for fast sensitivity analysis and improved design from different perspectives.

  6. Impurity diagnosis of a KSTAR graphite divertor tile using laser induced breakdown spectroscopy technique

    NASA Astrophysics Data System (ADS)

    Kim, Minju; Cho, Min Sang; Cho, Byoung Ick

    2017-04-01

    Laser induced breakdown spectroscopy (LIBS) has been tested to diagnose impurity elements on a Korea Superconducting Tokamak Advanced Research (KSTAR) divertor tile. Spectral lines of various impurity elements such as iron, chromium, and nickel were detected from the divertor surface. The variation of spectra with consecutive laser pulses demonstrates the potential for depth profiling analysis for the deposited impurity layer. The LIBS plasma parameters have been qualitatively determined from analysis of the relative line intensities and linewidths for each element. The validity of this analysis has been checked with atomic spectral simulations.

  7. Finger somatotopy in human motor cortex.

    PubMed

    Beisteiner, R; Windischberger, C; Lanzenberger, R; Edward, V; Cunnington, R; Erdler, M; Gartus, A; Streibl, B; Moser, E; Deecke, L

    2001-06-01

    Although qualitative reports about somatotopic representation of fingers in the human motor cortex exist, up to now no study could provide clear statistical evidence. The goal of the present study was to reinvestigate finger motor somatotopy by means of a thorough investigation of standardized movements of the index and little finger of the right hand. Using high resolution fMRI at 3 Tesla, blood oxygenation level-dependent (BOLD) responses in a group of 26 subjects were repeatedly measured to achieve reliable statistical results. The center of mass of all activated voxels within the primary motor cortex was calculated for each finger and each run. Results of all runs were averaged to yield an individual index and little finger representation for each subject. The mean center of mass localizations for all subjects were then submitted to a paired t test. Results show a highly significant though small scale somatotopy of fingerspecific activation patterns in the order indicated by Penfields motor homunculus. In addition, considerable overlap of finger specific BOLD responses was found. Comparing various methods of analysis, the mean center of mass distance for the two fingers was 2--3 mm with overlapping voxels included and 4--5 mm with overlapping voxels excluded. Our data may be best understood in the context of the work of Schieber (1999) who recently described overlapping somatotopic gradients in lesion studies with humans. Copyright 2001 Academic Press.

  8. Trigger finger, tendinosis, and intratendinous gene expression.

    PubMed

    Lundin, A-C; Aspenberg, P; Eliasson, P

    2014-04-01

    The pathogenesis of trigger finger has generally been ascribed to primary changes in the first annular ligament. In contrast, we recently found histological changes in the tendons, similar to the findings in Achilles tendinosis or tendinopathy. We therefore hypothesized that trigger finger tendons would show differences in gene expression in comparison to normal tendons in a pattern similar to what is published for Achilles tendinosis. We performed quantitative real-time polymerase chain reaction on biopsies from finger flexor tendons, 13 trigger fingers and 13 apparently healthy control tendons, to assess the expression of 10 genes which have been described to be differently expressed in tendinosis (collagen type 1a1, collagen 3a1, MMP-2, MMP-3, ADAMTS-5, TIMP-3, aggrecan, biglycan, decorin, and versican). In trigger finger tendons, collagen types 1a1 and 3a1, aggrecan and biglycan were all up-regulated, and MMP-3and TIMP-3 were down-regulated. These changes were statistically significant and have been previously described for Achilles tendinosis. The remaining four genes were not significantly altered. The changes in gene expression support the hypothesis that trigger finger is a form of tendinosis. Because trigger finger is a common condition, often treated surgically, it could provide opportunities for clinical research on tendinosis. © 2012 John Wiley & Sons A/S. Published by John Wiley & Sons Ltd.

  9. More efficient swimming by spreading your fingers

    NASA Astrophysics Data System (ADS)

    van de Water, Willem; van Houwelingen, Josje; Willemsen, Dennis; Breugem, Wim Paul; Westerweel, Jerry; Delfos, Rene; Grift, Ernst Jan

    2016-11-01

    A tantalizing question in free-style swimming is whether the stroke efficiency during the pull phase depends on spreading the fingers. It is a subtle effect-not more than a few percent-but it could make a big difference in a race. We measure the drag of arm models with increasing finger spreading in a wind tunnel and compare forces and moments to the results of immersed boundary simulations. Virtual arms were used in the simulations and their 3D-printed real versions in the experiment. We find an optimal finger spreading, accompanied by a marked increase of coherent vortex shedding. A simple actuator disk model explains this optimum.

  10. Optical flow based finger stroke detection

    NASA Astrophysics Data System (ADS)

    Zhu, Zhongdi; Li, Bin; Wang, Kongqiao

    2010-07-01

    Finger stroke detection is an important topic in hand based Human Computer Interaction (HCI) system. Few research studies have carried out effective solutions to this problem. In this paper, we present a novel approach for stroke detection based on mono vision. Via analyzing the optical flow field within the finger area, our method is able to detect finger stroke under various camera position and visual angles. We present a thorough evaluation for each component of the algorithm, and show its efficiency and effectiveness on solving difficult stroke detection problems.

  11. Finger blood flow in Antarctica

    PubMed Central

    Elkington, E. J.

    1968-01-01

    1. Finger blood flow was estimated, by strain-gauge plethysmography, before and during a 1 hr immersion in ice water, on twenty-five men throughout a year at Wilkes, Antarctica. A total of 121 satisfactory immersions were made. 2. Blood flow before and during immersion decreased significantly in the colder months of the year, and the increase caused by cold-induced vasodilatation (CIVD) became less as the year progressed. The time of onset, blood flow at onset, and frequency of the cycles of CIVD showed no significant relation to the coldness of the weather (as measured by mean monthly wind chill) or the time in months. Comparisons of blood flow before and after five field trips (average duration 42 days), on which cold exposure was more severe than at Wilkes station, gave similar results. 3. The results suggest that vasoconstrictor tone increased. This interpretation agrees with previous work on general acclimatization in Antarctica, but contrasts with work elsewhere on local acclimatization of the hands. PMID:5684034

  12. Trajectory of the index finger during grasping.

    PubMed

    Friedman, Jason; Flash, Tamar

    2009-07-01

    The trajectory of the index finger during grasping movements was compared to the trajectories predicted by three optimization-based models. The three models consisted of minimizing the integral of the weighted squared joint derivatives along the path (inertia-like cost), minimizing torque change, and minimizing angular jerk. Of the three models, it was observed that the path of the fingertip and the joint trajectories, were best described by the minimum angular jerk model. This model, which does not take into account the dynamics of the finger, performed equally well when the inertia of the finger was altered by adding a 20 g weight to the medial phalange. Thus, for the finger, it appears that trajectories are planned based primarily on kinematic considerations at a joint level.

  13. [Ligament injuries of fingers and thumbs].

    PubMed

    Schmitt, R

    2017-01-01

    Degenerative and traumatic ligament lesions of the carpometacarpal joints frequently occur at the thumb ray, whereas the carpometacarpal amphiarthrosis of other finger rays are rarely affected. The metacarpophalangeal and interphalangeal joints of the thumb and fingers are stabilized by bilaterally running collateral ligaments and palmar plates. At the base of the metacarpophalangeal joints, several ligaments of the extensor hoods guide the extensor tendons and coordinate the fine motoric skills of phalangeal flexing and extending. Several annular and cruciform ligaments hold the flexor tendons close to the finger skeleton. Other than at the wrist, differentiation between dynamic and static instability patterns is possible by physical examination. This review article presents the ligaments of the thumb and the fingers, the traumatic and degenerative lesions as well as the diagnostic capability of x‑rays, cinematography, magnetic resonance imaging (MRI) and MR arthrography.

  14. Spreading and fingering in spin coating

    NASA Astrophysics Data System (ADS)

    Holloway, Kristi E.; Habdas, Piotr; Semsarillar, Naeim; Burfitt, Kim; de Bruyn, John R.

    2007-04-01

    We study the spreading and fingering of drops of silicone oil on a rotating substrate for a range of rotation speeds and drop volumes. The spreading of the drop prior to the onset of fingering is found to follow the theoretically predicted time dependence, but with a large shift in time scale. For the full range of experimental parameters studied, the contact line becomes unstable and fingers develop when the radius of the drop becomes sufficiently large. We study the growth of perturbations around the perimeter of the drop and find the growth rate of the most unstable mode to agree well with the predictions of lubrication theory. The number of fingers which form around the perimeter of the drop is found to be a function of both rotation speed and drop volume, and is also in excellent agreement with theoretical predictions.

  15. Case reports: thumb reconstruction using amputated fingers.

    PubMed

    Hoang, Nguyen T; Staudenmaier, R; Hoehnke, C

    2008-08-01

    Reconstruction of an irreparably amputated thumb in multiple digit amputations using amputated fingers can considerably improve hand function and allows creation of a newly transplanted thumb with acceptable cosmetic and functional attributes. However, the surgery is challenging and rarely reported. We report six cases using this procedure in patients with crushed thumbs unsuitable for replantation. In four of the patients, the remnant of the index finger was replanted on the thumb stump and in another two patients, an amputated middle finger and ring finger were used. The patients had a minimum followup of 12 months (mean, 18 months; range, 12-45 months). All newly transplanted thumbs survived resulting in the patients having satisfactory postoperative hand function and appearance.

  16. Repair of webbed fingers - series (image)

    MedlinePlus

    Syndactyly is the abnormal development of the hand, such that the fingers are fused. The number of ... second surgery, depending on the complexity of the syndactyly. Hospital stays of 1 or 2 days are ...

  17. Finger prosthesis: a boon to handicapped

    PubMed Central

    Gupta, Ridhima; Kumar, Lakshya; Rao, Jitendra; Singh, Kamleshwar

    2013-01-01

    This is a clinical case report of a 52-year-old male patient with four partially missing fingers of the left hand. The article describes the clinical and laboratory procedure of making prosthesis with modern silicone material. A wax pattern was fabricated using the right hand of the patient. A special type of wax was formulated to make the pattern so that it can be easily moulded and carved. Intrinsic and extrinsic staining was also performed to match the adjacent skin colour. The patient was given the finger prosthesis and was asked to use a half glove (sports) to mask the junction between the prosthesis and the normal tissue. It also provides additional retention to the artificial fingers. The patient felt his social acceptance improved after wearing the finger prosthesis. PMID:23988821

  18. Salt-finger convection under reduced gravity

    NASA Technical Reports Server (NTRS)

    Chen, C. F.

    1990-01-01

    Salt-finger convection in a double-diffusive system is a motion driven by the release of gravitational potential due to differential diffusion rates. Because of the fact that the destabilizing effect of the concentration gradient is amplified by the Lewis number (the ratio of thermal diffusivity to solute diffusivity) salt-finger convection can be generated at very much reduced gravity levels. This effect may be of importance in the directional solidification of binary alloys carried out in space. The transport of solute and heat by salt-finger convection at microgravity conditions is considered; instability arising from surface tension gradients, the Marangoni instability, is discussed, and the possible consequences of combined salt-finger and Marangoni instability are considered.

  19. Coexistent Regime of H-mode with a Dense & Cold Divorter plasma in JFT-2M Closed Divertor

    NASA Astrophysics Data System (ADS)

    Sengoku, Seio; JFT-2m Group

    1996-11-01

    A possibility of stable coexistence of an H-mode with a dense & cold divertor plasma had been demonstrated using a strong gas puffing in divertor chamber at the lower density limit for the H-mode transition ( 2x10^19m-3: the regime of no spontaneous dense & cold state) by modifying the divertor shape of JFT-2M to a closed configuration.footnote S. Sengoku et al., Bull. Amer. Phys. Soc. 40, 1675 (1995) A build up of neutral pressure occurs only in the divertor chamber without degrading the H-factor, and the divertor plasma results in a dense & cold state (n_e=1.3 2.5x 10^19m-3, T_e=4 15eV). In order to improve baffling effect and to extend operational regime of the coexistence, the divertor baffle plates of JFT-2M had been modified from relatively wide baffle-opening to more closed one. Studies on fueling and exhaustion, particle control, neutral buildup scaling and SOL plasma behaviors are being carried out with the modified divertor shape.

  20. Thermal Analysis to Calculate the Vessel Temperature and Stress in Alcator C-Mod Due to the Divertor Upgrade

    SciTech Connect

    Han Zhang, Peter H. Titus, Robert Ellis, Soren Harrison and Rui Vieira

    2012-08-29

    Alcator C-Mod is planning an upgrade to its outer divertor. The upgrade is intended to correct the existing outer divertor alignment with the plasma, and to operate at elevated temperatures. Higher temperature operation will allow study of edge physics behavior at reactor relevant temperatures. The outer divertor and tiles will be capable of operating at 600oC. Longer pulse length, together with the plasma and RF heat of 9MW, and the inclusion of heater elements within the outer divertor produces radiative energy which makes the sustained operation much more difficult than before. An ANSYS model based on ref. 1 was built for the global thermal analysis of C-Mod. It models the radiative surfaces inside the vessel and between the components, and also includes plasma energy deposition. Different geometries have been simulated and compared. Results show that steady state operation with the divertor at 600oC is possible with no damage to major vessel internal components. The differential temperature between inner divertor structure, or "girdle" and inner vessel wall is ~70oC. This differential temperature is limited by the capacity of the studs that hold the inner divertor backing plates to the vessel wall. At a 70oC temperature differential the stress on the studs is within allowable limits. The thermal model was then used for a stress pass to quantify vessel shell stresses where thermal gradients are significant.

  1. Interpretations of the impact of cross-field drifts on divertor flows in DIII-D with UEDGE

    DOE PAGES

    Jaervinen, Aaro E.; Allen, Steve L.; Groth, Mathias; ...

    2017-01-27

    Simulations using the multi-fluid code UEDGE indicates that, in low confinement (Lmode) plasmas in DIII-D, recycling driven flows dominate poloidal particle flows in the divertor, whereas E×B drift flows dominate the radial particle flows. In contrast, in high confinement (H-mode) conditions E×B drift flows dominate both poloidal and radial particle flows in the divertor. UEDGE indicates that the toroidal C2+ flow velocities in the divertor plasma are entrained within 30% to the background deuterium flow in both Land H-mode plasmas in the plasma region where the CIII 465 nm emission is measured. Therefore, UEDGE indicates that the Carbon Doppler Coherencemore » Imaging System (CIS), measuring the toroidal velocity of the C2+ ions, can provide insight to the deuterium flows in the divertor. Parallel-to-B velocity dominates the toroidal divertor flow; direct drift impact being less than 1%. Toroidal divertor flow is predicted to reverse when the magnetic field is reversed. This is explained by the parallel-B flow towards the nearest divertor plate corresponding to opposite toroidal directions in opposite toroidal field configurations. Due to strong poloidal E×B flows in H-mode, net poloidal particle transport can be in opposite direction than the poloidal component of the parallel-B plasma flow.« less

  2. Finger Lake Region, NY State, USA

    NASA Technical Reports Server (NTRS)

    1992-01-01

    This view of the central portion of upstate New York, centers on the Finger Lakes. The large city on the shore of Lake Ontario, is Rochester. Although the city, being a business, educational and technical center, has no heavy industry, the outline of the city shows fairly well in the snow, but not as well as the outlines of industrial cities elsewhere in the world. The Finger Lakes are large linear lakes carved out by glaciers during the last ice age.

  3. Finger Cooling During Cold Air Exposure.

    NASA Astrophysics Data System (ADS)

    Tikuisis, Peter

    2004-05-01

    This paper presents a method for predicting the onset of finger freezing. It is an extension of a tissue-cooling model originally developed to predict the onset of cheek freezing. The extension to the finger is presented as a more conservative warning of wind chill. Indeed, guidance on the risk of finger freezing is important not only to safeguard the finger, but also because it pertains more closely to susceptible facial features, such as the nose, than if only the risk of cheek freezing was provided. The importance of blood flow to the finger and the modeling of vaso-constriction are demonstrated through cooling predictions that agree reasonably well with several reported observations. Differences in the prediction between the present physiologic-based model and the engineering model used to develop the wind chill index are also discussed. New wind chill charts are presented that tabulate the mean cooling rates and corresponding onset times to freezing of the finger for various combinations of air temperature and wind speed. Results indicate that the surface of the finger cools to its freezing point in approximately one-eighth of the time predicted for the cheek. For combinations that result in the same wind chill temperature (WCT), the rate of finger cooling is faster at the higher wind speed. This asymmetry was previously disclosed through the application of the model to cheek cooling, and it reiterates the ambiguity associated with the reporting of WCT. It is further emphasized that the reporting of onset times to freezing, or safe exposure limits, is a more logical and meaningful alternative to the WCT.

  4. Finger Lake Region, NY State, USA

    NASA Image and Video Library

    1992-04-02

    This view of the central portion of upstate New York, centers on the Finger Lakes. The large city on the shore of Lake Ontario, is Rochester. Although the city, being a business, educational and technical center, has no heavy industry, the outline of the city shows fairly well in the snow, but not as well as the outlines of industrial cities elsewhere in the world. The Finger Lakes are large linear lakes carved out by glaciers during the last ice age.

  5. Reverse Pressure Capable Finger Seal (Preprint)

    DTIC Science & Technology

    2012-01-01

    Currently, the typical solution for locations requiring reverse capable sealing are labyrinth seals which can exhibit significantly higher leakage...AFRL-RX-WP-TP-2012-0215 REVERSE PRESSURE CAPABLE FINGER SEAL (PREPRINT) Nathan Gibson and Joe Yanof Honeywell International, Inc...AND SUBTITLE REVERSE PRESSURE CAPABLE FINGER SEAL (PREPRINT) 5a. CONTRACT NUMBER FA8650-09-D-2925-0003 5b. GRANT NUMBER 5c. PROGRAM ELEMENT

  6. Contributions and co-ordination of individual fingers in multiple finger prehension.

    PubMed

    Kinoshita, H; Kawai, S; Ikuta, K

    1995-06-01

    The contributions and co-ordination of external finger grip forces were examined during a lifting task with a precision grip using multiple fingers. The subjects (n = 10) lifted a force transducer-equipped grip apparatus. Grip force from each of the five fingers was continuously measured under different object weight (200 g, 400 g and 800 g) and surface structure (plastic and sandpaper) conditions. The effect of five-, four-, and three-finger grip modes was also examined. It was found that variation of object weight or surface friction resulted in change of the total grip force magnitude; the largest change in finger force, was that for the index finger, followed by the middle, ring, and little fingers. Percentage contribution of static grip force to the total grip force for the index, middle, ring, and little fingers was 42.0%, 27.4%, 17.6% and 12.9%, respectively. These values were fairly constant for all object weight conditions, as well as for all surface friction conditions, suggesting that all individual finger force adjustments for light loads less than 800 g are controlled comprehensively simply by using a single common scaling value. A higher surface friction provided faster lifting initiation and required lesser grip force exertion, indicating advantageous effect of a non-slippery surface over a slippery surface. The results indicate that nearly 40% force reduction can be obtained when a non-slippery surface is used. Variation in grip mode changed the total grip force, i.e., the fewer the number of fingers, the greater the total grip force. The percent value of static grip force for the index, middle, and ring fingers in the four-finger grip mode was 42.7%, 32.5%, and 24.7%, respectively, and that for the index and middle fingers in the three-finger grip mode was 43.0% and 56.9%, respectively. Therefore, the grip mode was found to influence the force contributions of the middle and ring fingers, but not of the index finger.

  7. Improved DNA binding specificity from polyzinc finger peptides by using strings of two-finger units

    PubMed Central

    Moore, Michael; Klug, Aaron; Choo, Yen

    2001-01-01

    Multizinc finger peptides are likely to reach increased prominence in the search for the “ideal” designer transcription factor for in vivo applications such as gene therapy. However, for these treatments to be effective and safe, the peptides must bind with high affinity and, more importantly, with great specificity. Our previous research has shown that zinc finger arrays can be made to bind 18 bp of DNA with picomolar affinity, but also has suggested that arrays of fingers also may bind tightly to related sequences. This work addresses the question of zinc finger DNA binding specificity. We show that by changing the way in which zinc finger arrays are constructed—by linking three two-finger domains rather than two three-finger units—far greater target specificity can be achieved through increased discrimination against mutated or closely related sequences. These new peptides have the added capability of being able to span two short gaps of unbound DNA, although still binding with picomolar affinity to their target sites. We believe that this new method of constructing zinc finger arrays will offer greater efficacy in the fields of gene therapy and in the production of transgenic organisms than previously reported zinc finger arrays. PMID:11171969

  8. Monte Carlo simulations of tungsten redeposition at the divertor target

    NASA Astrophysics Data System (ADS)

    Chankin, A. V.; Coster, D. P.; Dux, R.

    2014-02-01

    Recent modeling of controlled edge-localized modes (ELMs) in ITER with tungsten (W) divertor target plates by the SOLPS code package predicted high electron temperatures (>100 eV) and densities (>1 × 1021 m-3) at the outer target. Under certain scenarios W sputtered during ELMs can penetrate into the core in quantities large enough to cause deterioration of the discharge performance, as was shown by coupled SOLPS5.0/STRAHL/ASTRA runs. The net sputtering yield, however, was expected to be dramatically reduced by the ‘prompt redeposition’ during the first Larmor gyration of W1+ (Fussman et al 1995 Proc. 15th Int. Conf. on Plasma Physics and Controlled Nuclear Fusion Research (Vienna: IAEA) vol 2, p 143). Under high ne/Te conditions at the target during ITER ELMs, prompt redeposition would reduce W sputtering by factor p-2 ˜ 104 (with p ≡ τionωgyro ˜ 0.01). However, this relation does not include the effects of multiple ionizations of sputtered W atoms and the electric field in the magnetic pre-sheath (MPS, or ‘Chodura sheath’) and Debye sheath (DS). Monte Carlo simulations of W redeposition with the inclusion of these effects are described in the paper. It is shown that for p ≪ 1, the inclusion of multiple W ionizations and the electric field in the MPS and DS changes the physics of W redeposition from geometrical effects of circular gyro-orbits hitting the target surface, to mainly energy considerations; the key effect is the electric potential barrier for ions trying to escape into the main plasma. The overwhelming majority of ions are drawn back to the target by a strong attracting electric field. It is also shown that the possibility of a W self-sputtering avalanche by ions circulating in the MPS can be ruled out due to the smallness of the sputtered W neutral energies, which means that they do not penetrate very far into the MPS before ionizing; thus the W ions do not gain a large kinetic energy as they are accelerated back to the surface by the

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

    NASA Astrophysics Data System (ADS)

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

    2016-02-01

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

  10. Lesson from Tungsten Leading Edge Heat Load Analysis in KSTAR Divertor

    NASA Astrophysics Data System (ADS)

    Hong, Suk-Ho; Pitts, Richard Anthony; Lee, Hyeong-Ho; Bang, Eunnam; Kang, Chan-Soo; Kim, Kyung-Min; Kim, Hong-Tack; ITER Organization Collaboration; Kstar Team Team

    2016-10-01

    An important design issue for the ITER tungsten (W) divertor and in fact for all such components using metallic plasma-facing elements and which are exposed to high parallel power fluxes, is the question of surface shaping to avoid melting of leading edges. We have fabricated a series of tungsten blocks with a variety of leading edge heights (0.3, 0.6, 1.0, and 2.0 mm), from the ITER worst case to heights even beyond the extreme value tested on JET. They are mounted into adjacent, inertially cooled graphite tile installed in the central divertor region of KSTAR, within the field of view of an infra-red (IR) thermography system with a spatial resolution to 0.4 mm/pixel. Adjustment of the outer divertor strike point position is used to deposit power on the different blocks in different discharges. The measured power flux density on flat regions of the surrounding graphite tiles is used to obtain the parallel power flux, q|| impinging on the various W blocks. Experiments have been performed in Type I ELMing H-mode with Ip = 600 kA, BT = 2 T, PNBI = 3.5 MW, leading to a hot attached divertor with typical pulse lengths of 10 s. Three dimensional ANSYS simulations using q|| and assuming geometric projection of the heat flux are found to be consistent with the observed edge loading. This research was partially supported by Ministry of Science, ICT, and Future Planning under KSTAR project.

  11. Thermal-hydraulic design issues and analysis for the ITER (International Thermonuclear Experimental Reactor) divertor

    SciTech Connect

    Koski, J.A.; Watson, R.D. ); Hassanien, A.M. ); Goranson, P.L. . Fusion Engineering Design Center); Salmonson, J.C. . Special Projects)

    1990-01-01

    Critical Heat Flux (CHF), also called burnout, is one of the major design limits for water-cooled divertors in tokamaks. Another important design issue is the correct thermal modeling of the divertor plate geometry where heat is applied to only one side of the plate and highly subcooled flow boiling in internal passages is used for heat removal. This paper discusses analytical techniques developed to address these design issues, and the experimental evidence gathered in support of the approach. Typical water-cooled divertor designs for the International Thermonuclear Experimental Reactor (ITER) are analyzed, and design margins estimated. Peaking of the heat flux at the tube-water boundary is shown to be an important issue, and design concerns which could lead to imposing large design safety margins are identified. The use of flow enhancement techniques such as internal twisted tapes and fins are discussed, and some estimates of the gains in the design margin are presented. Finally, unresolved issues and concerns regarding hydraulic design of divertors are summarized, and some experiments which could help the ITER final design process identified. 23 refs., 10 figs.

  12. Poloidal divertor experiment with applied E vector x B vector/B/sup 2/ drift

    SciTech Connect

    Strait, E J

    1980-05-01

    It has been proposed that the E vector x B vector/B/sup 2/ drift arising from an externally applied electric field could be used in a tokamak or other toroidal device to remove plasma and impurities from the region near the wall and to reduce the amount of plasma striking the wall, either assisting or replacing a conventional magnetic field divertor. A poloidal magnetic divertor (without pumping chamber) was added to the Wisconsin Levitated Toroidal Octupole, and the octupole was operated with a tokamak-like magnetic field configuration (q = 0.7). A radial electric field was applied in the scrape-off zone, causing an E vector x B vector/B/sup 2/ drift with a large poloidal component. This reduced plasma flux reaching the wall of the toroid by up to a factor of 5 beyond the effect of the magnetic divertor, for divertor configurations with both high and low magnetic mirror ratios, in good agreement with a simple theoretical model. Plasma density and density scale length were also reduced in the scrape-off zone, in qualitative agreement with the model. This was not accompanied by any new instabilities in the scrape-off zone, nor by any appreciable degradation of confinement of the central plasma.

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

    SciTech Connect

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

    2016-06-15

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

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

    SciTech Connect

    Fenstermacher, M.E.; Meyer, W.H.; Wood, R.D.

    1996-02-01

    A video camera system has been installed on the DIII-D tokamak for 2-D spatial studies of line emission in the lower divertor region. The system views the divertor tangentially from an outer port at approximately the height of the X-point. At the tangency plane the entire divertor from inner wall to outside the DIII-D bias ring is viewed with spatial resolution of approximately 1 cm. The image contains information from approximately 90 degrees of toroidal angle. In a recent upgrade, remotely controllable filter changers were added which have produced images from nominally identical shots using a series of spectral lines. Software was developed to calculate the response function matrix using distributed computing techniques and assuming toroidal symmetry. Standard sparse matrix algorithms are then used to invert the 3-D 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 shots 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 Partially Detached Divertor (PDD) operation.

  15. Outline of optical design and viewing geometry for divertor Thomson scattering on MAST upgrade

    NASA Astrophysics Data System (ADS)

    Hawke, J.; Scannell, R.; Harrison, J.; Huxford, R.; Bohm, P.

    2013-11-01

    The super-X divertor on MAST Upgrade will be diagnosed by a Thomson scattering diagnostic. A preliminary design of the collection optics and calculations of the diagnostic's performance are discussed in this paper. As part of the design the location and size of the collection cell were optimized to minimize vignetting, especially in the region of interest close to the divertor strike point. The design process was complicated by the limited access available in the closed divertor geometry. In the study of the diagnostic's performance, the radial resolution, projection of the laser image onto the fiber bundle, and impact of depth of field with a multiple laser system were investigated. In this design there is a trade-off between the resolution of the system and the lifetime of the beam dump. For this reason the beam has its focal point at the start of the viewing region and diverges in width to approximately five millimeters near the divertor tile. The effect of this large variation in beam width is examined primarily at the two extremes by means of ray trace modeling. This model takes an object with dimensions of the beam width imaged onto the fiber bundle to investigate the effect of misalignment for a narrow or broad laser image. In a similar manner ray tracing was performed to determine the effects of depth of field for four and two laser systems. As the electron density of the system may be low, performance analysis considers firing multiple lasers simultaneously to improve photon statistics.

  16. Effect of RMP spectrum on ELM suppression and the divertor plasma in KSTAR

    NASA Astrophysics Data System (ADS)

    Ahn, Joon-Wook; Park, J.-K.; in, Y.; Loarte, A.; Kim, J.; Jeon, Y. M.; Park, G. Y.; Choe, W.; Hong, J. H.; Hong, S. H.; Lee, H. H.; Kang, C. S.; Ko, W. H.; Yoon, S. W.

    2016-10-01

    ELM suppression by n =1 and n =2 magnetic perturbations have been robustly obtained in KSTAR, and effects of various coil configurations for applied magnetic perturbations (MPs) on ELM suppression as well as divertor plasma conditions have been investigated. The 4 toroidal and 3 poloidal sectors of internal coils allow to fully scan the phase difference (Δφ) of n =1 between different rows of coils, where it is shown that ideal plasma response can either shield or amplify applied MPs, depending on Δφ , which leads respectively to the weakening and strengthening of divertor footprint striations compared to the vacuum case. On the other hand, shielding is found to be the dominant plasma response for all possible cases of n =2 configuration (Δφ =0o and 90o, and mid-plane coil only), which weakens footprint striations. Spectra of applied MPs have been varied by changing Δφ as well as modifying the ratio of coil currents between different row of coils, e.g. IU/IL, in order to find optimal conditions for ELM suppression and divertor heat and particle flux dispersal. Effects of divertor conditions in various density and impurity levels on the ELM behavior and footprint striations are also being investigated. Work supported by the U.S. DOE, contract # DE-AC05-00OR22725.

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

    SciTech Connect

    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.

  18. Divertor heat load in ASDEX Upgrade L-mode in presence of external magnetic perturbation

    NASA Astrophysics Data System (ADS)

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

    2017-09-01

    Power exhaust is one of the major challenges for a future fusion device. Applying a non-axisymmetric external magnetic perturbation is one technique that is studied in order to mitigate or suppress large edge localized modes which accompany the high confinement regime in tokamaks. The external magnetic perturbation induces breaking in the axisymmetry of a tokamak and leads to a 2D heat flux pattern on the divertor target. The 2D heat flux pattern at the outer divertor target is studied on ASDEX Upgrade in stationary L-mode discharges. The amplitude of the 2D characteristic of the heat flux depends on the alignment between the field lines at the edge and the vacuum response of the applied magnetic perturbation spectrum. The 2D characteristic reduces with increasing density. The increasing divertor broadening, S, with increasing density is proposed as the main actuator. This is supported by a generic model using field line tracing and the vacuum field approach that is in quantitative agreement with the measured heat flux. The perturbed heat flux, averaged over a full toroidal rotation of the magnetic perturbation, is identical to the non-perturbed heat flux without magnetic perturbation. The transport qualifiers, power fall-off length {λ }q and divertor broadening, S, are the same within the uncertainty compared to the unperturbed reference. No additional cross field transport is observed.

  19. Density fluctuations at high density in the ergodic divertor configuration of Tore Supra

    NASA Astrophysics Data System (ADS)

    Devynck, P.; Gunn, J.; Ghendrih, Ph.; Garbet, X.; Antar, G.; Beyer, P.; Boucher, C.; Honore, C.; Gervais, F.; Hennequin, P.; Quémeneur, A.; Truc, A.

    2001-03-01

    The effect of the ergodic divertor on the plasma edge in Tore Supra is to enhance the perpendicular transport through ergodization of the magnetic field lines [Ph. Ghendrih et al., Contrib. Plasma Phys. 32 (3&4) (1992) 179]. Nevertheless, the hot spots observed on the divertor plates during ergodic divertor operation indicate that the cross-field transport driven by the fluctuations is still playing an important role, although measurements by CO 2 laser scattering and reflectometry show a decrease of the turbulence level [J. Payan, X. Garbet, J.H. Chatenet et al., Nucl. Fusion 35 (1995) 1357; P. Beyer, X. Garbet, P. Ghendrih, Phys. Plasmas 5 (12) (1998) 4271]. In order to gain more understanding, fluctuation level and poloidal velocity have been measured with a reciprocating Langmuir probe biased to collect the ion saturation current ( jsat) and with a CO 2 laser scattering diagnostic. Though the relative fluctuation level behaves as previously observed at low density, a new interesting result is that this picture is gradually modified when the density is increased. Both diagnostics observe an increase of δn/ n with density in the ergodic region, which is not the usual behavior observed in limiter configuration. This increase is detected on both sides of the Er inversion radius and is therefore also affecting the plasma bulk. Finally, the confinement time is found to follow an L-mode law at all densities indicating that the ergodic divertor does not change the global confinement properties of the plasma.

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

    NASA Astrophysics Data System (ADS)

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

    2014-02-01

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

  1. Divertor heat loads in RMP ELM controlled H-mode plasmas on DIII-D*

    SciTech Connect

    Jakubowski, M; Lasnier, C; Schmitz, O; Evans, T; Fenstermacher, M; Groth, M; Watkins, J; Eich, T; Moyer, R; Wolf, R; Baylor, L; Boedo, J; Burrell, K; Frerichs, H; deGrassie, J; Gohil, P; Joseph, I; Lehnen, M; Leonard, A; Petty, C; Pinsker, R; Reiter, D; Rhodes, T; Samm, U; Snyder, P; Stoschus, H; Osborne, T; Unterberg, B; West, W

    2008-10-13

    In this paper the manipulation of power deposition on divertor targets at DIII-D by application of resonant magnetic perturbations (RMPs) is analyzed. It has been found that heat transport shows a different reaction to the applied RMP depending on the plasma pedestal collisionality. At pedestal electron collisionality above 0.5 the heat flux during the ELM suppressed phase is of the same order as the inter-ELM in the non-RMP phase. Below this collisionality value we observe a slight increase of the total power flux to the divertor. This can be caused by much more negative potential at the divertor surface due to hot electrons reaching the divertor surface from the pedestal area and/or so called pump out effect. In the second part we discuss modification of ELM behavior due to the RMP. It is shown, that the width of the deposition pattern in ELMy H-mode depends linearly on the ELM deposited energy, whereas in the RMP phase of the discharge those patterns seem to be controlled by the externally induced magnetic perturbation. D{sub 2} pellets injected into the plasma bulk during ELM-free RMP H-mode lead in some cases to a short term small transients, which have very similar properties to ELMs in the initial RMP-on phase.

  2. Scaling of midplane separatrix density with power at divertor detachment onset

    NASA Astrophysics Data System (ADS)

    Leonard, A. W.; Makowski, M. A.; McLean, A. G.; Stangeby, P. C.

    2015-11-01

    The midplane separatrix density at divertor detachment onset is found to increase with higher parallel heat flux, q| |, flowing into the divertor, but at a slower rate than expected from simple scaling models. The separatrix density will be an important parameter in determining the compatibility of divertor heat flux control with robust pedestal operation and high core confinement in future devices. The parallel heat flux is examined by separately varying several parameters, including injected power, plasma current, toroidal field and injected impurities. Several methods are employed to locate the separatrix in this critical region of steep density gradients, including magnetic equilibrium reconstruction, power balance assumptions and spatial fiducials from other diagnostics. All methods exhibit a slower than the q|| 5 / 7 scaling predicted by a simple two point model. The nonlinear dependence of divertor radiation with power and density is one of several factors leading to this difference. Supported in part by the US DOE under DE-FC02-04ER54698 & DE-AC52-07NA27344.

  3. Development of ion source for simulation of edge localized mode in divertor plasma.

    PubMed

    Daibo, A; Okamoto, A; Takahashi, H; Kumagai, T; Takahashi, T; Tsubota, S; Kitajima, S

    2014-02-01

    A helium ion beam is injected into a linear plasma device for the development of an ion beam source simulating high energy particle flux in divertor plasma. Beam current density more than 10 mA/cm(2) is extracted. Measurement of beam currents indicates that the beam is transported along the linear device and reaches to the downstream end plate.

  4. Design of Divertor Scraper Elements for the W7-X Stellarator

    NASA Astrophysics Data System (ADS)

    Harris, Jeffrey; Lumsdaine, Arnold; Canik, John; Lore, Jeremy; McGinnis, Dean; Peacock, Alan; Hurd, Fred; Boscary, Jean; Geiger, Joachim; Tipton, Joseph

    2011-10-01

    A PPPL/ORNL/LANL team is partnering with the Max-Planck Institut für Plasmaphysik in the Wendelstein 7-X (W7-X) stellarator project. W7-X is a large superconducting, steady-state stellarator (R = 5.5, a = 0.5, B = 3T) with P =15-30 MW that will begin operation in 2015. The US team is focusing on control of the magnetic configuration and divertor heat flux. The W7-X divertor consists of cooled CFC plates arranged as a magnetic island divertor outside the last closed flux surface. While the W7-X configuration is optimized to minimize both Pfirsch-Schlüter and bootstrap currents, the ~30 sec evolution of the plasma to its final equilibrium drives bootstrap currents which transiently alter the distribution of divertor heat flux. This necessitates the addition of 10 actively cooled scraper elements (dimensions ~0.2 m x 1 m) capable of absorbing localized heat fluxes < 12 MW/m2. ORNL/IPP are developing an engineering design for the scraper elements using ITER CFC monoblock technology. Work supported by US Department of Energy.

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

    SciTech Connect

    WAMPLER,WILLIAM R.; BASTASZ,ROBERT J.; WHYTE,D.G.; WONG,C.P.C.; WEST,W.P.

    2000-05-25

    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{sub 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.

  6. Upstream Density for Plasma Detachment with Conventional and Lithium Vapor-Box Divertors

    NASA Astrophysics Data System (ADS)

    Goldston, Rj; Schwartz, Ja

    2016-10-01

    Fusion power plants are likely to require detachment of the divertor plasma from material targets. The lithium vapor box divertor is designed to achieve this, while limiting the flux of lithium vapor to the main plasma. We develop a simple model of near-detachment to evaluate the required upstream plasma density, for both conventional and lithium vapor-box divertors, based on particle and dynamic pressure balance between up- and down-stream, at near-detachment conditions. A remarkable general result is found, not just for lithium-induced detachment, that the upstream density divided by the Greenwald-limit density scales as (P 5 / 8 /B 3 / 8) Tdet1 / 2 / (ɛcool + γTdet) , with no explicit size scaling. Tdet is the temperature just before strong pressure loss, 1/2 of the ionization potential of the dominant recycling species, ɛcool is the average plasma energy lost per injected hydrogenic and impurity atom, and γ is the sheath heat transmission factor. A recent 1-D calculation agrees well with this scaling. The implication is that the plasma exhaust problem cannot be solved by increasing R. Instead significant innovation, such as the lithium vapor box divertor, will be required. This work supported by DOE Contract No. DE-AC02-09CH11466.

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

    SciTech Connect

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

    1991-01-01

    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. Achieving the criteria may require the development of new components and innovative configurations, specifically a new class of remote fasteners and electrically resistant material for mounts. The possible design of such components and an R D program to develop them are described, and issues specific to the high-aspect-ratio design (HARD) configuration are summarized. Analysis and experiments that will resolve these issues and concerns and lead to a final ITER design are identified. 2 refs., 2 figs.

  8. Viscous fingering in a microfluidic network

    NASA Astrophysics Data System (ADS)

    Budek, Agnieszka; Garstecki, Piotr; Samborski, Adam; Szymczak, Piotr

    2014-05-01

    We study experimentally and numerically two-phase flow in a rectangular network of microfluidic channels. If the pressure gradient is oriented along the lattice, growth of long and thin dendrites ('thin fingers') is promoted. The dynamics of thin finger growth is of interest due to their appearance in a variety of other pattern forming systems, such as the growth of dendrites in electrochemical deposition experiments, channeling in dissolving rocks or side-branches growth in crystallization. Due to their simplicity, thin finger models are also attractive for theoretical analysis. A characteristic feature of these systems is a strong competition between the fingers which is a reflection of Saffman-Taylor instability acting in a nonlinear regime. Surprisingly, the case of miscible fluids turns out to be different, with the competition between the fingers hindered due to the strong lateral currents of the displaced fluid, which eventually cut off the heads of the advancing fingers, thus preventing their further growth. The heads continue to move through the system, preserving their shapes, thus forming the 'miscible droplets'. In immiscible case this process is hindered by the presence of the surface tension. A detailed analysis of this phenomenon is given with a particular emphasis on the scaling properties of the system.

  9. New Finger Biometric Method Using Near Infrared Imaging

    PubMed Central

    Lee, Eui Chul; Jung, Hyunwoo; Kim, Daeyeoul

    2011-01-01

    In this paper, we propose a new finger biometric method. Infrared finger images are first captured, and then feature extraction is performed using a modified Gaussian high-pass filter through binarization, local binary pattern (LBP), and local derivative pattern (LDP) methods. Infrared finger images include the multimodal features of finger veins and finger geometries. Instead of extracting each feature using different methods, the modified Gaussian high-pass filter is fully convolved. Therefore, the extracted binary patterns of finger images include the multimodal features of veins and finger geometries. Experimental results show that the proposed method has an error rate of 0.13%. PMID:22163741

  10. Finger multibiometric cryptosystems: fusion strategy and template security

    NASA Astrophysics Data System (ADS)

    Peng, Jialiang; Li, Qiong; Abd El-Latif, Ahmed A.; Niu, Xiamu

    2014-03-01

    We address two critical issues in the design of a finger multibiometric system, i.e., fusion strategy and template security. First, three fusion strategies (feature-level, score-level, and decision-level fusions) with the corresponding template protection technique are proposed as the finger multibiometric cryptosystems to protect multiple finger biometric templates of fingerprint, finger vein, finger knuckle print, and finger shape modalities. Second, we theoretically analyze different fusion strategies for finger multibiometric cryptosystems with respect to their impact on security and recognition accuracy. Finally, the performance of finger multibiometric cryptosystems at different fusion levels is investigated on a merged finger multimodal biometric database. The comparative results suggest that the proposed finger multibiometric cryptosystem at feature-level fusion outperforms other approaches in terms of verification performance and template security.

  11. Scattering Removal for Finger-Vein Image Restoration

    PubMed Central

    Yang, Jinfeng; Zhang, Ben; Shi, Yihua

    2012-01-01

    Finger-vein recognition has received increased attention recently. However, the finger-vein images are always captured in poor quality. This certainly makes finger-vein feature representation unreliable, and further impairs the accuracy of finger-vein recognition. In this paper, we first give an analysis of the intrinsic factors causing finger-vein image degradation, and then propose a simple but effective image restoration method based on scattering removal. To give a proper description of finger-vein image degradation, a biological optical model (BOM) specific to finger-vein imaging is proposed according to the principle of light propagation in biological tissues. Based on BOM, the light scattering component is sensibly estimated and properly removed for finger-vein image restoration. Finally, experimental results demonstrate that the proposed method is powerful in enhancing the finger-vein image contrast and in improving the finger-vein image matching accuracy. PMID:22737028

  12. Scattering removal for finger-vein image restoration.

    PubMed

    Yang, Jinfeng; Zhang, Ben; Shi, Yihua

    2012-01-01

    Finger-vein recognition has received increased attention recently. However, the finger-vein images are always captured in poor quality. This certainly makes finger-vein feature representation unreliable, and further impairs the accuracy of finger-vein recognition. In this paper, we first give an analysis of the intrinsic factors causing finger-vein image degradation, and then propose a simple but effective image restoration method based on scattering removal. To give a proper description of finger-vein image degradation, a biological optical model (BOM) specific to finger-vein imaging is proposed according to the principle of light propagation in biological tissues. Based on BOM, the light scattering component is sensibly estimated and properly removed for finger-vein image restoration. Finally, experimental results demonstrate that the proposed method is powerful in enhancing the finger-vein image contrast and in improving the finger-vein image matching accuracy.

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

    SciTech Connect

    Groth, M; Porter, G D; Petrie, T W; Fenstermacher, M E; Brooks, N H

    2003-06-16

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

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

    DOE PAGES

    Wang, H. Q.; Guo, Houyang Y.; Petrie, Thomas W.; ...

    2017-03-18

    The impurity-seeded detached divertor is essential for heat exhaust in ITER and other reactor-relevant devices. Dedicated experiments with injection of N2, Ne and Ar have been performed in DIII-D to assess the impact of the different impurities on divertor detachment and confinement. Seeding with N2, Ne and Ar all promote divertor detachment, greatly reducing heat flux near the strike point. The upstream plasma density at the onset of detachment decreases with increasing impurity-puffing flow rates. For all injected impurity species, the confinement and pedestal pressure are correlated with the impurity content and the ratio of separatrix loss power to themore » L-H transition threshold power. As the divertor plasma approaches detachment, the high-Z impurity seeding tends to degrade the core confinement owing to the increased core radiation. In particular, Ar injection leads to an increase in core radiation, up to 50% of the injected power, and a reduction in pedestal temperature over 60%, thus significantly degrading the confinement, i.e., with H98 reducing from 1.1 to below 0.7. As for Ne seeding, H98 near 0.8 can be maintained during the detachment phase with the pedestal temperature being reduced by about 50%. In contrast, in the N2 seeded plasmas, radiation is predominately confined in the boundary plasma, with up to 50% of heating power being radiated in the divertor region and less than 25% in the core at the onset of detachment. In the case of strong N2 gas puffing, the confinement recovers during the detachment, from ~20% reduction at the onset of the detachment to greater than that before the seeding. The core and pedestal temperatures feature a reduction of 30% from the initial attached phase and remain nearly constant during the detachment phase. The improvement in confinement appears to arise from the increase in pedestal and core density despite the temperature reduction.« less

  15. Utilization of vanadium alloys in the DIII-D Radiative Divertor Program

    SciTech Connect

    Smith, J.P.; Johnson, W.R.; Stambaugh, R.D.; Trester, P.W.; Smith, D.; Bloom, E.

    1995-10-01

    Vanadium alloys are attractive candidate structural materials for fusion power plants because of their potential for minimum environmental impact due to low neutron activation and rapid activation decay. They also possess favorable material properties for operation in a fusion environment. General Atomics (GA), in conjunction with Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL), has developed a plan for the utilization of vanadium alloys as part of the Radiative Divertor (RD) upgrade for the DIII-D tokamak. The plan will be carried out in conjunction with General Atomics and the Materials Program of the US Department of Energy (DOE). This application of a vanadium alloy will provide a meaningful step in the development of advanced materials for fusion power devices by: (1) developing necessary materials processing technology for the fabrication of large vanadium alloy components, and (2) demonstrating the in-service behavior of a vanadium alloy (V-4Cr-4Ti) in a tokamak environment. The program consists of three phases: first, small vanadium alloy coupon samples will be exposed in DIII-D at positions in the vessel floor and within the pumping plenum region of the existing divertor structure; second, a small vanadium alloy component will be installed in the existing divertor, and third, during the forthcoming Radiative Divertor modification, scheduled for completion in mid-1997, the upper section of the new double-null, slotted divertor will be fabricated from vanadium alloy product forms. This program also includes research and development (R and D) efforts to support fabrication development and to resolve key issues related to environmental effects.

  16. Assessment of erosion and surface tritium inventory issues for the ITER divertor

    SciTech Connect

    Brooks, J.N.; Causey, R.; Federici, G.; Ruzic, D.N.

    1996-08-01

    The authors analyzed sputtering erosion and tritium codeposition for the ITER vertical target divertor design using erosion and plasma codes (WBC/REDEP/DEGAS+) coupled to available materials data. Computations were made for a beryllium, carbon, and tungsten coated divertor plate, and for three edged plasma regimes. New data on tritium codeposition in beryllium was obtained with the TPE facility. This shows codeposited H/Be ratios of the order of 10% for surface temperatures {le} 300 C, beryllium thereby being similar to carbon in this respect. Hydrocarbon transport calculations show significant loss (10--20%) of chemically sputtered carbon for detached conditions (T{sub e} {approx} 1 eV at the divertor), compared to essentially no loss (100% redeposition) for higher temperature plasmas. Calculations also show a high, non-thermal, D-T molecular flux for detached conditions. Tritium codeposition rates for carbon are very high for detached conditions ({approximately} 20g-T/1000 s discharge), due to buildup of chemically sputtered carbon on relatively cold surfaces of the divertor cassette. Codeposition is lower ({approximately} 10X) for higher edge temperatures ({approximately} 8--30 eV) and is primarily due to divertor plate buildup of physically sputtered carbon. Peak net erosion rates for carbon are of order 30 cm/burn-yr. Erosion and codeposition rates for beryllium are much lower than for carbon at detached conditions, but are similar to carbon for the higher temperatures. Both erosion and tritium codeposition are essentially nil for tungsten for the regimes studied.

  17. Effect of changes in separatrix magnetic geometry on divertor behaviour in DIII-D

    NASA Astrophysics Data System (ADS)

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

    2013-11-01

    Results and interpretation of recent experiments on DIII-D designed to evaluate divertor geometries favourable for radiative heat dispersal are presented. Two approaches examined here involved lengthening the parallel connection in the scrape-off layer, L‖, and increasing the radius of the outer divertor separatrix strike point, ROSP, with the goal of reducing target temperature, TTAR, and increasing target density, nTAR. From one-dimensional (1D) two-point modelling based on conducted parallel heat flux, it is expected that: n_{TAR} \\propto R_{OSP}^{2} L_{\\parallel}^{6/7} n_{SEP}^{3} and T_{TAR} \\propto R_{OSP}^{-2} L_{\\parallel}^{{-4}/7} n_{SEP}^{-2} , where nSEP is the midplane separatrix density. These scalings suggest that conditions conducive to a radiative divertor solution can be achieved at low nSEP by increasing either ROSP or L‖. Our data are consistent with the above L‖ scalings. On the other hand, the observed dependence of nTAR and TTAR on ROSP displayed a more complex behaviour, under certain conditions deviating from the above scalings. Our analysis indicates that deviations from the ROSP scaling were due to the presence of convected heat flux, driven by escaping neutrals, in the more open configurations of the larger ROSP cases. A comparison of ‘open’ versus ‘closed’ divertor configurations for the H-mode plasmas in this study show that the ‘closed’ case provides at least 30% reduction in the peaked heat flux at common density with the ‘open’ case and partial divertor detachment at lower plasma density.

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

    DOE PAGES

    Briesemeister, Alexis R.; Ahn, Joon -Wook; Canik, John M.; ...

    2017-06-07

    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 indicated non-axisymmetric magnetic field perturbations, created using in-vessel control coils, which have components which has at least one but typically many resonances with the rotational transform of the plasma. RMPs are found to alter inter-ELM heat flux to the divertormore » 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. Furthermore, 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

  19. Rehabilitation of single finger amputation with customized silicone prosthesis

    PubMed Central

    Yadav, Niharika; Chand, Pooran; Jurel, Sunit Kumar

    2016-01-01

    Finger amputations are common in accidents at home, work, and play. Apart from trauma, congenital disease and deformity also leads to finger amputation. This results in loss of function, loss of sensation as well as loss of body image. Finger prosthesis offers psychological support and social acceptance in such cases. This clinical report describes a method to fabricate ring retained silicone finger prosthesis in a patient with partial finger loss. PMID:28163487

  20. Recent Progress in the NSTX/NSTX-U Lithium Program and Prospects for Reactor-Relevant Liquid-Lithium Based Divertor Development

    SciTech Connect

    M. Ono, et al.

    2012-10-27

    Developing a reactor compatible divertor has been identified as a particularly challenging technology problem for magnetic confinement fusion. While tungsten has been identified as the most attractive solid divertor material, the NSTX/NSTX-U lithium (Li) program is investigating the viability of liquid lithium (LL) as a potential reactor compatible divertor plasma facing component (PFC) . In the near term, operation in NSTX-U is projected to provide reactor-like divertor heat loads < 40 MW/m^2 for 5 s. During the most recent NSTX campaign, ~ 0.85 kg of Li was evaporated onto the NSTX PFCs where a ~50% reduction in heat load on the Liquid Lithium Divertor (LLD) was observed, attributable to enhanced divertor bolometric radiation. This reduced divertor heat flux through radiation observed in the NSTX LLD experiment is consistent with the results from other lithium experiments and calculations. These results motivate an LL-based closed radiative divertor concept proposed here for NSTX-U and fusion reactors. With an LL coating, the Li is evaporated from the divertor strike point surface due to the intense heat. The evaporated Li is readily ionized by the plasma due to its low ionization energies, and the ionized Li ions can radiate strongly, resulting in a significant reduction in the divertor heat flux. Due to the rapid plasma transport in divertor plasma, the radiation values can be significantly enhanced up to ~ 11 MJ/cc of LL. This radiative process has the desired function of spreading the focused divertor heat load to the entire divertor chamber facilitating the divertor heat removal. The LL divertor surface can also provide a "sacrificial" surface to protect the substrate solid material from transient high heat flux such as the ones caused by the ELMs. The closed radiative LLD concept has the advantages of providing some degree of partition in terms of plasma disruption forces on the LL, Li particle divertor retention, and strong divertor pumping action from the