WorldWideScience

Sample records for nstx graphite divertor

  1. Divertor scenario development for NSTX Upgrade

    Science.gov (United States)

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

    2012-10-01

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

  2. NSTX Tangential Divertor Camera

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  3. Snowflake Divertor Configuration in NSTX

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  4. 'Snowflake' divertor configuration in NSTX

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  5. "Snowflake" divertor configuration in NSTX

    Science.gov (United States)

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

    2011-08-01

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

  6. NSTX Plasma Response to Lithium Coated Divertor

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-01-21

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

  7. NSTX plasma response to lithium coated divertor

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  8. Transition to ELM-free Improved H-mode by Lithium Deposition on NSTX Graphite Divertor Surfaces

    International Nuclear Information System (INIS)

    Mansfield, D.K.; Kugel, H.W.; Maingi, R.; Bell, M.G.; Bell, R.; Kaita, R.; Kallman, J.; Kaye, S.; LeBlanc, B.; Mueller, D.; Paul, S.; Raman, R.; Roquemore, L.; Sabbagh, S.; Schneider, H.; Skinner, C.H.; Soukhanovskii, V.; Timberlake, J.; Wilgen, J.; Zakharov, L.

    2009-01-01

    Lithium evaporated onto plasma facing components in the NSTX lower divertor has made dramatic improvements in discharge performance. As lithium accumulated, plasmas previously exhibiting robust Type 1 ELMs gradually transformed into discharges with intermittent ELMs and finally into continuously evolving ELM-free discharges. During this sequence, other discharge parameters changed in a complicated manner. As the ELMs disappeared, energy confinement improved and remarkable changes in edge and scrape-off layer plasma properties were observed. These results demonstrate that active modification of plasma surface interactions can preempt large ELMs.

  9. Snowflake divertor configuration studies for NSTX-Upgrade

    International Nuclear Information System (INIS)

    Soukhanovskii, V.A.

    2011-01-01

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

  10. Mechanical Design of the NSTX Liquid Lithium Divertor

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. Mechanical Design of the NSTX Liquid Lithium Divertor

    International Nuclear Information System (INIS)

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

    2009-01-01

    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, vacuum compatible 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.

  12. NSTX plasma operation with a Liquid Lithium Divertor

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  13. Modeling detachment physics in the NSTX snowflake divertor

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

  14. Operational Characteristics of Liquid Lithium Divertor in NSTX

    Science.gov (United States)

    Kaita, R.; Kugel, H.; Abrams, T.; Bell, M. G.; Bell, R. E.; Gerhardt, S.; Jaworski, M. A.; Kallman, J.; Leblanc, B.; Mansfield, D.; Mueller, D.; Paul, S.; Roquemore, A. L.; Scotti, F.; Skinner, C. H.; Timberlake, J.; Zakharov, L.; Maingi, R.; Nygren, R.; Raman, R.; Sabbagh, S.; Soukhanovskii, V.

    2010-11-01

    Lithium coatings on plasma-facing components (PFC's) have resulted in improved plasma performance on NSTX in deuterium H-mode plasmas with neutral beam heating.^ Salient results included improved electron confinement and ELM suppression. In CDX-U, the use of lithium-coated PFC's and a large-area liquid lithium limiter resulted in a six-fold increase in global energy confinement time. A Liquid Lithium Divertor (LLD) has been installed in NSTX for the 2010 run campaign. The LLD PFC consists of a thin film of lithium on a temperature-controlled substrate to keep the lithium liquefied between shots, and handle heat loads during plasmas. This capability was demonstrated when the LLD withstood a strike point on its surface during discharges with up to 4 MW of neutral beam heating.

  15. Toroidal asymmetries in divertor impurity influxes in NSTX

    Directory of Open Access Journals (Sweden)

    F. Scotti

    2017-08-01

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

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

    Science.gov (United States)

    Chen, Zhongping; Kotschenreuther, Mike; Mahajan, Swadesh

    2017-10-01

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

  17. Surface chemistry analysis of lithium conditioned NSTX graphite tiles correlated to plasma performance

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, C.N., E-mail: chase.taylor@inl.gov [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47906 (United States); Birck Nanotechnology Center, Discovery Park, West Lafayette, IN 47907 (United States); Luitjohan, K.E. [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47906 (United States); Heim, B. [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47906 (United States); Birck Nanotechnology Center, Discovery Park, West Lafayette, IN 47907 (United States); Kollar, L. [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47906 (United States); Allain, J.P. [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47906 (United States); Birck Nanotechnology Center, Discovery Park, West Lafayette, IN 47907 (United States); Skinner, C.H.; Kugel, H.W.; Kaita, R.; Roquemore, A.L. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Maingi, R. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

    2013-12-15

    Lithium wall conditioning in NSTX has resulted in reduced divertor recycling, improved energy confinement, and reduced frequency of edge-localized modes (ELMs), up to the point of complete ELM suppression. NSTX tiles were removed from the vessel following the 2008 campaign and subsequently analyzed using X-ray photoelectron spectroscopy as well as nuclear reaction ion beam analysis. In this paper we relate surface chemistry to deuterium retention/recycling, develop methods for cleaning of passivated NSTX tiles, and explore a method to effectively extract bound deuterium from lithiated graphite. Li–O–D and Li–C–D complexes characteristic of deuterium retention that form during NSTX operations are revealed by sputter cleaning and heating. Heating to ∼850 °C desorbed all deuterium complexes observed in the O 1s and C 1s photoelectron energy ranges. Tile locations within approximately ±2.5 cm of the lower vertical/horizontal divertor corner appear to have unused Li-O bonds that are not saturated with deuterium, whereas locations immediately outboard of this region indicate high deuterium recycling. X-ray photo electron spectra of a specific NSTX tile with wide ranging lithium coverage indicate that a minimum lithium dose, 100–500 nm equivalent thickness, is required for effective deuterium retention. This threshold is suspected to be highly sensitive to surface morphology. The present analysis may explain why plasma discharges in NSTX continue to benefit from lithium coating thickness beyond the divertor deuterium ion implantation depth, which is nominally <10 nm.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-11-15

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

  19. Characterization of fueling NSTX H-mode plasmas diverted to a liquid lithium divertor

    Energy Technology Data Exchange (ETDEWEB)

    Kaita, R., E-mail: kaita@pppl.gov [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Kugel, H.W.; Abrams, T. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Allain, J.P. [Purdue University, West Lafayette, IN 47907 (United States); Bell, M.G.; Bell, R.E.; Diallo, A.; Gerhardt, S.P. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Heim, B. [Purdue University, West Lafayette, IN 47907 (United States); Jaworski, M.A., E-mail: mjaworsk@pppl.gov [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Kallman, J. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Kaye, S.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Maingi, R. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Mansfield, D. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); McLean, A. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Menard, J.; Mueller, D. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Nygren, R. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Ono, M. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); and others

    2013-07-15

    Deuterium fueling experiments were conducted with the NSTX Liquid Lithium Divertor (LLD). Lithium evaporation recoated the LLD surface to approximate flowing liquid Li to sustain D retention. In the first experiment with the diverted outer strike point on the LLD, the difference between the applied D gas input and the plasma D content reached very high values without disrupting the plasma, as would normally occur in the absence of Li pumping, and there was also little change in plasma D content. In the second experiment, constant fueling was applied, as the LLD temperature was varied to change the surface from solid to liquid. The D retention was relatively constant, and about the same as that for solid Li coatings on graphite, or twice that achieved without Li PFC coatings. Contamination of the LLD surface was also possible due to compound formation and erosion and redeposition from carbon PFCs.

  20. Diagnostics for Evaluating Performance of NSTX Liquid Lihium Divertor

    Science.gov (United States)

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

    2009-11-01

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

  1. Response of NSTX liquid lithium divertor to high heat loads

    Energy Technology Data Exchange (ETDEWEB)

    Abrams, T., E-mail: tabrams@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Jaworski, M.A. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Kallman, J. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Foley, E.L. [Nova Photonics, Inc., Princeton, NJ 08543 (United States); Gray, T.K. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Kugel, H. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Levinton, F. [Nova Photonics, Inc., Princeton, NJ 08543 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

    2013-07-15

    Samples of the NSTX Liquid Lithium Divertor (LLD) with and without an evaporative Li coating were directly exposed to a neutral beam ex-situ at a power of ∼1.5 MW/m{sup 2} for 1–3 s. Measurements of front face and bulk sample temperature were obtained. Predictions of temperature evolution were derived from a 1D heat flux model. No macroscopic damage occurred when the “bare” sample was exposed to the beam but microscopic changes to the surface were observed. The Li-coated sample developed a lithium hydroxide (LiOH) coating, which did not change even when the front face temperature exceeded the pure Li melting point. These results are consistent with the lack of damage to the LLD surface and imply that heating alone may not expose pure liquid Li if the melting point of surface impurities is not exceeded. This suggests that flow and heat are needed for future PFCs requiring a liquid Li surface.

  2. Measurement and modeling of surface temperature dynamics of the NSTX liquid lithium divertor

    Science.gov (United States)

    McLean, A. G.; Gan, K. F.; Ahn, J.-W.; Gray, T. K.; Maingi, R.; Abrams, T.; Jaworski, M. A.; Kaita, R.; Kugel, H. W.; Nygren, R. E.; Skinner, C. H.; Soukhanovskii, V. A.

    2013-07-01

    Dual-band infrared (IR) measurements of the National Spherical Torus eXperiment (NSTX) Liquid Lithium Divertor (LLD) are reported that demonstrate liquid Li is more effective at removing plasma heat flux than Li-conditioned graphite. Extended dwell of the outer strike point (OSP) on the LLD caused an incrementally larger area to be heated above the Li melting point through the discharge leading to enhanced D retention and plasma confinement. Measurement of Tsurface near the OSP demonstrates a significant reduction of the LLD surface temperature compared to that of Li-coated graphite at the same major radius. Modeling of these data with a 2-D simulation of the LLD structure in the DFLUX code suggests that the structure of the LLD was successful at handling up to q⊥,peak = 5 MW/m2 inter-ELM and up to 10 MW/m2 during ELMs from its plasma-facing surface as intended, and provide an innovative method for inferring the Li layer thickness.

  3. Measurement and modeling of surface temperature dynamics of the NSTX liquid lithium divertor

    Energy Technology Data Exchange (ETDEWEB)

    McLean, A.G., E-mail: mclean@fusion.gat.com [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Gan, K.F. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Ahn, J.-W.; Gray, T.K.; Maingi, R. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Abrams, T.; Jaworski, M.A.; Kaita, R.; Kugel, H.W. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Nygren, R.E. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States)

    2013-07-15

    Dual-band infrared (IR) measurements of the National Spherical Torus eXperiment (NSTX) Liquid Lithium Divertor (LLD) are reported that demonstrate liquid Li is more effective at removing plasma heat flux than Li-conditioned graphite. Extended dwell of the outer strike point (OSP) on the LLD caused an incrementally larger area to be heated above the Li melting point through the discharge leading to enhanced D retention and plasma confinement. Measurement of T{sub surface} near the OSP demonstrates a significant reduction of the LLD surface temperature compared to that of Li-coated graphite at the same major radius. Modeling of these data with a 2-D simulation of the LLD structure in the DFLUX code suggests that the structure of the LLD was successful at handling up to q{sub ⊥,peak} = 5 MW/m{sup 2} inter-ELM and up to 10 MW/m{sup 2} during ELMs from its plasma-facing surface as intended, and provide an innovative method for inferring the Li layer thickness.

  4. Impact of ELM filaments on divertor heat flux dynamics in NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, J.-W., E-mail: jahn@pppl.gov [Oak Ridge National Laboratory, Oak Ridge (United States); Maingi, R. [Princeton Plasma Physics Laboratory, Princeton (United States); Canik, J.M. [Oak Ridge National Laboratory, Oak Ridge (United States); Gan, K.F. [Institute of Plasma Physics, Chinese Academy of Science, Hefei (China); Gray, T.K. [Oak Ridge National Laboratory, Oak Ridge (United States); McLean, A.G. [Lawrence Livermore National Laboratory, Livermore (United States)

    2015-08-15

    The ELM induced change in wetted area (A{sub wet}) and peak heat flux (q{sub peak}) of divertor heat flux is investigated as a function of the number of striations, which represent ELM filaments, observed in the heat flux profile in NSTX. More striations are found to lead to larger A{sub wet} and lower q{sub peak}. The typical number of striations observed in NSTX is 0–9, while 10–15 striations are normally observed in other machines such as JET, and the ELM contracts heat flux profile when the number of striations is less than 3–4 but broadens it with more of them. The smaller number of striations in NSTX is attributed to the fact that NSTX ELMs are against kink/peeling boundary with lower toroidal mode number (n = 1–5), while typical peeling–ballooning ELMs have higher mode number of n = 10–20. For ELMs with smaller number of striations, relative A{sub wet} change is rather constant and q{sub peak} change rapidly increases with increasing ELM size, while A{sub wet} change slightly increases leading to a weaker increase of q{sub peak} change for ELMs with larger number of striations, both of which are unfavourable trend for the material integrity of divertor tiles.

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

    Science.gov (United States)

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

    2018-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Eden, G. G. van; Morgan, T. W. [Dutch Institute for Fundamental Energy Research, 5612 AJ Eindhoven (Netherlands); Reinke, M. L.; Gray, T. K.; Lore, J. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Peterson, B. J.; Mukai, K. [National Institute for Fusion Science, Toki 509-5292 (Japan); Delgado-Aparicio, L. F.; Jaworski, M. A. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, New Jersey 08543 (United States); Sano, R. [National Institutes for Quantum and Radiological Science and Technology, Naka 311-0193 (Japan); Pandya, S. N. [Institute for Plasma Research, Bhat Village, Gandhinagar, 382428 Gujarat (India)

    2016-11-15

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  8. Advanced divertor configurations with large flux expansion

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-15

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

  9. NSTX Disruption Simulations of Detailed Divertor and Passive Plate Models by Vector Potential Transfer from OPERA Global Analysis Results

    International Nuclear Information System (INIS)

    Titus, P.H.; Avasaralla, S.; Brooks, A.; Hatcher, R.

    2010-01-01

    The National Spherical Torus Experiment (NSTX) project is planning upgrades to the toroidal field, plasma current and pulse length. This involves the replacement of the center-stack, including the inner legs of the TF, OH, and inner PF coils. A second neutral beam will also be added. The increased performance of the upgrade requires qualification of the remaining components including the vessel, passive plates, and divertor for higher disruption loads. The hardware needing qualification is more complex than is typically accessible by large scale electromagnetic (EM) simulations of the plasma disruptions. The usual method is to include simplified representations of components in the large EM models and attempt to extract forces to apply to more detailed models. This paper describes a more efficient approach of combining comprehensive modeling of the plasma and tokamak conducting structures, using the 2D OPERA code, with much more detailed treatment of individual components using ANSYS electromagnetic (EM) and mechanical analysis. This capture local eddy currents and resulting loads in complex details, and allows efficient non-linear, and dynamic structural analyses.

  10. NSTX Disruption Simulations of Detailed Divertor and Passive Plate Models by Vector Potential Transfer from OPERA Global Analysis Results

    Energy Technology Data Exchange (ETDEWEB)

    P. H. Titus, S. Avasaralla, A.Brooks, R. Hatcher

    2010-09-22

    The National Spherical Torus Experiment (NSTX) project is planning upgrades to the toroidal field, plasma current and pulse length. This involves the replacement of the center-stack, including the inner legs of the TF, OH, and inner PF coils. A second neutral beam will also be added. The increased performance of the upgrade requires qualification of the remaining components including the vessel, passive plates, and divertor for higher disruption loads. The hardware needing qualification is more complex than is typically accessible by large scale electromagnetic (EM) simulations of the plasma disruptions. The usual method is to include simplified representations of components in the large EM models and attempt to extract forces to apply to more detailed models. This paper describes a more efficient approach of combining comprehensive modeling of the plasma and tokamak conducting structures, using the 2D OPERA code, with much more detailed treatment of individual components using ANSYS electromagnetic (EM) and mechanical analysis. This capture local eddy currents and resulting loads in complex details, and allows efficient non-linear, and dynamic structural analyses.

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  12. Lithium Surface Coatings for Improved Plasma Performance in NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Kugel, H W; Ahn, J -W; Allain, J P; Bell, R; Boedo, J; Bush, C; Gates, D; Gray, T; Kaye, S; Kaita, R; LeBlanc, B; Maingi, R; Majeski, R; Mansfield, D; Menard, J; Mueller, D; Ono, M; Paul, S; Raman, R; Roquemore, A L; Ross, P W; Sabbagh, S; Schneider, H; Skinner, C H; Soukhanovskii, V; Stevenson, T; Timberlake, J; Wampler, W R

    2008-02-19

    NSTX high-power divertor plasma experiments have shown, for the first time, significant and frequent benefits from lithium coatings applied to plasma facing components. Lithium pellet injection on NSTX introduced lithium pellets with masses 1 to 5 mg via He discharges. Lithium coatings have also been applied with an oven that directed a collimated stream of lithium vapor toward the graphite tiles of the lower center stack and divertor. Lithium depositions from a few mg to 1 g have been applied between discharges. Benefits from the lithium coating were sometimes, but not always seen. These improvements sometimes included decreases plasma density, inductive flux consumption, and ELM frequency, and increases in electron temperature, ion temperature, energy confinement and periods of MHD quiescence. In addition, reductions in lower divertor D, C, and O luminosity were measured.

  13. Status of National Spherical Torus Experiment Liquid Lithium Divertor

    Science.gov (United States)

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

    2009-11-01

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

  14. Infrared Camera Diagnostic for Heat Flux Measurements on NSTX

    International Nuclear Information System (INIS)

    D. Mastrovito; R. Maingi; H.W. Kugel; A.L. Roquemore

    2003-01-01

    An infrared imaging system has been installed on NSTX (National Spherical Torus Experiment) at the Princeton Plasma Physics Laboratory to measure the surface temperatures on the lower divertor and center stack. The imaging system is based on an Indigo Alpha 160 x 128 microbolometer camera with 12 bits/pixel operating in the 7-13 (micro)m range with a 30 Hz frame rate and a dynamic temperature range of 0-700 degrees C. From these data and knowledge of graphite thermal properties, the heat flux is derived with a classic one-dimensional conduction model. Preliminary results of heat flux scaling are reported

  15. NSTX High Temperature Sensor Systems

    International Nuclear Information System (INIS)

    McCormack, B.; Kugel, H.W.; Goranson, P.; Kaita, R.

    1999-01-01

    The design of the more than 300 in-vessel sensor systems for the National Spherical Torus Experiment (NSTX) has encountered several challenging fusion reactor diagnostic issues involving high temperatures and space constraints. This has resulted in unique miniature, high temperature in-vessel sensor systems mounted in small spaces behind plasma facing armor tiles, and they are prototypical of possible high power reactor first-wall applications. In the Center Stack, Divertor, Passive Plate, and vessel wall regions, the small magnetic sensors, large magnetic sensors, flux loops, Rogowski Coils, thermocouples, and Langmuir Probes are qualified for 600 degrees C operation. This rating will accommodate both peak rear-face graphite tile temperatures during operations and the 350 degrees C bake-out conditions. Similar sensor systems including flux loops, on other vacuum vessel regions are qualified for 350 degrees C operation. Cabling from the sensors embedded in the graphite tiles follows narrow routes to exit the vessel. The detailed sensor design and installation methods of these diagnostic systems developed for high-powered ST operation are discussed

  16. Studies of lithiumization and boronization of ATJ graphite PFCs for NSTX-U

    Science.gov (United States)

    Dominguez, Javier; Bedoya, Felipe; Krstic, Predrag; Allain, Jean Paul; Neff, Anton; Luitjohan, Kara

    2016-10-01

    We examine and compare the effects of boron and lithium conditioning on ATJ graphite surfaces bombarded by low-energy deuterium atoms on deuterium retention and chemical sputtering. We use atomistic simulations and compare them with experimental in-situ ex-tempore studies with X-ray photoelectron spectroscopy (XPS), to understand the effects of deuterium exposure on the chemistry in lithiated, boronized and oxidized amorphous carbon surfaces. Our results are validated qualitatively by comparison with experiments and with classical-quantum molecular dynamic simulations. We explain the important role of oxygen in D retention for lithiated surfaces and the suppression of the oxygen role by boron in boronized surfaces. The calculated increase of the oxygen role in deuterium uptake after D accumulation in a B-C-O surface configuration is discussed. The sputtering yield per low-energy D impact is significantly smaller in boronized surfaces than in lithiated surfaces. This work was supported by the USDOE Grants DE-SC0013752 (PSK), DE-SC0010717 (JPA and FB) and DE-SC0010719 (AN) and by National council for Science and Technology of Mexico (CONACyT) through postdoctoral fellowship # 267898 (JD).

  17. Lithium Coatings on NSTX Plasma Facing Components and Its Effects On Boundary Control, Core Plasma Performance, and Operation

    Energy Technology Data Exchange (ETDEWEB)

    H.W.Kugel, M.G.Bell, H.Schneider, J.P.Allain, R.E.Bell, R Kaita, J.Kallman, S. Kaye, B.P. LeBlanc, D. Mansfield, R.E. Nygen, R. Maingi, J. Menard, D. Mueller, M. Ono, S. Paul, S.Gerhardt, R.Raman, S.Sabbagh, C.H.Skinner, V.Soukhanovskii, J.Timberlake, L.E.Zakharov, and the NSTX Research Team

    2010-01-25

    NSTX high-power divertor plasma experiments have used in succession lithium pellet injection (LPI), evaporated lithium, and injected lithium powder to apply lithium coatings to graphite plasma facing components. In 2005, following wall conditioning and LPI, discharges exhibited edge density reduction and performance improvements. Since 2006, first one, and now two lithium evaporators have been used routinely to evaporate lithium onto the lower divertor region at total rates of 10-70 mg/min for periods 5-10 min between discharges. Prior to each discharge, the evaporators are withdrawn behind shutters. Significant improvements in the performance of NBI heated divertor discharges resulting from these lithium depositions were observed. These evaporators are now used for more than 80% of NSTX discharges. Initial work with injecting fine lithium powder into the edge of NBI heated deuterium discharges yielded comparable changes in performance. Several operational issues encountered with lithium wall conditions, and the special procedures needed for vessel entry are discussed. The next step in this work is installation of a Liquid Lithium Divertor surface on the outer part of the lower divertor.

  18. Lithium coatings on NSTX plasma facing components and its effects on boundary control, core plasma performance, and operation

    Energy Technology Data Exchange (ETDEWEB)

    Kugel, H.W., E-mail: hkugel@pppl.gov [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Bell, M.G.; Schneider, H. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Allain, J.P. [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47907 (United States); Bell, R.E.; Kaita, R.; Kallman, J.; Kaye, S.; LeBlanc, B.P.; Mansfield, D. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Nygren, R.E. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Maingi, R. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Menard, J.; Mueller, D.; Ono, M.; Paul, S.; Gerhardt, S. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Raman, R. [University of Washington, Seattle, WA 98195 (United States); Sabbagh, S. [Columbia University, New York, NY 10027 (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States)

    2010-11-15

    NSTX high power divertor plasma experiments have used in succession lithium pellet injection (LPI), evaporated lithium, and injected lithium powder to apply lithium coatings to graphite plasma facing components. In 2005, following the wall conditioning and LPI, discharges exhibited edge density reduction and performance improvements. Since 2006, first one, and now two lithium evaporators have been used routinely to evaporate lithium onto the lower divertor region at total rates of 10-70 mg/min for periods 5-10 min between discharges. Prior to each discharge, the evaporators are withdrawn behind shutters. Significant improvements in the performance of NBI heated divertor discharges resulting from these lithium depositions were observed. These evaporators are now used for more than 80% of NSTX discharges. Initial work with injecting fine lithium powder into the edge of NBI heated deuterium discharges yielded comparable changes in performance. Several operational issues encountered with lithium wall conditions, and the special procedures needed for vessel entry are discussed. The next step in this work is installation of a liquid lithium divertor surface on the outer part of the lower divertor.

  19. Lithium Coatings on NSTX Plasma Facing Components and Its Effects On Boundary Control, Core Plasma Performance, and Operation

    International Nuclear Information System (INIS)

    Kugel, H.W.; Bell, M.G.; Schneider, H.; Allain, J.P.; Bell, R.E.; Kaita, R.; Kallman, J.; Kaye, S.; LeBlanc, B.P.; Mansfield, D.; Nygen, R.E.; Maingi, R.; Menard, J.; Mueller, D.; Ono, M.; Paul, S.; Gerhardt, S.; Raman, R.; Sabbagh, S.; Skinner, C.H.; Soukhanovskii, V.; Timberlake, J.; Zakharov, L.E.; NSTX Research Team

    2010-01-01

    NSTX high-power divertor plasma experiments have used in succession lithium pellet injection (LPI), evaporated lithium, and injected lithium powder to apply lithium coatings to graphite plasma facing components. In 2005, following wall conditioning and LPI, discharges exhibited edge density reduction and performance improvements. Since 2006, first one, and now two lithium evaporators have been used routinely to evaporate lithium onto the lower divertor region at total rates of 10-70 mg/min for periods 5-10 min between discharges. Prior to each discharge, the evaporators are withdrawn behind shutters. Significant improvements in the performance of NBI heated divertor discharges resulting from these lithium depositions were observed. These evaporators are now used for more than 80% of NSTX discharges. Initial work with injecting fine lithium powder into the edge of NBI heated deuterium discharges yielded comparable changes in performance. Several operational issues encountered with lithium wall conditions, and the special procedures needed for vessel entry are discussed. The next step in this work is installation of a Liquid Lithium Divertor surface on the outer part of the lower divertor.

  20. An In-situ materials analysis particle probe (MAPP) diagnostic to study particle density control and hydrogenic fuel retention in NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Allain, Jean-Paul [Purdue Univ., West Lafayette, IN (United States)

    2014-09-05

    A new materials analysis particle probe (MAPP) was designed, constructed and tested to develop understanding of particle control and hydrogenic fuel retention in lithium-based plasma-facing surfaces in NSTX. The novel feature of MAPP is an in-situ tool to probe the divertor NSTX floor during LLD and lithium-coating shots with subsequent transport to a post-exposure in-vacuo surface analysis chamber to measure D retention. In addition, the implications of a lithiated graphite-dominated plasma-surface environment in NSTX on LLD performance, operation and ultimately hydrogenic pumping and particle control capability are investigated in this proposal. MAPP will be an invaluable tool for erosion/redeposition simulation code validation.

  1. Impurity diagnosis of a KSTAR graphite divertor tile using laser induced breakdown spectroscopy technique

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minju; Cho, Min Sang; Cho, Byoung Ick, E-mail: bicho@gist.ac.kr

    2017-04-15

    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.

  2. An overview of recent physics results from NSTX

    Science.gov (United States)

    Kaye, S. M.; Abrams, T.; Ahn, J.-W.; Allain, J. P.; Andre, R.; Andruczyk, D.; Barchfeld, R.; Battaglia, D.; Bhattacharjee, A.; Bedoya, F.; Bell, R. E.; Belova, E.; Berkery, J.; Berry, L.; Bertelli, N.; Beiersdorfer, P.; Bialek, J.; Bilato, R.; Boedo, J.; Bonoli, P.; Boozer, A.; Bortolon, A.; Boyer, M. D.; Boyle, D.; Brennan, D.; Breslau, J.; Brooks, J.; Buttery, R.; Capece, A.; Canik, J.; Chang, C. S.; Crocker, N.; Darrow, D.; Davis, W.; Delgado-Aparicio, L.; Diallo, A.; D'Ippolito, D.; Domier, C.; Ebrahimi, F.; Ethier, S.; Evans, T.; Ferraro, N.; Ferron, J.; Finkenthal, M.; Fonck, R.; Fredrickson, E.; Fu, G. Y.; Gates, D.; Gerhardt, S.; Glasser, A.; Gorelenkov, N.; Gorelenkova, M.; Goumiri, I.; Gray, T.; Green, D.; Guttenfelder, W.; Harvey, R.; Hassanein, A.; Heidbrink, W.; Hirooka, Y.; Hooper, E. B.; Hosea, J.; Humphreys, D.; Jaeger, E. F.; Jarboe, T.; Jardin, S.; Jaworski, M. A.; Kaita, R.; Kessel, C.; Kim, K.; Koel, B.; Kolemen, E.; Kramer, G.; Ku, S.; Kubota, S.; LaHaye, R. J.; Lao, L.; LeBlanc, B. P.; Levinton, F.; Liu, D.; Lore, J.; Lucia, M.; Luhmann, N., Jr.; Maingi, R.; Majeski, R.; Mansfield, D.; Maqueda, R.; McKee, G.; Medley, S.; Meier, E.; Menard, J.; Mueller, D.; Munsat, T.; Muscatello, C.; Myra, J.; Nelson, B.; Nichols, J.; Ono, M.; Osborne, T.; Park, J.-K.; Peebles, W.; Perkins, R.; Phillips, C.; Podesta, M.; Poli, F.; Raman, R.; Ren, Y.; Roszell, J.; Rowley, C.; Russell, D.; Ruzic, D.; Ryan, P.; Sabbagh, S. A.; Schuster, E.; Scotti, F.; Sechrest, Y.; Shaing, K.; Sizyuk, T.; Sizyuk, V.; Skinner, C.; Smith, D.; Snyder, P.; Solomon, W.; Sovenic, C.; Soukhanovskii, V.; Startsev, E.; Stotler, D.; Stratton, B.; Stutman, D.; Taylor, C.; Taylor, G.; Tritz, K.; Walker, M.; Wang, W.; Wang, Z.; White, R.; Wilson, J. R.; Wirth, B.; Wright, J.; Yuan, X.; Yuh, H.; Zakharov, L.; Zweben, S. J.

    2015-10-01

    flux handling. Simulations indicate that snowflake and impurity seeded radiative divertors are candidates for heat flux mitigation in NSTX-U. Studies of lithium evaporation on graphite surfaces indicate that lithium increases oxygen surface concentrations on graphite, and deuterium-oxygen affinity, which increases deuterium pumping and reduces recycling. In situ and test-stand experiments of lithiated graphite and molybdenum indicate temperature-enhanced sputtering, although that test-stand studies also show the potential for heat flux reduction through lithium vapour shielding. Non-linear gyro kinetic simulations have indicated that ion transport can be enhanced by a shear-flow instability, and that non-local effects are necessary to explain the observed rapid changes in plasma turbulence. Predictive simulations have shown agreement between a microtearing-based reduced transport model and the measured electron temperatures in a microtearing unstable regime. Two Alfvén eigenmode-driven fast ion transport models have been developed and successfully benchmarked against NSTX data. Upgrade construction is moving on schedule with initial physics research operation of NSTX-U planned for mid-2015.

  3. Experimental evaluation of brazed molybdenum-graphite bonds for the divertor of the NET/ITER nuclear fusion device

    International Nuclear Information System (INIS)

    Smid, I.; Linke, J.; Nickel, H.; Kny, E.; Reheis, N.; Kneringer, G.; Bolt, H.

    1995-01-01

    Composites consisting of plasma-facing carbon material brazed to molybdenum (TZM) substrates are a promising system for the divertor of the Next European Torus (NET) and the International Thermonuclear Experimental Reactor (ITER). Isotropic graphite and a refractory metal (molybdenum or TZM, a high temperature alloy of molybdenum), two dissimilar substrate materials, yet closely matched in their thermal expansivities, were joined with the use of four different high-temperature brazes: Zr, 90Ni-10Ti, 90Cu- 10Ti, and 70Ag-27Cu-3Ti (compositions in wt%). A summary is given of experiments on mechanical strength, heat transfer capability, structural changes, and failure modes under high heat loads of brazed bonds. Tensile-strength tests on the brazing interface prove the suitability of the brazes up to their melting point. The expected enhancement in thermal contact compared with graphite is confirmed. Passively cooled tiles of dimensions 25 mm x 25 mm were subjected to thermal cycling in electron-beam simulations. Heat fluxes of up to 10 MW m -2 were applied. (author)

  4. Experimental evaluation of brazed molybdenum-graphite bonds for the divertor of the NET/ITER nuclear fusion device

    International Nuclear Information System (INIS)

    Smid, Ivica; Linke, Jochen; Nickel, Hubertus; Kny, Erich; Reheis, Nikolaus; Kneringer, Guenther; Bolt, Harald

    1990-01-01

    Composites consisting of plasma-facing carbon material brazed to molybdenum (TZM) substrates are a promising system for the divertor of the Next European Torus (NET) and the International Thermonuclear Experimental Reactor (ITER). Isotropic graphite and a refractory metal (molybdenum or TZM, a high temperature alloy of molybdenum), two dissimilar substrate materials, yet closely matched in their thermal expansivities, were joined with the use of four different high-temperature brazes: Zr,90Ni-10Ti,90Cu-10Ti, and 70Ag-27Cu-3Ti(compositions in wt%). A summary is given of experiments on mechanical strength, heat transfer capability, structural changes, and failure modes under high heat loads of brazed bonds. Tensile-strength tests on the brazing interface prove the suitability of the brazes up to their melting point. The expected enhancement in thermal contact compared with graphite is confirmed. Passively cooled tiles of dimensions 25 mm x 25 mm were subjected to thermal cycling in electron-beam simulations. Heat fluxes of up to 10 MW m -2 were applied. (author)

  5. Fast Neutral Pressure Gauges in NSTX

    International Nuclear Information System (INIS)

    Raman, R.; Kugel, H.W.; Gernhardt, R.; Provost, T.; Jarboe, T.R.; Soukhanovskii, V.

    2004-01-01

    Successful operation in NSTX of two prototype fast-response micro ionization gauges during plasma operations has motivated us to install five gauges at different toroidal and poloidal locations to measure the edge neutral pressure and its dependence on the type of discharge (L-mode, H-mode, CHI) and the fueling method and location. The edge neutral pressure is also used as an input to the transport analysis codes TRANSP and DEGAS-2. The modified PDX-type Penning gauges are well suited for pressure measurements in the NSTX divertor where the toroidal field is relatively high. Behind the NSTX outer divertor plates where the field is lower, an unshielded fast ion gauge of a new design has been installed. This gauge was developed after laboratory testing of several different designs in a vacuum chamber with applied magnetic fields

  6. Fast Neutral Pressure Measurements in NSTX

    International Nuclear Information System (INIS)

    R. Raman; H.W. Kugel; T. Provost; R. Gernhardt; T.R. Jarboe; M.G. Bell

    2002-01-01

    Several fast neutral pressure gauges have been installed on NSTX [National Spherical Torus Experiment] to measure the vessel and divertor pressure during inductive and coaxial helicity injected (CHI) plasma operations. Modified, PDX [Poloidal Divertor Experiment]-type Penning gauges have been installed on the upper and lower divertors. Neutral pressure measurements during plasma operations from these and from two shielded fast Micro ion gauges at different toroidal locations on the vessel mid-plane are described. A new unshielded ion gauge, referred to as the In-vessel Neutral Pressure (INP) gauge is under development

  7. Feasibility study for an engineering concept of a stainless steel/copper divertor plate protected by W-5 Re alloy or graphite armor

    International Nuclear Information System (INIS)

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

    1988-01-01

    The latest Joint Research Centre (JRC)-Ispra proposal is presented to support the design of a divertor concept that has long been considered the most crucial component of the plasma impurity control system for the Next Europen Torus (NET) tokamak fusion reactor. Because of the harsh tokamak environment, the divertor panel is the plasma facing component that suffers the most severe loading conditions, such as high thermal stresses, thermal fatigue, severe erosion rates and neutron damage. An analysis of a new divertor panel concept has evolved from the previous studies carried out at JRC-Ispra. The materials considered in this study are AISI 316 stainless steel for the cooling tubes, pure copper for the heat sink, and W-5 Re alloy or graphite for the protective armor. The panel is cooled by pressurized water circulation in U-tubes. A preliminary thermo-hydraulic analysis has been carried out to evaluate a set of reference parameters, such as optimum coolant velocity, maximum outlet water temperature, convective heat exchange coefficient, and the expected pressure drops in the channels. Thermal and mechanical calculations, performed by using the finite element technique, showed encouraging results about the engineering feasibility of the pressure boundary of the divertor for loading conditions similar to those of NET double null, assumed as the reference mainframe

  8. Deposition Measurements in NSTX

    Science.gov (United States)

    Skinner, C. H.; Kugel, H. W.; Hogan, J. T.; Wampler, W. R.

    2004-11-01

    Two quartz microbalances have been used to record deposition on the National Spherical Torus Experiment. The experimental configuration mimics a typical diagnostic window or mirror. An RS232 link was used to acquire the quartz crystal frequency and the deposited thickness was recorded continuously with 0.01 nm resolution. Nuclear Reaction Analysis of the deposit was consistent with the measurement of the total deposited mass from the change in crystal frequency. We will present measurements of the variation of deposition with plasma conditions. The transport of carbon impurities in NSTX has been modelled with the BBQ code. Preliminary calculations indicated a negligible fraction of carbon generated at the divertor plates in quiescent discharges directly reaches the outer wall, and that transient events are responsible for the deposition.

  9. NSTX Overview

    International Nuclear Information System (INIS)

    M. Ono; M. Bell; R.E. Bell; M. Bitter; C. Bourdelle; D. Darrow; D. Gates; J. Hosea; S.M. Kaye; R. Kaita; H. Kugel; D. Johnson; B. LeBlanc; S. Medley

    2001-01-01

    The National Spherical Torus Experiment (NSTX) has had a very productive period of plasma operations since the last ST Workshop in Seattle, WA, in November 1999. A number of new research tools have become available and the plasma parameters have improved significantly. These advances are describe in this paper

  10. Edge Recycling and Heat Fluxes in L- and H-mode NSTX Plasmas

    International Nuclear Information System (INIS)

    Soukhanovskii, V.A.; Maingi, R.; Raman, R.; Kugel, H.; LeBlanc, B.; Roquemore, A.L.; Lasnier, C.J.

    2003-01-01

    Introduction Edge characterization experiments have been conducted in NSTX to provide an initial survey of the edge particle and heat fluxes and their scaling with input power and electron density. The experiments also provided a database of conditions for the analyses of the NSTX global particle sources, core fueling, and divertor operating regimes

  11. Overview of innovative PMI research on NSTX-U and associated PMI facilities at PPPL

    International Nuclear Information System (INIS)

    Ono, M.; Jaworski, M.; Kaita, R.; Skinner, C. N.; Allain, J. P.; Maingi, R.; Scotti, F.; Soukhanovskii, V. A.

    2013-01-01

    Developing a reactor compatible divertor and managing the associated plasma material interaction (PMI) has been identified as a high priority research area for magnetic confinement fusion. Accordingly on NSTX-U, the PMI research has received a strong emphasis. Moreover, with ∼15 MW of auxiliary heating power, NSTX-U will be able to test the PMI physics with the peak divertor plasma facing component (PFC) heat loads of up to 40-60 MW/m 2

  12. Snowflake divertor configuration studies in National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-08-15

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

  13. Hydrogen isotope inventory in the graphite divertor tiles of ASDEX Upgrade as measured by thermal desorption spectroscopy

    International Nuclear Information System (INIS)

    Franzen, P.; Behrisch, R.; Garcia-Rosales, C.; Schleussner, D.; Roesler, D.; Becker, J.; Knapp, W.; Edelmann, C.

    1997-01-01

    The hydrogen and deuterium inventories of the ASDEX Upgrade divertor tiles were measured after the experimental period from December 1994 to July 1995 by thermal desorption spectroscopy (TDS) of samples cut out of the divertor tiles. The samples were heated by electron bombardment up to 2100 K; the released gases were measured by means of a calibrated quadrupole mass spectrometer. The measured hydrogen or deuterium inventories are of the order of 10 23 m -2 . They are larger for samples of the inner divertor than of the outer divertor by a factor of about 2. The largest inventory was found at the separatrix position of the inner divertor. Most of the released hydrogen (H) can be attributed to water adsorbed in the near surface region during the air exposure prior to the TDS measurements. The total inventories measured by TDS exceed the inventories in the near surface region (< 25 μm) measured by ion beam analysis methods by a factor of up to 10. Hence, the total hydrogen retention is governed by the diffusion out of the near surface region deep into the material. The hydrogen and deuterium inventories decreased with increasing surface temperature. (author). 64 refs, 12 figs, 2 tabs

  14. Dependence of recycling and edge profiles on lithium evaporation in high triangularity, high performance NSTX H-mode discharges

    Energy Technology Data Exchange (ETDEWEB)

    Maingi, R., E-mail: rmaingi@pppl.gov [Princeton Plasma Physics Laboratory, Receiving 3, Route 1 North, Princeton, NJ 08543 (United States); Osborne, T.H. [General Atomics, 3550 General Atomics Ct., San Diego, CA 92121 (United States); Bell, M.G.; Bell, R.E.; Boyle, D.P. [Princeton Plasma Physics Laboratory, Receiving 3, Route 1 North, Princeton, NJ 08543 (United States); Canik, J.M. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States); Diallo, A.; Kaita, R.; Kaye, S.M.; Kugel, H.W.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, Receiving 3, Route 1 North, Princeton, NJ 08543 (United States); Sabbagh, S.A. [Applied Physics and Applied Math Dept., Columbia University, New York, NY 10027 (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, Receiving 3, Route 1 North, Princeton, NJ 08543 (United States); Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, 7000 East Ave, PO Box 808, Livermore, CA 94551 (United States)

    2015-08-15

    In this paper, the effects of a pre-discharge lithium evaporation variation on highly shaped discharges in the National Spherical Torus Experiment (NSTX) are documented. Lithium wall conditioning (‘dose’) was routinely applied onto graphite plasma facing components between discharges in NSTX, partly to reduce recycling. Reduced D{sub α} emission from the lower and upper divertor and center stack was observed, as well as reduced midplane neutral pressure; the magnitude of reduction increased with the pre-discharge lithium dose. Improved energy confinement, both raw τ{sub E} and H-factor normalized to scalings, with increasing lithium dose was also observed. At the highest doses, we also observed elimination of edge-localized modes. The midplane edge plasma profiles were dramatically altered, comparable to lithium dose scans at lower shaping, where the strike point was farther from the lithium deposition centroid. This indicates that the benefits of lithium conditioning should apply to the highly shaped plasmas planned in NSTX-U.

  15. Analyses of erosion and re-deposition layers on graphite tiles used in the W-shaped divertor region of JT-60U

    International Nuclear Information System (INIS)

    Gotoh, Y.; Yagyu, J.; Masaki, K.; Kizu, K.; Kaminaga, A.; Kodama, K.; Arai, T.; Tanabe, T.; Miya, N.

    2003-01-01

    Erosion and re-deposition profiles were studied on graphite tiles used in the W-shaped divertor of JT-60U in June 1997-October 1998 periods, operated with all-carbon walls with boronizations and inner-private flux pumping. Continuous re-deposition layers were found neither on the dome top nor on the outer wing, while re-deposition layers of around 20 μm thickness were found on the inner wing, in the region close to the dome top. On the outer divertor target, erosion was found to be dominant: maximum erosion depth of around 20 μm was measured, while on the inner target, re-deposition was dominant: columnar structure layers of maximum thickness at around 30 μm on the inner zone while laminar/columnar-layered structures of maximum thickness around 60 μm were found on the outer zone. Poloidal distributions of the erosion depth/re-deposition layer thickness were well correlated with the frequency histograms of strike point position, which were weighted with total power of neutral beam injection, on both the outer and inner targets. Through X-ray photoelectron spectroscopy, composition of the re-deposition layers at a mid zone on the inner target were 3-4 at.% B and <0.6 at.% O, Fe, Cr, and Ni with remaining C. Boron atoms are mostly bound to C atoms but some may precipitated as boron

  16. Graphite

    Science.gov (United States)

    Robinson, Gilpin R.; Hammarstrom, Jane M.; Olson, Donald W.; Schulz, Klaus J.; DeYoung,, John H.; Seal, Robert R.; Bradley, Dwight C.

    2017-12-19

    Graphite is a form of pure carbon that normally occurs as black crystal flakes and masses. It has important properties, such as chemical inertness, thermal stability, high electrical conductivity, and lubricity (slipperiness) that make it suitable for many industrial applications, including electronics, lubricants, metallurgy, and steelmaking. For some of these uses, no suitable substitutes are available. Steelmaking and refractory applications in metallurgy use the largest amount of produced graphite; however, emerging technology uses in large-scale fuel cell, battery, and lightweight high-strength composite applications could substantially increase world demand for graphite.Graphite ores are classified as “amorphous” (microcrystalline), and “crystalline” (“flake” or “lump or chip”) based on the ore’s crystallinity, grain-size, and morphology. All graphite deposits mined today formed from metamorphism of carbonaceous sedimentary rocks, and the ore type is determined by the geologic setting. Thermally metamorphosed coal is the usual source of amorphous graphite. Disseminated crystalline flake graphite is mined from carbonaceous metamorphic rocks, and lump or chip graphite is mined from veins in high-grade metamorphic regions. Because graphite is chemically inert and nontoxic, the main environmental concerns associated with graphite mining are inhalation of fine-grained dusts, including silicate and sulfide mineral particles, and hydrocarbon vapors produced during the mining and processing of ore. Synthetic graphite is manufactured from hydrocarbon sources using high-temperature heat treatment, and it is more expensive to produce than natural graphite.Production of natural graphite is dominated by China, India, and Brazil, which export graphite worldwide. China provides approximately 67 percent of worldwide output of natural graphite, and, as the dominant exporter, has the ability to set world prices. China has significant graphite reserves, and

  17. Tungsten covered graphite and copper elements and ITER-like actively cooled tungsten divertor plasma facing units for the WEST project

    International Nuclear Information System (INIS)

    Guilhem, D; Bucalossi, J; Burles, S; Corre, Y; Ferlay, F; Firdaouss, M; Languille, P; Lipa, M; Martinez, A; Missirlian, M; Proust, M; Richou, M; Samaille, F; Tsitrone, E

    2016-01-01

    After a brief introduction giving some insight of the WEST project, we present the three types of plasma facing units (PFUs) developed for the WEST project taking into account the envisaged main scenarios: (1) high power short pulse scenario (a few seconds) where the objective is to maximize the power handling of the PFUs, up to 20 MW m −2 , (2) high fluence scenario (a few 100 s) on actively cooled ITER-like tungsten (W) PFUs, up to 10 MW m −2 during 1000 s. For the graphite PFUs, the high heat flux tests have been done at GLADIS (ion beam test facility), and for the CuCrZr PFUs on the JUDITH (electron beam test facility). The tests were successful, as no damage occurred for the different load cases. This confirms that the modelling done during the design phase is appropriate to describe these PFUs. Series productions are expected to be achieved by the end of 2015 for the graphite and CuCrZr PFUs, and few ITER-like W PFUs are expected at the beginning of 2016. The lower divertor will be complemented with ITER-like W PFUs as soon as available from our partners so that different fabrication procedures could be evaluated in a real industrial process and a real tokamak environment. (paper)

  18. Implications of NSTX lithium results for magnetic fusion research

    Energy Technology Data Exchange (ETDEWEB)

    Ono, M., E-mail: mono@pppl.gov [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Bell, M.G.; Bell, R.E.; Kaita, R.; Kugel, H.W.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Canik, J.M.; Diem, S. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States); Gerhardt, S.P.; Hosea, J.; Kaye, S.; Mansfield, D. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Maingi, R. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States); Menard, J.; Paul, S.F. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Raman, R. [University of Washington at Seattle, Seattle, WA (United States); Sabbagh, S.A. [Columbia University, New York, NY (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Soukhanovskii, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Taylor, G. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States)

    2010-11-15

    Lithium wall coating techniques have been experimentally explored on National Spherical Torus Experiment (NSTX) for the last five years. The lithium experimentation on NSTX started with a few milligrams of lithium injected into the plasma as pellets and it has evolved to a lithium evaporation system which can evaporate up to {approx}100 g of lithium onto the lower divertor plates between lithium re-loadings. The unique feature of the lithium research program on NSTX is that it can investigate the effects of lithium in H-mode divertor plasmas. This lithium evaporation system thus far has produced many intriguing and potentially important results; the latest of these are summarized in a companion paper by H. Kugel. In this paper, we suggest possible implications and applications of the NSTX lithium results on the magnetic fusion research which include electron and global energy confinement improvements, MHD stability enhancement at high beta, edge localized mode (ELM) control, H-mode power threshold reduction, improvements in radio frequency heating and non-inductive plasma start-up performance, innovative divertor solutions and improved operational efficiency.

  19. Implications of NSTX Lithium Results for Magnetic Fusion Research

    International Nuclear Information System (INIS)

    Ono, M.; Bell, M.G.; Bell, R.E.; Kaita, R.; Kugel, H.W.; LeBlanc, B.P.; Canik, J.M.; Diem, S.; Gerhardt, S.P.; Hosea, J.; Kaye, S.; Mansfield, D.; Maingi, R.; Menard, J.; Paul, S.F.; Raman, R.; Sabbagh, S.A.; Skinner, C.H.; Soukhanovskii, V.; Taylor, G.

    2010-01-01

    Lithium wall coating techniques have been experimentally explored on NSTX for the last five years. The lithium experimentation on NSTX started with a few milligrams of lithium injected into the plasma as pellets and it has evolved to a lithium evaporation system which can evaporate up to ∼ 100 g of lithium onto the lower divertor plates between lithium reloadings. The unique feature of the lithium research program on NSTX is that it can investigate the effects of lithium in H-mode divertor plasmas. This lithium evaporation system thus far has produced many intriguing and potentially important results; the latest of these are summarized in a companion paper by H. Kugel. In this paper, we suggest possible implications and applications of the NSTX lithium results on the magnetic fusion research which include electron and global energy confinement improvements, MHD stability enhancement at high beta, ELM control, H-mode power threshold reduction, improvements in radio frequency heating and non-inductive plasma start-up performance, innovative divertor solutions and improved operational efficiency.

  20. Implications of NSTX lithium results for magnetic fusion research

    International Nuclear Information System (INIS)

    Ono, M.; Bell, M.G.; Bell, R.E.; Kaita, R.; Kugel, H.W.; LeBlanc, B.P.; Canik, J.M.; Diem, S.; Gerhardt, S.P.; Hosea, J.; Kaye, S.; Mansfield, D.; Maingi, R.; Menard, J.; Paul, S.F.; Raman, R.; Sabbagh, S.A.; Skinner, C.H.; Soukhanovskii, V.; Taylor, G.

    2010-01-01

    Lithium wall coating techniques have been experimentally explored on National Spherical Torus Experiment (NSTX) for the last five years. The lithium experimentation on NSTX started with a few milligrams of lithium injected into the plasma as pellets and it has evolved to a lithium evaporation system which can evaporate up to ∼100 g of lithium onto the lower divertor plates between lithium re-loadings. The unique feature of the lithium research program on NSTX is that it can investigate the effects of lithium in H-mode divertor plasmas. This lithium evaporation system thus far has produced many intriguing and potentially important results; the latest of these are summarized in a companion paper by H. Kugel. In this paper, we suggest possible implications and applications of the NSTX lithium results on the magnetic fusion research which include electron and global energy confinement improvements, MHD stability enhancement at high beta, edge localized mode (ELM) control, H-mode power threshold reduction, improvements in radio frequency heating and non-inductive plasma start-up performance, innovative divertor solutions and improved operational efficiency.

  1. Rapidly Moving Divertor Plates In A Tokamak

    International Nuclear Information System (INIS)

    Zweben, S.

    2011-01-01

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

  2. Fabrication and characterization of tungsten and graphite based PFC for divertor target elements of ITER like tokamak application

    Energy Technology Data Exchange (ETDEWEB)

    Khirwadkar, S.S., E-mail: sameer@ipr.res.in [Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Singh, K.P.; Patil, Y.; Khan, M.S.; Buch, J.J.U.; Patel, Alpesh; Tripathi, Sudhir [Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Jaman, P.M.; Rangaraj, L.; Divakar, C. [Materials Science Division, National Aerospace Laboratories, CSIR, Bangalore, Karnataka (India)

    2011-10-15

    The development of the fabrication technology of macro-brush configuration of tungsten (W) and carbon (graphite and CFC) plasma facing components (PFCs) for ITER like tokamak application is presented. The fabrication of qualified joint of PFC is a requirement for fusion tokamak. Vacuum brazing method has been employed for joining of W/CuCrZr and C/CuCrZr. Oxygen free high conductivity (OFHC) copper casting on W tiles was performed followed by machining, polishing and ultrasonic cleaning of the samples prior to vacuum brazing. The W/CuCrZr and graphite/CuCrZr based test mockups were vacuum brazed using silver free alloys. The mechanical shear and tensile strengths were evaluated for the W/CuCrZr and graphite/CuCrZr brazed joint samples. The micro-structural examination of the joints showed smooth interface. The details of fabrication and characterization procedure for macro-brush tungsten and carbon based PFC test mockups are presented.

  3. Recent progress of NSTX lithium program and opportunities for magnetic fusion research

    Energy Technology Data Exchange (ETDEWEB)

    Ono, M., E-mail: mono@pppl.gov [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Bell, M.G.; Kaita, R.; Kugel, H.W. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Ahn, J.-W. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States); Allain, J.P.; Battaglia, D. [Purdue University, West Lafayette, IN 47907 (United States); Bell, R.E. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Canik, J.M. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States); Ding, S. [Academy of Science Institute of Plasma Physics, Hefei (China); Gerhardt, S. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Gray, T.K. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States); Guttenfelder, W.; Hosea, J.; Jaworski, M.A.; Kallman, J.; Kaye, S.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Maingi, R. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States); Mansfield, D.K. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); and others

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer In this paper, we review the recent progress on the NSTX lithium research. Black-Right-Pointing-Pointer We summarize positive features of lithium effects on plasma. Black-Right-Pointing-Pointer We also point out unresolved issues and unanswered questions on the lithium research. Black-Right-Pointing-Pointer We describe a possible closed liquid lithium divertor tray concept. Black-Right-Pointing-Pointer We note opportunities and challenges of lithium applications for magnetic fusion. - Abstract: Lithium wall coating techniques have been experimentally explored on National Spherical Torus Experiment (NSTX) for the last six years. The lithium experimentation on NSTX started with a few milligrams of lithium injected into the plasma as pellets and it has evolved to a dual lithium evaporation system which can evaporate up to {approx}160 g of lithium onto the lower divertor plates between re-loadings. The unique feature of the NSTX lithium research program is that it can investigate the effects of lithium coated plasma-facing components in H-mode divertor plasmas. This lithium evaporation system has produced many intriguing and potentially important results. In 2010, the NSTX lithium program has focused on the effects of liquid lithium divertor (LLD) surfaces including the divertor heat load, deuterium pumping, impurity control, electron thermal confinement, H-mode pedestal physics, and enhanced plasma performance. To fill the LLD with lithium, 1300 g of lithium was evaporated into the NSTX vacuum vessel during the 2010 operations. The routine use of lithium in 2010 has significantly improved the plasma shot availability resulting in a record number of plasma shots in any given year. In this paper, as a follow-on paper from the 1st lithium symposium [1], we review the recent progress toward developing fundamental understanding of the NSTX lithium experimental observations as well as the opportunities and associated R and D required

  4. Advanced divertor experiments on DIII-D

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    Science.gov (United States)

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

    2009-02-01

    Steady-state handling of divertor heat flux is a critical issue for both ITER and spherical torus-based devices with compact high power density divertors. Significant reduction of heat flux to the divertor plate has been achieved simultaneously with favorable core and pedestal confinement and stability properties in a highly shaped lower single null configuration in the National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 2000] using high magnetic flux expansion at the divertor strike point and the radiative divertor technique. A partial detachment of the outer strike point was achieved with divertor deuterium injection leading to peak flux reduction from 4-6MWm-2to0.5-2MWm-2 in small-ELM 0.8-1.0MA, 4-6MW neutral beam injection-heated H-mode discharges. A self-consistent picture of the outer strike point partial detachment was evident from divertor heat flux profiles and recombination, particle flux and neutral pressure measurements. Analytic scrape-off layer parallel transport models were used for interpretation of NSTX detachment experiments. The modeling showed that the observed peak heat flux reduction and detachment are possible with high radiated power and momentum loss fractions, achievable with divertor gas injection, and nearly impossible to achieve with main electron density, divertor neutral density or recombination increases alone.

  6. Theory of Advanced Magnetic Divertors

    Science.gov (United States)

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

    2013-10-01

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

  7. Advanced divertor experiments on DIII-D

    International Nuclear Information System (INIS)

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

    1991-04-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  9. Recent Physics Results from NSTX

    International Nuclear Information System (INIS)

    Menard, J E; Bell, M G; Bell, R E; Bialek, J M; Boedo, J A; Bush, C E; Crocker, N A; Diem, S; Ferron, J R; Fredrickson, E D; Gates, D A; Hill, K W; Hosea, J C; Kaye, S M; Kessel, C E; Kubota, S; Kugel, H W; LeBlanc, B P; Lee, K C; Levinton, F M; Maingi, R; Mansfield, D K; Majeski, R P; Maqueda, R J; Mazzucato, E; Medley, S S; Mueller, D; Park, H K; Paul, S F; Peebles, W A; Raman, R; Sabbagh, S A; Skinner, C H; Smith, D R; Sontag, A C; Soukhanovskii, V A; Stratton, B C; Stutman, D; Taylor, G; Tritz, K; Wilson, J R; Yuh, H; Zhu, W; Zweben, S J

    2006-01-01

    The National Spherical Torus Experiment (NSTX) has made considerable progress in advancing the scientific understanding of high performance long-pulse plasmas needed for ITER and future low-aspect-ratio Spherical Torus (ST) devices. Plasma durations up to 1.6s (5 current redistribution times) have been achieved at plasma currents of 0.7 MA with non-inductive current fractions above 65% while achieving β T and β N values of 16% and 5.7 (%mT/MA), respectively. Newly available Motional Stark Effect data has allowed systematic study and validation of current drive sources and improved the understanding of ''hybrid''-like scenarios. In MHD research, six mid-plane ex-vessel radial field coils have been utilized to infer and correct intrinsic error fields, provide rotation control, and actively stabilize the n=1 resistive wall mode at ITER-relevant low plasma rotation values. In transport and turbulence, the low aspect ratio and wide range of achievable β in NSTX provide unique data for confinement scaling studies. A new high-k scattering diagnostic is investigating turbulent density fluctuations with wavenumbers extending from ion to electron gyro-scales. In the area of energetic particle research, cyclic neutron rate drops have been associated with the destabilization of multiple large Toroidal Alfven Eigenmodes (TAEs) similar to the ''sea-of-TAE'' modes predicted for ITER. Three wave coupling processes between energetic particle modes and TAEs have also been observed for the first time. In boundary physics, advanced shape control has been utilized to study the role of magnetic balance in H-mode access and ELM stability. Peak divertor heat flux has been reduced by a factor of 5 using an H-mode compatible radiative divertor, and Lithium conditioning has demonstrated particle pumping and improved thermal confinement. Finally, non-solenoidal plasma start-up research is particularly important for the ST, and Coaxial Helicity Injection has now produced 160kA plasma

  10. Advantages and Challenges of Radiative Liquid Lithium Divertor

    Science.gov (United States)

    Ono, Masayuki

    2017-10-01

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

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

    International Nuclear Information System (INIS)

    O'NEIL, RC; STAMBAUGH, RD

    2002-01-01

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

  12. NSTX Electrical Power Systems

    International Nuclear Information System (INIS)

    A. Ilic; E. Baker; R. Hatcher; S. Ramakrishnan; et al

    1999-01-01

    The National Spherical Torus Experiment (NSTX) has been designed and installed in the existing facilities at Princeton Plasma Physic Laboratory (PPPL). Most of the hardware, plant facilities, auxiliary sub-systems, and power systems originally used for the Tokamak Fusion Test Reactor (TFTR) have been used with suitable modifications to reflect NSTX needs. The design of the NSTX electrical power system was tailored to suit the available infrastructure and electrical equipment on site. Components were analyzed to verify their suitability for use in NSTX. The total number of circuits and the location of the NSTX device drove the major changes in the Power system hardware. The NSTX has eleven (11) circuits to be fed as compared to the basic three power loops for TFTR. This required changes in cabling to insure that each cable tray system has the positive and negative leg of cables in the same tray. Also additional power cabling had to be installed to the new location. The hardware had to b e modified to address the need for eleven power loops. Power converters had to be reconnected and controlled in anti-parallel mode for the Ohmic heating and two of the Poloidal Field circuits. The circuit for the Coaxial Helicity Injection (CHI) System had to be carefully developed to meet this special application. Additional Protection devices were designed and installed for the magnet coils and the CHI. The thrust was to making the changes in the most cost-effective manner without compromising technical requirements. This paper describes the changes and addition to the Electrical Power System components for the NSTX magnet systems

  13. Divertor detachment

    Science.gov (United States)

    Krasheninnikov, Sergei

    2015-11-01

    The heat exhaust is one of the main conceptual issues of magnetic fusion reactor. In a standard operational regime the large heat flux onto divertor target reaches unacceptable level in any foreseeable reactor design. However, about two decades ago so-called ``detached divertor'' regimes were found. They are characterized by reduced power and plasma flux on divertor targets and look as a promising solution for heat exhaust in future reactors. In particular, it is envisioned that ITER will operate in a partly detached divertor regime. However, even though divertor detachment was studied extensively for two decades, still there are some issues requiring a new look. Among them is the compatibility of detached divertor regime with a good core confinement. For example, ELMy H-mode exhibits a very good core confinement, but large ELMs can ``burn through'' detached divertor and release large amounts of energy on the targets. In addition, detached divertor regimes can be subject to thermal instabilities resulting in the MARFE formation, which, potentially, can cause disruption of the discharge. Finally, often inner and outer divertors detach at different plasma conditions, which can lead to core confinement degradation. Here we discuss basic physics of divertor detachment including different mechanisms of power and momentum loss (ionization, impurity and hydrogen radiation loss, ion-neutral collisions, recombination, and their synergistic effects) and evaluate the roles of different plasma processes in the reduction of the plasma flux; detachment stability; and an impact of ELMs on detachment. We also evaluate an impact of different magnetic and divertor geometries on detachment onset, stability, in- out- asymmetry, and tolerance to the ELMs. Supported by the U.S. Department of Energy Office of Science, Office of Fusion Energy Sciences under Award Number DE-DE-FG02-04ER54739 at UCSD.

  14. The MAST improved divertor

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  15. ECH on NSTX

    International Nuclear Information System (INIS)

    Bigelow, T.S.; Batchelor, D.B.; Carter, M.D.; Peng, M.; Wilson, J.R.

    1997-01-01

    Electron Cyclotron Heating has been proposed for plasma initiation, startup assistance and non-inductive startup on NSTX. One physics goal of NSTX will be to establish entirely non-inductive plasma operation by utilizing ECH to provide a sufficient start-up plasma to support further current drive from other heating systems. Scaling of previous ECH-only startup experiments on CDX-U and DIII-D indicate that 400 kW of ECH should be capable of driving 42 kA of pressure driven current on NSTX and possibly higher levels after optimizing the process. Due to the low NSTX magnetic field, over-dense plasmas exist during most of the discharge so conventional ECH operation is limited to the low density startup phase. To extend the useful operating range for ECH, a scheme involving mode conversion to the electron Bernstein Wave (EBW) from either O or X mode launch is being investigated for bulk heating and current drive applications at higher density. Microwave equipment, including 18 GHz klystrons and 28 GHz gyrotrons are available at ORNL and appear ideal for use on NSTX. Preliminary pre-ionization and start-up system configurations are presented here along with discussions on various operation modes. copyright 1997 American Institute of Physics

  16. ECH on NSTX

    International Nuclear Information System (INIS)

    Bigelow, T.S.; Batchelor, D.B.; Carter, M.D.; Peng, M.; Wilson, J.R.

    1997-01-01

    Electron Cyclotron Heating has been proposed for plasma initiation, startup assistance and non-inductive startup on NSTX. One physics goal of NSTX will be to establish entirely non-inductive plasma operation by utilizing ECH to provide a sufficient start-up plasma to support further current drive from other heating systems. Scaling of previous ECH-only startup experiments on CDX-U and DIII-D indicate that 400 kW of ECH should be capable of driving 42 kA of pressure driven current on NSTX and possibly higher levels after optimizing the process. Due to the low NSTX magnetic field, over-dense plasmas exist during most of the discharge so conventional ECH operation is limited to the low density startup phase. To extend the useful operating range for ECH, a scheme involving mode conversion to the electron Bernstein Wave (EBW) from either O r X mode launch is being investigated for bulk heating and current drive applications at higher density. Microwave equipment, including 18 GHz klystrons and 28 GHz gyrotrons are available at ORNL and appear ideal for use on NSTX. Preliminary pre-ionization and start-up system configurations are presented here along with discussions on various operation modes

  17. Power exhaust scenarios and control for projected high-power NSTX-U operation

    Science.gov (United States)

    Menard, Jonathan; Gerhardt, S. P.; Myers, C. E.; Reinke, M. L.; Brooks, A.; Mardenfeld, M.; NSTX Upgrade Team

    2017-10-01

    An important goal of the NSTX Upgrade (NSTX-U) research program is to characterize energy confinement in the low-aspect-ratio spherical tokamak configuration over a significantly expanded range of plasma current, toroidal field, and heating power, while increasing flattop durations up to 5 seconds. However, the narrowing of the scrape-off layer at higher current combined with an improved understanding of expected halo-current loads has motivated a significant re-design of NSTX-U plasma facing components in the high-heat-flux regions of the divertor. In order to reduce the expected divertor heat flux to acceptable levels, a combination of mitigation techniques will be used: increased divertor poloidal flux expansion, increased divertor radiation, and controlled strike-point sweeping. The machine requirements for these various mitigation techniques are studied here using a newly implemented reduced heat-flux model. Systematic equilibrium scans are used to quantify the required divertor coil currents and to verify vertical stability for a range of plasma shapes. Free-boundary control schemes to constrain the strike-point location and field-line angle-of-incidence will also be discussed. Work supported by DOE contract DE-AC02- 09CH11466.

  18. Power and Particle Balance Calculations with Impurities in NSTX

    Science.gov (United States)

    Holland, C. G.; Maingi, R.; Owen, L. W.; Kaye, S. M.

    1998-11-01

    We reported the development C. Holland, et. al., Bull. Am. Phys. Soc. 42 (1997) 1927. and application R. Maingi et al., Proc. 3rd International Workshop on Spherical Tori, Sept. 3-5, 1997, St. Petersburg, Russia. of a Graphical User Interface to assess the important terms for edge and divertor plasma calculations for NSTX with the b2.5 edge plasma transport code B. Braams, Contrib. Plasma Phys. 36 (1996) 276.. The goals of those calculations were to estimate the worst case peak heat flux for plasma-facing component design, and the radiation requirements to reduce the peak heat flux. In this study we present the first simulations with intrinsic carbon impurity radiation. We find in general that the intrinsic carbon radiation should be sufficient to provide a wide operation window for the NSTX device. Details of the relative importance of heat flux transport mechanisms as determined with the GUI will be presented.

  19. CIT divertor conceptual design

    International Nuclear Information System (INIS)

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

    1988-06-01

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

  20. Raman Spectroscopy of Carbon Dust Samples from NSTX

    International Nuclear Information System (INIS)

    Raitses, Y.; Skinner, C.H.; Jiang, F.; Duffy, T.S.

    2008-01-01

    The Raman spectrum of dust particles exposed to the NSTX plasma is different from the spectrum of unexposed particles scraped from an unused graphite tile. For the unexposed particles, the high energy G-mode peak (Raman shift ∼1580 cm -1 ) is much stronger than the defect-induced D-mode peak (Raman shift ∼1350 cm -1 ), a pattern that is consistent with Raman spectrum for commercial graphite materials. For dust particles exposed to the plasma, the ratio of G-mode to D-mode peaks is lower and becomes even less than 1. The Raman measurements indicate that the production of carbon dust particles in NSTX involves modifications of the physical and chemical structure of the original graphite material. These modifications are shown to be similar to those measured for carbon deposits from atmospheric pressure helium arc discharge with an ablating anode electrode made from a graphite tile material. We also demonstrate experimentally that heating to 2000-2700 K alone can not explain the observed structural modifications indicating that they must be due to higher temperatures needed for graphite vaporization, which is followed either by condensation or some plasma-induced processes leading to the formation of more disordered forms of carbon material than the original graphite.

  1. Evaporated Lithium Surface Coatings in NSTX

    International Nuclear Information System (INIS)

    Kugel, H.W.; Mansfield, D.; Maingi, Rajesh; Bell, M.G.; Bell, R.E.; Allain, J.P.; Gates, D.; Gerhardt, S.P.; Kaita, R.; Kallman, J.; Kaye, S.; LeBlanc, B.P.; Majeski, R.; Menard, J.; Mueller, D.; Ono, M.; Paul, S.; Raman, R.; Roquemore, A.L.; Ross, P.W.; Sabbagh, S.A.; Schneider, H.; Skinner, C.H.; Soukhanovskii, V.; Stevenson, T.; Timberlake, J.; Wampler, W.R.; Wilgen, John B.; Zakharov, L.E.

    2009-01-01

    Two lithium evaporators were used to evaporate more than 100 g of lithium on to the NSTX lower divertor region. Prior to each discharge, the evaporators were withdrawn behind shutters, where they also remained during the subsequent HeGDC applied for periods up to 9.5 min. After the HeGDC, the shutters were opened and the LITERs were reinserted to deposit lithium on the lower divertor target for 10 min, at rates of 10-70 mg/min, prior to the next discharge. The major improvements in plasma performance from these lithium depositions include: (1) plasma density reduction as a result of lithium deposition; (2) suppression of ELMs; (3) improvement of energy confinement in a low-triangularity shape; (4) improvement in plasma performance for standard, high-triangularity discharges: (5) reduction of the required HeGDC time between discharges; (6) increased pedestal electron and ion temperature; (7) reduced SOL plasma density; and (8) reduced edge neutral density.

  2. Evaporated Lithium Surface Coatings in NSTX

    International Nuclear Information System (INIS)

    Kugel, H.W.; Mansfield, D.; Maingi, R.; Bel, M.G.; Bell, R.E.; Allain, J.P.; Gates, D.; Gerhardt, S.; Kaita, R.; Kallman, J.; Kaye, S.; LeBlanc, B.; Majeski, R.; Menard, J.; Mueller, D.; Ono, M.

    2009-01-01

    Two lithium evaporators were used to evaporate more than 100 g of lithium on to the NSTX lower divertor region. Prior to each discharge, the evaporators were withdrawn behind shutters, where they also remained during the subsequent HeGDC applied for periods up to 9.5 min. After the HeGDC, the shutters were opened and the LITERs were reinserted to deposit lithium on the lower divertor target for 10 min, at rates of 10-70 mg/min, prior to the next discharge. The major improvements in plasma performance from these lithium depositions include: (1) plasma density reduction as a result of lithium deposition; (2) suppression of ELMs; (3) improvement of energy confinement in a low-triangularity shape; (4) improvement in plasma performance for standard, high-triangularity discharges; (5) reduction of the required HeGDC time between discharges; (6) increased pedestal electron and ion temperature; (7) reduced SOL plasma density; and (8) reduced edge neutral density

  3. Diagnostic Development on NSTX

    International Nuclear Information System (INIS)

    A.L. Roquemore; D. Johnson; R. Kaita; et al

    1999-01-01

    Diagnostics are described which are currently installed or under active development for the newly commissioned NSTX device. The low aspect ratio (R/a less than or equal to 1.3) and low toroidal field (0.1-0.3T) used in this device dictate adaptations in many standard diagnostic techniques. Technical summaries of each diagnostic are given, and adaptations, where significant, are highlighted

  4. The NSTX Trouble Reporting System

    International Nuclear Information System (INIS)

    Sengupta, S.; Oliaro, G.

    2002-01-01

    An online Trouble Reporting System (TRS) has been introduced at the National Spherical Torus Experiment (NSTX). The TRS is used by NSTX operators to report problems that affect NSTX operations. The purpose of the TRS is to enhance NSTX reliability and maintainability by identifying components, occurrences, and trends that contribute to machine downtime. All NSTX personnel have access to the TRS. The user interface is via a web browser, such as Netscape or Internet Explorer. This web-based feature permits any X-terminal, PC, or MAC access to the TRS. The TRS is based upon a trouble reporting system developed at the DIII-D Tokamak, at General Atomics Technologies. This paper will provide a detailed description of the TRS software architecture, user interface, MS SQL server interface and operational experiences. In addition, sample data from the TRS database will be summarized and presented

  5. Mass changes in NSTX Surface Layers with Li Conditioning as Measured by Quartz Microbalances

    International Nuclear Information System (INIS)

    Skinner, C.H.; Kugel, H.W.; Roquemore, A.L.; Krstic, P.S.; Beste, A.

    2008-01-01

    Dynamic retention, lithium deposition, and the stability of thick deposited layers were measured by three quartz crystal microbalances (QMB) deployed in plasma shadowed areas at the upper and lower divertor and outboard midplane in the National Spherical Torus Experiment (NSTX). Deposition of 185 (micro)/g/cm 2 over 3 months in 2007 was measured by a QMB at the lower divertor while a QMB on the upper divertor, that was shadowed from the evaporator, received an order of magnitude less deposition. During helium glow discharge conditioning both neutral gas collisions and the ionization and subsequent drift of Li + interrupted the lithium deposition on the lower divertor. We present calculations of the relevant mean free paths. Occasionally strong variations in the QMB frequency were observed of thick lithium films suggesting relaxation of mechanical stress and/or flaking or peeling of the deposited layers.

  6. OEDGE modeling of outer wall erosion in NSTX and the effect of changes in neutral pressure

    Energy Technology Data Exchange (ETDEWEB)

    Nichols, J.H., E-mail: jnichols@pppl.gov; Jaworski, M.A.; Kaita, R.; Abrams, T.; Skinner, C.H.; Stotler, D.P.

    2015-08-15

    Gross erosion from the outer wall is expected to be a major source of impurities for high power fusion devices due to the low redeposition fraction. Scaling studies of sputtering from the all-carbon outer wall of NSTX are reported. It is found that wall erosion decreases with divertor plasma pressure in low/mid temperature regimes, due to increasing divertor neutral opacity. Wall erosion is found to consistently decrease with reduced recycling coefficient, with outer target recycling providing the largest contribution. Upper and lower bounds are calculated for the increase in wall erosion due to a low-field-side gas puff.

  7. Design of DIII-D advanced divertor

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  8. Design of DIII-D Advanced Divertor

    International Nuclear Information System (INIS)

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

    1989-11-01

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

  9. Fast wave power flow along SOL field lines in NSTX

    Science.gov (United States)

    Perkins, R. J.; Bell, R. E.; Diallo, A.; Gerhardt, S.; Hosea, J. C.; Jaworski, M. A.; Leblanc, B. P.; Kramer, G. J.; Phillips, C. K.; Roquemore, L.; Taylor, G.; Wilson, J. R.; Ahn, J.-W.; Gray, T. K.; Green, D. L.; McLean, A.; Maingi, R.; Ryan, P. M.; Jaeger, E. F.; Sabbagh, S.

    2012-10-01

    On NSTX, a major loss of high-harmonic fast wave (HHFW) power can occur along open field lines passing in front of the antenna over the width of the scrape-off layer (SOL). Up to 60% of the RF power can be lost and at least partially deposited in bright spirals on the divertor floor and ceiling [1,2]. The flow of HHFW power from the antenna region to the divertor is mostly aligned along the SOL magnetic field [3], which explains the pattern of heat deposition as measured with infrared (IR) cameras. By tracing field lines from the divertor back to the midplane, the IR data can be used to estimate the profile of HHFW power coupled to SOL field lines. We hypothesize that surface waves are being excited in the SOL, and these results should benchmark advanced simulations of the RF power deposition in the SOL (e.g., [4]). Minimizing this loss is critical optimal high-power long-pulse ICRF heating on ITER while guarding against excessive divertor erosion.[4pt] [1] J.C. Hosea et al., AIP Conf Proceedings 1187 (2009) 105. [0pt] [2] G. Taylor et al., Phys. Plasmas 17 (2010) 056114. [0pt] [3] R.J. Perkins et al., to appear in Phys. Rev. Lett. [0pt] [4] D.L. Green et al., Phys. Rev. Lett. 107 (2011) 145001.

  10. Making of the NSTX Facility

    International Nuclear Information System (INIS)

    Neumeyer, C.; Ono, M.; Kaye, S.M.; Peng, Y.-K.M.

    1999-01-01

    The NSTX (National Spherical Torus Experiment) facility located at Princeton Plasma Physics Laboratory is the newest national fusion science experimental facility for the restructured US Fusion Energy Science Program. The NSTX project was approved in FY 97 as the first proof-of-principle national fusion facility dedicated to the spherical torus research. On Feb. 15, 1999, the first plasma was achieved 10 weeks ahead of schedule. The project was completed on budget and with an outstanding safety record. This paper gives an overview of the NSTX facility construction and the initial plasma operations

  11. A solid tungsten divertor for ASDEX Upgrade

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  12. Lithium Wall Conditioning And Surface Dust Detection On NSTX

    International Nuclear Information System (INIS)

    Skinner, C.H.; Allain, J.P.; Bell, M.G.; Friesen, F.Q.L.; Heim, B.; Jaworski, M.A.; Kugel, H.; Maingi, R.; Rais, B.; Taylor, C.N.

    2011-01-01

    Lithium evaporation onto NSTX plasma facing components (PFC) has resulted in improved energy confinement, and reductions in the number and amplitude of edge-localized modes (ELMs) up to the point of complete ELM suppression. The associated PFC surface chemistry has been investigated with a novel plasma material interface probe connected to an in-vacuo surface analysis station. Analysis has demonstrated that binding of D atoms to the polycrystalline graphite material of the PFCs is fundamentally changed by lithium - in particular deuterium atoms become weakly bonded near lithium atoms themselves bound to either oxygen or the carbon from the underlying material. Surface dust inside NSTX has been detected in real-time using a highly sensitive electrostatic dust detector. In a separate experiment, electrostatic removal of dust via three concentric spiral-shaped electrodes covered by a dielectric and driven by a high voltage 3-phase waveform was evaluated for potential application to fusion reactors

  13. Liquid lithium applications for solving challenging fusion reactor issues and NSTX-U contributions

    Energy Technology Data Exchange (ETDEWEB)

    Ono, M., E-mail: mono@pppl.gov [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Jaworski, M.A.; Kaita, R. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Hirooka, Y. [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Gray, T.K. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States)

    2017-04-15

    Steady-state fusion reactor operation presents major divertor technology challenges, including high divertor heat flux both steady-state and transients. In addition, there are unresolved issues of long term dust accumulation and associated tritium inventory and safety concerns (Federici et al., 2001) . It has been suggested that radiative liquid lithium divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving reactor plasma performance (Ono et al., 2013, 2014) . The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium (LL) divertor (RLLD) concept (Ono et al., 2013) and its variant, the active liquid lithium divertor concept (ARLLD) (Ono et al., 2014) , taking advantage of the enhanced non-coronal Li radiation in relatively poorly confined divertor plasmas. It was estimated that only a few moles/s of lithium injection would be needed to significantly reduce the divertor heat flux in a tokamak fusion power plant. By operating at lower temperatures ≤450 °C than the first wall ∼600–700 °C, the LL-covered divertor chamber wall surfaces can serve as an effective particle pump, as impurities generally migrate toward lower temperature LL divertor surfaces. To maintain the LL purity, a closed LL loop system with a modest circulating capacity of ∼1 l/s (l/s) is envisioned to sustain the steady-state operation of a 1 GW-electric class fusion power plant. By running the Li loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to outside where the dust/impurities are removed by relatively simple filter and cold/hot trap systems. Using a

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  16. Advanced divertor concepts

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    Science.gov (United States)

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

    2018-03-01

    divertor was found. The results complement the initial SF divertor studies conducted in high-power H-mode discharges in the NSTX and DIII-D tokamaks, and, along with snowflake divertor results from TCV and other tokamaks, contribute to the physics basis of the SF divertor as a power exhaust concept for future high power density tokamaks.

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  20. Quantitative AMS depth profiling of the hydrogen isotopes collected in graphite divertor and wall tiles of the tokamak ASDEX-Upgrade

    International Nuclear Information System (INIS)

    Sun, G.Y.; Friedrich, M.; Groetzschel, R.; Buerger, W.; Behrisch, R.; Garcia-Rosales, C.

    1997-01-01

    The accelerator mass spectrometry (AMS) facility at the 3 MV Tandetron in Rossendorf has been applied for quantitative depth profiling of deuterium and tritium in samples cut from graphite protection tiles at the vessel walls of the fusion experiment ASDEX-Upgrade at the Max-Planck-Institut fuer Plasmaphysik in Garching. The tritium originates from D(d,p)T fusion reactions in the plasma and it is implanted in the vessel walls together with deuterium atoms and ions from the plasma. The T concentrations in the surface layers down to the analyzing depth of about 25 μm are in the range of 10 11 to 5 x 10 15 T-atoms/cm 3 corresponding to a tritium retention of 3 x 10 10 to 3.5 x 10 12 T-atoms/cm 2 . The much higher deuterium concentrations in the samples were simultaneously measured by calibrated conventional SIMS. In the surface layers down to the analyzing depth of about 25 μm the deuterium concentrations are between 3 x 10 18 and 8 x 10 21 atoms/cm 3 , corresponding to a deuterium retention of 2.5 x 10 16 to 2.5 x 10 18 atoms/cm 2 The estimated total amount of tritium in the vessel walls is of the same order of magnitude as the total number of neutrons produced in D(d,n) 3 He reactions. (orig.)

  1. Overview of Results from the National Spherical Torus Experiment (NSTX)

    International Nuclear Information System (INIS)

    Gates, D.A.; Ahn, J.; Allain, R.; Andre, R.; Bastasz, R.; Bell, M.; Bell, R.; Belova, E.; Berkery, J.; Betti, R.

    2009-01-01

    The mission of NSTX is the demonstration of the physics basis required to extrapolate to the next steps for the spherical torus (ST), such as a plasma facing component test facility (NHTX) or an ST based component test facility (ST-CTF), and to support ITER. Key issues for the ST are transport, and steady state high β operation. To better understand electron transport, a new high-k scattering diagnostic was used extensively to investigate electron gyro-scale fluctuations with varying electron temperature gradient scale-length. Results from n = 3 braking studies are consistent with the flow shear dependence of ion transport. New results from electron Bernstein wave emission measurements from plasmas with lithium wall coating applied indicate transmission efficiencies near 70% in H-mode as a result of reduced collisionality. Improved coupling of High Harmonic Fast-Waves has been achieved by reducing the edge density relative to the critical density for surface wave coupling. In order to achieve high bootstrap current fraction, future ST designs envision running at very high elongation. Plasmas have been maintained on NSTX at very low internal inductance l i ∼ 0.4 with strong shaping (κ ∼ 2.7, (delta) ∼ 0.8) with β N approaching the with-wall beta limit for several energy confinement times. By operating at lower collisionality in this regime, NSTX has achieved record non-inductive current drive fraction f NI ∼ 71%. Instabilities driven by super-Alfvenic ions will be an important issue for all burning plasmas, including ITER. Fast ions from NBI on NSTX are super-Alfvenic. Linear TAE thresholds and appreciable fast-ion loss during multi-mode bursts are measured and these results are compared to theory. The impact of n > 1 error fields on stability is a important result for ITER. RWM/RFA feedback combined with n=3 error field control was used on NSTX to maintain plasma rotation with β above the no-wall limit. Other highlights are: results of lithium coating

  2. Accounting of the Power Balance for Neutral-beam heated H-Mode Plasmas in NSTX

    International Nuclear Information System (INIS)

    Paul, S.F.; Maingi, R.; Soukhanovskii, V.; Kaye, S.M.; Kugel, H.

    2004-01-01

    A survey of the dependence of power balance on input power, shape, and plasma current was conducted for neutral-beam-heated plasmas in the National Spherical Torus Experiment (NSTX). Measurements of heat to the divertor strike plates and divertor and core radiation were taken over a wide range of plasma conditions. The different conditions were obtained by inducing a L-mode to H-mode transition, changing the divertor configuration [lower single null (LSN) vs. double-null (DND)] and conducting a NBI power scan in H-mode. 60-70% of the net input power is accounted for in the LSN discharges with 20% of power lost as fast ions, 30-45% incident on the divertor plates, up to 10% radiated in the core, and about 12% radiated in the divertor. In contrast, the power accountability in DND is 85-90%. A comparison of DND and LSN data show that the remaining power in the LSN is likely to be directed to the upper divertor

  3. Improvement in Plasma Performance with Lithium Coatings in NSTX

    International Nuclear Information System (INIS)

    Kaita, R.

    2009-01-01

    Lithium as a plasma-facing material has attractive features, including a reduction in the recycling of hydrogenic species and the potential for withstanding high heat and neutron fluxes in fusion reactors. Dramatic effects on plasma performance with lithium-coated plasma-facing components (PFC's) have been demonstrated on many fusion devices, including TFTR, T-11M, and FT-U. Using a liquid-lithium-filled tray as a limiter, the CDX-U device achieved very significant enhancement in the confinement time of ohmically heated plasmas. The recent NSTX experiments reported here have demonstrated, for the first time, significant and recurring benefits of lithium PFC coatings on divertor plasma performance in both L- and H- mode regimes heated by neutral beams.

  4. Status and Plans for NSTX-U Recovery

    Science.gov (United States)

    Hawryluk, R. J.; Gerhardt, S.; Menard, J.; Neumeyer, C.

    2017-10-01

    The NSTX-U device experienced a series of technical problems; the most recent of which was the failure of one of the poloidal magnetic field coils, which has rendered the device inoperable and in need of significant repair. As a result of these incidents, the Laboratory performed a very comprehensive analysis of all of the systems on NSTX-U. Through an integrated system's analysis approach, this process identified which actions need to be taken to form a corrective action plan to ensure reliable and predictable operation. The actions required to address the deficiencies were reviewed by external experts who made recommendations on four high-level programmatic decisions regarding the inner poloidal field coils, limitations to the required bakeout temperature needed for conditioning of the vacuum vessel, divertor and wall protection tiles and coaxial helicity injection. The plans for addressing the recommendations from the external review panels will be presented. This research was sponsored by the U.S. Dept. of Energy under contract DE-AC02-09CH11466.

  5. Kinetic Profiles in NSTX Plasmas

    International Nuclear Information System (INIS)

    Bell, R.E.; LeBlanc, B.P.; Bourdelle, C.; Ernst, D.R.; Fredrickson, E.D.; Gates, D.A.; Hosea, J.C.; Johnson, D.W.; Kaye, S.M.; Maingi, R.; Medley, S.; Menard, J.E.; Mueller, D.; Ono, M.; Paoletti, F.; Peng, M.; Sabbagh, S.A.; Stutman, D.; Swain, D.W.; Synakowski, E.J.; Wilson, J.R.

    2001-01-01

    The National Spherical Torus Experiment (NSTX) is a low aspect ratio (R/a approximately 1.3) device with auxiliary heating from neutral-beam injection (NBI) and high-harmonic fast-wave heating (HHFW). Typical NSTX parameters are R(subscript ''0'') = 85 cm, a = 67 cm, I(subscript ''p'') = 0.7-1.4 MA, B(subscript ''phi'') = 0.25-0.45 T. Three co-directed deuterium neutral-beam sources have injected P(subscript ''NB'') less than or equal to 4.7 MW. HHFW plasmas typically have delivered P(subscript ''RF'') less than or equal to 3 MW. Important to the understanding of NSTX confinement are the new kinetic profile diagnostics: a multi-pulse Thomson scattering system (MPTS) and a charge-exchange recombination spectroscopy (CHERS) system. The MPTS diagnostic currently measures electron density and temperature profiles at 30 Hz at ten spatial locations. The CHERS system has recently become available to measure carbon ion temperature and toroidal flow at 17 radial positions spanning the outer half of the minor radius with 20 msec time resolution during NBI. Experiments conducted during the last year have produced a wide range of kinetic profiles in NSTX. Some interesting examples are presented below

  6. Two-dimensional AXUV-based radiated power density diagnostics on NSTX-U.

    Science.gov (United States)

    Faust, I; Delgado-Aparicio, L; Bell, R E; Tritz, K; Diallo, A; Gerhardt, S P; LeBlanc, B; Kozub, T A; Parker, R R; Stratton, B C

    2014-11-01

    A new set of radiated-power-density diagnostics for the National Spherical Torus Experiment Upgrade (NSTX-U) tokamak have been designed to measure the two-dimensional poloidal structure of the total photon emissivity profile in order to perform power balance, impurity transport, and magnetohydrodynamic studies. Multiple AXUV-diode based pinhole cameras will be installed in the same toroidal angle at various poloidal locations. The local emissivity will be obtained from several types of tomographic reconstructions. The layout and response expected for the new radially viewing poloidal arrays will be shown for different impurity concentrations to characterize the diagnostic sensitivity. The radiated power profile inverted from the array data will also be used for estimates of power losses during transitions from various divertor configurations in NSTX-U. The effect of in-out and top/bottom asymmetries in the core radiation from high-Z impurities will be addressed.

  7. Two-dimensional AXUV-based radiated power density diagnostics on NSTX-Ua)

    Science.gov (United States)

    Faust, I.; Delgado-Aparicio, L.; Bell, R. E.; Tritz, K.; Diallo, A.; Gerhardt, S. P.; LeBlanc, B.; Kozub, T. A.; Parker, R. R.; Stratton, B. C.

    2014-11-01

    A new set of radiated-power-density diagnostics for the National Spherical Torus Experiment Upgrade (NSTX-U) tokamak have been designed to measure the two-dimensional poloidal structure of the total photon emissivity profile in order to perform power balance, impurity transport, and magnetohydrodynamic studies. Multiple AXUV-diode based pinhole cameras will be installed in the same toroidal angle at various poloidal locations. The local emissivity will be obtained from several types of tomographic reconstructions. The layout and response expected for the new radially viewing poloidal arrays will be shown for different impurity concentrations to characterize the diagnostic sensitivity. The radiated power profile inverted from the array data will also be used for estimates of power losses during transitions from various divertor configurations in NSTX-U. The effect of in-out and top/bottom asymmetries in the core radiation from high-Z impurities will be addressed.

  8. Two-dimensional AXUV-based radiated power density diagnostics on NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Faust, I.; Parker, R. R. [MIT - Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Delgado-Aparicio, L.; Bell, R. E.; Diallo, A.; Gerhardt, S. P.; LeBlanc, B.; Kozub, T. A. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Tritz, K. [The Johns Hopkins University, Baltimore, Maryland 21209 (United States); Stratton, B. C. [MIT - Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States)

    2014-11-15

    A new set of radiated-power-density diagnostics for the National Spherical Torus Experiment Upgrade (NSTX-U) tokamak have been designed to measure the two-dimensional poloidal structure of the total photon emissivity profile in order to perform power balance, impurity transport, and magnetohydrodynamic studies. Multiple AXUV-diode based pinhole cameras will be installed in the same toroidal angle at various poloidal locations. The local emissivity will be obtained from several types of tomographic reconstructions. The layout and response expected for the new radially viewing poloidal arrays will be shown for different impurity concentrations to characterize the diagnostic sensitivity. The radiated power profile inverted from the array data will also be used for estimates of power losses during transitions from various divertor configurations in NSTX-U. The effect of in-out and top/bottom asymmetries in the core radiation from high-Z impurities will be addressed.

  9. Two-dimensional AXUV-based radiated power density diagnostics on NSTX-U

    International Nuclear Information System (INIS)

    Faust, I.; Parker, R. R.; Delgado-Aparicio, L.; Bell, R. E.; Diallo, A.; Gerhardt, S. P.; LeBlanc, B.; Kozub, T. A.; Tritz, K.; Stratton, B. C.

    2014-01-01

    A new set of radiated-power-density diagnostics for the National Spherical Torus Experiment Upgrade (NSTX-U) tokamak have been designed to measure the two-dimensional poloidal structure of the total photon emissivity profile in order to perform power balance, impurity transport, and magnetohydrodynamic studies. Multiple AXUV-diode based pinhole cameras will be installed in the same toroidal angle at various poloidal locations. The local emissivity will be obtained from several types of tomographic reconstructions. The layout and response expected for the new radially viewing poloidal arrays will be shown for different impurity concentrations to characterize the diagnostic sensitivity. The radiated power profile inverted from the array data will also be used for estimates of power losses during transitions from various divertor configurations in NSTX-U. The effect of in-out and top/bottom asymmetries in the core radiation from high-Z impurities will be addressed

  10. Effect of lithium PFC coatings on NSTX density control

    International Nuclear Information System (INIS)

    Kugel, H.W.; Bell, M.G.; Bell, R.; Bush, C.; Gates, D.; Gray, T.; Kaita, R.; Leblanc, B.; Maingi, R.; Majeski, R.; Mansfield, D.; Mueller, D.; Paul, S.; Raman, R.; Roquemore, A.L.; Sabbagh, S.; Skinner, C.H.; Soukhanovskii, V.; Stevenson, T.; Zakharov, L.

    2007-01-01

    Lithium coatings on the graphite plasma facing components (PFCs) in NSTX are being investigated as a tool for density profile control and reducing the recycling of hydrogen isotopes. Repeated lithium pellet injection into Center Stack Limited and Lower Single Null ohmic helium discharges were used to coat graphite surfaces that had been pre-conditioned with ohmic helium discharges of the same shape to reduce their contribution to hydrogen isotope recycling. The following deuterium NBI reference discharges exhibited a reduction in density by a factor of about 3 for limited and 2 for diverted plasmas, respectively, and peaked density profiles. Recently, a lithium evaporator has been used to apply thin coatings on conditioned and unconditioned PFCs. Effects on the plasma density and the impurities were obtained by pre-conditioning the PFCs with ohmic helium discharges, and performing the first deuterium NBI discharge as soon as possible after applying the lithium coating

  11. Upgrades toward high-heat flux, liquid lithium plasma-facing components in the NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Jaworski, M.A., E-mail: mjaworsk@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Brooks, A.; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Lopes-Cardozo, N. [TU/Eindhoven, Eindhoven (Netherlands); Menard, J.; Ono, M. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Rindt, P. [TU/Eindhoven, Eindhoven (Netherlands); Tresemer, K. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

    2016-11-15

    Highlights: • An upgrade path for the NSTX-U tokamak is proposed that maintains scientific productivity while enabling exploration of novel, liquid metal PFC. • Pre-filled liquid metal divertor targets are proposed as an intermediate step that mitigates technical and scientific risks associated with liquid metal PFC. • Analysis of leading edge features show a strong link between engineering design considerations and expected performance as a PFC. • A method for optimizing porous liquid metal targets restrained by capillary forces is provided indicating pore-sizes well within current technical capabilities. - Abstract: Liquid metal plasma-facing components (PFCs) provide numerous potential advantages over solid-material components. One critique of the approach is the relatively less developed technologies associated with deploying these components in a fusion plasma-experiment. Exploration of the temperature limits of liquid lithium PFCs in a tokamak divertor and the corresponding consequences on core operation are a high priority informing the possibilities for future liquid lithium PFCs. An all-metal NSTX-U is envisioned to make direct comparison between all high-Z wall operation and liquid lithium PFCs in a single device. By executing the all-metal upgrades incrementally, scientific productivity will be maintained while enabling physics and engineering-science studies to further develop the solid- and liquid-metal components. Six major elements of a flowing liquid-metal divertor system are described and a three-step program for implementing this system is laid out. The upgrade steps involve the first high-Z divertor target upgrade in NSTX-U, pre-filled liquid metal targets and finally, an integrated, flowing liquid metal divertor target. Two example issues are described where the engineering and physics experiments are shown to be closely related in examining the prospects for future liquid metal PFCs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-11-21

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

  13. LHD helical divertor

    International Nuclear Information System (INIS)

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

    1993-07-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  16. Hydrogen retention in lithium on metallic walls from “in vacuo” analysis in LTX and implications for high-Z plasma-facing components in NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Kaita, R., E-mail: kaita@pppl.gov [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Lucia, M. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Allain, J.P.; Bedoya, F. [Department of Nuclear, Plasma, & Radiological Engineering, University of Illinois at Urbana-Champaign, Urbana, IL (United States); Bell, R.; Boyle, D. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Capece, A. [Department of Physics, The College of New Jersey, Ewing, NJ (United States); Jaworski, M. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Koel, B.E. [Department of Chemical & Biological Engineering, Princeton University, Princeton, NJ (United States); Majeski, R. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Roszell, J. [Department of Chemical & Biological Engineering, Princeton University, Princeton, NJ (United States); Schmitt, J. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Scotti, F. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Soukhanovskii, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States)

    2017-04-15

    The application of lithium to plasma-facing components (PFCs) has long been used as a technique for wall conditioning in magnetic confinement devices to improve plasma performance. Determining the characteristics of PFCs at the time of exposure to the plasma, however, is difficult because they can only be analyzed after venting the vacuum vessel and removing them at the end of an operational period. The Materials Analysis and Particle Probe (MAPP) addresses this problem by enabling PFC samples to be exposed to plasmas, and then withdrawn into an analysis chamber without breaking vacuum. The MAPP system was used to introduce samples that matched the metallic PFCs of the Lithium Tokamak Experiment (LTX). Lithium that was subsequently evaporated onto the walls also covered the MAPP samples, which were then subject to LTX discharges. In vacuo extraction and analysis of the samples indicated that lithium oxide formed on the PFCs, but improved plasma performance persisted in LTX. The reduced recycling this suggests is consistent with separate surface science experiments that demonstrated deuterium retention in the presence of lithium oxide films. Since oxygen decreases the thermal stability of the deuterium in the film, the release of deuterium was observed below the lithium deuteride dissociation temperature. This may explain what occurred when lithium was applied to the surface of the NSTX Liquid Lithium Divertor (LLD). The LLD had segments with individual heaters, and the deuterium-alpha emission was clearly lower in the cooler regions. The plan for NSTX-U is to replace the graphite tiles with high-Z PFCs, and apply lithium to their surfaces with lithium evaporation. Experiments with lithium coatings on such PFCs suggest that deuterium could still be retained if lithium compounds form, but limiting their surface temperatures may be necessary.

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

    International Nuclear Information System (INIS)

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

    1996-10-01

    New divertor hardware is being designed and fabricated for the Radiative Divertor modification of the DIII-D tokamak. The installation of the hardware has been separated into two phases, the first phase starting in October of 1996 and the second and final phase, in 1998. The phased approach enables the continuation of the divertor characterization research in the lower divertor while providing pumping for density control in high triangularity, single- or double-null advanced tokamak discharges. When completed, the Radiative Divertor Project hardware will provide pumping at all four strike points of a double-null, high triangularity discharge and provide baffling of the neutral particles from transport back to the core plasma. By puffing neutral gas into the divertor region, a reduction in the heat flux on the target plates will be be demonstrated without a large rise in core density. This reduction in heat flux is accomplished by dispersing the power with radiation in the divertor region. Experiments and modeling have formed the basis for the new design. The capability of the DIII-D cryogenic system is being upgraded as part of this project. The increased capability of the cryogenic system will allow delivery of liquid helium and nitrogen to three new cryopumps. Physics studies on the effects of slot width and length can be accomplished easily with the design of the Radiative Divertor. The slot width can be varied by installing graphite tiles of different geometry. The change in slot length, the distance from the X-point to the target plate, requires relocating the structure vertically and can be completed in about 6-8 weeks. Radiative Divertor diagnostics are being designed to provide comprehensive measurements for diagnosing the divertor. Required diagnostic modifications will be minimal for Phase 1, but extensive for Phase 2 installation. These Phase 2 diagnostics will be required to fully diagnose the high triangularity discharges in the divertor slots

  18. Brazing graphite to graphite

    International Nuclear Information System (INIS)

    Peterson, G.R.

    1976-01-01

    Graphite is joined to graphite by employing both fine molybdenum powder as the brazing material and an annealing step that together produce a virtually metal-free joint exhibiting properties similar to those found in the parent graphite. Molybdenum powder is placed between the faying surfaces of two graphite parts and melted to form molybdenum carbide. The joint area is thereafter subjected to an annealing operation which diffuses the carbide away from the joint and into the graphite parts. Graphite dissolved by the dispersed molybdenum carbide precipitates into the joint area, replacing the molybdenum carbide to provide a joint of graphite

  19. Erosion of lithium coatings on TZM molybdenum and graphite during high-flux plasma bombardment

    NARCIS (Netherlands)

    Abrams, T.; Jaworski, M. A.; Kaita, R.; Stotler, D. P.; De Temmerman, G.; Morgan, T. W.; van den Berg, M. A.; van der Meiden, H. J.

    2014-01-01

    Abstract The rate at which Li films will erode under plasma bombardment in the NSTX-U divertor is currently unknown. It is important to characterize this erosion rate so that the coatings can be replenished before they are completely depleted. An empirical formula for the Li erosion rate as a

  20. National Spherical Torus Experiment (NSTX)

    International Nuclear Information System (INIS)

    Masayuki Ono

    2000-01-01

    The main aim of National Spherical Torus Experiment (NSTX) is to establish the fusion physics principles of the innovative spherical torus (ST) concept. Physics outcome of the NSTX research program is relevant to near-term applications such as the Volume Neutron Source (VNS) and burning plasmas, and future applications such as the pilot and power plants. The NSTX device began plasma operations in February 1999 and the plasma current was successfully ramped up to the design value of 1 million amperes (MA) on December 14, 1999. The CHI (Coaxial Helicity Injection) and HHFW (High Harmonic Fast Wave) experiments have also started. Stable CHI discharges of up to 133 kA and 130-msec duration have been produced using 20 kA of injected current. Using eight antennas connected to two transmitters, up to 2 MW of HHFW power was successfully coupled to the plasma. The Neutral-beam Injection (NBI) heating system and associated NBI-based diagnostics such as the Charge-exchange Recombination Spectrometer (CHERS) will be operational in October 2000

  1. Special graphites; Graphites speciaux

    Energy Technology Data Exchange (ETDEWEB)

    Leveque, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    A large fraction of the work undertaken jointly by the Commissariat a l'Energie Atomique (CEA) and the Pechiney Company has been the improvement of the properties of nuclear pile graphite and the opening up of new fields of graphite application. New processes for the manufacture of carbons and special graphites have been developed: forged graphite, pyro-carbons, high density graphite agglomeration of graphite powders by cracking of natural gas, impervious graphites. The physical properties of these products and their reaction with various oxidising gases are described. The first irradiation results are also given. (authors) [French] Ameliorer les proprietes du graphite nucleaire pour empilements et ouvrir de nouveaux domaines d'application au graphite constituent une part importante de l'effort entrepris en commun par le Commissariat a l'Energie Atomique (CEA) et la compagnie PECHINEY. Des procedes nouveaux de fabrication de carbones et graphites speciaux ont ete mis au point: graphite forge, pyrocarbone, graphite de haute densite, agglomeration de poudres de graphite par craquage de gaz naturel, graphites impermeables. Les proprietes physiques de ces produits ainsi que leur reaction avec differents gaz oxydants sont decrites. Les premiers resultats d'irradiation sont aussi donnes. (auteurs)

  2. The NSTX Trouble Reporting System; TOPICAL

    International Nuclear Information System (INIS)

    S. Sengupta; G. Oliaro

    2002-01-01

    An online Trouble Reporting System (TRS) has been introduced at the National Spherical Torus Experiment (NSTX). The TRS is used by NSTX operators to report problems that affect NSTX operations. The purpose of the TRS is to enhance NSTX reliability and maintainability by identifying components, occurrences, and trends that contribute to machine downtime. All NSTX personnel have access to the TRS. The user interface is via a web browser, such as Netscape or Internet Explorer. This web-based feature permits any X-terminal, PC, or MAC access to the TRS. The TRS is based upon a trouble reporting system developed at the DIII-D Tokamak, at General Atomics Technologies. This paper will provide a detailed description of the TRS software architecture, user interface, MS SQL server interface and operational experiences. In addition, sample data from the TRS database will be summarized and presented

  3. Operating windows of pebble divertor

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  4. Development of liquid lithium divertor for fusion reactor

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  5. The impact of lithium wall coatings on NSTX discharges and the engineering of the Lithium Tokamak eXperiment (LTX)

    International Nuclear Information System (INIS)

    Majeski, R.; Kugel, H.; Kaita, R.; Avasarala, S.; Bell, M.G.; Bell, R.E.; Berzak, L.; Beiersdorfer, P.; Gerhardt, S.P.; Gransted, E.; Gray, T.; Jacobson, C.; Kallman, J.; Kaye, S.; Kozub, T.; LeBlanc, B.P.; Lepson, J.; Lundberg, D.P.; Maingi, R.; Mansfield, D.; Paul, S.F.; Pereverzev, G.V.; Schneider, H.; Soukhanovskii, V.; Strickler, T.; Stotler, D.; Timberlake, J.; Zakharov, L.E.

    2010-01-01

    Recent experiments on the National Spherical Torus eXperiment (NSTX) have shown the benefits of solid lithium coatings on carbon PFC's to diverted plasma performance, in both L- and H-mode confinement regimes. Better particle control, with decreased inductive flux consumption, and increased electron temperature, ion temperature, energy confinement time, and DD neutron rate were observed. Successive increases in lithium coverage resulted in the complete suppression of ELM activity in H-mode discharges. A liquid lithium divertor (LLD), which will employ the porous molybdenum surface developed for the LTX shell, is being installed on NSTX for the 2010 run period, and will provide comparisons between liquid walls in the Lithium Tokamak eXperiment (LTX) and liquid divertor targets in NSTX. LTX, which recently began operations at the Princeton Plasma Physics Laboratory, is the world's first confinement experiment with full liquid metal plasma-facing components (PFCs). All materials and construction techniques in LTX are compatible with liquid lithium. LTX employs an inner, heated, stainless steel-faced liner or shell, which will be lithium-coated. In order to ensure that lithium adheres to the shell, it is designed to operate at up to 500-600 degrees C to promote wetting of the stainless by the lithium, providing the first hot wall in a tokamak to Operate at reactor-relevant temperatures. The engineering of LTX will be discussed.

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

    Science.gov (United States)

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

    2012-10-01

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

  7. Divertor design for the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    Hill, D.N.; Braams, B.

    1994-05-01

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

  8. Models for poloidal divertors

    International Nuclear Information System (INIS)

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

    1982-07-01

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

  9. Models for poloidal divertors

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-07-01

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

  10. RF Rectification on LAPD and NSTX: the relationship between rectified currents and potentials

    Science.gov (United States)

    Perkins, R. J.; Carter, T.; Caughman, J. B.; van Compernolle, B.; Gekelman, W.; Hosea, J. C.; Jaworski, M. A.; Kramer, G. J.; Lau, C.; Martin, E. H.; Pribyl, P.; Tripathi, S. K. P.; Vincena, S.

    2017-10-01

    RF rectification is a sheath phenomenon important in the fusion community for impurity injection, hot spot formation on plasma-facing components, modifications of the scrape-off layer, and as a far-field sink of wave power. The latter is of particular concern for the National Spherical Torus eXperiment (NSTX), where a substantial fraction of the fast-wave power is lost to the divertor along scrape-off layer field lines. To assess the relationship between rectified currents and rectified voltages, detailed experiments have been performed on the Large Plasma Device (LAPD). An electron current is measured flowing out of the antenna and into the limiters, consistent with RF rectification with a higher RF potential at the antenna. The scaling of this current with RF power will be presented. The limiters are also floated to inhibit this DC current; the impact of this change on plasma-potential and wave-field measurements will be shown. Comparison to data from divertor probes in NSTX will be made. These experiments on a flexible mid-sized experiment will provide insight and guidance into the effects of ICRF on the edge plasma in larger fusion experiments. Funded by the DOE OFES (DE-FC02-07ER54918 and DE-AC02-09CH11466), NSF (NSF- PHY 1036140), and the Univ. of California (12-LR- 237124).

  11. High-resolution Tangential AXUV Arrays for Radiated Power Density Measurements on NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Delgado-Aparicio, L [PPPL; Bell, R E [PPPL; Faust, I [MIT; Tritz, K [The Johns Hopkins University, Baltimore, MD, 21209, USA; Diallo, A [PPPL; Gerhardt, S P [PPPL; Kozub, T A [PPPL; LeBlanc, B P [PPPL; Stratton, B C [PPPL

    2014-07-01

    Precise measurements of the local radiated power density and total radiated power are a matter of the uttermost importance for understanding the onset of impurity-induced instabilities and the study of particle and heat transport. Accounting of power balance is also needed for the understanding the physics of various divertor con gurations for present and future high-power fusion devices. Poloidal asymmetries in the impurity density can result from high Mach numbers and can impact the assessment of their flux-surface-average and hence vary the estimates of P[sub]rad (r, t) and (Z[sub]eff); the latter is used in the calculation of the neoclassical conductivity and the interpretation of non-inductive and inductive current fractions. To this end, the bolometric diagnostic in NSTX-U will be upgraded, enhancing the midplane coverage and radial resolution with two tangential views, and adding a new set of poloidally-viewing arrays to measure the 2D radiation distribution. These systems are designed to contribute to the near- and long-term highest priority research goals for NSTX-U which will integrate non-inductive operation at reduced collisionality, with high-pressure, long energy-confinement-times and a divertor solution with metal walls.

  12. Module of lithium divertor for KTM tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  13. Analysis Efforts Supporting NSTX Upgrades

    International Nuclear Information System (INIS)

    Zhang, H.; Titus, P.; Rogoff, P.; Zolfaghari, A.; Mangra, D.; Smith, M.

    2010-01-01

    The National Spherical Torus Experiment (NSTX) is a low aspect ratio, spherical torus (ST) configuration device which is located at Princeton Plasma Physics Laboratory (PPPL) This device is presently being updated to enhance its physics by doubling the TF field to 1 Tesla and increasing the plasma current to 2 Mega-amperes. The upgrades include a replacement of the centerstack and addition of a second neutral beam. The upgrade analyses have two missions. The first is to support design of new components, principally the centerstack, the second is to qualify existing NSTX components for higher loads, which will increase by a factor of four. Cost efficiency was a design goal for new equipment qualification, and reanalysis of the existing components. Showing that older components can sustain the increased loads has been a challenging effort in which designs had to be developed that would limit loading on weaker components, and would minimize the extent of modifications needed. Two areas representing this effort have been chosen to describe in more details: analysis of the current distribution in the new TF inner legs, and, second, analysis of the out-of-plane support of the existing TF outer legs.

  14. Rogowski Loop design for NSTX

    International Nuclear Information System (INIS)

    McCormack, B.; Kaita, R.; Kugel, H.; Hatcher, R.

    2000-01-01

    The Rogowski Loop is one of the most basic diagnostics for tokamak operations. On the National Spherical Torus Experiment (NSTX), the plasma current Rogowski Loop had the constraints of the very limited space available on the center stack, 5,000 volt isolation, flexibility requirements as it remained a part of the Center Stack assembly after the first phase of operation, and a +120 C temperature requirement. For the second phase of operation, four Halo Current Rogowski Loops under the Center Stack tiles will be installed having +600 C and limited space requirements. Also as part of the second operational phase, up to ten Rogowski Loops will installed to measure eddy currents in the Passive Plate support structures with +350 C, restricted space, and flexibility requirements. This presentation will provide the details of the material selection, fabrication techniques, testing, and installation results of the Rogowski Loops that were fabricated for the high temperature operational and bakeout requirements, high voltage isolation requirements, and the space and flexibility requirements imposed upon the Rogowski Loops. In the future operational phases of NSTX, additional Rogowski Loops could be anticipated that will measure toroidal plasma currents in the vacuum vessel and in the Passive Plate assemblies

  15. Exploratory studies of flowing liquid metal divertor options for fusion-relevant magnetic fields in the MTOR facility

    International Nuclear Information System (INIS)

    Ying, A.Y.; Abdou, M.A.; Morley, N.; Sketchley, T.; Woolley, R.; Burris, J.; Kaita, R.; Fogarty, P.; Huang, H.; Lao, X.; Narula, M.; Smolentsev, S.; Ulrickson, M.

    2004-01-01

    This paper reports on experimental findings on liquid metal (LM) free surface flows crossing complex magnetic fields. The experiments involve jet and film flows using GaInSn and are conducted at the UCLA MTOR facility. The goal of this study is to understand the magnetohydrodynamics (MHD) features associated with such a free surface flow in a fusion-relevant magnetic field environment, and determine what LM free surface flow option is most suitable for lithium divertor particle pumping and surface heat removal applications in a near-term experimental plasma device, such as NSTX. Experimental findings indicate that a steady transverse magnetic field, even with gradients typical of NSTX outer divertor conditions, stabilizes a LM jet flow--reducing turbulent disturbances and delaying jet breakup. Important insights into the MHD behavior of liquid metal films under NSTX-like environments are also presented. It is possible to establish an uphill liquid metal film flow on a conducting substrate, although the MHD drag experienced by the flow could be strong and cause the flow to pile-up under simulated NSTX magnetic field conditions. The magnetic field changes the turbulent film flow so that wave structures range from 2D column-type surface disturbances at regions of high magnetic field, to ordinary hydrodynamic turbulence wave structures at regions of low field strength at the outboard. Plans for future work are also presented

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

    International Nuclear Information System (INIS)

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

    2001-01-01

    The DIII-D tokamak has demonstrated an operational scenario where the graphite-covered divertor is free of net erosion. Reduction of divertor carbon erosion is accomplished using a low temperature (detached) divertor plasma that eliminates physical sputtering. Likewise, the carbon source rate arising from chemical erosion is found to be very low in the detached divertor. Near strikepoint regions, the rate of carbon deposition is ∼3 cm/burn-year, with a corresponding hydrogenic codeposition rate >1kg/m 2 /burn-year; rates both problematic for steady-state fusion reactors. The carbon net deposition rate in the divertor is consistent with carbon arriving from the core plasma region. Carbon influx from the main wall is measured to be relatively large in the high-density detached regime and is of sufficient magnitude to account for the deposition rate in the divertor. Divertor redeposition is therefore determined by non-divertor erosion and transport. Despite the success in reducing divertor erosion on DIII-D with detachment, no significant reduction is found in the core plasma carbon density, illustrating the importance of non-divertor erosion and the complex coupling between erosion/redeposition and impurity plasma transport. (author)

  17. Tokamak Simulation Code modeling of NSTX

    International Nuclear Information System (INIS)

    Jardin, S.C.; Kaye, S.; Menard, J.; Kessel, C.; Glasser, A.H.

    2000-01-01

    The Tokamak Simulation Code [TSC] is widely used for the design of new axisymmetric toroidal experiments. In particular, TSC was used extensively in the design of the National Spherical Torus eXperiment [NSTX]. The authors have now benchmarked TSC with initial NSTX results and find excellent agreement for plasma and vessel currents and magnetic flux loops when the experimental coil currents are used in the simulations. TSC has also been coupled with a ballooning stability code and with DCON to provide stability predictions for NSTX operation. TSC has also been used to model initial CHI experiments where a large poloidal voltage is applied to the NSTX vacuum vessel, causing a force-free current to appear in the plasma. This is a phenomenon that is similar to the plasma halo current that sometimes develops during a plasma disruption

  18. Textor bundle divertor

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  19. TEXTOR bundle divertor

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  20. Divertor cooling device

    International Nuclear Information System (INIS)

    Nakayama, Tadakazu; Hayashi, Katsumi; Handa, Hiroyuki

    1993-01-01

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

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

    Science.gov (United States)

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

    2016-02-01

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

  2. 'EU divertor celebration day'

    International Nuclear Information System (INIS)

    Merola, M.

    2002-01-01

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

  3. T-12 divertor experiment

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-01-01

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

  4. Diagnostic Development for ST Plasmas on NSTX

    International Nuclear Information System (INIS)

    Johnson, D.

    2003-01-01

    Spherical tokamaks (STs) have much lower aspect ratio (a/R) and lower toroidal magnetic field, relative to tokamaks and stellarators. This paper will highlight some of the challenges and opportunities these features pose in the diagnosis of ST plasmas on the National Spherical Torus Experiment (NSTX), and discuss some of the corresponding diagnostic development that is underway. The low aspect ratio necessitates a small center stack, with tight space constraints and large thermal excursions, complicating the design of magnetic sensors in this region. The toroidal magnetic field on NSTX is less than or equal to 0.6 T, making it impossible to use ECE as a good monitor of electron temperature. A promising new development for diagnosing electron temperature is electron Bernstein wave (EBW) radiometry, which is currently being pursued on NSTX. A new high-resolution charge exchange recombination spectroscopy system is being installed. Since non-inductive current initiation and sustainment ar e top-level NSTX research goals, measurements of the current profile J(R) are essential to many planned experiments. On NSTX several modifications are planned to adapt the MSE technique to lower field, and two novel MSE systems are being prototyped. Several high speed 2-D imaging techniques are being developed, for viewing both visible and x-ray emission. The toroidal field is comparable to the poloidal field at the outside plasma edge, producing a large field pitch (>50 o ) at the outer mid-plane. The large shear in pitch angle makes some fluctuation diagnostics like beam emission spectroscopy very difficult, while providing a means of achieving spatial localization for microwave scattering investigations of high-k turbulence, which are predicted to be virulent for NSTX plasmas. A brief description of several of these techniques will be given in the context of the current NSTX diagnostic set

  5. Plans of LHD divertor experiment

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  6. Innovative divertor concepts for LHD

    International Nuclear Information System (INIS)

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

    1994-07-01

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

  7. Towards identifying the mechanisms underlying field-aligned edge-loss of HHFW power on NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Perkins, R. J.; Bell, R. E.; Bertelli, N.; Diallo, A.; Gerhardt, S.; Hosea, J. C.; Jaworski, M. A.; LeBlanc, B. P.; Kramer, G. J.; Maingi, R.; Phillips, C. K.; Podestà, M.; Roquemore, L.; Scotti, F.; Taylor, G.; Wilson, J. R. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Ahn, J-W.; Gray, T. K.; Green, D. L.; McLean, A. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); and others

    2014-02-12

    Fast-wave heating will be a major heating scheme on ITER, as it can heat ions directly and is relatively unaffected by the large machine size unlike neutral beams. However, fast-wave interactions with the plasma edge can lead to deleterious effects such as, in the case of the high-harmonic fast-wave (HHFW) system on NSTX, large losses of fast-wave power in the scrape off layer (SOL) under certain conditions. In such scenarios, a large fraction of the lost HHFW power is deposited on the upper and lower divertors in bright spiral shapes. The responsible mechanism(s) has not yet been identified but may include fast-wave propagation in the scrape off layer, parametric decay instability, and RF currents driven by the antenna reactive fields. Understanding and mitigating these losses is important not only for improving the heating and current-drive on NSTX-Upgrade but also for understanding fast-wave propagation across the SOL in any fast-wave system. This talk summarizes experimental results demonstrating that the flow of lost HHFW power to the divertor regions largely follows the open SOL magnetic field lines. This lost power flux is relatively large close to both the antenna and the last closed flux surface with a reduced level in between, so the loss mechanism cannot be localized to the antenna. At the same time, significant losses also occur along field lines connected to the inboard edge of the bottom antenna plate. The power lost within the spirals is roughly estimated, showing that these field-aligned losses to the divertor are significant but may not account for the total HHFW loss. To elucidate the role of the onset layer for perpendicular fast-wave propagation with regards to fast-wave propagation in the SOL, a cylindrical cold-plasma model is being developed. This model, in addition to advanced RF codes such as TORIC and AORSA, is aimed at identifying the underlying mechanism(s) behind these SOL losses, to minimize their effects in NSTX-U, and to predict

  8. Towards identifying the mechanisms underlying field-aligned edge-loss of HHFW power on NSTX

    International Nuclear Information System (INIS)

    Perkins, R. J.; Bell, R. E.; Bertelli, N.; Diallo, A.; Gerhardt, S.; Hosea, J. C.; Jaworski, M. A.; LeBlanc, B. P.; Kramer, G. J.; Maingi, R.; Phillips, C. K.; Podestà, M.; Roquemore, L.; Scotti, F.; Taylor, G.; Wilson, J. R.; Ahn, J-W.; Gray, T. K.; Green, D. L.; McLean, A.

    2014-01-01

    Fast-wave heating will be a major heating scheme on ITER, as it can heat ions directly and is relatively unaffected by the large machine size unlike neutral beams. However, fast-wave interactions with the plasma edge can lead to deleterious effects such as, in the case of the high-harmonic fast-wave (HHFW) system on NSTX, large losses of fast-wave power in the scrape off layer (SOL) under certain conditions. In such scenarios, a large fraction of the lost HHFW power is deposited on the upper and lower divertors in bright spiral shapes. The responsible mechanism(s) has not yet been identified but may include fast-wave propagation in the scrape off layer, parametric decay instability, and RF currents driven by the antenna reactive fields. Understanding and mitigating these losses is important not only for improving the heating and current-drive on NSTX-Upgrade but also for understanding fast-wave propagation across the SOL in any fast-wave system. This talk summarizes experimental results demonstrating that the flow of lost HHFW power to the divertor regions largely follows the open SOL magnetic field lines. This lost power flux is relatively large close to both the antenna and the last closed flux surface with a reduced level in between, so the loss mechanism cannot be localized to the antenna. At the same time, significant losses also occur along field lines connected to the inboard edge of the bottom antenna plate. The power lost within the spirals is roughly estimated, showing that these field-aligned losses to the divertor are significant but may not account for the total HHFW loss. To elucidate the role of the onset layer for perpendicular fast-wave propagation with regards to fast-wave propagation in the SOL, a cylindrical cold-plasma model is being developed. This model, in addition to advanced RF codes such as TORIC and AORSA, is aimed at identifying the underlying mechanism(s) behind these SOL losses, to minimize their effects in NSTX-U, and to predict

  9. Radiation transport effects in divertor plasmas generated during a tokamak reactor disruption

    International Nuclear Information System (INIS)

    Peterson, R.R.; MacFarlane, J.J.; Wang, P.

    1994-01-01

    Vaporization of material from tokamak divertors during disruptions is a critical issue for tokamak reactors from ITER to commercial power plants. Radiation transport from the vaporized material onto the remaining divertor surface plays an important role in the total mass loss to the divertor. Radiation transport in such a vapor is very difficult to calculate in full detail, and this paper quantifies the sensitivity of the divertor mass loss to uncertainties in the radiation transport. Specifically, the paper presents the results of computer simulations of the vaporization of a graphite coated divertor during a tokamak disruption with ITER CDA parameters. The results show that a factor of 100 change in the radiation conductivity changes the mass loss by more than a factor of two

  10. Overview of the NSTX Control System

    International Nuclear Information System (INIS)

    Sichta, P.; Dong, J.; Oliaro, G.; Roney, P.

    2001-01-01

    The National Spherical Torus Experiment (NSTX) is an innovative magnetic fusion device that was constructed by the Princeton Plasma Physics Laboratory (PPPL) in collaboration with the Oak Ridge National Laboratory, Columbia University, and the University of Washington at Seattle. Since achieving first plasma in 1999, the device has been used for fusion research through an international collaboration of more than twenty institutions. The NSTX is operated through a collection of control systems that encompass a wide range of technology, from hardwired relay controls to real-time control systems with giga-FLOPS of capability. This paper presents a broad introduction to the control systems used on NSTX, with an emphasis on the computing controls, data acquisition, and synchronization systems

  11. Versator divertor experiment: preliminary designs

    International Nuclear Information System (INIS)

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

    1984-08-01

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

  12. Analysis of NSTX TF Joint Voltage Measurements

    International Nuclear Information System (INIS)

    Woolley R

    2005-01-01

    This report presents findings of analyses of recorded current and voltage data associated with 72 electrical joints operating at high current and high mechanical stress. The analysis goal was to characterize the mechanical behavior of each joint and thus evaluate its mechanical supports. The joints are part of the toroidal field (TF) magnet system of the National Spherical Torus Experiment (NSTX) pulsed plasma device operating at the Princeton Plasma Physics Laboratory (PPPL). Since there is not sufficient space near the joints for much traditional mechanical instrumentation, small voltage probes were installed on each joint and their voltage monitoring waveforms have been recorded on sampling digitizers during each NSTX ''shot''

  13. Stabilizing effect of resistivity towards ELM-free H-mode discharge in lithium-conditioned NSTX

    Science.gov (United States)

    Banerjee, Debabrata; Zhu, Ping; Maingi, Rajesh

    2017-07-01

    Linear stability analysis of the national spherical torus experiment (NSTX) Li-conditioned ELM-free H-mode equilibria is carried out in the context of the extended magneto-hydrodynamic (MHD) model in NIMROD. The purpose is to investigate the physical cause behind edge localized mode (ELM) suppression in experiment after the Li-coating of the divertor and the first wall of the NSTX tokamak. Besides ideal MHD modeling, including finite-Larmor radius effect and two-fluid Hall and electron diamagnetic drift contributions, a non-ideal resistivity model is employed, taking into account the increase of Z eff after Li-conditioning in ELM-free H-mode. Unlike an earlier conclusion from an eigenvalue code analysis of these equilibria, NIMROD results find that after reduced recycling from divertor plates, profile modification is necessary but insufficient to explain the mechanism behind complete ELMs suppression in ideal two-fluid MHD. After considering the higher plasma resistivity due to higher Z eff, the complete stabilization could be explained. A thorough analysis of both pre-lithium ELMy and with-lithium ELM-free cases using ideal and non-ideal MHD models is presented, after accurately including a vacuum-like cold halo region in NIMROD to investigate ELMs.

  14. Edge Plasma Simulations in NSTX and CTF: Synergy of Lithium Coating, Non-Diffusive Anomalous Transport and Drifts. Final Technical Report

    International Nuclear Information System (INIS)

    Pigarov, Alexander

    2012-01-01

    This is the final report for the Research Grant DE-FG02-08ER54989 'Edge Plasma Simulations in NSTX and CTF: Synergy of Lithium Coating, Non-Diffusive Anomalous Transport and Drifts'. The UCSD group including: A.Yu. Pigarov (PI), S.I. Krasheninnikov and R.D. Smirnov, was working on modeling of the impact of lithium coatings on edge plasma parameters in NSTX with the multi-species multi-fluid code UEDGE. The work was conducted in the following main areas: (i) improvements of UEDGE model for plasma-lithium interactions, (ii) understanding the physics of low-recycling divertor regime in NSTX caused by lithium pumping, (iii) study of synergistic effects with lithium coatings and non-diffusive ballooning-like cross-field transport, (iv) simulation of experimental multi-diagnostic data on edge plasma with lithium pumping in NSTX via self-consistent modeling of D-Li-C plasma with UEDGE, and (v) working-gas balance analysis. The accomplishments in these areas are given in the corresponding subsections in Section 2. Publications and presentations made under the Grant are listed in Section 3.

  15. Biasing, acquisition, and interpretation of a dense Langmuir probe array in NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Jaworski, M. A.; Kallman, J.; Kaita, R.; Kugel, H.; LeBlanc, B.; Marsala, R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Ruzic, D. N. [Department of Nuclear, Plasma, and Radiological Engineering, University of Illinois at Urbana-Champaign, Urbana, Illinois 60181 (United States)

    2010-10-15

    A dense array of 99 Langmuir probes has been installed in the lower divertor region of the National Spherical Torus Experiment (NSTX). This array is instrumented with a system of electronics that allows flexibility in the choice of probes to bias as well as the type of measurement (including standard swept, single probe, triple probe, and operation as passive floating potential and scrape-off-layer SOL current monitors). The use of flush-mounted probes requires careful interpretation. The time dependent nature of the SOL makes swept-probe traces difficult to interpret. To overcome these challenges, the single- and triple-Langmuir probe signals are used in complementary fashion to determine the temperature and density at the probe location. A comparison to midplane measurements is made.

  16. Progress towards Steady State at Low Aspect Ratio on the National Spherical Torus Experiment (NSTX)

    International Nuclear Information System (INIS)

    Gates, D.A.; Menard, J.; Maingi, R.; Kaye, S.; Sabbagh, S.A.; Diem, S.; Wilson, J.R.; Bell, M.G.; Bell, R.E.; Ferron, J.; Fredrickson, E.D.; Kessel, C.E.; LeBlanc, B.P.; Levinton, F.; Manickam, J.; Mueller, D.; Raman, R.; Stevenson, T.; Stutman, D.; Taylor, G.; Tritz, K.; Yu, H.

    2007-01-01

    Modifications to the plasma control capabilities and poloidal field coils of the National Spherical Torus Experiment (NSTX) have enabled a significant enhancement in shaping capability which has led to the transient achievement of a record shape factor (S (triple b ond) q 95 (I p /aB t )) of ∼ 41 (MA m -1 T -1 ) simultaneous with a record plasma elongation of κ (triple b ond) b/a ∼ 3. This result was obtained using isoflux control and real-time equilibrium reconstruction. Achieving high shape factor together with tolerable divertor loading is an important result for future ST burning plasma experiments as exemplified by studies for future ST reactor concepts, as well as neutron producing devices, which rely on achieving high shape factors in order to achieve steady state operation while maintaining MHD stability. Statistical evidence is presented which demonstrates the expected correlation between increased shaping and improved plasma performance.

  17. Biasing, Acquisition and Interpretation of a Dense Langmuir Probe Array in NSTX

    International Nuclear Information System (INIS)

    Jaworski, M.A.; Kallman, J.; Kaita, R.; Kugel, H.; LeBlanc, B.; Marsala, R.; Ruzic, D.

    2010-01-01

    A dense array of 99 Langmuir probes has been installed in the lower divertor region of the National Spherical Torus Experiments (NSTX). This array is instrumented with a system of elec- tronics that allows flexibility in the choice of probes to bias as well as the type of measurement (including standard swept, single probe, triple probe and operation as passive floating potential and scrape-off-layer (SOL) current monitors). The use of flush-mounted probes requires careful inter- pretation. The time dependent nature of the SOL makes swept-probe traces difficult to interpret. To overcome these challenges, the single- and triple-Langmuir probe signals are used in comple- mentary fashion to determine the temperature and density at the probe location. A comparison to mid-plane measurements is made.

  18. Recent Fast Wave Coupling and Heating Studies on NSTX, with Possible Implications for ITER

    International Nuclear Information System (INIS)

    Hosea, J.C.; Bell, R.E.; Feibush, E.; Harvey, R.W.; Jaeger, E.F.; LeBlanc, B.P; Maingi, R.; Phillips, C.K.; Roquemore, L.; Ryan, P.M.; Taylor, G.; Tritz, K.; Valeo, E.J.; Wilgen, J.; Wilson, J.R.

    2009-01-01

    The goal of the high harmonic fast wave (HHFW) research on NSTX is to maximize the coupling of RF power to the core of the plasma by minimizing the coupling of RF power to edge loss processes. HHFW core plasma heating efficiency in helium and deuterium L-mode discharges is found to improve markedly on NSTX when the density 2 cm in front of the antenna is reduced below that for the onset of perpendicular wave propagation (n onset ∝ B*k # parallel# 2 /ω). In NSTX, the observed RF power losses in the plasma edge are driven in the vicinity of the antenna as opposed to resulting from multi-pass edge damping. PDI surface losses through ion-electron collisions are estimated to be significant. Recent spectroscopic measurements suggest that additional PDI losses could be caused by the loss of energetic edge ions on direct loss orbits and perhaps result in the observed clamping of the edge rotation. Initial deuterium H-mode heating studies reveal that core heating is degraded at lower k φ (- 8 m -1 relative to 13 m -1 ) as for the Lmode case at elevated edge density. Fast visible camera images clearly indicate that a major edge loss process is occurring from the plasma scrape off layer (SOL) in the vicinity of the antenna and along the magnetic field lines to the lower outer divertor plate. Large type I ELMs, which are observed at both k φ values, appear after antenna arcs caused by precursor blobs, low level ELMs, or dust. For large ELMs without arcs, the source reflection coefficients rise on a 0.1 ms time scale, which indicates that the time derivative of the reflection coefficient can be used to discriminate between arcs and ELMs.

  19. The ITER divertor concept

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  20. Special graphites

    International Nuclear Information System (INIS)

    Leveque, P.

    1964-01-01

    A large fraction of the work undertaken jointly by the Commissariat a l'Energie Atomique (CEA) and the Pechiney Company has been the improvement of the properties of nuclear pile graphite and the opening up of new fields of graphite application. New processes for the manufacture of carbons and special graphites have been developed: forged graphite, pyro-carbons, high density graphite agglomeration of graphite powders by cracking of natural gas, impervious graphites. The physical properties of these products and their reaction with various oxidising gases are described. The first irradiation results are also given. (authors) [fr

  1. NSTX-U Control System Upgrades

    International Nuclear Information System (INIS)

    Erickson, K.G.; Gates, D.A.; Gerhardt, S.P.; Lawson, J.E.; Mozulay, R.; Sichta, P.; Tchilinguirian, G.J.

    2014-01-01

    The National Spherical Tokamak Experiment (NSTX) is undergoing a wealth of upgrades (NSTX-U). These upgrades, especially including an elongated pulse length, require broad changes to the control system that has served NSTX well. A new fiber serial Front Panel Data Port input and output (I/O) stream will supersede the aging copper parallel version. Driver support for the new I/O and cyber security concerns require updating the operating system from Redhat Enterprise Linux (RHEL) v4 to RedHawk (based on RHEL) v6. While the basic control system continues to use the General Atomics Plasma Control System (GA PCS), the effort to forward port the entire software package to run under 64-bit Linux instead of 32-bit Linux included PCS modifications subsequently shared with GA and other PCS users. Software updates focused on three key areas: (1) code modernization through coding standards (C99/C11), (2) code portability and maintainability through use of the GA PCS code generator, and (3) support of 64-bit platforms. Central to the control system upgrade is the use of a complete real time (RT) Linux platform provided by Concurrent Computer Corporation, consisting of a computer (iHawk), an operating system and drivers (RedHawk), and RT tools (NightStar). Strong vendor support coupled with an extensive RT toolset influenced this decision. The new real-time Linux platform, I/O, and software engineering will foster enhanced capability and performance for NSTX-U plasma control

  2. NSTX-U Control System Upgrades

    Energy Technology Data Exchange (ETDEWEB)

    Erickson, K.G., E-mail: kerickso@pppl.gov; Gates, D.A.; Gerhardt, S.P.; Lawson, J.E.; Mozulay, R.; Sichta, P.; Tchilinguirian, G.J.

    2014-06-15

    The National Spherical Tokamak Experiment (NSTX) is undergoing a wealth of upgrades (NSTX-U). These upgrades, especially including an elongated pulse length, require broad changes to the control system that has served NSTX well. A new fiber serial Front Panel Data Port input and output (I/O) stream will supersede the aging copper parallel version. Driver support for the new I/O and cyber security concerns require updating the operating system from Redhat Enterprise Linux (RHEL) v4 to RedHawk (based on RHEL) v6. While the basic control system continues to use the General Atomics Plasma Control System (GA PCS), the effort to forward port the entire software package to run under 64-bit Linux instead of 32-bit Linux included PCS modifications subsequently shared with GA and other PCS users. Software updates focused on three key areas: (1) code modernization through coding standards (C99/C11), (2) code portability and maintainability through use of the GA PCS code generator, and (3) support of 64-bit platforms. Central to the control system upgrade is the use of a complete real time (RT) Linux platform provided by Concurrent Computer Corporation, consisting of a computer (iHawk), an operating system and drivers (RedHawk), and RT tools (NightStar). Strong vendor support coupled with an extensive RT toolset influenced this decision. The new real-time Linux platform, I/O, and software engineering will foster enhanced capability and performance for NSTX-U plasma control.

  3. Fast-wave power flow along SOL field lines in NSTX and the associated power deposition profile across the SOL in front of the antenna

    International Nuclear Information System (INIS)

    Perkins, R.J.; Bell, R.E.; Diallo, A.; Gerhardt, S.; Hosea, J.C.; Jaworski, M.A.; LeBlanc, B.P.; Kramer, G.J.; Maingi, R.; Phillips, C.K.; Podestà, M.; Roquemore, L.; Scotti, F.; Ahn, J.-W.; Gray, T.K.; Green, D.L.; McLean, A.; Ryan, P.M.; Jaeger, E.F.; Sabbagh, S.

    2013-01-01

    Fast-wave heating and current drive efficiencies can be reduced by a number of processes in the vicinity of the antenna and in the scrape-off layer (SOL). On NSTX from around 25% to more than 60% of the high-harmonic fast-wave power can be lost to the SOL regions, and a large part of this lost power flows along SOL magnetic field lines and is deposited in bright spirals on the divertor floor and ceiling. We show that field-line mapping matches the location of heat deposition on the lower divertor, albeit with a portion of the heat outside of the predictions. The field-line mapping can then be used to partially reconstruct the profile of lost fast-wave power at the midplane in front of the antenna, and the losses peak close to the last closed flux surface as well as the antenna. This profile suggests a radial standing-wave pattern formed by fast-wave propagation in the SOL, and this hypothesis will be tested on NSTX-U. RF codes must reproduce these results so that such codes can be used to understand this edge loss and to minimize RF heat deposition and erosion in the divertor region on ITER. (paper)

  4. A Neutral Beam Injector Upgrade for NSTX

    International Nuclear Information System (INIS)

    Stevenson, T.; McCormack, B.; Loesser, G.D.; Kalish, M.; Ramakrishnan, S.; Grisham, L.; Edwards, J.; Cropper, M.; Rossi, G.; Halle, A. von; Williams, M.

    2002-01-01

    The National Spherical Torus Experiment (NSTX) capability with a Neutral Beam Injector (NBI) capable of 80 kiloelectronvolt (keV), 5 Megawatt (MW), 5 second operation. This 5.95 million dollar upgrade reused a previous generation injector and equipment for technical, cost, and schedule reasons to obtain these specifications while retaining a legacy capability of 120 keV neutral particle beam delivery for shorter pulse lengths for possible future NSTX experiments. Concerns with NBI injection included power deposition in the plasma, aiming angles from the fixed NBI fan array, density profiles and beam shine through, orbit losses of beam particles, and protection of the vacuum vessel wall against beam impingement. The upgrade made use of the beamline and cryo panels from the Neutral Beam Test Stand facility, existing power supplies and controls, beamline components and equipment not contaminated by tritium during DT [deuterium-tritium] experiments, and a liquid Helium refrigerator plant to power and cryogenically pump a beamline and three ion sources. All of the Tokamak Fusion Test Reactor (TFTR) ion sources had been contaminated with tritium, so a refurbishment effort was undertaken on selected TFTR sources to rid the three sources destined for the NSTX NBI of as much tritium as possible. An interconnecting duct was fabricated using some spare and some new components to attach the beamline to the NSTX vacuum vessel. Internal vacuum vessel armor using carbon tiles was added to protect the stainless steel vacuum vessel from beam impingement in the absence of plasma and interlock failure. To date, the NBI has operated to 80 keV and 5 MW and has injected requested power levels into NSTX plasmas with good initial results, including high beta and strong heating characteristics at full rated plasma current

  5. NSTX-U Digital Coil Protection System Software Detailed Design

    Energy Technology Data Exchange (ETDEWEB)

    None

    2014-06-01

    The National Spherical Torus Experiment (NSTX) currently uses a collection of analog signal processing solutions for coil protection. Part of the NSTX Upgrade (NSTX-U) entails replacing these analog systems with a software solution running on a conventional computing platform. The new Digital Coil Protection System (DCPS) will replace the old systems entirely, while also providing an extensible framework that allows adding new functionality as desired.

  6. Conceptual design for the NSTX Central Instrumentation and Control System

    International Nuclear Information System (INIS)

    Bashore, D.; Oliaro, G.; Roney, P.; Sichta, P.; Tindall, K.

    1997-01-01

    The design and construction phase for the National Spherical Torus Experiment (NSTX) is under way at the Princeton Plasma Physics Laboratory (PPPL). Operation is scheduled to begin on April 30, 1999. This paper describes the conceptual design for the NSTX Central Instrumentation and Control (I and C) System. Major elements of the Central I and C System include the Process Control System, Plasma Control System, Network System, Data Acquisition System, and Synchronization System to support the NSTX experimental device

  7. Divertor characterization experiments

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  8. The JET divertor coil

    International Nuclear Information System (INIS)

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

    1989-01-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    International Nuclear Information System (INIS)

    Jaksic, Nikola; Greuner, Henri; Herrmann, Albrecht

    2013-01-01

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

  11. ELM induced divertor heat loads on TCV

    Science.gov (United States)

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

    2009-06-01

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

  12. ELM induced divertor heat loads on TCV

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-06-15

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

  13. The magnetic vapour shield effect at divertor plates during plasma disruptions

    International Nuclear Information System (INIS)

    Piazza, G.; Goel, B.; Hoebel, W.; Wuerz, H.; Landman, I.

    1995-01-01

    Hard disruptions in a TOKAMAK cause a large thermal load on the divertor plates with an instantaneous ablation of a part of the heated material. The produced vapour cloud screens the plasma facing component from the direct interaction with the disrupting plasma (vapour shield effect). In order to quantify the damage to the divertor the magneto-hydrodynamic behaviour of the expanding vapour cloud has been investigated using an extended version of the 1-dimensional Lagrangian hydrodynamic code KATACO. Modelling of the magnetic field effects on the expanding plasma takes into account that the magnetic field is oblique to the divertor (1 1/2 dimensional model). The ''Radiation Heat Conduction Approximation'' has been used for describing the radiative energy transport. In this paper results are presented assuming graphite as divertor material, irradiated with a proton beam of an energy density of 12MJ/m 2 and a duration of 100μs. (orig.)

  14. EBW simulation for MAST and NSTX experiments

    International Nuclear Information System (INIS)

    Preinhaelter, J.; Urban, J.; Pavlo, P.; Taylor, G.; Shevchenko, V.; Valovic, M.; Vahala, L.; Vahala, G.

    2005-01-01

    The interpretation of EBW emission from spherical tokamaks is nontrivial. We report on a 3D simulation model of this process that incorporates Gaussian beams for the antenna, a full wave solution of EBW-X and EBW-X-O conversions using adaptive finite elements, and EBW ray tracing to determine the radiative temperature. This model is then used to interpret the experimental results from MAST and NSTX. EBW for ELM free H-modes in MAST suggests that the magnetic equilibrium determined by the EFIT code does not adequately represent the B-field within the transport barrier. Using the EBW signal for the reconstruction of the radial profile of the magnetic field, we determine a new equilibrium and see that the EBW simulation now yields better agreement with experimental results. EBW simulations yield excellent results for the time development of the plasma temperature as measured by the EBW radiometer on NSTX

  15. Neutral Particle Analyzer Diagnostic on NSTX

    International Nuclear Information System (INIS)

    Medley, S.S.; Roquemore, A.L.

    2004-01-01

    The Neutral Particle Analyzer (NPA) diagnostic on the National Spherical Torus Experiment (NSTX) utilizes a PPPL-designed E||B spectrometer that measures the energy spectra of minority hydrogen and bulk deuterium species simultaneously with 39 energy channels per mass specie and a time resolution of 1 ms. The calibrated energy range is E = 0.5-150 keV and the energy resolution varies from AE/E = 3-7% over the surface of the microchannel plate detector

  16. Progress towards Steady State on NSTX

    International Nuclear Information System (INIS)

    Gates, D.A.; Kessel, C.; Menard, J.; Taylor, G.; Wilson, J.R.

    2005-01-01

    In order to reduce recirculating power fraction to acceptable levels, the spherical torus concept relies on the simultaneous achievement of high toroidal β and high bootstrap fraction in steady state. In the last year, as a result of plasma control system improvements, the achievable plasma elongation on the National Spherical Torus Experiment (NSTX) has been raised from κ ∼ 2.1 to κ ∼ 2.6--approximately a 25% increase. This increase in elongation has lead to a doubling increase in the toroidal β for long-pulse discharges. The increase in β is associated with an increase in plasma current at nearly fixed poloidal β, which enables higher β t with nearly constant bootstrap fraction. As a result, for the first time in a spherical torus, a discharge with a plasma current of 1 MA has been sustained for 1 second. Data is presented from NSTX correlating the increase in performance with increased plasma shaping capability. In addition to improved shaping, H-modes induced during the current ramp phase of the plasma discharge have been used to reduce flux consumption during and to delay the onset of MHD instabilities. A modeled integrated scenario, which has 100% non-inductive current drive with very high toroidal β, will also be presented. The NSTX poloidal field coils are currently being modified to produce the plasma shape which is required for this scenario, which requires high triangularity ((delta) ∼ 0.8) at elevated elongation (κ ∼ 2.5). The other main requirement for steady state on NSTX is the ability to drive a fraction of the total plasma current with radio-frequency waves. The results of High Harmonic Fast Wave heating and current drive studies as well as electron Bernstein Wave emission studies will be presented

  17. Neutral Particle Analyzer Diagnostic on NSTX

    Energy Technology Data Exchange (ETDEWEB)

    S.S. Medley; A.L. Roquemore

    2004-03-16

    The Neutral Particle Analyzer (NPA) diagnostic on the National Spherical Torus Experiment (NSTX) utilizes a PPPL-designed E||B spectrometer that measures the energy spectra of minority hydrogen and bulk deuterium species simultaneously with 39 energy channels per mass specie and a time resolution of 1 ms. The calibrated energy range is E = 0.5-150 keV and the energy resolution varies from AE/E = 3-7% over the surface of the microchannel plate detector.

  18. Images of Edge Turbulence in NSTX

    International Nuclear Information System (INIS)

    Zweben, S.J.; Bush, C.E.; Maqueda, R.; Munsat, T.; Stotler, D.; Lowrance, J.; Mastracola, V.; Renda, G.

    2004-01-01

    The 2-D structure of edge plasma turbulence has been measured in the National Spherical Torus Experiment (NSTX) by viewing the emission of the Da spectral line of deuterium. Images have been made at framing rates of up to 250,000 frames/sec using an ultra-high speed CCD camera developed by Princeton Scientific Instruments. A sequence of images showing the transition between L-mode and H-mode states is shown

  19. The NSTX Central Instrumentation and Control System

    International Nuclear Information System (INIS)

    G. Oliaro; J. Dong; K. Tindall; P. Sichta

    1999-01-01

    Earlier this year the National Spherical Torus Experiment (NSTX) at the Princeton Plasma Physics Laboratory achieved ''first plasma''. The Central Instrumentation and Control System was used to support plasma operations. Major elements of the system include the Process Control System, Plasma Control System, Network System, Data Acquisition System, and Synchronization System. This paper will focus on the Process Control System. Topics include the architecture, hardware interface, operator interface, data management, and system performance

  20. Recent EBW Emission Results on NSTX

    Czech Academy of Sciences Publication Activity Database

    Diem, S.J.; Taylor, G.; Efthimion, P.C.; LeBlanc, B.P.; Caughman, J.B.; Bigelow, T.S.; Wilgen, J.B.; Harvey, R.W.; Preinhaelter, Josef; Urban, Jakub; Sabbagh, S.A.

    2007-01-01

    Roč. 52, č. 16 (2007), s. 63-63 ISSN 0003-0503. [Annual Meeting of the Division of Plasma Physics/49th./. Orlando , Florida, 12.11.2007-16.11.2007] Institutional research plan: CEZ:AV0Z20430508 Keywords : Conversion * Emission * Tokamaks * Electron Bernstein waves * Simulation * NSTX Subject RIV: BL - Plasma and Gas Discharge Physics http://meetings.aps.org/Meeting/DPP07/Content/901

  1. CHI Research on NSTX-U

    Science.gov (United States)

    Lay, W.-S.; Raman, R.; Jarboe, T. R.; Nelson, B. A.; Mueller, D.; Ebrahimi, F.; Ono, M.; Jardin, S. C.; Taylor, G.

    2017-10-01

    At present about 20% of the total plasma current required for sustained operation has been generated by transient CHI. The present understanding suggests that it may be possible to generate all of the needed current in a ST / tokamak using transient CHI. In such a scenario, one could transition directly from a CHI produced plasma to a non-inductively sustained plasma, without the difficult intermediate step that involves non-inductive current ramp-up. STs based on this new configuration would take advantage of evolving developments in high-temperature superconductor technology to develop a simpler design ST that relies primarily on CHI for plasma current generation. Motivated by the very good results from NSTX and HIT-II, we are examining the potential application of transient CHI for reactor configurations through these studies. (1) Study of the maximum levels of start-up currents that could be generated on NSTX-U, (2) application of a single biased electrode configuration on QUEST to protect the insulator from neutron damage in a CHI reactor installation, and (3) QUEST-like, but a double biased electrode configuration for PEGASUS and NSTX-U. Results from these on-going studies will be described. This work is supported by U.S. DOE Contracts: DE-AC02-09CH11466, DE-FG02-99ER54519 AM08, and DE-SC0006757.

  2. NSTX Diagnostics for Fusion Plasma Science Studies

    International Nuclear Information System (INIS)

    Kaita, R.; Johnson, D.; Roquemore, L.; Bitter, M.; Levinton, F.; Paoletti, F.; Stutman, D.

    2001-01-01

    This paper will discuss how plasma science issues are addressed by the diagnostics for the National Spherical Torus Experiment (NSTX), the newest large-scale machine in the magnetic confinement fusion (MCF) program. The development of new schemes for plasma confinement involves the interplay of experimental results and theoretical interpretations. A fundamental requirement, for example, is a determination of the equilibria for these configurations. For MCF, this is well established in the solutions of the Grad-Shafranov equation. While it is simple to state its basis in the balance between the kinetic and magnetic pressures, what they are as functions of space and time are often not easy to obtain. Quantities like the plasma pressure and current density are not directly measurable. They are derived from data that are themselves complex products of more basic parameters. The same difficulties apply to the understanding of plasma instabilities. Not only are the needs for spatial and temporal resolution more stringent, but the wave parameters which characterize the instabilities are difficult to resolve. We will show how solutions to the problems of diagnostic design on NSTX, and the physics insight the data analysis provides, benefits both NSTX and the broader scientific community

  3. Simulation Of Microtearing Turbulence In NSTX

    International Nuclear Information System (INIS)

    Guttenfelder, W.; Candy, J.; Kaye, S.M.; Nevins, W.M.; Wanag, E.; Zhang, J.; Bell, R.E.; Crocker, N.A.; Hammett, G.W.; LeBlanc, B.P.; Mikkelsen, D.R.; Ren, Y.; Yuh, H.

    2012-01-01

    Thermal energy confinement times in NSTX dimensionless parameter scans increase with decreasing collisionality. While ion thermal transport is neoclassical, the source of anomalous electron thermal transport in these discharges remains unclear, leading to considerable uncertainty when extrapolating to future ST devices at much lower collisionality. Linear gyrokinetic simulations find microtearing modes to be unstable in high collisionality discharges. First non-linear gyrokinetic simulations of microtearing turbulence in NSTX show they can yield experimental levels of transport. Magnetic flutter is responsible for almost all the transport (∼98%), perturbed field line trajectories are globally stochastic, and a test particle stochastic transport model agrees to within 25% of the simulated transport. Most significantly, microtearing transport is predicted to increase with electron collisionality, consistent with the observed NSTX confinement scaling. While this suggests microtearing modes may be the source of electron thermal transport, the predictions are also very sensitive to electron temperature gradient, indicating the scaling of the instability threshold is important. In addition, microtearing turbulence is susceptible to suppression via sheared E-B flows as experimental values of E-B shear (comparable to the linear growth rates) dramatically reduce the transport below experimental values. Refinements in numerical resolution and physics model assumptions are expected to minimize the apparent discrepancy. In cases where the predicted transport is strong, calculations suggest that a proposed polarimetry diagnostic may be sensitive to the magnetic perturbations associated with the unique structure of microtearing turbulence.

  4. Boronization on NSTX using Deuterated Trimethylboron

    International Nuclear Information System (INIS)

    Blanchard, W.R.; Gernhardt, R.C.; Kugel, H.W.; LaMarche, P.H.

    2002-01-01

    Boronization on the National Spherical Torus Experiment (NSTX) has proved to be quite beneficial with increases in confinement and density, and decreases in impurities observed in the plasma. The boron has been applied to the interior surfaces of NSTX, about every 2 to 3 weeks of plasma operation, by producing a glow discharge in the vacuum vessel using deuterated trimethylboron (TMB) in a 10% mixture with helium. Special NSTX requirements restricted the selection of the candidate boronization method to the use of deuterated boron compounds. Deuterated TMB met these requirements, but is a hazardous gas and special care in the execution of the boronization process is required. This paper describes the existing GDC, Gas Injection, and Torus Vacuum Pumping System hardware used for this process, the glow discharge process, and the automated control system that allows for remote operation to maximize both the safety and efficacy of applying the boron coating. The administrative requirements and the detailed procedure for the setup, operation and shutdown of the process are also described

  5. Initial operation of NSTX with plasma control

    International Nuclear Information System (INIS)

    Gates, D.; Bell, M.; Ferron, J.; Kaye, S.; Menard, J.; Mueller, D.; Neumeyer, C.; Sabbagh, S.

    2000-01-01

    First plasma, with a maximum current of 300kA, was achieved on NSTX in February 1999. These results were obtained using preprogrammed coil currents. The first controlled plasmas on NSTX were made starting in August 1999 with the full 1MA plasma current achieved in December 1999. The controlled quantities were plasma position (R, Z) and current (Ip). Variations in the plasma shape are achieved by adding preprogrammed currents to those determined by the control parameters. The control system is fully digital, with plasma position and current control, data acquisition, and power supply control all occurring in the same four-processor real time computer. The system uses the PCS (Plasma Control Software) system designed at General Atomics. Modular control algorithms, specific to NSTX, were written and incorporated into the PCS. The application algorithms do the actual control calculations, with the PCS handling data passing. The control system, including planned upgrades, will be described, along with results of the initial controlled plasma operations. Analysis of the performance of the control system will also be presented

  6. Development of NSTX Particle Control Techniques

    International Nuclear Information System (INIS)

    Kugel, H.W.; Maingi, R.; Bell, M.; Gates, D.; Hill, K.; LeBlanc, B.; Mueller, D.; Kaita, R.; Paul, S.; Sabbagh, S.; Skinner, C.H.; Soukhanovskii, V.; Stratton, B.; Raman, R.

    2004-01-01

    The National Spherical Torus Experiment (NSTX) High Harmonic Fast Wave (HHFW) current-drive discharges will require density control for acceptable efficiency. In NSTX, this involves primarily controlling impurity influxes and recycling. We have compared boronization on hot and cold surfaces, varying helium glow discharge conditioning (HeGDC) durations, helium discharge cleaning, brief daily boronization, and between discharge boronization to reduce and control spontaneous density rises. Access to Ohmic H-modes was enabled by boronization on hot surfaces, however, the duration of the effectiveness of hot and cold boronization was comparable. A 15 minute HeGDC between discharges was needed for reproducible L-H transitions. Helium discharge conditioning yielded slower density rises than 15 minutes of HeGDC. Brief daily boronization followed by a comparable duration of applied HeGDC restored and enhanced good conditions. Additional brief boronizations between discharges did not improve plasma performance (reduced recycling, reduced impurity luminosities, earlier L-H transitions, longer plasma current flattops, higher stored energies) if conditions were already good. Between discharge boronization required increases in the NSTX duty cycle due to the need for additional HeGDC to remove codeposited D

  7. Status of National Spherical Torus Experiment (NSTX)*

    Science.gov (United States)

    Ono, Masayuki

    2001-10-01

    The main aim of National Spherical Torus Experiment (NSTX) is to establish the fusion physics principles of the innovative spherical torus (ST) concept. The NSTX experimental facility has been operating reliably and its capabilities steadily improving. Due to relatively efficient ohmic current drive and benign halo current behavior, the plasma current was increased to 1.4 MA, which is well above the design value of 1 MA. The plasmas at 1 MA are now routinely heated by NBI to the average toroidal beta value of 20 percent range at 3 kG with electrons and ions in the 1-2 keV range. Even with the “L-mode” edge, the energy confinement time can well exceed the so-called L-mode (and even H-mode) scaling values. As a part of ST tool development, High Harmonic Fast Wave (HHFW) heating has demonstrated efficient electron heating with the central electron temperatures reaching 3.7 keV. HHFW induced H-modes have been also observed. For CHI (Coaxial Helicity Injection) non-inductive start-up, CHI discharges of up to 300 kA of toroidal current and 300 msec duration have been produced from zero current using = 25 kA of injected current. The poster presentation will also include the near term NSTX facility upgrade plan.

  8. Numerical studies on divertor experiments

    International Nuclear Information System (INIS)

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

    1988-04-01

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

  9. Electron Bernstein Wave Research on NSTX and CDX-U

    International Nuclear Information System (INIS)

    Taylor, G.; Efthimion, P.C.; Jones, B.; Bell, G.L.; Bers, A.; Bigelow, T.S.; Carter, M.D.; Harvey, R.W.; Ram, A.K.; Rasmussen, D.A.; Smirnov, A.P.; Wilgen, J.B.; Wilson, J.R.

    2003-01-01

    Studies of thermally emitted electron Bernstein waves (EBWs) on CDX-U and NSTX, via mode conversion (MC) to electromagnetic radiation, support the use of EBWs to measure the Te profile and provide local electron heating and current drive (CD) in overdense spherical torus plasmas. An X-mode antenna with radially adjustable limiters successfully controlled EBW MC on CDX-U and enhanced MC efficiency to ∼ 100%. So far the X-mode MC efficiency on NSTX has been increased by a similar technique to 40-50% and future experiments are focused on achieving * 80% MC. MC efficiencies on both machines agree well with theoretical predictions. Ray tracing and Fokker-Planck modeling for NSTX equilibria are being conducted to support the design of a 3 MW, 15 GHz EBW heating and CD system for NSTX to assist non-inductive plasma startup, current ramp up, and to provide local electron heating and CD in high beta NSTX plasmas

  10. Innovations in the LHD divertor program

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  11. High temperature divertor plasma operation

    International Nuclear Information System (INIS)

    Ohyabu, Nobuyoshi.

    1991-02-01

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

  12. Metal/graphite - composites in fusion engineering

    International Nuclear Information System (INIS)

    Staffler, R.; Kneringer, G.; Kny, E.; Reheis, N.

    1989-01-01

    Metal/graphite composites have been well known in medical industry for many years. X-ray tubes used in modern radiography, particularly in computerized tomography are equipped with rotating targets able to absorb a maximum of heat in a given time. Modern rotating targets consist of a refractory metal/graphite composite. Today the use of graphite as a plasma facing material is one predominant concept in fusion engineering. Depending on the thermal load, the graphite components have to be directly cooled (i.e. divertor plates) or inertially cooled (i.e. firstwall tiles). In case of direct cooling a metallurgical joining such as high temperature brazing between graphite and a metallic cooling structure shows the most promising results /1/. Inertially cooled graphite tiles have to be joined to a metallic backing plate in order to get a stable attachment to the supporting structure. The main requirements on the metallic partner of a metal/graphite composite used in the first wall area are: high melting point, high thermal strength, high thermal conductivity, low vapor pressure and a thermal expansion matching that of graphite. These properties are typical for the refractory metals such as molybdenum, tungsten and their alloys. 4 refs., 13 figs., 1 tab

  13. Metal/graphite - composites in fusion engineering

    International Nuclear Information System (INIS)

    Staffler, R.; Kneringer, G.; Kny, E.; Reheis, N.

    1995-01-01

    Metal/graphite composites have been well known in medical industry for many years. X-ray tubes used in modern radiography, particulary in computerized tomography are equipped with rotating targets able to absorb a maximum of heat in a given time. Modern rotating targets consist of a refractory metal/graphite composite. Today the use of graphite as a plasma facing material is one predominant concept in fusion engineering. Depending on the thermal load, the graphite components have to be directly cooled (i.e. divertor plates) or inertially cooled (i.e. firstwall tiles). In case of direct cooling a metallurgical joining such as high temperature brazing between graphite and a metalic cooling structure shows the most promising results /1/. Inertially cooled graphite tiles have to be joined to a metallic backing plate in order to get a stable attachment to the supporting structure. The main requirements on the metallic partner of a metal/graphite composite and in the first wall area are: high melting point, high thermal strength, high thermal conductivity, low vapour pressure and a thermal expansion matching that of graphite. These properties are typical for the refractory metals such as molybdenum, tungsten and their alloys. (author)

  14. Divertor development for ITER

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  15. The divertor remote maintenance project

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  16. The divertor remote maintenance project

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  17. Heating and current drive on NSTX

    Science.gov (United States)

    Wilson, J. R.; Batchelor, D.; Carter, M.; Hosea, J.; Ignat, D.; LeBlanc, B.; Majeski, R.; Ono, M.; Phillips, C. K.; Rogers, J. H.; Schilling, G.

    1997-04-01

    Low aspect ratio tokamaks pose interesting new challenges for heating and current drive. The NSTX (National Spherical Tokamak Experiment) device to be built at Princeton is a low aspect ratio toroidal device that has the achievement of high toroidal beta (˜45%) and non-inductive operation as two of its main research goals. To achieve these goals significant auxiliary heating and current drive systems are required. Present plans include ECH (Electron cyclotron heating) for pre-ionization and start-up assist, HHFW (high harmonic fast wave) for heating and current drive and eventually NBI (neutral beam injection) for heating, current drive and plasma rotation.

  18. Heating and current drive on NSTX

    International Nuclear Information System (INIS)

    Wilson, J.R.; Batchelor, D.; Carter, M.; Hosea, J.; Ignat, D.; LeBlanc, B.; Majeski, R.; Ono, M.; Phillips, C.K.; Rogers, J.H.; Schilling, G.

    1997-01-01

    Low aspect ratio tokamaks pose interesting new challenges for heating and current drive. The NSTX (National Spherical Tokamak Experiment) device to be built at Princeton is a low aspect ratio toroidal device that has the achievement of high toroidal beta (∼45%) and non-inductive operation as two of its main research goals. To achieve these goals significant auxiliary heating and current drive systems are required. Present plans include ECH (Electron cyclotron heating) for pre-ionization and start-up assist, HHFW (high harmonic fast wave) for heating and current drive and eventually NBI (neutral beam injection) for heating, current drive and plasma rotation. copyright 1997 American Institute of Physics

  19. Visible imaging of edge turbulence in NSTX

    International Nuclear Information System (INIS)

    Zweben, S.; Maqueda, R.; Hill, K.; Johnson, D.

    2000-01-01

    Edge plasma turbulence in tokamaks and stellarators is believed to cause the radical heat and particle flux across the separatrix and into the scrape-off-layers of these devices. This paper describes initial measurements of 2-D space-time structure of the edge density turbulence made using a visible imaging diagnostic in the National Spherical Torus Experiment (NSTX). The structure of the edge turbulence is most clearly visible using a method of gas puff imaging to locally illuminate the edge density turbulence

  20. Overview of physics results from NSTX

    Czech Academy of Sciences Publication Activity Database

    Raman, R.; Ahn, J-W.; Allain, J.P.; Andre, R.; Bastasz, R.; Battaglia, D.; Beiersdorfer, P.; Bell, M.; Bell, R.; Belova, E.; Berkery, J.; Betti, R.; Bialek, J.; Bigelow, T.; Bitter, M.; Boedo, J.; Bonoli, P.; Boozer, A.; Bortolon, A.; Brennan, D.; Breslau, J.; Buttery, R.; Canik, J.; Caravelli, G.; Chang, C.; Crocker, N.A.; Darrow, D.; Davis, W.; Delgado-Aparicio, L.; Diallo, A.; Ding, S.; D’Ippolito, D.; Domier, C.; Dorland, W.; Ethier, S.; Evans, T.; Ferron, J.; Finkenthal, M.; Foley, J.; Fonck, R.; Frazin, R.; Fredrickson, E.; Fu, G.; Gates, D.; Gerhardt, S.; Glasser, A.; Gorelenkov, N.; Gray, T.; Guo, Y.; Guttenfelder, W.; Hahm, T.; Harvey, R.; Hassanein, A.; Heidbrink, W.; Hill, K.; Hirooka, Y.; Hooper, E.B.; Hosea, J.; Hu, B.; Humphreys, D.; Indireshkumar, K.; Jaeger, F.; Jarboe, T.; Jardin, S.; Jaworski, M.; Kaita, R.; Kallman, J.; Katsuro-Hopkins, O.; Kaye, S.; Kessel, C.; Kim, J.; Kolemen, E.; Krasheninnikov, S.; Kubota, S.; Kugel, H.; La Haye, R.; Lao, L.; LeBlanc, B.; Lee, W.; Lee, K.; Leuer, J.; Levinton, F.; Liang, Y.; Liu, D.; Luhmann Jr, N.; Maingi, R.; Majeski, R.; Manickam, J.; Mansfield, D.; Maqueda, R.; Mazzucato, E.; McLean, A.; McCune, D.; McGeehan, B.; McKee, G.; Medley, S.; Menard, J.; Menon, M.; Meyer, H.; Mikkelsen, D.; Miloshevsky, G.; Mueller, D.; Munsat, T.; Myra, J.; Nelson, B.; Nishino, N.; Nygren, R.; Ono, M.; Osborne, T.; Park, H.; Park, J.; Paul, S.; Peebles, W.; Penaflor, B.; Phillips, C.; Pigarov, A.; Podesta, M.; Preinhaelter, Josef; Ren, Y.; Reimerdes, H.; Ross, P.; Rowley, C.; Ruskov, E.; Russell, D.; Ruzic, D.; Ryan, P.; Sabbagh, S.A.; Schaffer, M.; Schuster, E.; Scotti, F.; Shaing, K.; Shevchenko, V.; Shinohara, K.; Sizyuk, V.; Skinner, C.H.; Smirnov, A.; Smith, D.; Snyder, P.; Solomon, W.; Sontag, A.; Soukhanovskii, V.; Stoltzfus-Dueck, T.; Stotler, D.; Stratton, B.; Stutman, D.; Takahashi, H.; Takase, Y.; Tamura, N.; Tang, X.; Taylor, C.N.; Taylor, G.; Taylor, C.; Tritz, K.; Tsarouhas, D.; Umansky, M.; Urban, Jakub; Walker, M.; Wampler, W.; Wang, W.; Whaley, J.; White, R.; Wilgen, J.; Wilson, R.; Wong, K.L.; Wright, J.; Xia, Z.; Youchison, D.; Yu, H.; Yuh, H.; Zakharov, L.; Zemlyanov, D.; Zimmer, G.; Zweben, S.J.

    2011-01-01

    Roč. 51, č. 9 (2011), 094011-094011 ISSN 0029-5515. [Fusion Energy Conference (FEC 2010)/23rd./. Daejon, 11.10.2010-16.10.2010] R&D Projects: GA ČR GA202/08/0419; GA MŠk 7G09042 Institutional research plan: CEZ:AV0Z20430508 Keywords : NSTX * Spherical tokamaks * Overdense plasma * Conversion * Emission * Tokamaks * Elektron Bernstein waves Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.090, year: 2011 http://iopscience.iop.org/0029-5515/51/9/094011/pdf/0029-5515_51_9_094011.pdf

  1. Transport in Auxiliary Heated NSTX Discharges

    International Nuclear Information System (INIS)

    LeBlanc, B.P.; Bell, M.G.; Bell, R.E.; Bitte, M.L.; Bourdelle, C.; Gates, D.A.; Kaye, S.M.; Maingi, R.; Menard, J.E.; Mueller, D.; Ono, M.; Paul, S.F.; Redi, M.H.; Roquemore, A.L.; Rosenberg, A.; Sabbagh, S.A.; Stutman, D.; Synakowski, E.J.; Soukhanovskii, V.A.; Wilson, J.R.

    2003-01-01

    The NSTX spherical torus (ST) provides a unique platform to investigate magnetic confinement in auxiliary-heated plasmas at low aspect ratio. Auxiliary power is routinely coupled to ohmically heated plasmas by deuterium neutral-beam injection (NBI) and by high-harmonic fast waves (HHFW) launch. While theory predicts both techniques to preferentially heat electrons, experiment reveals the electron temperature is greater than the ion temperature during HHFW, but the electron temperature is less than the ion temperature during NBI. In the following we present the experimental data and the results of transport analyses

  2. Operation of the NSTX Thomson Scattering System

    International Nuclear Information System (INIS)

    LeBlanc, B.P.; Bell, R.E.; Johnson, D.W.; Hoffman, D.E.; Long, D.C.; Palladino, R.W.

    2002-01-01

    The NSTX multi-point Thomson scattering system has been in operation for nearly two years and provides routine Te(R,t) and ne(R,t) measurements. The laser beams from two 30-Hz Nd:YAG lasers are imaged by a spherical mirror onto 36 fiber-optics bundles. In the present configuration, the output ends of 20 of these bundles are instrumented with filter polychromators and avalanche photodiode detectors. In this paper, we discuss the laser implementation and the installed collection optics. We follow with examples of raw and analyzed data. We close with some comments about calibration

  3. Visible imaging of edge turbulence in NSTX

    International Nuclear Information System (INIS)

    S. Zweben; R. Maqueda; K. Hill; D. Johnson; S. Kaye; H. Kugel; F. Levinton; R. Maingi; L. Roquemore; S. Sabbagh; G. Wurden

    2000-01-01

    Edge plasma turbulence in tokamaks and stellarators is believed to cause the radial heat and particle flux across the separatrix and into the scrape-off-layers of these devices. This paper describes initial measurements of 2-D space-time structure of the edge density turbulence made using a visible imaging diagnostic in the National Spherical Torus Experiment (NSTX). The structure of the edge turbulence is most clearly visible using a method of ''gas puff imaging'' to locally illuminate the edge density turbulence

  4. Some problems of brazing technology for the divertor plate manufacturing

    Science.gov (United States)

    Prokofiev, Yu. G.; Barabash, V. R.; Khorunov, V. F.; Maksimova, S. V.; Gervash, A. A.; Fabritsiev, S. A.; Vinokurov, V. F.

    1992-09-01

    Among the different design options of the ITER reactor divertor, the joints of the carbon-based materials and molybdenum alloys and joints of tungsten and copper alloys are considered. High-temperature brazing is one of the most promising joining methods for the plasma facing and heat sink materials. The use of brazing for creation of W-Cu and graphite-Mo joints are given here. In addition, the investigation results of microstructure, microhardness and mechanical properties of the joints are presented. For W-Cu samples an influence of the neutron irradiation on the joining strength was studied.

  5. Some problems of brazing technology for the divertor plate manufacturing

    International Nuclear Information System (INIS)

    Prokofiev, Yu.G.; Barabash, V.R.; Gervash, A.A.; Khorunov, V.F.; Maksimova, S.V.; Vinokurov, V.F.; Fabritsiev, S.A.

    1992-01-01

    Among the different design options of the ITER reactor divertor, the joints of the carbon-based materials and molybdenum alloys and joints of tungsten and copper alloys are considered. High-temperature brazing is one of the most promising joining methods for the plasma facing and heat sink materials. The use of brazing for creation of W-Cu and graphite-Mo joints are given here. In addition, the investigation results of microstructure, microhardness and mechanical properties of the joints are presented. For W-Cu samples an influence of the neutron irradiation on the joining strength was studied. (orig.)

  6. Some problems of brazing technology for the divertor plate manufacturing

    Energy Technology Data Exchange (ETDEWEB)

    Prokofiev, Yu.G.; Barabash, V.R.; Gervash, A.A. (D.V. Efremov Scientific Research Inst. of Electrophysical Apparatus, St. Petersburg (Russia)); Khorunov, V.F.; Maksimova, S.V. (E.O. Paton Inst. of Electronwelding, Kiev (Ukraine)); Vinokurov, V.F. (Central Scientific Research Inst. of Structural Materials ' Prometey' , St. Petersburg (Russia)); Fabritsiev, S.A.

    1992-09-01

    Among the different design options of the ITER reactor divertor, the joints of the carbon-based materials and molybdenum alloys and joints of tungsten and copper alloys are considered. High-temperature brazing is one of the most promising joining methods for the plasma facing and heat sink materials. The use of brazing for creation of W-Cu and graphite-Mo joints are given here. In addition, the investigation results of microstructure, microhardness and mechanical properties of the joints are presented. For W-Cu samples an influence of the neutron irradiation on the joining strength was studied. (orig.).

  7. Control System for the NSTX Lithium Pellet Injector

    International Nuclear Information System (INIS)

    Sichta, P.; Dong, J.; Gernhardt, R.; Gettelfinger, G.; Kugel, H.

    2003-01-01

    The Lithium Pellet Injector (LPI) is being developed for the National Spherical Torus Experiment (NSTX). The LPI will inject ''pellets'' of various composition into the plasma in order to study wall conditioning, edge impurity transport, liquid limiter simulations, and other areas of research. The control system for the NSTX LPI has incorporated widely used advanced technologies, such as LabVIEW and PCI bus I/O boards, to create a low-cost control system which is fully integrated into the NSTX computing environment. This paper will present the hardware and software design of the computer control system for the LPI

  8. The use of MDSplus on NSTX at PPPL

    International Nuclear Information System (INIS)

    Davis, W.; Roney, P.; Carroll, T.; Gibney, T.; Mastrovito, D.

    2002-01-01

    The MDSplus data acquisition system has been used successfully since the 1999 startup of NSTX for control, data acquisition and analysis for diagnostic subsystems. For each plasma 'shot' on NSTX about 75 MBs of data is acquired and loaded into MDSplus hierarchical data structures in 2-3 min. Physicists adapted to the MDSplus software tools with no real difficulty. Some locally developed tools are described. The support from the developers at MIT was timely and insightful. The use of MDSplus has resulted in significant cost savings for NSTX

  9. The Use of MDSplus on NSTX at PPPL; TOPICAL

    International Nuclear Information System (INIS)

    W. Davis; P. Roney; T. Carroll; T. Gibney; D. Mastrovito

    2002-01-01

    The MDSplus data acquisition system has been used successfully since the 1999 startup of NSTX[National Spherical Torus Experiment] for control, data acquisition, and analysis for diagnostic subsystems. For each plasma ''shot'' on NSTX about 75 MBs of data is acquired and loaded into MDSplus hierarchical data structures in 2-3 minutes. Physicists adapted to the MDSplus software tools with no real difficulty. Some locally developed tools are described. The support from the developers at MIT[Massachusetts Institute of Technology] was timely and insightful. The use of MDSplus has resulted in a significant cost savings for NSTX

  10. The Use of MDSplus on NSTX at PPPL

    International Nuclear Information System (INIS)

    Davis, W.; Roney, P.; Carroll, T.; Gibney, T.; Mastrovito, D.

    2002-01-01

    The MDSplus data acquisition system has been used successfully since the 1999 startup of NSTX [National Spherical Torus Experiment] for control, data acquisition, and analysis for diagnostic subsystems. For each plasma ''shot'' on NSTX about 75 MBs of data is acquired and loaded into MDSplus hierarchical data structures in 2-3 minutes. Physicists adapted to the MDSplus software tools with no real difficulty. Some locally developed tools are described. The support from the developers at MIT [Massachusetts Institute of Technology] was timely and insightful. The use of MDSplus has resulted in a significant cost savings for NSTX

  11. Study of high-Z target plate materials in the divertor of ASDEX-Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Hirsch, S; Asmussen, K; Engelhardt, W; Field, A R; Fussmann, G; Lieder, G; Naujoks, D; Neu, R; Radtke, R; Wenzel, U [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

    1994-12-31

    The reduction of divertor tile erosion is a challenging problem in present and future tokamaks. Until now, graphite has almost exclusively been used for divertor plates, and it is estimated that unacceptable amounts of material would be eroded under reactor relevant conditions where power fluxes to the target plates as high as 20 MW/m{sup 2} are expected. In a high-recycling divertor with relatively low temperature (5 eVdivertor plates, in-situ studies of the erosion of various divertor target materials have been performed by means of passive spectroscopy. From our spectroscopic observations we infer that under high density divertor conditions the advantages of high-Z materials become fully efficient. (author) 6 refs., 2 figs.

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

    International Nuclear Information System (INIS)

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

    2009-01-01

    Recent National Spherical Tokamak Experiment (NSTX) high-power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components (PFCs) to the performance of divertor plasmas in both L- and H-mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15 25% ne decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, ne/nGW 1), to enable ne scan capability (2) in the H-mode, to test the ability to operate at significantly lower density (e.g., ne/nGW = 0.25), for future reactor designs based on the Spherical Tokamak, and eventually to investigate high heat-flux power handling (10 MW/m2) with long pulse discharges (>1.5 s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-09-26

    Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on PFC's to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15-25% ne decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, ne/nGW~1), to enable ne scan capability (x2) in the H-mode, to test the ability to operate at significantly lower density for future ST-CTF reactor designs (e.g., ne/nGW = 0.25), and eventually to investigate high heat-flux power handling (10 MW/m2) with longpulse discharges (>1.5s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

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

    International Nuclear Information System (INIS)

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

    2008-01-01

    Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on PFC's to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15-25% ne decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, ne/nGW∼1), to enable ne scan capability (x2) in the H-mode, to test the ability to operate at significantly lower density for future ST-CTF reactor designs (e.g., ne/nGW = 0.25), and eventually to investigate high heat-flux power handling (10 MW/m2) with long pulse discharges (>1.5s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization

  15. Overview of the initial NSTX experimental results

    International Nuclear Information System (INIS)

    Ono, M.; Bell, M.G.; Bell, R.E.

    2001-01-01

    The main aim of the National Spherical Torus Experiment (NSTX) is to establish the fusion physics principles of the spherical torus (ST) concept. The NSTX device began plasma operations in February 1999 and the plasma current I p was successfully brought up to the design value of 1 MA on 14 December 1999. The planned plasma shaping parameters, elongation κ=1:6-2.2 and triangularity δ=0:2-0.4, were achieved in inner wall limited, and single null and double null diverted configurations. The coaxial helicity injection (CHI) and high harmonic fast wave (HHFW) experiments were also initiated. CHI current of 27 kA produced up to 260 kA toroidal current without using an ohmic solenoid. With the injection of 2.3 MW of HHFW power, using 12 antennas connected to six transmitters, electrons were heated from a central temperature of 400 eV to 900 eV at a central density of 3.5x10 13 cm 3 , increasing the plasma energy to 59 kJ and the toroidal β, β T , to 10%. The NBI system commenced operation in September 2000. The initial results with two ion sources (P NBI =2:8 MW) show good heating, producing a total plasma stored energy of 90 kJ corresponding to β T ∼18% at a plasma current of 1.1 MA. (author)

  16. Lithium Pellet Injector Development for NSTX

    International Nuclear Information System (INIS)

    Gettelfinger, G.; Dong, J.; Gernhardt, R.; Kugel, H.; Sichta, P.; Timberlake, J.

    2003-01-01

    A pellet injector suitable for the injection of lithium and other low-Z pellets of varying mass into plasmas at precise velocities from 5 to 500 m/s is being developed for use on NSTX (National Spherical Torus Experiment). The ability to inject low-Z impurities will significantly expand NSTX experimental capability for a broad range of diagnostic and operational applications. The architecture employs a pellet-carrying cartridge propelled through a guide tube by deuterium gas. Abrupt deceleration of the cartridge at the end of the guide tube results in the pellet continuing along its intended path, thereby giving controlled reproducible velocities for a variety of pellets materials and a reduced gas load to the torus. The planned injector assembly has four hundred guide tubes contained in a rotating magazine with eight tubes provided for injection into plasmas. A PC-based control system is being developed as well and will be described elsewhere in these Proceedings. The development path and mechanical performance of the injector will be described

  17. Overview of the initial NSTX experimental results

    International Nuclear Information System (INIS)

    Ono, M.; Bell, M.; Bell, R.

    2001-01-01

    The main aim of the National Spherical Torus Experiment (NSTX) is to establish the fusion physics principles of the spherical torus (ST) concept. The NSTX device began plasma operations in February 1999 and the plasma current I p was successfully brought up to the design value of 1 million amperes on December 14, 1999. The planned plasma shaping parameters, κ=1.6-2.2 and δ=0.2-0.4, were achieved in inner limited, single null and double null configurations. The CHI (Coaxial Helicity Injection) and HHFW (High Harmonic Fast Wave) experiments were also initiated. A CHI injected current of 27 kA produced up to 260 kA of toroidal current without using an ohmic solenoid. With an injection of 2.3 MW of HHFW power, using twelve antennas connected to six transmitters, electrons were heated from a central temperature of 400 eV to 900 eV at a central density of 3.5x10 13 cm -3 increasing the plasma energy to 59 kJ and the toroidal beta, β T to 10 %. Finally, the NBI system commenced operation in Sept. 2000. The initial results with two ion sources (P NBI =2.8MW) shows good heating, producing a total plasma stored energy of 90 kJ corresponding to β T ∼18% at a plasma current of 1.1 MA. (author)

  18. Parametric Decay during HHFW on NSTX

    International Nuclear Information System (INIS)

    Wilson, J.R.; Bernabei, S.; Biewer, T.; Diem, S.; Hosea, J.; LeBlanc, B.; Phillips, C.K.; Ryan, P.; Swain, D.W.

    2005-01-01

    High Harmonic Fast Wave (HHFW) heating experiments on NSTX have been observed to be accompanied by significant edge ion heating (T i >> T e ). This heating is found to be anisotropic with T perp > T par . Simultaneously, coherent oscillations have been detected with an edge Langmuir probe. The oscillations are consistent with parametric decay of the incident fast wave (ω > 13ω ci ) into ion Bernstein waves and an unobserved ion-cyclotron quasi-mode. The observation of anisotropic heating is consistent with Bernstein wave damping, and the Bernstein waves should completely damp in the plasma periphery as they propagate toward a cyclotron harmonic resonance. The number of daughter waves is found to increase with rf power, and to increase as the incident wave's toroidal wavelength increases. The frequencies of the daughter wave are separated by the edge ion cyclotron frequency. Theoretical calculations of the threshold for this decay in uniform plasma indicate an extremely small value of incident power should be required to drive the instability. While such decays are commonly observed at lower harmonics in conventional ICRF heating scenarios, they usually do not involve the loss of significant wave power from the pump wave. On NSTX an estimate of the power loss can be found by calculating the minimum power required to support the edge ion heating (presumed to come from the decay Bernstein wave). This calculation indicates at least 20-30% of the incident rf power ends up as decay waves

  19. Advances in boronization on NSTX-Upgrade

    Directory of Open Access Journals (Sweden)

    C. H Skinner

    2017-08-01

    Full Text Available Boronization has been effective in reducing plasma impurities and enabling access to higher density, higher confinement plasmas in many magnetic fusion devices. The National Spherical Torus eXperiment, NSTX, has recently undergone a major upgrade to NSTX-U in order to develop the physics basis for a ST-based Fusion Nuclear Science Facility (FNSF with capability for double the toroidal field, plasma current, and NBI heating power and increased pulse duration from 1–1.5s to 5–8s. A new deuterated tri-methyl boron conditioning system was implemented together with a novel surface analysis diagnostic. We report on the spatial distribution of the boron deposition versus discharge pressure, gas injection and electrode location. The oxygen concentration of the plasma facing surface was measured by in-vacuo XPS and increased both with plasma exposure and with exposure to trace residual gases. This increase correlated with the rise of oxygen emission from the plasma.

  20. Overview of the Initial NSTX Experimental Results

    International Nuclear Information System (INIS)

    Ono, M.; Bell, M.; Bell, R. E.; Bigelow, T.; Bitter, M.

    2000-01-01

    The main aim of the National Spherical Torus Experiment (NSTX) is to establish the fusion physics principles of the spherical torus (ST) concept. The NSTX device began plasma operations in February 1999 and the plasma current Ip was successfully brought up to the design value of 1 million amperes on December 14, 1999. The planned plasma shaping parameters, k = 1.6 ± 2.2 and d = 0.2 ± 0.4, were achieved in inner limited, single null and double null configurations. The CHI (Coaxial Helicity Injection) and HHFW (High Harmonic Fast Wave) experiments were also initiated. A CHI injected current of 27 kA produced up to 260 kA of toroidal current without using an ohmic solenoid. With an injection of 2.3 MW of HHFW power, using twelve antennas connected to six transmitters, electrons were heated from a central temperature of 400 eV to 900 eV at a central density of 3.5 x 1013 cm-3 increasing the plasma energy to 59 kJ and the toroidal beta, bT to 10 %. Finally, the NBI system commenced operation in Sept. 2000. The initial results with two ion sources (PNBI = 2.8 MW) shows good heating, producing a total plasma stored energy of 90 kJ corresponding to bT = 18 % at a plasma current of 1.1 MA

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

    International Nuclear Information System (INIS)

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

    1995-10-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  3. Exploration of spherical torus physics in the NSTX device

    Science.gov (United States)

    Ono, M.; Kaye, S. M.; Peng, Y.-K. M.; Barnes, G.; Blanchard, W.; Carter, M. D.; Chrzanowski, J.; Dudek, L.; Ewig, R.; Gates, D.; Hatcher, R. E.; Jarboe, T.; Jardin, S. C.; Johnson, D.; Kaita, R.; Kalish, M.; Kessel, C. E.; Kugel, H. W.; Maingi, R.; Majeski, R.; Manickam, J.; McCormack, B.; Menard, J.; Mueller, D.; Nelson, B. A.; Nelson, B. E.; Neumeyer, C.; Oliaro, G.; Paoletti, F.; Parsells, R.; Perry, E.; Pomphrey, N.; Ramakrishnan, S.; Raman, R.; Rewoldt, G.; Robinson, J.; Roquemore, A. L.; Ryan, P.; Sabbagh, S.; Swain, D.; Synakowski, E. J.; Viola, M.; Williams, M.; Wilson, J. R.; NSTX Team

    2000-03-01

    The National Spherical Torus Experiment (NSTX) is being built at Princeton Plasma Physics Laboratory to test the fusion physics principles for the spherical torus concept at the MA level. The NSTX nominal plasma parameters are R0 = 85 cm, a = 67 cm, R/a >= 1.26, Bt = 3 kG, Ip = 1 MA, q95 = 14, elongation κ The plasma heating/current drive tools are high harmonic fast wave (6 MW, 5 s), neutral beam injection (5 MW, 80 keV, 5 s) and coaxial helicity injection. Theoretical calculations predict that NSTX should provide exciting possibilities for exploring a number of important new physics regimes, including very high plasma β, naturally high plasma elongation, high bootstrap current fraction, absolute magnetic well and high pressure driven sheared flow. In addition, the NSTX programme plans to explore fully non-inductive plasma startup as well as a dispersive scrape-off layer for heat and particle flux handling.

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

    International Nuclear Information System (INIS)

    Suzuki, Satoshi

    2011-01-01

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

  5. Overview of impurity control and wall conditioning in NSTX

    International Nuclear Information System (INIS)

    Kugel, H.W.; Maingi, R.; Wampler, W.; Barry, R.E.; Bell, M.; Blanchard, W.; Gates, D.; Johnson, D.; Kaita, R.; Kaye, S.; Maqueda, R.; Menard, J.; Menon, M.M.; Mueller, D.; Ono, M.; Paul, S.; Peng, Y-K.M.; Raman, R.; Roquemore, A.; Skinner, C. H.; Sabbagh, S.; Stratton, B.; Stutman, D.; Wilson, J. R.; Zweben, S.

    2000-01-01

    The National Spherical Torus Experiment (NSTX) started plasma operations i n February 1999. In the first extended period of experiments, NSTX achieved high current, inner wall limited, double null, and single null plasma discharges, initial Coaxial Helicity Injection, and High Harmonic Fast Wave results. As expected, discharge reproducibility and performance were strongly affected by wall conditions. In this paper, the authors describe the internal geometry, and initial plasma discharge, impurity control, wall conditioning, erosion, and deposition results

  6. National Spherical Torus Experiment (NSTX) Center Stack Upgrade

    International Nuclear Information System (INIS)

    Neumeyer, C.; Avasarala, S.; Chrzanowski, J.; Dudek, L.; Fan, H.; Hatcher, H.; Heitzenroeder, P.; Menard, J.; Ono, M.; Ramakrishnan, S.; Titus, P.; Woolley, R.; Zhan, H.

    2009-01-01

    The purpose of the NSTX Center Stack Upgrade project is to expand the NSTX operational space and thereby the physics basis for next-step ST facilities. The plasma aspect ratio (ratio of plasma major to minor radius) of the upgrade is increased to 1.5 from the original value of 1.26, which increases the cross sectional area of the center stack by a factor of ∼ 3 and makes possible higher levels of performance and pulse duration.

  7. TCV divertor upgrade for alternative magnetic configurations

    Directory of Open Access Journals (Sweden)

    H. Reimerdes

    2017-08-01

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

  8. Artificial graphites

    International Nuclear Information System (INIS)

    Maire, J.

    1984-01-01

    Artificial graphites are obtained by agglomeration of carbon powders with an organic binder, then by carbonisation at 1000 0 C and graphitization at 2800 0 C. After description of the processes and products, we show how the properties of the various materials lead to the various uses. Using graphite enables us to solve some problems, but it is not sufficient to satisfy all the need of the application. New carbonaceous material open application range. Finally, if some products are becoming obsolete, other ones are being developed in new applications [fr

  9. The ITER divertor cassette project meeting

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-15

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

  11. Divertor design through shape optimization

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  12. Detached divertor plasmas in JET

    Energy Technology Data Exchange (ETDEWEB)

    Horton, L D; Borrass, K; Corrigan, G; Gottardi, N; Lingertat, J; Loarte, A; Simonini, R; Stamp, M F; Taroni, A [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking; Stangeby, P C [Toronto Univ., ON (Canada). Inst. for Aerospace Studies

    1994-07-01

    In simulations with high radiated power fractions, it is possible to produce the drop in ion current to the divertor targets typical of detached plasmas. Despite the fact that these experiments are performed on beryllium target tiles, radiation from deuterium and beryllium cannot account for the measured power losses. The neutral deuterium levels in the SOL in these plasmas are higher than the model predicts. This may be due to leakage from the divertor or to additional wall sources related to the non-steady nature of these plasmas. In contrast, a surprisingly high level of carbon is present in these discharges; higher even than would be predicted are the divertor target tiles pure carbon. This level may well be large enough to produce the measured radiation. (authors). 6 refs., 2 figs., 1 tab.

  13. Fast ion loss diagnostic plans for NSTX

    International Nuclear Information System (INIS)

    Darrow, D. S.; Bell, R.; Johnson, R.; Kugel, H.; Wilson, J. R.; Cecil, F. E.; Maingi, R.; Krasilnikov, A.; Alekseyev, A.

    2000-01-01

    The prompt loss of neutral beam ions from the National Spherical Torus Experiment (NSTX) is expected to be between 12% and 42% of the total 5 MW of beam power. There may, in addition, be losses of fast ions arising from high harmonic fast wave (HHFW) heating. Most of the lost ions will strike the HHFW antenna or the neutral beam dump. To measure these losses in the 2000 experimental campaign, thermocouples in the antenna, several infrared camera views, and a Faraday cup lost ion probe will be employed. The probe will measure loss of fast ions with E > 1 keV at three radial locations, giving the scrape-off length of the fast ions

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

    International Nuclear Information System (INIS)

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

    1997-11-01

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

  15. Actively convected liquid metal divertor

    International Nuclear Information System (INIS)

    Shimada, Michiya; Hirooka, Yoshi

    2014-01-01

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

  16. Design and Calibration of a Dispersive Imaging Spectrometer Adaptor for a Fast IR Camera on NSTX-U

    Science.gov (United States)

    Reksoatmodjo, Richard; Gray, Travis; Princeton Plasma Physics Laboratory Team

    2017-10-01

    A dispersive spectrometer adaptor was designed, constructed and calibrated for use on a fast infrared camera employed to measure temperatures on the lower divertor tiles of the NSTX-U tokamak. This adaptor efficiently and evenly filters and distributes long-wavelength infrared photons between 8.0 and 12.0 microns across the 128x128 pixel detector of the fast IR camera. By determining the width of these separated wavelength bands across the camera detector, and then determining the corresponding average photon count for each photon wavelength, a very accurate measurement of the temperature, and thus heat flux, of the divertor tiles can be calculated using Plank's law. This approach of designing an exterior dispersive adaptor for the fast IR camera allows accurate temperature measurements to be made of materials with unknown emissivity. Further, the relative simplicity and affordability of this adaptor design provides an attractive option over more expensive, slower, dispersive IR camera systems. This work was made possible by funding from the Department of Energy for the Summer Undergraduate Laboratory Internship (SULI) program. This work is supported by the US DOE Contract No. DE-AC02-09CH11466.

  17. Advanced ST Plasma Scenario Simulations for NSTX

    International Nuclear Information System (INIS)

    Kessel, C.E.; Synakowski, E.J.; Gates, D.A.; Harvey, R.W.; Kaye, S.M.; Mau, T.K.; Menard, J.; Phillips, C.K.; Taylor, G.; Wilson, R.

    2004-01-01

    Integrated scenario simulations are done for NSTX [National Spherical Torus Experiment] that address four primary milestones for developing advanced ST configurations: high β and high β N inductive discharges to study all aspects of ST physics in the high-beta regime; non-inductively sustained discharges for flattop times greater than the skin time to study the various current-drive techniques; non-inductively sustained discharges at high β for flattop times much greater than a skin time which provides the integrated advanced ST target for NSTX; and non-solenoidal start-up and plasma current ramp-up. The simulations done here use the Tokamak Simulation Code (TSC) and are based on a discharge 109070. TRANSP analysis of the discharge provided the thermal diffusivities for electrons and ions, the neutral-beam (NB) deposition profile, and other characteristics. CURRAY is used to calculate the High Harmonic Fast Wave (HHFW) heating depositions and current drive. GENRAY/CQL3D is used to establish the heating and CD [current drive] deposition profiles for electron Bernstein waves (EBW). Analysis of the ideal-MHD stability is done with JSOLVER, BALMSC, and PEST2. The simulations indicate that the integrated advanced ST plasma is reachable, obtaining stable plasmas with β ∼ 40% at β N 's of 7.7-9, I P = 1.0 MA, and B T = 0.35 T. The plasma is 100% non-inductive and has a flattop of 4 skin times. The resulting global energy confinement corresponds to a multiplier of H 98(y,2) 1.5. The simulations have demonstrated the importance of HHFW heating and CD, EBW off-axis CD, strong plasma shaping, density control, and early heating/H-mode transition for producing and optimizing these plasma configurations

  18. Advanced ST plasma scenario simulations for NSTX

    International Nuclear Information System (INIS)

    Kessel, C.E.; Synakowski, E.J.; Gates, D.A.; Kaye, S.M.; Menard, J.; Phillips, C.K.; Taylor, G.; Wilson, R.; Harvey, R.W.; Mau, T.K.

    2005-01-01

    Integrated scenario simulations are done for NSTX that address four primary milestones for developing advanced ST configurations: high β and high β N inductive discharges to study all aspects of ST physics in the high beta regime; non-inductively sustained discharges for flattop times greater than the skin time to study the various current drive techniques; non-inductively sustained discharges at high βfor flattop times much greater than a skin time which provides the integrated advanced ST target for NSTX; and non-solenoidal startup and plasma current rampup. The simulations done here use the Tokamak Simulation Code (TSC) and are based on a discharge 109070. TRANSP analysis of the discharge provided the thermal diffusivities for electrons and ions, the neutral beam (NB) deposition profile and other characteristics. CURRAY is used to calculate the High Harmonic Fast Wave (HHFW) heating depositions and current drive. GENRAY/CQL3D is used to establish the heating and CD deposition profiles for electron Bernstein waves (EBW). Analysis of the ideal MHD stability is done with JSOLVER, BALMSC, and PEST2. The simulations indicate that the integrated advanced ST plasma is reachable, obtaining stable plasmas with β ∼ 40% at β N 's of 7.7-9, I P = 1.0 MA and B T = 0.35 T. The plasma is 100% non-inductive and has a flattop of 4 skin times. The resulting global energy confinement corresponds to a multiplier of H 98(y,2 ) = 1.5. The simulations have demonstrated the importance of HHFW heating and CD, EBW off-axis CD, strong plasma shaping, density control, and early heating/H-mode transition for producing and optimizing these plasma configurations (author)

  19. Snowflake divertor plasmas on TCV

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  1. Retention of Hydrogen Isotopes in Divertor Tiles Used in JT-60U

    International Nuclear Information System (INIS)

    Hirohata, Y.; Shibahara, T.; Tanabe, T.; Oya, Y.; Arai, T.; Gotoh, Y.; Masaki, K.; Yagyu, J.; Oyaidzu, M.; Okuno, K.; Nishikawa, M.; Miya, N.

    2005-01-01

    Retention characteristics of deuterium and hydrogen retained in graphite tiles placed in the divertor region of JT-60U were investigated by thermal desorption spectroscopy (TDS). The deuterium retained in the near surface of all graphite tiles was mostly replaced by hydrogen due to exposure to hydrogen plasma at the final stage operations, resulting in main deuterium retention in the deeper region. The dominant species desorbed from the divertor tiles were H 2 , HD, D 2 and CH 4 . The smallest retention of hydrogen isotopes (H+D) was observed in the outer divertor tile which was eroded with maximum of 20 μm depth. The amount of H+D retained in the inner divertor tiles covered by the re-deposited layers increased with the thickness of the re-deposited layers. Hydrogen isotopes concentration ((H+D)/C) in the re-deposited layers was ∼0.02, which was much smaller than those observed in JET and other devices

  2. Thermal properties of redeposition layers in the JT-60U divertor region

    International Nuclear Information System (INIS)

    Ishimoto, Y.; Gotoh, Y.; Arai, T.; Masaki, K.; Miya, N.; Oyama, N.; Asakura, N.

    2006-01-01

    Thermal properties of the redeposition layer on the inner plate of the W-shaped divertor of JT-60U have been measured with laser flash method so as to estimate transient heat loads onto the divertor. Morphology analysis of the redeposition layer was conducted with a scanning electron microscope. Measurement of a redeposition layer sample of more than 200 μm thick, which had been produced near the most frequent striking point, showed following results: (1) the bulk density of the redeposition layer is about half of that of carbon fiber composite material; (2) the specific heat of the layer is roughly equal to that of the isotropic graphite; (3) the thermal conductivity of the redeposition layer is two orders of magnitude smaller than that of the carbon fiber composite. This low thermal conductivity of the redeposition layer is considered to be caused by a low graphitization degree of the redeposition layer. The difference between the divertor heat loads and the loss of the plasma stored energy becomes smaller taking account of thermal properties of the redeposition layer on the inner divertor, whereas estimated heat loads due to the ELMs is still larger than the loss. This is probably caused by the poloidal distribution of the thermal properties

  3. A review of progress towards radiative divertor

    International Nuclear Information System (INIS)

    Zagorski, Roman

    1997-07-01

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

  4. Continuum Gyrokinetic Simulations of Turbulence in a Helical Model SOL with NSTX-type parameters

    Science.gov (United States)

    Hammett, G. W.; Shi, E. L.; Hakim, A.; Stoltzfus-Dueck, T.

    2017-10-01

    We have developed the Gkeyll code to carry out 3D2V full- F gyrokinetic simulations of electrostatic plasma turbulence in open-field-line geometries, using special versions of discontinuous-Galerkin algorithms to help with the computational challenges of the edge region. (Higher-order algorithms can also be helpful for exascale computing as they reduce the ratio of communications to computations.) Our first simulations with straight field lines were done for LAPD-type cases. Here we extend this to a helical model of an SOL plasma and show results for NSTX-type parameters. These simulations include the basic elements of a scrape-off layer: bad-curvature/interchange drive of instabilities, narrow sources to model plasma leaking from the core, and parallel losses with model sheath boundary conditions (our model allows currents to flow in and out of the walls). The formation of blobs is observed. By reducing the strength of the poloidal magnetic field, the heat flux at the divertor plate is observed to broaden. Supported by the Max-Planck/Princeton Center for Plasma Physics, the SciDAC Center for the Study of Plasma Microturbulence, and DOE Contract DE-AC02-09CH11466.

  5. VUV Spectroscopy in DIII-D Divertor

    International Nuclear Information System (INIS)

    Alkesh Punjabi; Nelson Jalufka

    2004-01-01

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

  6. Thermomechanical behavior of graphite and coating materials subjected to a high heat flux

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K.; Yamada, M.; Nishikawa, M.; Uchikawa, T.; Onozuka, M.; Yamao, H.

    1987-07-01

    This study has been performed for the development of limiter and divertor plates. Their thermal and thermomechanical behavior were examined in heat load experiments with an electron beam facility, and were compared with analysis results. Graphite was proven to have a high thermal shock resistance. Its erosion thickness and thermal contact conductance were also studied. Copper alloy with coating and graphite brazed to metal were tested, and their feasibility was demonstrated for use as limiter and divertor plates of an advanced-type concept.

  7. Thermomechanical behavior of graphite and coating materials subjected to a high heat flux

    International Nuclear Information System (INIS)

    Ioki, K.; Yamada, M.; Nishikawa, M.; Uchikawa, T.; Onozuka, M.; Yamao, H.

    1987-01-01

    This study has been performed for the development of limiter and divertor plates. Their thermal and thermomechanical behavior were examined in heat load experiments with an electron beam facility, and were compared with analysis results. Graphite was proven to have a high thermal shock resistance. Its erosion thickness and thermal contact conductance were also studied. Copper alloy with coating and graphite brazed to metal were tested, and their feasibility was demonstrated for use as limiter and divertor plates of an advanced-type concept. (orig.)

  8. ECRH/EBWH system for NSTX-U

    Directory of Open Access Journals (Sweden)

    Hosea J.C.

    2012-09-01

    Full Text Available The National Spherical Torus Experiment Upgrade (NSTX-U will operate at an axial toroidal field of up to 1 T, about twice the field available on NSTX. A 28 GHz electron cylotron resonance heating (ECRH system is currently being planned for NSTX-U. A 1 MW 28 GHz gyrotron will be employed. Intially the system will use short, 10-50 ms, 1 MW pulses for ECRH-assisted discharge start-up. Later the pulse length will be extended to 1-5 s to study electron Bernstein wave heating (EBWH during the plasma current flat top. A mirror launcher will be used to couple microwave power to the plasma via O-mode to the slow X-mode to EBW (O-X-B double mode conversion. This paper presents a pre-conceptual design for the ECRH/EBWH system proposed for NSTX-U and includes ray tracing and Fokker-Planck modeling results for 28 GHz ECRH during plasma start-up and EBW heating and current drive during the plasma current flattop of a NSTX-U advanced H-mode plasma scenario.

  9. Improvement of thermal shock resistance of isotropic graphite by Ti-doping

    International Nuclear Information System (INIS)

    Lopez-Galilea, I.; Ordas, N.; Garcia-Rosales, C.; Lindig, S.

    2009-01-01

    Ti-doped isotropic graphite is a promising candidate material for the strike point area of the ITER divertor due to its reduced chemical erosion by hydrogen bombardment and its high thermal shock resistance, mainly due the catalytic effect of TiC on the graphitization leading to an increase of thermal conductivity and to higher mechanical strength. Several manufacturing parameters such as oxidative stabilization treatment, carbonization cycle, graphitization temperature and dwell time during graphitization have been investigated in order to establish a relationship between these parameters and the final properties.

  10. Improvement of thermal shock resistance of isotropic graphite by Ti-doping

    Energy Technology Data Exchange (ETDEWEB)

    Lopez-Galilea, I. [Inmaculada Lopez-Galilea, CEIT and Tecnun (University of Navarra), Po de Manuel Lardizabal, 15 E-20018 San Sebastian (Spain)], E-mail: ilopez@ceit.es; Ordas, N.; Garcia-Rosales, C. [Inmaculada Lopez-Galilea, CEIT and Tecnun (University of Navarra), Po de Manuel Lardizabal, 15 E-20018 San Sebastian (Spain); Lindig, S. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, D-85748 Garching (Germany)

    2009-04-30

    Ti-doped isotropic graphite is a promising candidate material for the strike point area of the ITER divertor due to its reduced chemical erosion by hydrogen bombardment and its high thermal shock resistance, mainly due the catalytic effect of TiC on the graphitization leading to an increase of thermal conductivity and to higher mechanical strength. Several manufacturing parameters such as oxidative stabilization treatment, carbonization cycle, graphitization temperature and dwell time during graphitization have been investigated in order to establish a relationship between these parameters and the final properties.

  11. A graphite foam reinforced by graphite particles

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, J.J.; Wang, X.Y.; Guo, L.F.; Wang, Y.M.; Wang, Y.P.; Yu, M.F.; Lau, K.T.T. [DongHua University, Shanghai (China). College of Material Science and Engineering

    2007-11-15

    Graphite foam was obtained after carbonization and graphitization of a pitch foam formed by the pyrolysis of coal tar based mesophase pitch mixed with graphite particles in a high pressure and temperature chamber. The graphite foam possessed high mechanical strength and exceptional thermal conductivity after adding the graphite particles. Experimental results showed that the thermal conductivity of modified graphite foam reached 110W/m K, and its compressive strength increased from 3.7 MPa to 12.5 MPa with the addition of 5 wt% graphite particles. Through the microscopic observation, it was also found that fewer micro-cracks were formed in the cell wall of the modified foam as compared with pure graphite foam. The graphitization degree of modified foam reached 84.9% and the ligament of graphite foam exhibited high alignment after carbonization at 1200{sup o}C for 3 h and graphitization at 3000{sup o}C for 10 min.

  12. Startup of the experimental physics industrial control system at NSTX

    International Nuclear Information System (INIS)

    Sichta, P.; Dong, J.

    1999-01-01

    The Experimental Physics Industrial Control System (EPICS) is a set of software which is being used as the basis of the National Spherical Torus Experiment's (NSTX) Process Control System, a major element of the NSTX's Central Instrumentation and Control System. EPICS is a result of a co-development effort started by several US Department of Energy National Laboratories. EPICS is actively supported through an international collaboration made up of government and industrial users. EPICS' good points include portability, scalability, and extensibility. A drawback for small experiments is that a wide range of software skills are necessary to get the software tools running for the process engineers. The authors' experience in designing, developing, operating, and maintaining NSTX's EPICS (system) will be reviewed

  13. Development of a Universal Networked Timer at NSTX

    International Nuclear Information System (INIS)

    Sichta, P.; Dong, J.; Lawson, J.E.; Oliaro, G.; Wertenbaker, J.

    2005-01-01

    A new Timing and Synchronization System component, the Universal Networked Timer (UNT), is under development at the National Spherical Torus Experiment (NSTX). The UNT is a second-generation multifunction timing device that emulates the timing functionality and electrical interfaces originally provided by various CAMAC modules. Using Field Programmable Gate Array (FPGA) technology, each of the UNT's eight channels can be dynamically programmed to emulate a specific CAMAC module type. The timer is compatible with the existing NSTX timing and synchronization system and will also support a (future) clock system with extended performance. To assist system designers and collaborators, software will be written to integrate the UNT with EPICS, MDSplus, and LabVIEW. This paper will describe the timing capabilities, hardware design, programming/software support, and the current status of the Universal Networked Timer at NSTX

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  17. Progress toward commissioning and plasma operation in NSTX-U

    Science.gov (United States)

    Ono, M.; Chrzanowski, J.; Dudek, L.; Gerhardt, S.; Heitzenroeder, P.; Kaita, R.; Menard, J. E.; Perry, E.; Stevenson, T.; Strykowsky, R.; Titus, P.; von Halle, A.; Williams, M.; Atnafu, N. D.; Blanchard, W.; Cropper, M.; Diallo, A.; Gates, D. A.; Ellis, R.; Erickson, K.; Hosea, J.; Hatcher, R.; Jurczynski, S. Z.; Kaye, S.; Labik, G.; Lawson, J.; LeBlanc, B.; Maingi, R.; Neumeyer, C.; Raman, R.; Raftopoulos, S.; Ramakrishnan, R.; Roquemore, A. L.; Sabbagh, S. A.; Sichta, P.; Schneider, H.; Smith, M.; Stratton, B.; Soukhanovskii, V.; Taylor, G.; Tresemer, K.; Zolfaghari, A.; The NSTX-U Team

    2015-07-01

    The National Spherical Torus Experiment-Upgrade (NSTX-U) is the most powerful spherical torus facility at PPPL, Princeton USA. The major mission of NSTX-U is to develop the physics basis for an ST-based Fusion Nuclear Science Facility (FNSF). The ST-based FNSF has the promise of achieving the high neutron fluence needed for reactor component testing with relatively modest tritium consumption. At the same time, the unique operating regimes of NSTX-U can contribute to several important issues in the physics of burning plasmas to optimize the performance of ITER. NSTX-U further aims to determine the attractiveness of the compact ST for addressing key research needs on the path toward a fusion demonstration power plant (DEMO). The upgrade will nearly double the toroidal magnetic field BT to 1 T at a major radius of R0 = 0.93 m, plasma current Ip to 2 MA and neutral beam injection (NBI) heating power to 14 MW. The anticipated plasma performance enhancement is a quadrupling of the plasma stored energy and near doubling of the plasma confinement time, which would result in a 5-10 fold increase in the fusion performance parameter nτ T. A much more tangential 2nd NBI system, with 2-3 times higher current drive efficiency compared to the 1st NBI system, is installed to attain the 100% non-inductive operation needed for a compact FNSF design. With higher fields and heating powers, the NSTX-U plasma collisionality will be reduced by a factor of 3-6 to help explore the favourable trend in transport towards the low collisionality FNSF regime. The NSTX-U first plasma is planned for the Summer of 2015, at which time the transition to plasma operations will occur.

  18. Time Resolved Deposition Measurements in NSTX

    International Nuclear Information System (INIS)

    Skinner, C.H.; Kugel, H.; Roquemore, A.L.; Hogan, J.; Wampler, W.R.

    2004-01-01

    Time-resolved measurements of deposition in current tokamaks are crucial to gain a predictive understanding of deposition with a view to mitigating tritium retention and deposition on diagnostic mirrors expected in next-step devices. Two quartz crystal microbalances have been installed on NSTX at a location 0.77m outside the last closed flux surface. This configuration mimics a typical diagnostic window or mirror. The deposits were analyzed ex-situ and found to be dominantly carbon, oxygen, and deuterium. A rear facing quartz crystal recorded deposition of lower sticking probability molecules at 10% of the rate of the front facing one. Time resolved measurements over a 4-week period with 497 discharges, recorded 29.2 (micro)g/cm 2 of deposition, however surprisingly, 15.9 (micro)g/cm 2 of material loss occurred at 7 discharges. The net deposited mass of 13.3 (micro)g/cm 2 matched the mass of 13.5 (micro)g/cm 2 measured independently by ion beam analysis. Monte Carlo modeling suggests that transient processes are likely to dominate the deposition

  19. Fabrication of divertor cassette for ITER

    International Nuclear Information System (INIS)

    Sanguinetti, G.P.

    2008-01-01

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

  20. A Lithium Vapor Box Divertor Similarity Experiment

    Science.gov (United States)

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

    2017-10-01

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

  1. Development of divertor remote maintenance system

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-04-01

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

  2. Development of divertor remote maintenance system

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  3. Internal Kink Mode Dynamics in High-β NSTX Plasmas

    International Nuclear Information System (INIS)

    Menard, J.E.; Bell, R.E.; Fredrickson, E.D.; Gates, D.A.; Kaye, S.M.; LeBlanc, B.P.; Medley, S.S.; Park, W.; Sabbagh, S.A.; Sontag, A.; Stutman, D.; Tritz, K.; Zhu, W.

    2004-01-01

    Saturated internal kink modes have been observed in many of the highest toroidal beta discharges of the National Spherical Torus Experiment (NSTX). These modes often cause rotation flattening in the plasma core, can degrade energy confinement, and in some cases contribute to the complete loss of plasma angular momentum and stored energy. Characteristics of the modes are measured using soft X-ray, kinetic profile, and magnetic diagnostics. Toroidal flows approaching Alfvenic speeds, island pressure peaking, and enhanced viscous and diamagnetic effects associated with high-beta may contribute to mode nonlinear stabilization. These saturation mechanisms are investigated for NSTX parameters and compared to experimental data

  4. Internal kink mode dynamics in high-β NSTX plasmas

    International Nuclear Information System (INIS)

    Menard, J.E.; Bell, R.E.; Fredrickson, E.D.; Gates, D.A.; Kaye, S.M.; LeBlanc, B.P.; Medley, S.S.; Park, W.; Sabbagh, S.A.; Sontag, A.; Zhu, W.; Stutman, D.; Tritz, K.

    2005-01-01

    Saturated internal kink modes have been observed in many of the highest toroidal beta discharges of the National Spherical Torus Experiment (NSTX). These modes often cause rotation flattening in the plasma core, can degrade energy confinement, and in some cases contribute to the complete loss of plasma angular momentum and stored energy. Characteristics of the modes are measured using soft X-ray, kinetic profile, and magnetic diagnostics. Toroidal flows approaching Alfvenic speeds, island pressure peaking, and enhanced viscous and diamagnetic effects associated with high-beta may contribute to mode non-linear stabilization. These saturation mechanisms are investigated for NSTX parameters and compared to experiment. (author)

  5. Precision metrology of NSTX surfaces using coherent laser radar ranging

    International Nuclear Information System (INIS)

    Kugel, H.W.; Loesser, D.; Roquemore, A. L.; Menon, M. M.; Barry, R. E.

    2000-01-01

    A frequency modulated Coherent Laser Radar ranging diagnostic is being used on the National Spherical Torus Experiment (NSTX) for precision metrology. The distance (range) between the 1.5 microm laser source and the target is measured by the shift in frequency of the linearly modulated beam reflected off the target. The range can be measured to a precision of < 100microm at distances of up to 22 meters. A description is given of the geometry and procedure for measuring NSTX interior and exterior surfaces during open vessel conditions, and the results of measurements are elaborated

  6. Controlling marginally detached divertor plasmas

    Science.gov (United States)

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

    2017-06-01

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

  7. The ITER divertor cassette project

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  8. The ITER divertor cassette project

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  9. Divertor plate for thermonuclear reactor

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  10. The lithium vapor box divertor

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  11. Effective Thermal Conductivity of Graphite Materials with Cracks

    Science.gov (United States)

    Pestchaanyi, S. E.; Landman, I. S.

    The dependence of effective thermal diffusivity on temperature caused by volumetric cracks is modelled for macroscopic graphite samples using the three-dimensional thermomechanics code Pegasus-3D. At high off-normal heat loads typical of the divertor armour, thermostress due to the anisotropy of graphite grains is much larger than that due to the temperature gradient. Numerical simulation demonstrated that the volumetric crack density both in fine grain graphites and in the CFC matrix depends mainly on the local sample temperature, not on the temperature gradient. This allows to define an effective thermal diffusivity for graphite with cracks. The results obtained are used to explain intense cracking and particle release from carbon based materials under electron beam heat load. Decrease of graphite thermal diffusivity with increase of the crack density explains particle release mechanism in the experiments with CFC where a clear energy threshold for the onset of particle release has been observed in J. Linke et al. Fusion Eng. Design, in press, Bazyler et al., these proceedings. Surface temperature measurement is necessary to calibrate the Pegasus-3D code for simulation of ITER divertor armour brittle destruction.

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

    International Nuclear Information System (INIS)

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

    1991-02-01

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

  13. Thermally exfoliated graphite oxide

    Science.gov (United States)

    Prud'Homme, Robert K. (Inventor); Aksay, Ilhan A. (Inventor); Abdala, Ahmed (Inventor)

    2011-01-01

    A modified graphite oxide material contains a thermally exfoliated graphite oxide with a surface area of from about 300 sq m/g to 2600 sq m/g, wherein the thermally exfoliated graphite oxide displays no signature of the original graphite and/or graphite oxide, as determined by X-ray diffraction.

  14. Power transport to the poloidal divertor experiment scoop limiter

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  15. Engineering conceptual design of CFETR divertor

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  16. A large divertor manipulator for ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    International Nuclear Information System (INIS)

    Fantz, U.

    2002-01-01

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

  18. Ramp-up of CHI Initiated Plasmas on NSTX

    International Nuclear Information System (INIS)

    Mueller, D.; Bell, M.G.; Bell, R.E.; LeBlanc, B.; Roquemore, A.L.; Raman, R.; Jarboe, T.R.; Nelson, B.A.; Soukhanovskii, V.

    2009-01-01

    Experiments on the National Spherical Torus (NSTX) have now demonstrated flux savings using transient coaxial helicity injection (CHI). In these discharges, the discharges initiated by CHI are ramped up with an inductive transformer and exhibit higher plasma current than discharges without the benefit of CHI initiation.

  19. Initial Results from Coaxial Helicity Injection Experiments in NSTX

    International Nuclear Information System (INIS)

    Raman, R.; Jarboe, T.R.; Mueller, D.; Schaffer, M.J.; Maqueda, R.; Nelson, B.A.; Sabbagh, S.; Bell, M.; Ewig, R.; Fredrickson, E.; Gates, D.; Hosea, J.; Ji, H.; Kaita, R.; Kaye, S.M.; Kugel, H.; Maingi, R.; Menard, J.; Ono, M.; Orvis, D.; Paolette, F.; Paul, S.; Peng, M.; Skinner, C.H.; Wilgen, W.; Zweben, S.

    2001-01-01

    Coaxial Helicity Injection (CHI) has been investigated on the National Spherical Torus Experiment (NSTX). Initial experiments produced 130 kA of toroidal current without the use of the central solenoid. The corresponding injector current was 20 kA. Discharges with pulse lengths up to 130 ms have been produced

  20. Beryllium and graphite performance in ITER during a disruption

    International Nuclear Information System (INIS)

    Hassanein, A.; Ehst, D.A.; Gahl, J.

    1993-09-01

    Plasma disruptions are considered one of the most limiting factors for successful operation of magnetic fusion reactors. During a disruption, a sharp, rapid release of energy strikes components such as the divertor or limiter plates. Severe surface erosion and melting of these components may then occur. The amount of material eroded from both ablation and melting is important to the reactor design and component lifetime. The anticipated performance of both beryllium and graphite as plasma-facing materials during such abnormal events is analyzed and compared. Recent experimental data obtained with both plasma guns and electron beams are carefully evaluated and compared to results of analytical modeling, including vapor shielding effect. Initial results from plasma gun experiments indicate that the Be erosion rate is about five times larger than that for a graphite material under the same disruption conditions. Key differences between simulation experiments and reactor disruption on the net erosion rate, and consequently on the lifetime of the divertor plate, are discussed in detail. The advantages and disadvantages of Be over graphite as a divertor plasma-facing material are discussed

  1. Beryllium and graphite performance in ITER during a disruption

    International Nuclear Information System (INIS)

    Hassanein, A.; Ehst, D.A.; Gahl, J.

    1994-01-01

    Plasma disruptions are considered one of the most limiting factors for successful operation of magnetic fusion reactors. During a disruption, a sharp, rapid release of energy strikes components such as the divertor or limiter plates. Severe surface erosion and melting of these components may then occur. The amount of material eroded from both ablation and melting is important to the reactor design and component lifetime. The anticipated performance of both beryllium and graphite as plasma-facing materials during such abnormal events is analyzed and compared. Recent experimental data obtained with both plasma guns and electron beams are carefully evaluated and compared to results of analytical modeling, including vapor shielding effect. Initial results from plasma gun experiments indicate that the Be erosion rate is about five times larger than that for a graphite material under the same disruption conditions. Key differences between simulation experiments and reactor disruption on the net erosion rate, and consequently on the lifetime of the divertor plate, are discussed in detail. The advantages and disadvantages of Be over graphite as a divertor plasma-facing material are discussed. ((orig.))

  2. Discharge power dependence of carbon dust flux in a divertor simulator

    International Nuclear Information System (INIS)

    Nishiyama, Katsushi; Morita, Yasuhiko; Uchida, Giichiro; Yamashita, Daisuke; Kamataki, Kunihiro; Seo, Hyunwoong; Itagaki, Naho; Koga, Kazunori; Shiratani, Masaharu; Ashikawa, Naoko; Masuzaki, Suguru; Nishimura, Kiyohiko; Sagara, Akio; Bornholdt, Sven; Kersten, Holger

    2013-01-01

    In fusion devices, dust particles are generated due to plasma–wall interactions and may cause safety or operational problems. Therefore it is necessary to clarify the generation and transport mechanisms of dust particles. Here we have measured energy influx from H 2 plasmas toward a graphite target using a calorimetric probe and compared the results with the dust flux toward a dust collecting substrate set on the reactor wall. The dust flux decreases with increasing the energy influx. For the higher discharge power, the more number of dust particles tend to redeposit onto the graphite target due to the higher ion drag force and hence the dust flux toward the reactor wall becomes smaller. The results show that dust inventory depends strongly on energy influx to graphite divertor plates in fusion devices

  3. Moving Divertor Plates in a Tokamak

    International Nuclear Information System (INIS)

    Zweben, S.J.; Zhang, H.

    2009-01-01

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

  4. Optimization of a bundle divertor for FED

    International Nuclear Information System (INIS)

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

    1982-01-01

    Optimal double-T bundle divertor configurations have been obtained for the Fusion Engineering Device (FED). On-axis ripple is minimized, while satisfying a series of engineering constraints. The ensuing non-linear optimization problem is solved via a sequence of quadratic programming subproblems, using the VMCON algorithm. The resulting divertor designs are substantially improved over previous configurations

  5. Moving Divertor Plates in a Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S.J. Zweben, H. Zhang

    2009-02-12

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

  6. In situ measurement of erosion/deposition in the DIII-D divertor by colorimetry

    International Nuclear Information System (INIS)

    Weschenfelder, F.; Wienhold, P.; Winter, J.

    1996-01-01

    Colorimetry was introduced into the DIII-D tokamak to measure in situ the growth and erosion of transparent wall coatings (a-C:H) on the divertor. The colorimetric measurement system consisting of a halogen light source, a set of three filters and a black/white camera is described together with a first erosion measurement. An insertable graphite sample with a diameter of 4.7 cm was precoated with a 300 nm thick amorphous carbon film and was exposed in the divertor for several discharges with its surface coplanar to the surrounding graphite tiles. For each of the discharges the plasma strike point was moved onto the sample for 1 s to erode the coating. Between the discharges a camera signal with each filter was recorded and the film thickness was evaluated along a radial line across the DIMES sample. Thus it has been possible for the first time to measure erosion and deposition of divertor material in situ and shot-by-shot. The average peak heat flux with the strike point on DIMES was about 110 W cm -2 . The measurement shows a strong decrease in the film thickness almost over the entire sample with an average erosion rate of ∼ 9 nm s -1 . (Author)

  7. Bridged graphite oxide materials

    Science.gov (United States)

    Herrera-Alonso, Margarita (Inventor); McAllister, Michael J. (Inventor); Aksay, Ilhan A. (Inventor); Prud'homme, Robert K. (Inventor)

    2010-01-01

    Bridged graphite oxide material comprising graphite sheets bridged by at least one diamine bridging group. The bridged graphite oxide material may be incorporated in polymer composites or used in adsorption media.

  8. Reactor application of an improved bundle divertor

    International Nuclear Information System (INIS)

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

    1978-11-01

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

  9. Evaluating Stellarator Divertor Designs with EMC3

    Science.gov (United States)

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

    2013-10-01

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

  10. Comparison between stellarator and tokamak divertor transport

    International Nuclear Information System (INIS)

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

    2010-11-01

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

  11. Recent Developments in High-Harmonic Fast Wave Physics in NSTX

    International Nuclear Information System (INIS)

    LeBlanc, B.P.; Bell, R.E.; Bonoli, P.; Harvey, R.; Heidbrink, W.W.; Hosea, J.C.; Kaye, S.M.; Liu, D.; Maingi, R.; Medley, S.S.; Ono, M.; Podesta, M.; Phillips, C.K.; Ryan, P.M.; Roquemore, A.L.; Taylor, G.; Wilson, J.R.

    2010-01-01

    Understanding the interaction between ion cyclotron range of frequency (ICRF) fast waves and the fast-ions created by neutral beam injection (NBI) is critical for future devices such as ITER, which rely on a combination ICRF and NBI. Experiments in NSTX which use 30 MHz High-Harmonic Fast-Wave (HHFW) ICRF and NBI heating show a competition between electron heating via Landau damping and transit-time magnetic pumping, and radio-frequency wave acceleration of NBI generated fast ions. Understanding and mitigating some of the power loss mechanisms outside the last closed flux surface (LCFS) has resulted in improved HHFW heating inside the LCFS. Nevertheless a significant fraction of the HHFW power is diverted away from the enclosed plasma. Part of this power is observed locally on the divertor. Experimental observations point toward the radio-frequency (RF) excitation of surface waves, which disperse wave power outside the LCFS, as a leading loss mechanism. Lithium coatings lower the density at the antenna, thereby moving the critical density for perpendicular fast-wave propagation away from the antenna and surrounding material surfaces. Visible and infrared imaging reveal flows of RF power along open field lines into the divertor region. In L-mode -- low average NBI power -- conditions, the fast-ion D-alpha (FIDA) diagnostic measures a near doubling and broadening of the density profile of the upper energetic level of the fast ions concurrent with the presence of HHFW power launched with k// = -8m-1. We are able to heat NBI-induced H-mode plasmas with HHFW. The captured power is expected to be split between absorption by the electrons and absorption by the fast ions, based on TORIC calculation. In the case discussed here the Te increases over the whole profile when ∼2MW of HHFW power with antenna k// = 13m-1 is applied after the H-mode transition. But somewhat unexpectedly fast-ion diagnostics do not observe a change between the HHFW heated NBI discharge and the

  12. Snowflake divertor experiments on TCV

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  13. Process for purifying graphite

    International Nuclear Information System (INIS)

    Clausius, R.A.

    1985-01-01

    A process for purifying graphite comprising: comminuting graphite containing mineral matter to liberate at least a portion of the graphite particles from the mineral matter; mixing the comminuted graphite particles containing mineral matter with water and hydrocarbon oil to form a fluid slurry; separating a water phase containing mineral matter and a hydrocarbon oil phase containing grahite particles; and separating the graphite particles from the hydrocarbon oil to obtain graphite particles reduced in mineral matter. Depending upon the purity of the graphite desired, steps of the process can be repeated one or more times to provide a progressively purer graphite

  14. Modelling of NSTX hot vertical displacement events using M 3 D -C 1

    Science.gov (United States)

    Pfefferlé, D.; Ferraro, N.; Jardin, S. C.; Krebs, I.; Bhattacharjee, A.

    2018-05-01

    The main results of an intense vertical displacement event (VDE) modelling activity using the implicit 3D extended MHD code M3D-C1 are presented. A pair of nonlinear 3D simulations are performed using realistic transport coefficients based on the reconstruction of a so-called NSTX frozen VDE where the feedback control was purposely switched off to trigger a vertical instability. The vertical drift phase is solved assuming axisymmetry until the plasma contacts the first wall, at which point the intricate evolution of the plasma, decaying to large extent in force-balance with induced halo/wall currents, is carefully resolved via 3D nonlinear simulations. The faster 2D nonlinear runs allow to assess the sensitivity of the simulations to parameter changes. In the limit of perfectly conducting wall, the expected linear relation between vertical growth rate and wall resistivity is recovered. For intermediate wall resistivities, the halo region contributes to slowing the plasma down, and the characteristic VDE time depends on the choice of halo temperature. The evolution of the current quench and the onset of 3D halo/eddy currents are diagnosed in detail. The 3D simulations highlight a rich structure of toroidal modes, penetrating inwards from edge to core and cascading from high-n to low-n mode numbers. The break-up of flux-surfaces results in a progressive stochastisation of field-lines precipitating the thermalisation of the plasma with the wall. The plasma current then decays rapidly, inducing large currents in the halo region and the wall. Analysis of normal currents flowing in and out of the divertor plate reveals rich time-varying patterns.

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

    International Nuclear Information System (INIS)

    1994-03-01

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

  16. Multiple equilibria of divertor plasmas

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  18. Divertor radiation in the ASDEX upgrade tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-15

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

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

    International Nuclear Information System (INIS)

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

    1995-10-01

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

  1. An Asdex-type divertor for ITER

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1989-01-01

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

  2. High Harmonic Fast Wave Heating Experiments on NSTX

    International Nuclear Information System (INIS)

    Wilson, J.R.; Bell, R.; Bitter, M.; Bonoli, P.

    2000-01-01

    A radio frequency (rf) system has been installed on the National Spherical Torus Experiment (NSTX) with the aim of heating the plasma and driving plasma current. The system consists of six rf transmitters, a twelve element antenna and associated transmission line components to distribute and couple the power from the transmitters to the antenna elements in a fashion to allow control of the antenna toroidal wavenumber spectrum. To date, power levels up to 3.85 MW have been applied to the NSTX plasmas. The frequency and spectrum of the rf waves has been selected to heat electrons via Landau damping and transit time magnetic pumping. The electron temperature has been observed to increase from 400 to 900 eV with little change in plasma density resulting in a plasma stored energy of 59 kJ and a toroidal beta, bT , =10% and bn = 2.7

  3. High harmonic fast wave heating experiments on NSTX

    International Nuclear Information System (INIS)

    Wilson, J.R.; Bell, R.; Bitter, M.

    2001-01-01

    A radio frequency (rf) system has been installed on the National Spherical Torus Experiment (NSTX) with the aim of heating the plasma and driving plasma current. The system consists of six rf transmitters, a twelve element antenna and associated transmission line components to distribute and couple the power from the transmitters to the antenna elements in a fashion to allow control of the antenna toroidal wavenumber spectrum. To date, power levels up to 3.85 MW have been applied to the NSTX plasmas. The frequency and spectrum of the rf waves has been selected to heat electrons via Landau damping and transit time magnetic pumping. The electron temperature has been observed to increase from 400 to 900 eV with little change in plasma density resulting in a plasma stored energy of 59 kJ , a toroidal beta, β T =10% and a normalized beta, β n =2.7. (author)

  4. Temperature gradient driven electron transport in NSTX and Tore Supra

    International Nuclear Information System (INIS)

    Horton, W.; Wong, H.V.; Morrison, P.J.; Wurm, A.; Kim, J.H.; Perez, J.C.; Pratt, J.; Hoang, G.T.; LeBlanc, B.P.; Ball, R.

    2005-01-01

    Electron thermal fluxes are derived from the power balance for Tore Supra (TS) and NSTX discharges with centrally deposited fast wave electron heating. Measurements of the electron temperature and density profiles, combined with ray tracing computations of the power absorption profiles, allow detailed interpretation of the thermal flux versus temperature gradient. Evidence supporting the occurrence of electron temperature gradient turbulent transport in the two confinement devices is found. With control of the magnetic rotational transform profile and the heating power, internal transport barriers are created in TS and NSTX discharges. These partial transport barriers are argued to be a universal feature of transport equations in the presence of invariant tori that are intrinsic to non-monotonic rotational transforms in dynamical systems

  5. Mechanical Design of the NSTX High-k Scattering Diagnostic

    International Nuclear Information System (INIS)

    Feder, R.; Mazzucato, E.; Munsat, T.; Park, H.; Smith, D.R.; Ellis, R.; Labik, G.; Priniski, C.

    2005-01-01

    The NSTX High-k Scattering Diagnostic measures small-scale density fluctuations by the heterodyne detection of waves scattered from a millimeter wave probe beam at 280 GHz and λ = 1.07 mm. To enable this measurement, major alterations were made to the NSTX vacuum vessel and Neutral Beam armor. Close collaboration between the PPPL physics and engineering staff resulted in a flexible system with steerable launch and detection optics that can position the scattering volume either near the magnetic axis (ρ ∼ .1) or near the edge (ρ ∼ .8). 150 feet of carefully aligned corrugated waveguide was installed for injection of the probe beam and collection of the scattered signal in to the detection electronics

  6. Solid State Neutral Particle Analyzer Array on NSTX

    International Nuclear Information System (INIS)

    Shinohara, K.; Darrow, D.S.; Roquemore, A.L.; Medley, S.S.; Cecil, F.E.

    2004-01-01

    A Solid State Neutral Particle Analyzer (SSNPA) array has been installed on the National Spherical Torus Experiment (NSTX). The array consists of four chords viewing through a common vacuum flange. The tangency radii of the viewing chords are 60, 90, 100, and 120 cm. They view across the three co-injection neutral beam lines (deuterium, 80 keV (typ.) with tangency radii 48.7, 59.2, and 69.4 cm) on NSTX and detect co-going energetic ions. A silicon photodiode used was calibrated by using a mono-energetic deuteron beam source. Deuterons with energy above 40 keV can be detected with the present setup. The degradation of the performance was also investigated. Lead shots and epoxy are used for neutron shielding to reduce handling any hazardous heavy metal. This method also enables us to make an arbitrary shape to be fit into the complex flight tube

  7. ELMs and the H-mode pedestal in NSTX

    International Nuclear Information System (INIS)

    Maingi, R.; Sabbagh, S.A.; Bush, C.E.; Fredrickson, E.D.; Menard, J.E.; Stutman, D.; Tritz, K.; Bell, M.G.; Bell, R.E.; Boedo, J.A.; Gates, D.A.; Johnson, D.W.; Kaita, R.; Kaye, S.M.; Kugel, H.W.; LeBlanc, B.P.; Mueller, D.; Raman, R.; Roquemore, A.L.; Soukhanovskii, V.A.; Stevenson, T.

    2005-01-01

    We report on the behavior of ELMs in NBI-heated H-mode plasmas in NSTX. It is observed that the size of Type I ELMs, characterized by the change in plasma energy, decreases with increasing line-average density, as observed at conventional aspect ratio. It is also observed that the Type I ELM size decreases as the plasma equilibrium is shifted from a symmetric double-null toward a lower single-null configuration. Type II/III ELMs have also been observed in NSTX, as well as a high-performance regime with small ELMs which we designate Type V. The Type V ELMs are characterized by an intermittent n 1 magnetic pre-cursor oscillation rotating counter to the plasma current; the mode vanishes between Type V ELMs crashes. Without active pumping, the density rises continuously through the Type V phase, albeit at a slower rate than ELM-free discharges

  8. Mechanical Design of the NSTX High-k Scattering Diagnostic

    Energy Technology Data Exchange (ETDEWEB)

    Feder, R.; Mazzucato, E.; Munsat, T.; Park, H,; Smith, D. R.; Ellis, R.; Labik, G.; Priniski, C.

    2005-09-26

    The NSTX High-k Scattering Diagnostic measures small-scale density fluctuations by the heterodyne detection of waves scattered from a millimeter wave probe beam at 280 GHz and {lambda}=1.07 mm. To enable this measurement, major alterations were made to the NSTX vacuum vessel and Neutral Beam armor. Close collaboration between the PPPL physics and engineering staff resulted in a flexible system with steerable launch and detection optics that can position the scattering volume either near the magnetic axis ({rho} {approx} .1) or near the edge ({rho} {approx} .8). 150 feet of carefully aligned corrugated waveguide was installed for injection of the probe beam and collection of the scattered signal in to the detection electronics.

  9. Overview of impurity control and wall conditioning in NSTX

    Energy Technology Data Exchange (ETDEWEB)

    KUGEL,H.W.; MAINGI,R.; BELL,M.; BLANCHARD,W.; GATES,D.; JOHNSON,D.; KAITA,R.; KAYE,S.; MARQUEDA,R.; MENARD,J.; MUELLER,D.; ONO,M.; PENG,Y-K.M.; RAMAN,R.; RAMSEY,A.; ROQUEMORE,A.; SKINNER,C.; SABBAGH,S.; STUTMAN,D.; WAMPLER,WILLIAM R.; WILSON,J.R.; ZWEBEN,S.

    2000-05-25

    The National Spherical Torus Experiment (NSTX) started plasma operations in February 1999, and promptly achieved high current, inner wall limited, double null, and single null plasma discharges, initial Coaxial Helicity Injection, and High Harmonic Fast Wave results. NSTX is designed to study the physics of Spherical Tori (ST) in a device that can produce non-inductively sustained high-{beta} discharges in the 1 MA regime and to explore approaches toward a small, economical high power density ST reactor core. As expected, discharge reproducibility and performance were strongly affected by wall conditions. In this paper, the authors describe the internal geometry, and initial plasma discharge, impurity control, wall conditioning, erosion, and deposition results.

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  12. Simulation of the time development of EBW emission from NSTX

    Czech Academy of Sciences Publication Activity Database

    Preinhaelter, Josef; Urban, Jakub; Taylor, G.; Diem, S.; Vahala, L.; Vahala, G.

    2006-01-01

    Roč. 51, č. 4 (2006), K1.00024 ISSN 0003-0503. [International Sherwood Fusion Theory Conference/2006./. Dallas, Texas , 22.4.2006-25.4.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : Conversion * Emission * Tokamaks * Electron Bernstein waves * Simulation * MAST * NSTX Subject RIV: BL - Plasma and Gas Discharge Physics http://www.aps.org/meet/APR06/baps/all_APR06.pdf http://meetings.aps.org/Meeting/APR06/Event/47670

  13. Diagnostics of ST Plasmas in NSTX: Challenges and Opportunities

    International Nuclear Information System (INIS)

    Johnson, D.; Efthimion, P.; Foley, J.; Jones, B.; Mazzucato, E.; Park, H.; Taylor, G.; Levinton, F.; Luhmann, N.

    2001-01-01

    This paper will highlight some of the challenges and opportunities present in the diagnosis of spherical torus (ST) plasmas on the National Spherical Torus Experiment (NSTX) and discuss the corresponding diagnostic development that is presently underway. After a brief description of diagnostic systems currently installed, examples of ST-specific diagnostic challenges will be highlighted, as will another case, where the ST configuration offers opportunities for new measurements

  14. Electron Bernstein Wave Coupling and Emission Measurements on NSTX

    Czech Academy of Sciences Publication Activity Database

    Taylor, G.; Diem, S.J.; Caughman, J.; Efthimion, P.; Harvey, R.W.; LeBlanc, B.P.; Philips, C.K.; Preinhaelter, Josef; Urban, Jakub

    2006-01-01

    Roč. 51, č. 7 (2006), s. 177 ISSN 0003-0503. [Annual Meeting of the Division of Plasma Physics/48th./. Philadelphia, Pennsylvania , 30.10.2006-3.11.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : Conversion * Emission * Tokamaks * Electron Bernstein waves * Simulation * MAST * NSTX Subject RIV: BL - Plasma and Gas Discharge Physics http://www.aps.org/meet/DPP06/baps/all_DPP06.pdf

  15. Effect of Various EFIT NSTX Equilibria on EBW Simulations

    Czech Academy of Sciences Publication Activity Database

    Urban, Jakub; Preinhaelter, Josef; Sabbagh, S.; Pavlo, Pavol; Vahala, L.; Vahala, G.

    2006-01-01

    Roč. 51, č. 7 (2006), QPI.00027 ISSN 0003-0503. [Annual Meeting of the Division of Plasma Physics/48th./. Philadelphia, Pennsylvania , 30.10.2006-3.11.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : Conversion * Emission * Tokamaks * Electron Bernstein waves * Simulation * MAST * NSTX Subject RIV: BL - Plasma and Gas Discharge Physics http://www.aps.org/meet/DPP06/baps/all_DPP06.pdf

  16. Solenoid-free plasma startup in NSTX using transient CHI

    International Nuclear Information System (INIS)

    Raman, R.; Jarboe, T.R.; Nelson, B.A.; Mueller, D.; Bell, M.G.; Bell, R.; Gates, D.; Gerhardt, S.; Hosea, J.; Kaita, R.; Kugel, H.; LeBlanc, B.; Menard, J.; Ono, M.; Paul, S.; Roquemore, L.; Maingi, R.; Maqueda, R.; Nagata, M.; Sabbagh, S.

    2009-01-01

    Experiments in NSTX have now demonstrated the coupling of toroidal plasmas produced by the technique of coaxial helicity injection (CHI) to inductive sustainment and ramp-up of the toroidal plasma current. In these discharges, the central Ohmic transformer was used to apply an inductive loop voltage to discharges with a toroidal current of about 100 kA created by CHI. The coupled discharges have ramped up to >700 kA and transitioned into an H-mode demonstrating compatibility of this startup method with conventional operation. The electron temperature in the coupled discharges reached over 800 eV and the resulting plasma had low inductance, which is preferred for long-pulse high-performance discharges. These results from NSTX in combination with the previously obtained record 160 kA non-inductively generated startup currents in an ST or tokamak in NSTX demonstrate that CHI is a viable solenoid-free plasma startup method for future STs and tokamaks.

  17. Easy web interfaces to IDL code for NSTX Data Analysis

    International Nuclear Information System (INIS)

    Davis, W.M.

    2012-01-01

    Highlights: ► Web interfaces to IDL code can be developed quickly. ► Dozens of Web Tools are used effectively on NSTX for Data Analysis. ► Web interfaces are easier to use than X-window applications. - Abstract: Reusing code is a well-known Software Engineering practice to substantially increase the efficiency of code production, as well as to reduce errors and debugging time. A variety of “Web Tools” for the analysis and display of raw and analyzed physics data are in use on NSTX [1], and new ones can be produced quickly from existing IDL [2] code. A Web Tool with only a few inputs, and which calls an IDL routine written in the proper style, can be created in less than an hour; more typical Web Tools with dozens of inputs, and the need for some adaptation of existing IDL code, can be working in a day or so. Efficiency is also increased for users of Web Tools because of the familiar interface of the web browser, and not needing X-windows, or accounts and passwords, when used within our firewall. Web Tools were adapted for use by PPPL physicists accessing EAST data stored in MDSplus with only a few man-weeks of effort; adapting to additional sites should now be even easier. An overview of Web Tools in use on NSTX, and a list of the most useful features, is also presented.

  18. Stability, divertors and innovative concepts

    International Nuclear Information System (INIS)

    Mirnov, S.

    2003-01-01

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

  19. Divertor cassette movers prototypes for ITER

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  20. Stochasticity about a poloidal divertor separatrix

    International Nuclear Information System (INIS)

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

    1986-10-01

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

  1. The H-mode Pedestal and Edge Localized Modes in NSTX

    International Nuclear Information System (INIS)

    Maingi, R.; Fredrickson, E.D.; Menard, J.E.; Nishino, N.; Roquemore, A.L.; Sabbagh, S.A.; Tritz, K.

    2004-01-01

    The research program of the National Spherical Torus Experiment (NSTX) routinely utilizes the H-mode confinement regime to test and extend beta and pulse length limits. As in conventional aspect ratio tokamaks, NSTX observes a variety of edge localized modes (ELMs) in H-mode. Hence a significant part of the research program is dedicated to ELMs studies

  2. ARIES-III divertor engineering design

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  3. ARIES-III divertor engineering design

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-01-01

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

  6. Hydrogen isotopes retention in divertor tiles of DIII-D tokamak

    International Nuclear Information System (INIS)

    Skorodumov, B.G.; Buzhinskij, O.I.; West, W.P.; Ulanov, V.G.

    1996-01-01

    The absolute concentration of hydrogen isotopes in graphite divertor tiles coated with boron carbide after the exposure in DIII-D during 16 operational weeks of the 1993 campaign was obtained using the 14 MeV neutron-induced recoil detection (NERD) method. It is shown that the absolute concentration of H in tile's surface layers correlates with thickness of the deposited layers. The graphite tile without boron carbide coating had a H concentration similar to that of the tile with the thickest deposited layer. Deuterium and tritium were not detected in any of the investigated tiles. The proposed method can be used for the determination of the thickness of coatings without sample destruction. Thus, the thickness of boron carbide coatings on the tiles obtained with this method varied from 80 to 115 μm, which corresponded well to electron microscope data. (orig.)

  7. Hydrogen isotope distributions and retentions in the inner divertor tile of JT-60U

    Energy Technology Data Exchange (ETDEWEB)

    Oya, Y. [Radioisotope Center, University of Tokyo, 2-11-16 Yayoi, Bunkyo-ku, Tokyo 113-0032 (Japan)]. E-mail: yoya@ric.u-tokyo.ac.jp; Hirohata, Y. [Graduate School of Engineering, Hokkaido University, Sapporo 060-8628 (Japan); Tanabe, T. [Graduate School of Engineering, Nagoya University, Nagoya 464-8603 (Japan); Shibahara, T. [Graduate School of Engineering, Nagoya University, Nagoya 464-8603 (Japan); Kimura, H. [Faculty of Science, Shizuoka University, Shizuoka 422-8529 (Japan); Oyaidzu, M. [Faculty of Science, Shizuoka University, Shizuoka 422-8529 (Japan); Arai, T. [Naka Fusion Establishment, Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Masaki, K. [Naka Fusion Establishment, Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Gotoh, Y. [Naka Fusion Establishment, Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Okuno, K. [Faculty of Science, Shizuoka University, Shizuoka 422-8529 (Japan); Miya, N. [Naka Fusion Establishment, Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Hino, T. [Graduate School of Engineering, Hokkaido University, Sapporo 060-8628 (Japan); Tanaka, S. [Graduate School of Engineering, University of Tokyo, Tokyo 113-8656 (Japan)

    2005-11-15

    Retention profiles of hydrogen and deuterium in graphite tiles placed in the inner divertor region of JT-60U were analyzed by secondary ion mass spectroscopy (SIMS) and thermal desorption spectroscopy (TDS). The difference in hydrogen and deuterium retention behaviour is discussed considering the frequency of the strike-point hit and history of NBI heating power. It was found that most of hydrogen/deuterium was retained in the deposited layers, HH deposition layers/DD deposition layers or co-deposited with carbon. Owing to the higher heating power of DD discharges, the deuterium concentration in the DD deposition layers was much lower than that of hydrogen in the HH deposition layers. On the area showing no deposition, very shallow profile of deuterium dominated hydrogen profile. These results indicate that the tritium retention is strongly influenced by the history of discharge and temperatures. Tritium retention on graphite tiles and deposition layers could be significantly reduced with increasing the operation temperature.

  8. Hydrogen isotope distributions and retentions in the inner divertor tile of JT-60U

    International Nuclear Information System (INIS)

    Oya, Y.; Hirohata, Y.; Tanabe, T.; Shibahara, T.; Kimura, H.; Oyaidzu, M.; Arai, T.; Masaki, K.; Gotoh, Y.; Okuno, K.; Miya, N.; Hino, T.; Tanaka, S.

    2005-01-01

    Retention profiles of hydrogen and deuterium in graphite tiles placed in the inner divertor region of JT-60U were analyzed by secondary ion mass spectroscopy (SIMS) and thermal desorption spectroscopy (TDS). The difference in hydrogen and deuterium retention behaviour is discussed considering the frequency of the strike-point hit and history of NBI heating power. It was found that most of hydrogen/deuterium was retained in the deposited layers, HH deposition layers/DD deposition layers or co-deposited with carbon. Owing to the higher heating power of DD discharges, the deuterium concentration in the DD deposition layers was much lower than that of hydrogen in the HH deposition layers. On the area showing no deposition, very shallow profile of deuterium dominated hydrogen profile. These results indicate that the tritium retention is strongly influenced by the history of discharge and temperatures. Tritium retention on graphite tiles and deposition layers could be significantly reduced with increasing the operation temperature

  9. Integrated core-edge-divertor modeling studies

    International Nuclear Information System (INIS)

    Stacey, W.M.

    2001-01-01

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

  10. High heat flux tests of mock-ups for ITER divertor application

    International Nuclear Information System (INIS)

    Giniatulin, R.; Gervash, A.; Komarov, V.L.; Makhankov, A.; Mazul, I.; Litunovsky, N.; Yablokov, N.

    1998-01-01

    One of the most difficult tasks in fusion reactor development is the designing, fabrication and high heat flux testing of actively cooled plasma facing components (PFCs). At present, for the ITER divertor project it is necessary to design and test components by using mock-ups which reflect the real design and fabrication technology. The cause of failure of the PFCs is likely to be through thermo-cycling of the surface with heat loads in the range 1-15 MW m -2 . Beryllium, tungsten and graphite are considered as the most suitable armour materials for the ITER divertor application. This work presents the results of the tests carried out with divertor mock-ups clad with beryllium and tungsten armour materials. The tests were carried out in an electron beam facility. The results of high heat flux screening tests and thermo-cycling tests in the heat load range 1-9 MW m -2 are presented along with the results of metallographic analysis carried out after the tests. (orig.)

  11. First results from the dynamic ergodic divertor at TEXTOR

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  12. Divertor scaling laws for tokamaks

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  13. Phonon scattering in graphite

    International Nuclear Information System (INIS)

    Wagner, P.

    1976-04-01

    Effects on graphite thermal conductivities due to controlled alterations of the graphite structure by impurity addition, porosity, and neutron irradiation are shown to be consistent with the phonon-scattering formulation 1/l = Σ/sub i equals 1/sup/n/ 1/l/sub i/. Observed temperature effects on these doped and irradiated graphites are also explained by this mechanism

  14. Geometrical properties of a 'snowflake' divertor

    International Nuclear Information System (INIS)

    Ryutov, D. D.

    2007-01-01

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

  15. EU R and D on divertor components

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  16. Electron Bernstein Wave Research on CDX-U and NSTX

    International Nuclear Information System (INIS)

    Taylor, G.; Efthimion, P.C.; Jones, B.; Hosea, J.C.; Kaita, R.; LeBlanc, B.P.; Majeski, R.; Munsat, T.; Phillips, C.K.; Spaleta, J.; Wilson, J.R.; Rasmussen, D.; Bell, G.; Bigelow, T.S.; Carter, M.D.; Swain, D.W.; Wilgen, J.B.; Ram, A.K.; Bers, A.; Harvey, R.W.; Forest, C.B.

    2001-01-01

    Mode-converted electron Bernstein waves (EBWs) potentially allow the measurement of local electron temperature (Te) and the implementation of local heating and current drive in spherical torus (ST) devices, which are not directly accessible to low harmonic electron cyclotron waves. This paper reports on the measurement of X-mode radiation mode-converted from EBWs observed normal to the magnetic field on the midplane of the Current Drive Experiment-Upgrade (CDX-U) and the National Spherical Torus Experiment (NSTX) spherical torus plasmas. The radiation temperature of the EBW emission was compared to Te measured by Thomson scattering and Langmuir probes. EBW mode-conversion efficiencies of over 20% were measured on both CDX-U and NSTX. Sudden increases of mode-conversion efficiency, of over a factor of three, were observed at high-confinement-mode transitions on NSTX, when the measured edge density profile steepened. The EBW mode-conversion efficiency was found to depend on the density gradient at the mode-conversion layer in the plasma scrape-off, consistent with theoretical predictions. The EBW emission source was determined by a perturbation technique to be localized at the electron cyclotron resonance layer and was successfully used for radial transport studies. Recently, a new in-vessel antenna and Langmuir probe array were installed on CDX-U to better characterize and enhance the EBW mode-conversion process. The probe incorporates a local adjustable limiter to control and maximize the mode-conversion efficiency in front of the antenna by modifying the density profile in the plasma scrape-off where fundamental EBW mode conversion occurs. Initial results show that the mode-conversion efficiency can be increased to ∼100% when the local limiter is inserted near the mode-conversion layer. Plans for future EBW research, including EBW heating and current-drive studies, are discussed

  17. National Spherical Torus Experiment (NSTX) Torus Design, Fabrication and Assembly

    International Nuclear Information System (INIS)

    Neumeyer, C.; Barnes, G.; Chrzanowski, J.H.; Heitzenroeder, P.

    1999-01-01

    The National Spherical Torus Experiment (NSTX) is a low aspect ratio spherical torus (ST) located at Princeton Plasma Physics Laboratory (PPPL). Fabrication, assembly, and initial power tests were completed in February of 1999. The majority of the design and construction efforts were constructed on the Torus system components. The Torus system includes the centerstack assembly, external Poloidal and Toroidal coil systems, vacuum vessel, torus support structure and plasma facing components (PFC's). NSTX's low aspect ratio required that the centerstack be made with the smallest radius possible. This, and the need to bake NSTXs carbon-carbon composite plasma facing components at 350 degrees C, was major drivers in the design of NSTX. The Centerstack Assembly consists of the inner legs of the Toroidal Field (TF) windings, the Ohmic Heating (OH) solenoid and its associated tension cylinder, three inner Poloidal Field (PF) coils, thermal insulation, diagnostics and an Inconel casing which forms the inner wall of the vacuum vessel boundary. It took approximately nine months to complete the assembly of the Centerstack. The tight radial clearances and the extreme length of the major components added complexity to the assembly of the Centerstack components. The vacuum vessel was constructed of 304-stainless steel and required approximately seven months to complete and deliver to the Test Cell. Several of the issues associated with the construction of the vacuum vessel were control of dimensional stability following welding and controlling the permeability of the welds. A great deal of time and effort was devoted to defining the correct weld process and material selection to meet our design requirements. The PFCs will be baked out at 350 degrees C while the vessel is maintained at 150 degrees C. This required care in designing the supports so they can accommodate the high electromagnetic loads resulting from plasma disruptions and the resulting relative thermal expansions

  18. Concept of a charged fusion product diagnostic for NSTX.

    Science.gov (United States)

    Boeglin, W U; Valenzuela Perez, R; Darrow, D S

    2010-10-01

    The concept of a new diagnostic for NSTX to determine the time dependent charged fusion product emission profile using an array of semiconductor detectors is presented. The expected time resolution of 1-2 ms should make it possible to study the effect of magnetohydrodynamics and other plasma activities (toroidal Alfvén eigenmodes (TAE), neoclassical tearing modes (NTM), edge localized modes (ELM), etc.) on the radial transport of neutral beam ions. First simulation results of deuterium-deuterium (DD) fusion proton yields for different detector arrangements and methods for inverting the simulated data to obtain the emission profile are discussed.

  19. Edge Turbulence Imaging on NSTX and Alcator C-Mod

    International Nuclear Information System (INIS)

    S.J. Zweben; R.A. Maqueda; J.L. Terry; B. Bai; C.J. Boswell; C.E. Bush; D. D'Ippolito; E.D. Fredrickson; M. Greenwald; K. Hallatschek; S. Kaye; B. LaBombard; R. Maingi; J. Myra; W.M. Nevins; B.N. Rogers; D.P. Stotler; J. Wilgen; and X.Q. Xu

    2002-01-01

    Edge turbulence images have been made using an ultra-high speed CCD camera on both NSTX and Alcator C-Mod. In both cases, the D-alpha or HeI (587.6 nm) line emission from localized deuterium or helium gas puffs was viewed along a local magnetic field line near the outer midplane. Fluctuations in this line emission reflect fluctuations in electron density and/or electron temperature through the atomic excitation rates, which can be modeled using the DEGAS-2 code. The 2-D structure of the measured turbulence can be compared with theoretical simulations based on 3-D fluid models

  20. Electron Bernstein Wave Research on NSTX and PEGASUS

    International Nuclear Information System (INIS)

    Diem, S. J.; LeBlanc, B. P.; Taylor, G.; Caughman, J. B.; Bigelow, T.; Wilgen, J. B.; Garstka, G. D.; Harvey, R. W.; Preinhaelter, J.; Urban, J.; Sabbagh, S. A.

    2007-01-01

    Spherical tokamaks (STs) routinely operate in the overdense regime (ω pe >>ω ce ), prohibiting the use of standard ECCD and ECRH. However, the electrostatic electron Bernstein wave (EBW) can propagate in the overdense regime and is strongly absorbed and emitted at the electron cyclotron resonances. As such, EBWs offer the potential for local electron temperature measurements and local electron heating and current drive. A critical challenge for these applications is to establish efficient coupling between the EBWs and electromagnetic waves outside the cutoff layer. Two STs in the U.S., the National Spherical Tokamak Experiment (NSTX, at Princeton Plasma Physics Laboratory) and PEGASUS Toroidal Experiment (University of Wisconsin-Madison) are focused on studying EBWs for heating and current drive. On NSTX, two remotely steered, quad-ridged antennas have been installed to measure 8-40 GHz (fundamental, second and third harmonics) thermal EBW emission (EBE) via the oblique B-X-O mode conversion process. This diagnostic has been successfully used to map the EBW mode conversion efficiency as a function of poloidal and toroidal angles on NSTX. Experimentally measured mode conversion efficiencies of 70±20% have been measured for 15.5 GHz (fundamental) emission in L-mode discharges, in agreement with a numerical EBE simulation. However, much lower mode conversion efficiencies of 25±10% have been measured for 25 GHz (second harmonic) emission in L-mode plasmas. Numerical modeling of EBW propagation and damping on the very-low aspect ratio PEGASUS Toroidal Experiment has been performed using the GENRAY ray-tracing code and CQL3D Fokker-Planck code in support of planned EBW heating and current drive (EBWCD) experiments. Calculations were performed for 2.45 GHz waves launched with a 10 cm poloidal extent for a variety of plasma equilibrium configurations. Poloidal launch scans show that driven current is maximum when the poloidal launch angle is between 10 and 25 degrees

  1. An Edge Rotation and Temperature Diagnostic on NSTX

    International Nuclear Information System (INIS)

    Biewer, T.M.; Bell, R.E.; Feder, R.; Johnson, D.W.; Palladino, R.W.

    2003-01-01

    A new diagnostic for the National Spherical Torus Experiment (NSTX) is described whose function is to measure ion rotation and temperature at the plasma edge. The diagnostic is sensitive to C III, C IV, and He II intrinsic emission, covering a radial region of 15 cm at the extreme edge of the outboard midplane. Thirteen chords are distributed between toroidal and poloidal views, allowing the toroidal and poloidal rotation and temperature of the plasma edge to be simultaneously measured with 10 ms resolution. Combined with the local pressure gradient and the EFIT code reconstructed magnetic field profile, the edge flow gives a measure of the local radial electric field

  2. Simulation of Diffusive Lithium Evaporation Onto the NSTX Vessel Walls

    International Nuclear Information System (INIS)

    Stotler, D.P.; Skinner, C.H.; Blanchard, W.R.; Krstic, P.S.; Kugel, H.W.; Schneider, H.; Zakharov, L.E.

    2010-01-01

    A model for simulating the diffusive evaporation of lithium into a helium filled NSTX vacuum vessel is described and validated against an initial set of deposition experiments. The DEGAS 2 based model consists of a three-dimensional representation of the vacuum vessel, the elastic scattering process, and a kinetic description of the evaporated atoms. Additional assumptions are required to account for deuterium out-gassing during the validation experiments. The model agrees with the data over a range of pressures to within the estimated uncertainties. Suggestions are made for more discriminating experiments that will lead to an improved model.

  3. Overview of results from the National Spherical Torus Experiment (NSTX)

    Czech Academy of Sciences Publication Activity Database

    Gates, D.A.; Ahn, J.; Allain, J.; Andre, R.; Bastasz, R.; Bell, M.; Bell, R.; Belova, E.; Berkery, J.; Betti, R.; Bialek, J.; Biewer, T.; Bigelow, T.; Bitter, M.; Boedo, J.; Bonoli, P.; Boozer, A.; Brennan, D.; Breslau, J.; Brower, D.; Bush, C.; Canik, J.; Caravelli, G.; Carter, M.; Caughman, J.; Chang, C.; Crocker, N.; Darrow, D.; Delgado-Aparicio, L.; Diem, S.; D’Ippolito, D.; Domier, C.; Dorland, W.; Efthimion, P.; Ejiri, A.; Ershov, N.; Evans, T.; Feibush, E.; Fenstermacher, M.; Ferron, J.; Finkenthal, M.; Foley, J.; Frazin, R.; Fredrickson, E.; Fu, G.; Funaba, H.; Gerhardt, S.; Glasser, A.; Gorelenkov, N.; Grisham, L.; Hahm, T.; Harvey, R.; Hassanein, A.; Heidbrink, W.; Hill, K.; Hillesheim, J.; Hillis, D.; Hirooka, Y.; Hosea, J.; Hu, B.; Humphreys, D.; Idehara, T.; Indireshkumar, K.; Ishida, A.; Jaeger, F.; Jarboe, T.; Jardin, S.; Jaworski, M.; Ji, H.; Kaita, R.; Kallman, J.; Katsuro-Hopkins, O.; Kawahata, K.; Kawamori, E.; Kaye, S.; Kessel, C.; Kimura, H.; Kolemen, E.; Krasheninnikov, H.; Krstic, P.; Ku, S.; Kubota, S.; Kugel, H.; La Haye, R.; Lao, L.; LeBlanc, B.; Lee, K.; Leuer, J.; Levinton, F.; Liang, Y.; Liu, D.; Luhmann Jr, N.; Maingi, R.; Majeski, R.; Manickam, J.; Mansfield, D.; Maqueda, R.; Mazzucato, E.; McCune, D.; McGeehan, B.; McKee, G.; Medley, S.; Menard, J.; Menon, M.; Meyer, H.; Mikkelsen, D.; Miloshevsky, G.; Mitarai, O.; Mueller, D.; Mueller, S.; Munsat, T.; Myra, J.; Nagayama, Y.; Nelson, B.; Nguyen, X.; Nishino, N.; Nishiura, M.; Nygren, R.; Ono, M.; Osborne, T.; Pacella, D.; Park, J.; Paul, S.; Peebles, W.; Penaflor, B.; Peng, M.; Phillips, C.; Pigarov, A.; Podesta, M.; Preinhaelter, Josef; Ram, A.; Raman, R.; Rasmussen, D.; Redd, A.; Reimerdes, H.; Rewoldt, G.; Ross, P.; Rowley, C.; Ruskov, E.; Russell, D.; Ruzic, D.; Ryan, P.; Sabbagh, S.; Schaffer, M.; Schuster, E.; Scott, S.; Shaing, K.; Sharpe, P.; Shevchenko, V.; Shinohara, K.; Sizyuk, V.; Skinner, C.; Smirnov, A.; Smith, D.; Snyder, P.; Solomon, W.; Sontag, A.; Soukhanovskii, V.; Stoltzfus-Dueck, T.; Stotler, D.; Strait, T.; Stratton, B.; Stutman, D.; Takahashi, R.; Takase, Y.; Tamura, N.; Tang, X.; Taylor, G.; Taylor, C.; Ticos, C.; Tritz, K.; Tsarouhas, D.; Turrnbull, A.; Tynan, G.; Ulrickson, M.; Umansky, M.; Urban, Jakub; Utergberg, E.; Walker, M.; Wampler, M.; Wang, J.; Wang, W.; Welander, A.; Whaley, J.; White, R.; Wilgen, J.; Wilson, R.; Wong, K.; Wright, J.; Xia, Z.; Xu, X.; Youchison, D.; Yu, G.; Yuh, H.; Zakharov, L.; Zemlyanov, D.; Zweben, S.; Choe, W.; Jung, H.; Kim, J.; Lee, W.; Park, H.

    2009-01-01

    Roč. 49, č. 10 (2009), s. 104016-104016 ISSN 0029-5515. [IAEA Fusion Energy Conference/22nd./. Geneva, 13.10.2008-18.10.2008] R&D Projects: GA ČR GA202/08/0419 Institutional research plan: CEZ:AV0Z20430508 Keywords : NSTX * Spherical tokamaks * Overdense plasma * Conversion * Emission * Tokamaks * Elektron Bernstein waves Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.270, year: 2009 http://www.iop.org/EJ/article/0029-5515/49/10/104016/nf9_10_104016

  4. Noninductive Current Generation in NSTX using Coaxial Helicity Injection

    International Nuclear Information System (INIS)

    Raman, R.; Jarboe, T.R.; Mueller, D.; Schaffer, M.J.; Maqueda, R.; Nelson, B.A.; Sabbagh, S.; Bell, M.; Ewig, R.; Fredrickson, E.; Gates, D.; Hosea, J.; Jardin, S.; Ji, H.; Kaita, R.; Kaye, S.M.; Kugel, H.; Lao, L.; Maingi, R.; Menard, J.; Ono, M.; Orvis, D.; Paul, S.; Peng, M.; Skinner, C.H.; Wilgen, J.B.; Zweben, S.

    2001-01-01

    Coaxial Helicity Injection (CHI) on the National Spherical Torus Experiment (NSTX) has produced 240 kA of toroidal current without the use of the central solenoid. Values of the current multiplication ratio (CHI produced toroidal current/injector current) up to 10 were obtained, in agreement with predictions. The discharges which lasted for up to 200 ms, limited only by the programmed waveform, are more than an order of magnitude longer in duration that any CHI discharges previously produced in a Spheromak or a Spherical Torus (ST)

  5. DSMC simulations of vapor transport toward development of the lithium vapor box divertor concept

    Science.gov (United States)

    Jagoe, Christopher; Schwartz, Jacob; Goldston, Robert

    2016-10-01

    The lithium vapor divertor box concept attempts to achieve volumetric dissipation of the high heat efflux from a fusion power system. The vapor extracts the heat of the incoming plasma by ionization and radiation, while remaining localized in the vapor box due to differential pumping based on rapid condensation. Preliminary calculations with lithium vapor at densities appropriate for an NSTX-U-scale machine give Knudsen numbers between 0.01 and 1, outside both the range of continuum fluid dynamics and of collisionless Monte Carlo. The direct-simulation Monte Carlo (DSMC) method, however, can simulate rarefied gas flows in this regime. Using the solver contained in the OpenFOAM package, pressure-driven flows of water vapor will be analyzed. The use of water vapor in the relevant range of Knudsen number allows for a flexible similarity experiment to verify the reliability of the code before moving to tests with lithium. The simulation geometry consists of chains of boxes on a temperature gradient, connected by slots with widths that are a representative fraction of the dimensions of the box. We expect choked flow, sonic shocks, and order-of-magnitude pressure and density drops from box to box, but this expectation will be tested in the simulation and then experiment. This work is supported by the Princeton Environmental Institute.

  6. Analysis of particle transport in a gas target divertor

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

  7. Experimental studies of the snowflake divertor in TCV

    NARCIS (Netherlands)

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

    2017-01-01

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

  8. Neutral particle retention in the JET MK I divertor

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  9. Thermal effects of runaway electrons in an armoured divertor

    International Nuclear Information System (INIS)

    Stad, R.C.L. van der.

    1993-12-01

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

  10. Liquid metal cooled divertor for ARIES

    International Nuclear Information System (INIS)

    Muraviev, E.

    1995-01-01

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

  11. Constrained ripple optimization of Tokamak bundle divertors

    International Nuclear Information System (INIS)

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

    1983-02-01

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

  12. Design integration of liquid surface divertors

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    Science.gov (United States)

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

    2018-05-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  15. Towards the procurement of the ITER divertor

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  16. Local island divertor experiments on LHD

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  17. Control of divertor configuration in JT-60

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  18. Safety characteristics of the monolithic CFC divertor

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  19. Safety characteristics of the monolithic CFC divertor

    Science.gov (United States)

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

    1994-09-01

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

  20. Recent Progress on the National Spherical Torus Experiment (NSTX)

    International Nuclear Information System (INIS)

    Gates, D.A.; Bell, M.G.; Bell, R.E.; Bialek, J.; Bigelow, T.; Bitter, M.; Bonoli, P.; Darrow, D.; Efthimion, P.

    2002-01-01

    Recent upgrades to the NSTX facility have led to improved plasma performance. Using 5MW of neutral beam injection, plasmas with toroidal β T (= 2(micro) 0 /B T 2 where B T is the vacuum toroidal field at the plasma geometric center) > 30% have been achieved with normalized β N (= β T aB I /I p ) ∼ 6% · m · T/MA.. The highest β discharge exceeded the calculated no-wall β limit for several wall times. The stored energy has reached 390kJ at higher toroidal field (0.55T) corresponding to β T ∼ 20% and β N = 5.4. Long pulse (∼1s) high β p (∼1.5) discharges have also been obtained at higher β φ (0.5T) with up to 6MW NBI power. The highest energy confinement times, up to 120ms, were observed during H-mode operation which is now routine. Confinement times of ∼1.5 times ITER98pby2 for several τ E are observed during both H-Mode and non-H-Mode discharges. Calculations indicate that many NSTX discharges have very good ion confinement, approaching neoclassical levels. High Harmonic Fast Wave current drive has been demonstrated by comparing discharges with waves launched parallel and anti-parallel to the plasma current

  1. Testing Gyrokinetics on C-Mod and NSTX

    International Nuclear Information System (INIS)

    Redi, M.H.; Dorland, W.; Fiore, C.L.; Stutman, D.; Baumgaertel, J.A.; Davis, B.; Kaye, S.M.; McCune, D.C.; Menard, J.; Rewoldt, G.

    2005-01-01

    Quantitative benchmarks of computational physics codes against experiment are essential for the credible application of such codes. Fluctuation measurements can provide necessary critical tests of nonlinear gyrokinetic simulations, but such require extraordinary computational resources. Linear micro-stability calculations with the GS2 [1] gyrokinetic code have been carried out for tokamak and ST experiments which exhibit internal transport barriers (ITB) and good plasma confinement. Qualitative correlation is found for improved confinement before and during ITB plasmas on Alcator C-Mod [2] and NSTX [3], with weaker long wavelength micro-instabilities in the plasma core regions. Mixing length transport models are discussed. The NSTX L-mode is found to be near marginal stability for kinetic ballooning modes. Fully electromagnetic, linear, gyrokinetic calculations of the Alcator C-Mod ITB during off-axis rf heating, following four plasma species and including the complete electron response show ITG/TEM microturbulence is suppressed in the plasma core and in the barrier region before barrier formation, without recourse to the usual requirements of velocity shear or reversed magnetic shear [4-5]. No strongly growing long or short wavelength drift modes are found in the plasma core but strong ITG/TEM and ETG drift wave turbulence is found outside the barrier region. Linear microstability analysis is qualitatively consistent with the experimental transport analysis, showing low transport inside and high transport outside the ITB region before barrier formation, without consideration of ExB shear stabilization

  2. Plasma control system upgrade and increased plasma stability in NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Mastrovito, D., E-mail: dmastrovito@pppl.go [Princeton Plasma Physics Laboratory, P.O. Box 451 Princeton, NJ 08543 (United States); Gates, D.; Gerhard, S.; Lawson, J.; Ludescher-Furth, C.; Marsala, R. [Princeton Plasma Physics Laboratory, P.O. Box 451 Princeton, NJ 08543 (United States)

    2010-07-15

    Plasma control on the National Spherical Torus Experiment (NSTX) was previously accomplished using eight 333 MHz G4 processors built by Sky computers. Several planned improvements and additional control algorithms required significant upgrades to our real-time control computers and real-time data acquisition infrastructure. Several in-house modules have been designed and implemented including: the digital time stamp module (DITS) and for digital/analog front panel data port (FPDP) output, the FPDP output module digital/analog (FOMD/A). Standard Linux based Intel computers perform the real-time control tasks and InfiniBand as been employed for communication between a user-accessible 'host' server and the real-time computer. In addition to several independent real-time processes the General Atomics developed PCS (Bell (2006) ) system infrastructure continues to be used on NSTX. While maintaining previous functionality, improvements in the control system software include: an RWM feedback algorithm, beta feedback NBI control, more comprehensive error logging and trapping, more user-friendly interface, more complete archiving and restoring functionality, and better status reporting and diagnostic tools. Once completed, we succeeded in increasing overall plasma stability and decreasing control system latency by several times.

  3. Diagnostics for the Biased Electrode Experiment on NSTX

    International Nuclear Information System (INIS)

    Roquemore, A.L.; Zweben, S.J.; Bush, C.E.; Kaita, R.; Marsalsa, R.J.; Maqueda, R.J.

    2009-01-01

    A linear array of four small biased electrodes was installed in NSTX in an attempt to control the width of the scrape-off layer (SOL) by creating a strong local poloidal electric field. The set of electrodes were separated poloidally by a 1 cm gap between electrodes and were located slightly below the midplane of NSTX, 1 cm behind the RF antenna and oriented so that each electrode is facing approximately normal to the magnetic field. Each electrode can be independently biased to ± 100 volts. Present power supplies limit the current on two electrodes to 30 amps the other two to 10 amps each. The effect of local biasing was measured with a set of Langmuir probes placed between the electrodes and another set extending radially outward from the electrodes, and also by the gas puff imaging diagnostic (GPI) located 1 m away along the magnetic field lines intersecting the electrodes. Two fast cameras were also aimed directly at the electrode array. The hardware and controls of the biasing experiment will be presented and the initial effects on local plasma parameters will be discussed

  4. Initial Studies of Core and Edge Transport of NSTX Plasmas

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Bell, M.G.; Bell, R.E.; Bush, C.E.; Bourdelle, C.; Darrow, D.; Dorland, W.; Ejiri, A.; Fredrickson, E.D.; Gates, D.A.; Kaye, S.M.; Kubota, S.; Kugel, H.W.; LeBlanc, B.P.; Maingi, R.; Maqueda, R.J.; Menard, J.E.; Mueller, D.; Rosenberg, A.; Sabbagh, S.A.; Stutman, D.; Taylor, G.; Johnson, D.W.; Kaita, R.; Ono, M.; Paoletti, F.; Peebles, W.; Peng, Y-K.M.; Roquemore, A.L.; Skinner, C.H.; Soukhanovskii, V.A.

    2001-01-01

    Rapidly developing diagnostic, operational, and analysis capability is enabling the first detailed local physics studies to begin in high-beta plasmas of the National Spherical Torus Experiment (NSTX). These studies are motivated in part by energy confinement times in neutral-beam-heated discharges that are favorable with respect to predictions from the ITER-89P scaling expression. Analysis of heat fluxes based on profile measurements with neutral-beam injection (NBI) suggest that the ion thermal transport may be exceptionally low, and that electron thermal transport is the dominant loss channel. This analysis motivates studies of possible sources of ion heating not presently accounted for by classical collisional processes. Gyrokinetic microstability studies indicate that long wavelength turbulence with k(subscript ''theta'') rho(subscript ''i'') ∼ 0.1-1 may be suppressed in these plasmas, while modes with k(subscript ''theta'') rho(subscript ''i'') ∼ 50 may be robust. High-harmonic fast-wave (HHFW) heating efficiently heats electrons on NSTX, and studies have begun using it to assess transport in the electron channel. Regarding edge transport, H-mode [high-confinement mode] transitions occur with either NBI or HHFW heating. The power required for low-confinement mode (L-mode) to H-mode transitions far exceeds that expected from empirical edge-localized-mode-free H-mode scaling laws derived from moderate aspect ratio devices. Finally, initial fluctuation measurements made with two techniques are permitting the first characterizations of edge turbulence

  5. Plasma decontamination during ergodic divertor experiments in TORE SUPRA

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  6. Conceptual design of CFETR divertor remote handling compatible structure

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  7. Conceptual design of CFETR divertor remote handling compatible structure

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-15

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

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

    Science.gov (United States)

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

    2017-04-01

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

  9. The investigation of structure, chemical composition, hydrogen isotope trapping and release processes in deposition layers on surfaces exposed to DIII-D divertor plasma

    International Nuclear Information System (INIS)

    Buzhinskij, O.I.; Opimach, I.V.; Barsuk, V.A.; Arkhipov, I.I.; Whyte, D.; Wampler, W.R.

    1998-05-01

    The exposure of ATG graphite sample to DIII-D divertor plasma was provided by the DiMES (Divertor Material Evaluation System) mechanism. The graphite sample arranged to receive the parallel heat flux on a small region of the surface was exposed to 600ms of outer strike point plasma. The sample was constructed to collect the eroded material directed downward into a trapping zone onto s Si disk collector. The average heat flux onto the graphite sample during the exposure was about 200W/cm 2 , and the parallel heat flux was about 10 KW/cm 2 . After the exposure the graphite sample and Si collector disk were analyzed using SEM, NRA, RBS, Auger spectroscopy. IR and Raman spectroscopy. The thermal desorption was studied also. The deposited coating on graphite sample is amorphous carbon layer. Just upstream of the high heat flux zone the redeposition layer has a globular structure. The deposition layer on Si disk is composed also from carbon but has a diamond-like structure. The areal density of C and D in the deposited layer on Si disk varied in poloidal and toroidal directions. The maximum D/C areal density ratio is about 0.23, maximum carbon density is about 3.8 x 10 18 cm -2 , maximum D area density is about 3 x 10 17 cm 2 . The thermal desorption spectrum had a peak at 1,250K

  10. Observation of non-uniform erosion and deposition phenomena on graphite after plasma exposure

    International Nuclear Information System (INIS)

    Hildebrandt, D.; Grote, H.; Schneider, W.; Wienhold, P.; Seggern, J. von

    1999-01-01

    The modifications of fine grain isotropic graphite surfaces after plasma exposure have been investigated using surface analysis techniques with high spatial resolution in area and depth. The samples are graphite target tiles of ASDEX-upgrade and coated graphite collector samples exposed for special erosion/deposition experiments in the divertor plasma of ASDEX-upgrade or in the scrape-off plasma of TEXTOR-94. In addition, a graphite sample was exposed to a low temperature, clean deuterium plasma to study the modifications of the surface morphology during plasma exposure. The results give clear indications of non-uniform erosion and deposition processes. The change of the surface morphology during these processes is discussed. (orig.)

  11. An X-point ergodic divertor

    International Nuclear Information System (INIS)

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

    1991-10-01

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

  12. The ITER Divertor Cassette Project meeting

    International Nuclear Information System (INIS)

    Akiba, M.; Tivey, R.

    2000-01-01

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

  13. Compact poloidal divertor reference design for TNS

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  14. Impact of the wall conditioning program on plasma performance in NSTX

    International Nuclear Information System (INIS)

    Kugel, H.W.; Soukhanovskii, V.; Bell, M.; Blanchard, W.; Gates, D.; LeBlanc, B.; Maingi, R.; Mueller, D.; Na, H.K.; Paul, S.; Skinner, C.H.; Stutman, D.; Wampler, W.R.

    2003-01-01

    High performance operating regimes have been achieved on NSTX through impurity control and wall conditioning techniques. These techniques include HeGDC-aided boronization using deuterated trimethylboron, inter-discharge HeGDC, 350 deg. C PFC bake-out followed by D 2 and HeGDC, and experiments to test fueling discharges with either a He-trimethylboron mixture or pure trimethylboron. The impact of this impurity and density control program on recent advances in NSTX plasma performance is discussed

  15. Progress towards RF heated steady-state plasma operations on LHD by employing ICRF heating methods and improved divertor plates

    International Nuclear Information System (INIS)

    Kumazawa, R.; Mutoh, T.; Saito, K.

    2008-10-01

    A long pulse plasma discharge experiment was carried out using RF heating power in the Large Helical Device (LHD), a currentless magnetic confining system. Progress in long pulse operation is summarized since the 10th experimental campaign (2006). A scaling relation of the plasma duration time to the applied RF power has been derived from the experimental data so far collected. It indicates that there exists a critical divertor temperature and consequently a critical RF heating power P RFcrit =0.65 MW. The area on the graph of the duration time versus the RF heating power was extended over the scaling relation by replacing divertor plates with new ones with better heat conductivity. The cause of the plasma collapse at the end of the long pulse operation was found to be the penetration of metal impurities. Many thin flakes consisting of heavy metals and graphite in stratified layers were found on the divertor plates and it was thought that they were the cause of impurity metals penetrating into the plasma. In a simulation involving injecting a graphite-coated Fe pellet to the plasma it was found that 230 Eμm in the diameter of the Fe pellet sphere was the critical size which led the plasma to collapse. A mode-conversion heating method was examined in place of the minority ICRF heating which has been employed in almost all the long-pulse plasma discharges. It was found that this method was much better from the viewpoint of achieving uniformity of the plasma heat load to the divertors. It is expected that P RFcrit will be increased by using the mode-conversion heating method. (author)

  16. Divertor erosion in DIII-D

    International Nuclear Information System (INIS)

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

    1998-08-01

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

  17. Divertor erosion in DIII-D

    International Nuclear Information System (INIS)

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

    1998-05-01

    Net erosion rates of carbon target plates have been measured in situ for the DIII-D lower divertor. The principal method of obtaining this data is the DiMES sample probe. Recent experiments have focused on erosion at the outer strike-point of two divertor plasma conditions: (1) attached (Te > 40 eV) ELMing plasmas and (2) detached (Te 10 cm/year, even with incident heat flux 2 . In this case, measurements and modeling agree for both gross and net carbon erosion, showing the near-surface transport and redeposition of the carbon is well understood and that effective sputtering yields are > 10%. In ELM-free discharges, this erosion rate can account for the rate of carbon accumulation in the core plasma. Divertor plasma detachment eliminates physical sputtering, while spectroscopically measured chemical erosion yields are also found to be low (Y(C/D + ) ≤ 2.0 x 10 -3 ). This leads to suppression of net erosion at the outer strike-point, which becomes a region of net redeposition (∼ 4 cm/year). The private flux wall is measured to be a region of net redeposition with dense, high neutral pressure, attached divertor plasmas. Leading edges intercepting parallel heat flux (∼ 50 MW/m 2 ) have very high net erosion rates (∼ 10 microm/s) at the OSP of an attached plasma. Leading edge erosion, and subsequent carbon redeposition, caused by tile gaps can account for half of the deuterium codeposition in the DIII-D divertor

  18. Operating conditions of the BPX divertor

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  19. A graphite nanoeraser

    DEFF Research Database (Denmark)

    Liu, Ze; Bøggild, Peter; Yang, Jia-rui

    2011-01-01

    We present here a method for cleaning intermediate-size (up to 50 nm) contamination from highly oriented pyrolytic graphite and graphene. Electron-beam-induced deposition of carbonaceous material on graphene and graphite surfaces inside a scanning electron microscope, which is difficult to remove...... by conventional techniques, can be removed by direct mechanical wiping using a graphite nanoeraser, thus drastically reducing the amount of contamination. We discuss potential applications of this cleaning procedure....

  20. Control and data acquisition upgrades for NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Davis, W.M., E-mail: bdavis@pppl.gov; Tchilinguirian, G.J., E-mail: gtchilin@pppl.gov; Carroll, T., E-mail: tcarroll@pppl.gov; Erickson, K.G., E-mail: kerickson@pppl.gov; Gerhardt, S.P., E-mail: sgerhardt@pppl.gov; Henderson, P., E-mail: phenderson@pppl.gov; Kampel, S.H., E-mail: skampel@pppl.gov; Sichta, P., E-mail: psichta@pppl.gov; Zimmer, G.N., E-mail: gzimmer@pppl.gov

    2016-11-15

    Highlights: • The NSTX-U upgrade is nearing completion, and various control and data acquisition upgrades are needed. • The Digital Coil Protection System is a major addition which provides hardware and software to protect the magnetic coils from the complex, increased, stresses added from the upgrade. • The increased computational requirements for the upgrade have largely followed Moore’s Law, and enhancements to the infrastructure and computer hardware should maintain or exceed the previous functionality. • Data requirements for Fast 2-D cameras have exceeded those of “conventional” time-varying signals. There has been a particular emphasis and increase in data from IR cameras. - Abstract: The extensive NSTX Upgrade (NSTX-U) Project includes major components which allow a doubling of the toroidal field strength to 1 T, of the Neutral Beam heating power to 12 MW, and the plasma current to 2 MA, and substantial structural enhancements to withstand the increased electromagnetic loads. The maximum pulse length will go from 1.5 to 5 s. The larger and more complex forces on the coils will be protected by a Digital Coil Protection System, which requires demanding real-time data input rates, calculations and responses. The amount of conventional digitized data for a given pulse is expected to increase from 2.5 to 5 GB per second of pulse. 2-D Fast Camera data is expected to go from 2.5 GB/pulse to 10, and another 2 GB/pulse is expected from new IR cameras. Our network capacity will be increased by a factor of 10, with 10 Gb/s fibers used for the major trunks. 32-core Linux systems will be used for several functions, including between-shot data processing, MDSplus data serving, between-shot EFIT analysis, real-time processing, and for a new capability, between-shot TRANSP. Improvements to the MDSplus events subsystem will be made through the use of both UDP and TCP/IP based methods and the addition of a dedicated “event server”.

  1. Oxidation Resistant Graphite Studies

    Energy Technology Data Exchange (ETDEWEB)

    W. Windes; R. Smith

    2014-07-01

    The Very High Temperature Reactor (VHTR) Graphite Research and Development Program is investigating doped nuclear graphite grades exhibiting oxidation resistance. During a oxygen ingress accident the oxidation rates of the high temperature graphite core region would be extremely high resulting in significant structural damage to the core. Reducing the oxidation rate of the graphite core material would reduce the structural effects and keep the core integrity intact during any air-ingress accident. Oxidation testing of graphite doped with oxidation resistant material is being conducted to determine the extent of oxidation rate reduction. Nuclear grade graphite doped with varying levels of Boron-Carbide (B4C) was oxidized in air at nominal 740°C at 10/90% (air/He) and 100% air. The oxidation rates of the boronated and unboronated graphite grade were compared. With increasing boron-carbide content (up to 6 vol%) the oxidation rate was observed to have a 20 fold reduction from unboronated graphite. Visual inspection and uniformity of oxidation across the surface of the specimens were conducted. Future work to determine the remaining mechanical strength as well as graphite grades with SiC doped material are discussed.

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  5. Effect of Boronization on Ohmic Plasmas in NSTX

    International Nuclear Information System (INIS)

    Skinner, C.H.; Kugel, H.; Maingi, R.; Wampler, W.R.; Blanchard, W.; Bell, M.; Bell, R.; LeBlanc, B.; Gates, D.; Kaye, S.; LaMarche, P.; Menard, J.; Mueller, D.; Na, H.K.; Nishino, N.; Paul, S.; Sabbagh, S.; Soukhanovskii, V.

    2001-01-01

    Boronization of the National Spherical Torus Experiment (NSTX) has enabled access to higher density, higher confinement plasmas. A glow discharge with 4 mTorr helium and 10% deuterated trimethyl boron deposited 1.7 g of boron on the plasma facing surfaces. Ion beam analysis of witness coupons showed a B+C areal density of 10 to the 18 (B+C) cm to the -2 corresponding to a film thickness of 100 nm. Subsequent ohmic discharges showed oxygen emission lines reduced by x15, carbon emission reduced by two and copper reduced to undetectable levels. After boronization, the plasma current flattop time increased by 70% enabling access to higher density, higher confinement plasmas

  6. Characteristics of the First H-mode Discharges in NSTX

    International Nuclear Information System (INIS)

    Maingi, R.; Bell, M.G.; Bell, R.E.; Bush, C.E.; Fredrickson, E.D.; Gates, D.A.; Kaye, S.M.; Kugel, H.W.; LeBlanc, B.P.; Menard, J.E.; Mueller, D.; Sabbagh, S.A.; Stutman, D.; Taylor, G.; Johnson, D.W.; Kaita, R.; Maqueda, R.J.; Ono, M.; Paoletti, F.; Peng, Y-K.M.; Roquemore, A.L.; Skinner, C.H.; Soukhanovskii, V.A.; Synakowski, E.J.

    2001-01-01

    We report observations of the first low-to-high (L-H) confinement mode transitions in the National Spherical Torus Experiment (NSTX). The H-mode energy confinement time increased over reference L-mode discharges transiently by 100-300%, as high as ∼150 ms. This confinement time is ∼1.8-2.3 times higher than predicted by a multi-machine ELM-free H-mode scaling. This achievement extends the H-mode window of fusion devices down to a record low aspect ratio (R/a) ∼ 1.3, challenging both confinement and L-H power thresholds scalings based on conventional aspect ratio tokamaks

  7. Energy Exchange Dynamics across L-H transitions in NSTX

    Science.gov (United States)

    Diallo, Ahmed

    2017-10-01

    H-mode is planned for future devices such as ITER, and is preceded by a low (L) to high (H) transition. A key question remains. What is the mechanism behind the L-H transition? Most theoretical descriptions of the L-H transition are based on the shear of the radial electric field and coincident ExB poloidal flow shear, which is thought to be responsible for the onset of the anomalous transport suppression that leads to the L-H transition. This talk will focus on the analysis of the flow dynamics across the L-H transition in NSTX. We analyze the L-H transition dynamics using the velocimetry of 2D edge turbulence data from gas-puff imaging (GPI). We determine the velocity components at the edge across the L-H transition for 17 discharges with three types of heating power (NBI, ohmic, and RF). Using a reduced model equation of edge flows and turbulence, the energy transfer dynamics is compared with the turbulence depletion hypothesis of the predator-prey model. In order for Reynolds work to suppress the turbulence, it must deplete the total turbulent free energy, including the thermal free-energy term. For this to occur, the increase in kinetic energy in the mean flow over the L-H transition must be comparable to the pre-transition thermal free energy. However, this ratio was found to be of order 10-2. Although there are significant simplifications in the theoretical model, they are unlikely to cause inaccuracy by two orders of magnitude, suggesting that direct turbulence depletion by the Reynolds work may not be large enough to explain the L-H transition on NSTX, contrary to the predator-prey model. This work is supported by the US DOE Contract No. DE-AC02-09CH11466.

  8. Energy exchange dynamics across L-H transitions in NSTX

    Science.gov (United States)

    Diallo, A.; Banerjee, S.; Zweben, S. J.; Stoltzfus-Dueck, T.

    2017-06-01

    We studied the energy exchange dynamics across the low-to-high-confinement (L-H) transition in NSTX discharges using the gas-puff imaging (GPI) diagnostic. The investigation focused on the energy exchange between flows and turbulence to help clarify the mechanism of the L-H transition. We applied this study to three types of heating schemes, including a total of 17 shots from the NSTX 2010 campaign run. Results show that the edge fluctuation characteristics (fluctuation levels, radial and poloidal correlation lengths) measured using GPI do not vary just prior to the H-mode transition, but change after the transition. Using a velocimetry approach (orthogonal-dynamics programming), velocity fields of a 24× 30 cm GPI view during the L-H transition were obtained with good spatial (˜1 cm) and temporal (˜2.5 μs) resolutions. Analysis using these velocity fields shows that the production term is systematically negative just prior to the L-H transition, indicating a transfer from mean flows to turbulence, which is inconsistent with the predator-prey paradigm. Moreover, the inferred absolute value of the production term is two orders of magnitude too small to explain the observed rapid L-H transition. These discrepancies are further reinforced by consideration of the ratio between the kinetic energy in the mean flow to the thermal free energy, which is estimated to be much less than 1, suggesting again that the turbulence depletion mechanism may not play an important role in the transition to the H-mode. Although the Reynolds work therefore appears to be too small to directly deplete the turbulent free energy reservoir, order-of-magnitude analysis shows that the Reynolds stress may still make a non-negligible contribution to the observed poloidal flows.

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

    Directory of Open Access Journals (Sweden)

    J. Uljanovs

    2017-08-01

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

  10. Method for producing dustless graphite spheres from waste graphite fines

    Science.gov (United States)

    Pappano, Peter J [Oak Ridge, TN; Rogers, Michael R [Clinton, TN

    2012-05-08

    A method for producing graphite spheres from graphite fines by charging a quantity of spherical media into a rotatable cylindrical overcoater, charging a quantity of graphite fines into the overcoater thereby forming a first mixture of spherical media and graphite fines, rotating the overcoater at a speed such that the first mixture climbs the wall of the overcoater before rolling back down to the bottom thereby forming a second mixture of spherical media, graphite fines, and graphite spheres, removing the second mixture from the overcoater, sieving the second mixture to separate graphite spheres, charging the first mixture back into the overcoater, charging an additional quantity of graphite fines into the overcoater, adjusting processing parameters like overcoater dimensions, graphite fines charge, overcoater rotation speed, overcoater angle of rotation, and overcoater time of rotation, before repeating the steps until graphite fines are converted to graphite spheres.

  11. Graphite targets at LAMPF

    International Nuclear Information System (INIS)

    Brown, R.D.; Grisham, D.L.

    1983-01-01

    Rotating polycrystalline and stationary pyrolytic graphite target designs for the LAMPF experimental area are described. Examples of finite element calculations of temperatures and stresses are presented. Some results of a metallographic investigation of irradiated pyrolytic graphite target plates are included, together with a brief description of high temperature bearings for the rotating targets

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

    Science.gov (United States)

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

    2016-10-01

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

  13. Development of a radiative divertor for DIII-D

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  14. Characteristics of the Secondary Divertor on DIII-D

    Science.gov (United States)

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

    2009-11-01

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

  15. ITER tungsten divertor design development and qualification program

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  16. Role of molecular effects in divertor plasma recombination

    Directory of Open Access Journals (Sweden)

    A.S. Kukushkin

    2017-08-01

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

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

    International Nuclear Information System (INIS)

    Ikegami, Hideo.

    1995-01-01

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

  18. Electrochemical treatment of graphite

    Energy Technology Data Exchange (ETDEWEB)

    Podlovilin, V.I.; Egorov, I.M.; Zhernovoj, A.I.

    1983-01-01

    In the course of investigating various modes of electrochemical treatment (ECT) it has been found that graphite anode treatment begins under the ''glow mode''. A behaviour of some marks of graphite with the purpose of ECT technique development in different electrolytes has been tested. Electrolytes have been chosen of three types: highly alkaline (pH 13-14), neutral (pH-Z) and highly acidic (pH 1-2). For the first time parallel to mechanical electroerosion treatment, ECT of graphite and carbon graphite materials previously considered chemically neutral is proposed. ECT of carbon graphite materials has a number of advantages as compared with electroerrosion and mechanical ones with respect to the treatment rate and purity (ronghness) of the surface. A small quantity of sludge (6-8%) under ECT is in highly alkali electrolytes.

  19. Electrochemical treatment of graphite

    International Nuclear Information System (INIS)

    Podlovilin, V.I.; Egorov, I.M.; Zhernovoj, A.I.

    1983-01-01

    In the course of investigating various modes of electroche-- mical treatment (ECT) it has been found that graphite anode treatment begins under the ''glow mode''. A behaviour of some marks of graphite with the purpose of ECT technique development in different electrolytes has been tested. Electrolytes have been chosen of three types: highly alkaline (pH 13-14), neutral (pH-Z) and highly acidic (pH 1-2). For the first time parallel to mechanical electroerosion treatment ECT graphite and carbon graphite materials previously considered chemically neutral is proposed. ECT of carbon graphite materials has a number of advantages as compared with electroerrosion and mechanical ones this is treatment rate and purity (ronghness) of the surface. A sMall quantity of sludge (6-8%) under ECT is in highly alkali electrolytes

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-02-01

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

  1. SLAC divertor channel entrance thermal stress analysis

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  2. Thermal effects of divertor sweeping in ITER

    International Nuclear Information System (INIS)

    Wesley, J.C.

    1992-01-01

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

  3. He-cooled divertor development for DEMO

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  4. Boundary plasma control with the ergodic divertor

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  5. Boundary plasma control with the ergodic divertor

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  6. Electron beam facility for divertor target experiments

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  8. Comparative divertor-transport study for helical devices

    International Nuclear Information System (INIS)

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

    2008-10-01

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

  9. Experimental studies of the snowflake divertor in TCV

    Directory of Open Access Journals (Sweden)

    B. Labit

    2017-08-01

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

  10. Divertor experiment for impurity control in DIVA

    International Nuclear Information System (INIS)

    Nagami, Masayuki

    1979-04-01

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

  11. Two-point model for divertor transport

    International Nuclear Information System (INIS)

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

    1984-04-01

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

  12. Divertor and scoop limiter experiments on PDX

    International Nuclear Information System (INIS)

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

    1985-01-01

    Routine operation in the enhanced energy confinement (or H-mode) regime during neutral beam injection was achieved by modifying the PDX divertor hardware to inhibit the influx of neutral gas from the divertor region to the main plasma chamber. A particle scoop limiter has been studied as a mechanical means of controlling particles at the plasma edge, and neutral beam heated discharges with this limiter show similar confinement times (normalized to tau/sub E//I/sub p/) to average H-mode plasmas. Two new instabilities are observed near the plasma edge in PDX during H-mode operation. The first, a quasicoherent fluctuation, occurred in bursts at well-defined frequencies (Δω/ω less than or equal to 0.1) in the range 50 to 180 kHz, and had no obvious effects on confinement. The second instability, the edge relaxation phenomena (ERP), did cause deterioration in the global confinement time. The ERP's are characterized by sharp spikes in the divertor plasma density, H/sub α/ emission, and on the x-ray signals they appear as sawtoothlike relaxations at the plasma edge with an inversion radius near the separatrix. Attempts to obtain high β/sub T/ in the H-mode discharges were hampered by a deterioration in the H-mode confinement and major disruptions which limited the achievable β/sub T/. A study of the stability of both the limiter L-mode and divertor H-mode discharges close to the theoretical β boundary, showed that the major disruptions observed there are sometimes caused by a fast growing m/n = 1/1 mode with no observable external precursor oscillations

  13. Divertor and scoop limiter experiments on PDX

    International Nuclear Information System (INIS)

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

    1985-01-01

    Routine operation in the enhanced-energy-confinement (or H-mode) regime during neutral-beam injection was achieved by modifying the PDX divertor hardware to inhibit the influx of neutral gas from the divertor region to the main plasma chamber. A particle scoop limiter has been studied as a mechanical means of controlling particles at the plasma edge, and neutral-beam-heated discharges with this limiter show similar confinement times (normalized to tausub(E)/Isub(p)) to average H-mode plasma. Two new instabilities are observed near the plasma edge in PDX during H-mode operation. The first, a quasi-coherent fluctuation, occurred in bursts at well-defined frequencies (Δω/ω<=0.1) in the range 50 to 180 kHz, and had no obvious effects on confinement. The second instability, the edge relaxation phenomena (ERP), did cause deterioration in the global confinement time. The ERPs are characterized by sharp spikes in the divertor plasma density, Hsub(α) emission, and on the X-ray signals they appear as sawtooth-like relaxations at the plasma edge with an inversion radius near the separatrix. Attempts to obtain high βsub(T) in the H-mode discharges were hampered by a deterioration in the H-mode confinement and major disruptions which limited the achievable βsub(T). A study of the stability of both the limiter L-mode and divertor H-mode discharge close to the theoretical β boundary showed that the major disruptions observed there are sometimes caused by a fast growing m/n=1/1 mode with no observable external precursor oscillations. (author)

  14. The remote exchange of the JET divertor

    International Nuclear Information System (INIS)

    Pick, M.

    1999-01-01

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

  15. Low energy neutral particle fluxes in the JET divertor

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  16. Towards a physics-integrated view on divertor pumping

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    International Nuclear Information System (INIS)

    Ruchti, C.

    1981-01-01

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

  18. Influence of stray light for divertor spectroscopy in ITER

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  19. Technological development of the Monobloc Divertor Concept

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  20. Westinghouse compact poloidal divertor reference design

    International Nuclear Information System (INIS)

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

    1977-08-01

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

  1. THERMAL HYDRAULIC ANALYSIS OF FIRE DIVERTOR

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  2. A new scaling for divertor detachment

    Science.gov (United States)

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

    2017-05-01

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

  3. Island divertor studies on W7-AS

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  4. Asymptomatic Intracorneal Graphite Deposits following Graphite Pencil Injury

    OpenAIRE

    Philip, Swetha Sara; John, Deepa; John, Sheeja Susan

    2012-01-01

    Reports of graphite pencil lead injuries to the eye are rare. Although graphite is considered to remain inert in the eye, it has been known to cause severe inflammation and damage to ocular structures. We report a case of a 12-year-old girl with intracorneal graphite foreign bodies following a graphite pencil injury.

  5. On residual gas analysis during high temperature baking of graphite tiles

    International Nuclear Information System (INIS)

    Prakash, A A; Chaudhuri, P; Khirwadkar, S; Reddy, D Chenna; Saxena, Y C; Chauhan, N; Raole, P M

    2008-01-01

    Steady-state Super-conducting Tokamak-1 (SST-1) is a medium size tokamak with major radius of 1.1 m and minor radius of 0.20 m. It is designed for plasma discharge duration of 1000 seconds to obtain fully steady-state plasma operation. Plasma Facing Components (PFC), consisting of divertors, passive stabilizers, baffles and poloidal limiters are also designed to be UHV compatible for steady state operation. All PFC are made up of graphite tiles mechanically attached to the copper alloy substrate. Graphite is one of the preferred first wall armour material in present day tokamaks. High thermal shock resistance and low atomic number of carbon are the most important properties of graphite for this application. High temperature vacuum baking of graphite tiles is the standard process to remove the impurities. Residual Gas Analyzer (RGA) has been used for qualitative and quantitative measurements of released gases from graphite tiles during baking. Surface Analysis of graphite tiles has also been done before and after baking. This paper describes the residual gas analysis during baking and surface analysis of graphite tiles

  6. On residual gas analysis during high temperature baking of graphite tiles

    Energy Technology Data Exchange (ETDEWEB)

    Prakash, A A; Chaudhuri, P; Khirwadkar, S; Reddy, D Chenna; Saxena, Y C [Institute for Plasma Research, Bhat, Gandhinagar - 382 428 (India); Chauhan, N; Raole, P M [Facilitation Center for Industrial Plasma Technologies, IPR, Gandhinagar (India)], E-mail: arun@ipr.res.in

    2008-05-01

    Steady-state Super-conducting Tokamak-1 (SST-1) is a medium size tokamak with major radius of 1.1 m and minor radius of 0.20 m. It is designed for plasma discharge duration of 1000 seconds to obtain fully steady-state plasma operation. Plasma Facing Components (PFC), consisting of divertors, passive stabilizers, baffles and poloidal limiters are also designed to be UHV compatible for steady state operation. All PFC are made up of graphite tiles mechanically attached to the copper alloy substrate. Graphite is one of the preferred first wall armour material in present day tokamaks. High thermal shock resistance and low atomic number of carbon are the most important properties of graphite for this application. High temperature vacuum baking of graphite tiles is the standard process to remove the impurities. Residual Gas Analyzer (RGA) has been used for qualitative and quantitative measurements of released gases from graphite tiles during baking. Surface Analysis of graphite tiles has also been done before and after baking. This paper describes the residual gas analysis during baking and surface analysis of graphite tiles.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

  8. A carbon-metal brazing for divertor plates in fusion devices

    International Nuclear Information System (INIS)

    Matsuda, T.; Matsumoto, T.; Miki, S.; Sogabe, T.; Okada, M.; Kubota, Y.; Sagara, A.; Noda, N.; Motojima, O.; Hino, T.; Yamashina, T.

    1993-01-01

    A divertor unit, which consists of carbon armors brazed to a copper cooling channel, is under development for fusion devices. Isotropic graphite (IG-430U) and CFC (CX-2002U) are used for the armor, and a copper for the cooling tube. A technique named as dissolution and deposit of base metal was employed for brazing. The reliability of the brazed components was evaluated both by 4-point bending test and thermal shock test. According to the results of a 4-point bending test under the temperature ranged from RT to 800 C in a vacuum, it was found that the strength of the brazed surface at RT was maintained up to the higher temperature, 600 C. High heat load test has been also performed on the brazed sample in order to find whether the samples meet the requirement of the divertor plates of LHD (Large Helical Device). Active Cooling Teststand (ACT:NIFS) with electron beam power of 100kW was used. In LHD, it is presumed that the maximum heat flux is 10MW/m 2 . In addition, the surface temperature of divertor has to be kept below 1,200 C to avoid RES, by active cooling. The heat load test showed that the brazing components of CX-2002U (flat plate type CFC-Cu brazed) was stable at 1,300 C under a heat flux of 10MW/m 2 , when the flow velocity of cooling water was 6m/s. No damage nor deterioration was found at the brazed zone after the heat load test

  9. Analysis of vertical stability limits and vertical displacement event behavior on NSTX-U

    Science.gov (United States)

    Boyer, Mark; Battaglia, Devon; Gerhardt, Stefan; Menard, Jonathan; Mueller, Dennis; Myers, Clayton; Sabbagh, Steven; Smith, David

    2017-10-01

    The National Spherical Torus Experiment Upgrade (NSTX-U) completed its first run campaign in 2016, including commissioning a larger center-stack and three new tangentially aimed neutral beam sources. NSTX-U operates at increased aspect ratio due to the larger center-stack, making vertical stabilization more challenging. Since ST performance is improved at high elongation, improvements to the vertical control system were made, including use of multiple up-down-symmetric flux loop pairs for real-time estimation, and filtering to remove noise. Similar operating limits to those on NSTX (in terms of elongation and internal inductance) were achieved, now at higher aspect ratio. To better understand the observed limits and project to future operating points, a database of vertical displacement events and vertical oscillations observed during the plasma current ramp-up on NSTX/NSTX-U has been generated. Shots were clustered based on the characteristics of the VDEs/oscillations, and the plasma parameter regimes associated with the classes of behavior were studied. Results provide guidance for scenario development during ramp-up to avoid large oscillations at the time of diverting, and provide the means to assess stability of target scenarios for the next campaign. Results will also guide plans for improvements to the vertical control system. Work supported by U.S. D.O.E. Contract No. DE-AC02-09CH11466.

  10. Recent developments in graphite

    International Nuclear Information System (INIS)

    Cunningham, J.E.

    1983-01-01

    Overall, the HTGR graphite situation is in excellent shape. In both of the critical requirements, fuel blocks and support structures, adequate graphites are at hand and improved grades are sufficiently far along in truncation. In the aerospace field, GraphNOL N3M permits vehicle performance with confidence in trajectories unobtainable with any other existing material. For fusion energy applications, no other graphite can simultaneously withstand both extreme thermal shock and neutron damage. Hence, the material promises to create new markets as well as to offer a better candidate material for existing applications

  11. Graphite for fusion energy applications

    International Nuclear Information System (INIS)

    Eatherly, W.P.; Clausing, R.E.; Strehlow, R.A.; Kennedy, C.R.; Mioduszewski, P.K.

    1987-03-01

    Graphite is in widespread and beneficial use in present fusion energy devices. This report reflects the view of graphite materials scientists on using graphite in fusion devices. Graphite properties are discussed with emphasis on application to fusion reactors. This report is intended to be introductory and descriptive and is not intended to serve as a definitive information source

  12. Conceptual design of a high temperature water-cooled divertor for a fusion power reactor

    International Nuclear Information System (INIS)

    Giancarli, L.; Bonal, J.P.; Puma, A. Li; Michel, B.; Sardain, P.; Salavy, J.F.

    2005-01-01

    This paper presents the conceptual design of a water-cooled divertor target using EUROFER as structural material, water coolant pressure and outlet temperature, respectively, of 15.5 MPa and 325 o C, and W-alloy monoblocks as armour. Assuming an advanced interface, formed by a thermal barrier in the pipe front part and a compliance layer between W and steel, this concept is able to withstand an incident surface heat flux of 15 MW/m 2 . Both thermal barrier and compliance layer are made of carbon-based materials. The main issues are the manufacturing process of the steel/W interface, and the behaviour under irradiation of graphite materials

  13. Conceptual design of a high temperature water-cooled divertor for a fusion power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Giancarli, L. [CEA Saclay, Direction de l' Energie Nucleaire, F-91191 Gif-sur-Yvette (France)]. E-mail: luciano.giancarli@cea.fr; Bonal, J.P. [CEA Saclay, Direction de l' Energie Nucleaire, F-91191 Gif-sur-Yvette (France); Puma, A. Li [CEA Saclay, Direction de l' Energie Nucleaire, F-91191 Gif-sur-Yvette (France); Michel, B. [CEA Cadarache, Direction de l' Energie Nucleaire, F-13108 St. Paul-les-Durances (France); Sardain, P. [EFDA Close Support Unit, Boltzmannstr. 2, D-85748 Garching (Germany); Salavy, J.F. [CEA Saclay, Direction de l' Energie Nucleaire, F-91191 Gif-sur-Yvette (France)

    2005-11-15

    This paper presents the conceptual design of a water-cooled divertor target using EUROFER as structural material, water coolant pressure and outlet temperature, respectively, of 15.5 MPa and 325 {sup o}C, and W-alloy monoblocks as armour. Assuming an advanced interface, formed by a thermal barrier in the pipe front part and a compliance layer between W and steel, this concept is able to withstand an incident surface heat flux of 15 MW/m{sup 2}. Both thermal barrier and compliance layer are made of carbon-based materials. The main issues are the manufacturing process of the steel/W interface, and the behaviour under irradiation of graphite materials.

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

    International Nuclear Information System (INIS)

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

    1996-09-01

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

  15. Plasma flow in the DIII-D divertor

    International Nuclear Information System (INIS)

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

    1998-07-01

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

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

    Science.gov (United States)

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

    2016-02-01

    The primary goal of the WEST project is to be a test bed to characterize the fatigue and lifetime of ITER-like W divertor components subjected to relevant thermal loads. During the first phase of exploitation (S2 2016), these components (W monoblock plasma facing unit—W-PFU) will be installed in conjunction with graphite components (G-PFU). Since the G-PFU will not be actively cooled, it is necessary to ensure the expected pulse duration allows the W-PFU to reach its steady state without overheating the G-PFU assembly structure or the embedded stainless-steel diagnostics. High heat flux tests were performed at the GLADIS facility to assess the thermal behavior of the G-PFU. Some operational limits based on plasma parameters were determined. It was found that it is possible to operate at an injected power such that the maximal incident heat flux on the lower divertor is 10 MW m-2 for the required pulse length.

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

    International Nuclear Information System (INIS)

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

    2016-01-01

    The primary goal of the WEST project is to be a test bed to characterize the fatigue and lifetime of ITER-like W divertor components subjected to relevant thermal loads. During the first phase of exploitation (S2 2016), these components (W monoblock plasma facing unit—W-PFU) will be installed in conjunction with graphite components (G-PFU). Since the G-PFU will not be actively cooled, it is necessary to ensure the expected pulse duration allows the W-PFU to reach its steady state without overheating the G-PFU assembly structure or the embedded stainless-steel diagnostics. High heat flux tests were performed at the GLADIS facility to assess the thermal behavior of the G-PFU. Some operational limits based on plasma parameters were determined. It was found that it is possible to operate at an injected power such that the maximal incident heat flux on the lower divertor is 10 MW m −2 for the required pulse length. (paper)

  18. Deuterium depth profiling in JT-60U W-shaped divertor tiles by nuclear reaction analysis

    International Nuclear Information System (INIS)

    Hayashi, T.; Ochiai, K.; Masaki, K.; Gotoh, Y.; Kutsukake, C.; Arai, T.; Nishitani, T.; Miya, N.

    2006-01-01

    Deuterium concentrations and depth profiles in plasma-facing graphite tiles used in the divertor of JAERI Tokamak-60 Upgrade (JT-60U) were investigated by nuclear reaction analysis (NRA). The highest deuterium concentration of D/ 12 C of 0.053 was found in the outer dome wing tile, where the deuterium accumulated probably through the deuterium-carbon co-deposition. In the outer and inner divertor target tiles, the D/ 12 C data were lower than 0.006. Additionally, the maximum (H + D)/ 12 C in the dome top tile was estimated to be 0.023 from the results of NRA and secondary ion mass spectroscopy (SIMS). Orbit following Monte-Carlo (OFMC) simulation showed energetic deuterons caused by neutral beam injections (NBI) were implanted into the dome region with high heat flux. Furthermore, the surface temperature and conditions such as deposition and erosion significantly influenced the accumulation process of deuterium. The deuterium depth profile, scanning electron microscope (SEM) observation and OFMC simulation indicated the deuterium was considered to accumulate through three processes: the deuterium-carbon co-deposition, the implantation of energetic deuterons and the deuterium diffusion into the bulk

  19. Physics of integrated high-performance NSTX plasmas

    International Nuclear Information System (INIS)

    Menard, J. E.; Bell, M. G.; Bell, R. E.; Fredrickson, E. D.; Gates, D. A.; Heidbrink, W.; Kaita, R.; Kaye, S. M.; Kessel, C. E.; Kugel, H.; LeBlanc, B. P.; Lee, K. C.; Levinton, F. M.; Maingi, R.; Medley, S. S.; Mikkelsen, D. R.; Mueller, D.; Nishino, N.; Ono, M.; Park, H.; Park, W.; Paul, S. F.; Peebles, T.; Peng, M.; Raman, R.; Redi, M.; Roquemore, L.; Sabbagh, S. A.; Skiner, C. H.; Sontag, A.; Soukhanovskii, V.; Stratton, B.; Stutman, D.; Synakowski, E.; Takase, Y.; Taylor, G.; Tritz, K.; Wade, M.; Wilson, J. R.; Zhu, W.

    2005-01-01

    An overarching goal of magnetic fusion research is the integration of steady state operation with high fusion power density, high plasma β, good thermal and fast particle confinement, and manageable heat and particle fluxes to reactor internal components. NSTX has made significant progress in integrating and understanding the interplay between these competing elements. Sustained high elongation up to 2.5 and H-mode transitions during the I p ramp-up have increased β p and reduced l i at high current resulting in I p flat-top durations exceeding 0.8s for I p >0.8MA. These shape and profile changes delay the onset of deleterious global MHD activity yielding β N values >4.5 and β T ∼20% maintained for several current diffusion times. Higher ∫ N discharges operating above the non-wall limit are sustained via rotational stabilization of the RWM. H-mode confinement scaling factors relative to H98(y,2) span the range 1±0.4 for B T >4kG and show a stron (Nearly linear) residual scaling with B T . Power balance analysis indicates the electron thermal transport dominates the loss power in beam-heated H m ode discharges, but the core χ e can be significantly reduced through current profile modification consistent with reversed magnetic shear. Small ELM regimes have been obtained in high performance plasmas on NSTX, but the ELM type and associated pedestal energy loss are found to depend sensitively on the boundary elongation, magnetic balance, and edge collisionality. NPA data and TRANSP analysis suggest resonant interactions with mid-radius tearing modes may lead to large fast-ion transport. The associated fast-ion diffusion and/or loss likely impact(s) both the driven current and power deposition profiles from NBI heating. Results from experiments to initiate the plasma without the ohmic solenoid and integrated scenario with the TSC code will also be described. (Author)

  20. Electron Bernstein wave simulations and comparison to preliminary NSTX emission data

    International Nuclear Information System (INIS)

    Preinhaelter, Josef; Urban, Jakub; Pavlo, Pavol; Taylor, Gary; Diem, Steffi; Vahala, Linda; Vahala, George

    2006-01-01

    Simulations indicate that during flattop current discharges the optimal angles for the aiming of the National Spherical Torus Experiment (NSTX) antennae are quite rugged and basically independent of time. The time development of electron Bernstein wave emission (EBWE) at particular frequencies as well as the frequency spectrum of EBWE as would be seen by the recently installed NSTX antennae are computed. The simulation of EBWE at low frequencies (e.g., 16 GHz) agrees well with the recent preliminary EBWE measurements on NSTX. At high frequencies, the sensitivity of EBWE to magnetic field variations is understood by considering the Doppler broadened electron cyclotron harmonics and the cutoffs and resonances in the plasma. Significant EBWE variations are seen if the magnetic field is increased by as little as 2% at the plasma edge. The simulations for the low frequency antenna are compared to preliminary experimental data published separately by Diem et al. [Rev. Sci. Instrum.77 (2006)

  1. Status of the Experimental Physics and Industrial Control System at NSTX

    International Nuclear Information System (INIS)

    Sichta, P.; Dong, J.

    2002-01-01

    The NSTX achieved first plasma in 1999. The Experimental Physics and Industrial Control System (EPICS) is used to provide data-integration services for monitoring and control of all NSTX engineering subsystems. EPICS is a set of software initially developed at U.S. DOE laboratories. It is currently used and maintained through a global collaboration of hundreds of scientists and engineers. This paper will relate some of our experiences using and supporting the EPICS software. Topics include reliability and maintainability, lessons learned, recently added engineering subsystems, new EPICS software tools, and a review of our first EPICS software upgrade. Steps to modernize the technical infrastructure of EPICS to ensure effective support for NSTX will also be described

  2. Three new extreme ultraviolet spectrometers on NSTX-U for impurity monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Weller, M. E., E-mail: weller4@llnl.gov; Beiersdorfer, P.; Soukhanovskii, V. A.; Magee, E. W.; Scotti, F. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States)

    2016-11-15

    Three extreme ultraviolet (EUV) spectrometers have been mounted on the National Spherical Torus Experiment–Upgrade (NSTX-U). All three are flat-field grazing-incidence spectrometers and are dubbed X-ray and Extreme Ultraviolet Spectrometer (XEUS, 8–70 Å), Long-Wavelength Extreme Ultraviolet Spectrometer (LoWEUS, 190–440 Å), and Metal Monitor and Lithium Spectrometer Assembly (MonaLisa, 50–220 Å). XEUS and LoWEUS were previously implemented on NSTX to monitor impurities from low- to high-Z sources and to study impurity transport while MonaLisa is new and provides the system increased spectral coverage. The spectrometers will also be a critical diagnostic on the planned laser blow-off system for NSTX-U, which will be used for impurity edge and core ion transport studies, edge-transport code development, and benchmarking atomic physics codes.

  3. Operation of the ultrasoft x-ray system on NSTX (abstract)

    International Nuclear Information System (INIS)

    Stutman, D.; Iovea, M.; Finkenthal, M.; Kaita, R.; Johnson, D.; Roquemore, L.; Roney, P.

    2001-01-01

    The ultrasoft x-ray imaging system on National Spherical Torus Experiment (NSTX) became operational and provided the first data in the filtered diode slow bow tie configuration. Using different band pass filters on each of three arrays allows an approximate spectroscopic estimate of the plasma impurity content, as well as of the electron temperature. Magnetohydrodynamics (MHD) activity from different plasma regions is also observed. The soft x-ray emission profiles are well behaved until an Internal Reconnection Event occurs. Examples of NSTX MHD phenomena seen in the ultrasoft x-ray emission under different operational regimes will be presented. From a technical point of view, we point out that the industrial PC based data acquisition system was not adversely affected by stray magnetic fields due to its close proximity to the NSTX device. Also, the surface barrier diodes withstood baking to 100 o C relatively well

  4. NSTX Protection And Interlock Systems For Coil And Powers Supply Systems

    International Nuclear Information System (INIS)

    Zhao, X.; Ramakrishnan, S.; Lawson, J.; Neumeyer, C.; Marsala, R.; Schneider, H.

    2009-01-01

    NSTX at Princeton Plasma Physics Laboratory (PPPL) requires sophisticated plasma positioning control system for stable plasma operation. TF magnetic coils and PF magnetic coils provide electromagnetic fields to position and shape the plasma vertically and horizontally respectively. NSTX utilizes twenty six coil power supplies to establish and initiate electromagnetic fields through the coil system for plasma control. A power protection and interlock system is utilized to detect power system faults and protect the TF coils and PF coils against excessive electromechanical forces, overheating, and over current. Upon detecting any fault condition the power system is restricted, and it is either prevented from initializing or suppressed to de-energize coil power during pulsing. Power fault status is immediately reported to the computer system. This paper describes the design and operation of NSTX's protection and interlocking system and possible future expansion.

  5. Far-infrared tangential interferometer/polarimeter design and installation for NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Scott, E. R., E-mail: evrscott@ucdavis.edu [Department of Mechanical and Aerospace Engineering, University of California, Davis, California 95616 (United States); Barchfeld, R. [Department of Applied Science, University of California, Davis, California 95616 (United States); Riemenschneider, P.; Domier, C. W.; Sohrabi, M.; Luhmann, N. C. [Department of Electrical and Computer Engineering, University of California, Davis, California 95616 (United States); Muscatello, C. M. [General Atomics, San Diego, California 92121 (United States); Kaita, R.; Ren, Y. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States)

    2016-11-15

    The Far-infrared Tangential Interferometer/Polarimeter (FIReTIP) system has been refurbished and is being reinstalled on the National Spherical Torus Experiment—Upgrade (NSTX-U) to supply real-time line-integrated core electron density measurements for use in the NSTX-U plasma control system (PCS) to facilitate real-time density feedback control of the NSTX-U plasma. Inclusion of a visible light heterodyne interferometer in the FIReTIP system allows for real-time vibration compensation due to movement of an internally mounted retroreflector and the FIReTIP front-end optics. Real-time signal correction is achieved through use of a National Instruments CompactRIO field-programmable gate array.

  6. Flux consumption optimization and the achievement of 1 MA discharges on NSTX

    International Nuclear Information System (INIS)

    Menard, J.; LeBlanc, B.; Sabbagh, S.A.

    2001-01-01

    The spherical tokamak (ST), because of its slender central column, has very limited volt-second capability relative to a standard aspect ratio tokamak of similar plasma cross-section. Recent experiments on the National Spherical Torus Experiment (NSTX) have begun to quantify and optimize the ohmic current drive efficiency in a MA-class ST device. Sustainable ramp-rates in excess of 5MA/sec during the current rise phase have been achieved on NSTX, while faster ramps generate significant MHD activity. Discharges with I P exceeding 1MA have been achieved in NSTX with nominal parameters: aspect ratio A=1.3-1.4, elongation κ=2-2.2, triangularity δ=0.4, internal inductance l i =0.6, and Ejima coefficient C E =0.35. Flux consumption efficiency results, performance improvements associated with first boronization, and comparisons to neoclassical resistivity are described. (author)

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  8. Engineering design of the Aries-IV gaseous divertor

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    1996-10-01

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

  10. Divertor impurity monitor for the International Thermonuclear Experimental Reactor

    Science.gov (United States)

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

    1999-01-01

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

  11. Physics design and experimental study of tokamak divertor

    International Nuclear Information System (INIS)

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

    2007-06-01

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

  12. Carbon-14 Graphitization Chemistry

    Science.gov (United States)

    Miller, James; Collon, Philippe; Laverne, Jay

    2014-09-01

    Accelerator Mass Spectrometry (AMS) is a process that allows for the analysis of mass of certain materials. It is a powerful process because it results in the ability to separate rare isotopes with very low abundances from a large background, which was previously impossible. Another advantage of AMS is that it only requires very small amounts of material for measurements. An important application of this process is radiocarbon dating because the rare 14C isotopes can be separated from the stable 14N background that is 10 to 13 orders of magnitude larger, and only small amounts of the old and fragile organic samples are necessary for measurement. Our group focuses on this radiocarbon dating through AMS. When performing AMS, the sample needs to be loaded into a cathode at the back of an ion source in order to produce a beam from the material to be analyzed. For carbon samples, the material must first be converted into graphite in order to be loaded into the cathode. My role in the group is to convert the organic substances into graphite. In order to graphitize the samples, a sample is first combusted to form carbon dioxide gas and then purified and reduced into the graphite form. After a couple weeks of research and with the help of various Physics professors, I developed a plan and began to construct the setup necessary to perform the graphitization. Once the apparatus is fully completed, the carbon samples will be graphitized and loaded into the AMS machine for analysis.

  13. Melting temperature of graphite

    International Nuclear Information System (INIS)

    Korobenko, V.N.; Savvatimskiy, A.I.

    2001-01-01

    Full Text: Pulse of electrical current is used for fast heating (∼ 1 μs) of metal and graphite specimens placed in dielectric solid media. Specimen consists of two strips (90 μm in thick) placed together with small gap so they form a black body model. Quasy-monocrystal graphite specimens were used for uniform heating of graphite. Temperature measurements were fulfilled with fast pyrometer and with composite 2-strip black body model up to melting temperature. There were fulfilled experiments with zirconium and tungsten of the same black body construction. Additional temperature measurements of liquid zirconium and liquid tungsten are made. Specific heat capacity (c P ) of liquid zirconium and of liquid tungsten has a common feature in c P diminishing just after melting. It reveals c P diminishing after melting in both cases over the narrow temperature range up to usual values known from steady state measurements. Over the next wide temperature range heat capacity for W (up to 5000 K) and Zr (up to 4100 K) show different dependencies of heat capacity on temperature in liquid state. The experiments confirmed a high quality of 2-strip black body model used for graphite temperature measurements. Melting temperature plateau of tungsten (3690 K) was used for pyrometer calibration area for graphite temperature measurement. As a result, a preliminary value of graphite melting temperature of 4800 K was obtained. (author)

  14. Numerical modeling and experimental simulation of vapor shield formation and divertor material erosion for ITER typical plasma disruptions

    International Nuclear Information System (INIS)

    Wuerz, H.; Arkhipov, N.I.; Bakhin, V.P.; Goel, B.; Hoebel, W.; Konkashbaev, I.; Landman, I.; Piazza, G.; Safronov, V.M.; Sherbakov, A.R.; Toporkov, D.A.; Zhitlukhin, A.M.

    1994-01-01

    The high divertor heat load during a tokamak plasma disruption results in sudden evaporation of a thin layer of divertor plate material, which acts as vapor shield and protects the target from further excessive evaporation. Formation and effectiveness of the vapor shield are theoretically modeled and experimentally investigated at the 2MK-200 facility under conditions simulating the thermal quench phase of ITER tokamak plasma disruptions. In the optical wavelength range C II, C III, C IV emission lines for graphite, Cu I, Cu II lines for copper and continuum radiation for tungsten samples are observed in the target plasma. The plasma expands along the magnetic field lines with velocities of (4±1)x10 6 cm/s for graphite and 10 5 cm/s for copper. Modeling was done with a radiation hydrodynamics code in one-dimensional planar geometry. The multifrequency radiation transport is treated in flux limited diffusion and in forward reverse transport approximation. In these first modeling studies the overall shielding efficiency for carbon and tungsten defined as ratio of the incident energy and the vaporization energy for power densities of 10 MW/cm 2 exceeds a factor of 30. The vapor shield is established within 2 μs, the power fraction to the target after 10 μs is below 3% and reaches in the stationary state after about 20 μs a value of around 1.5%. ((orig.))

  15. Thermal analysis of W VII-AS limiter system and presentation of a graphite-block concept

    International Nuclear Information System (INIS)

    Mukherjee, S.; Grigull, P.

    1989-01-01

    A 2D-finite element thermal analysis of the initial W VII-AS limiter system has been performed and is discussed. Furhter to this analysis a graphite block concept is presented. This concept has been numerically analyzed for applications as a limiter in plasma and nuclear fusion experimental devices. The results are described in this paper. This block concept seems to be also applicable to first wall and divertor designs; the graphite elements could be replaced by ceramic ones. (author). 10 refs.; 13 figs

  16. Particle control in the DIII-D advanced divertor

    International Nuclear Information System (INIS)

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

    1991-11-01

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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  18. Structural analysis of the ITER Divertor toroidal rails

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  19. Alternative divertor target concepts for next step fusion devices

    Science.gov (United States)

    Mazul, I. V.

    2016-12-01

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

  20. Simulation of the ASDEX divertor performance after hardening

    International Nuclear Information System (INIS)

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

    1985-05-01

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

  1. Divertor heat flux control and plasma-material interaction

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  2. Initial results from the NSTX Real-Time Velocity diagnostic

    Science.gov (United States)

    Podesta, M.; Bell, R. E.

    2011-10-01

    A new diagnostic for fast measurements of plasma rotation through active charge-exchange recombination spectroscopy (CHERS) was installed on NSTX. The diagnostic infers toroidal rotation from carbon ions undergoing charge-exchange with neutrals from a heating Neutral Beam (NB). Each of the 4 channels, distributed along the outer major radius, includes active views intercepting the NB and background views missing the beam. Estimated uncertainties in the measured velocity are system. Signals are acquired on 2 CCD detectors, each controlled by a dedicated PC. Spectra are fitted in real-time through a C++ processing code and velocities are made available to the Plasma Control System for future implementation of feedback on velocity. Results from the initial operation during the 2011 run are discussed, emphasizing the fast dynamics of toroidal rotation, e . g . during L-H mode transition and breaking caused by instabilities and by externally-imposed magnetic perturbations. Work supported by USDOE Contract No. DE-AC02-09CH11466.

  3. Bifurcation to Enhanced Performance H-mode on NSTX

    Science.gov (United States)

    Battaglia, D. J.; Chang, C. S.; Gerhardt, S. P.; Kaye, S. M.; Maingi, R.; Smith, D. R.

    2015-11-01

    The bifurcation from H-mode (H98 Performance (EP)H-mode (H98 = 1.2 - 2.0) on NSTX is found to occur when the ion thermal (χi) and momentum transport become decoupled from particle transport, such that the ion temperature (Ti) and rotation pedestals increase independent of the density pedestal. The onset of the EPH-mode transition is found to correlate with decreased pedestal collisionality (ν*ped) and an increased broadening of the density fluctuation (dn/n) spectrum in the pedestal as measured with beam emission spectroscopy. The spectrum broadening at decreased ν*ped is consistent with GEM simulations that indicate the toroidal mode number of the most unstable instability increases as ν*ped decreases. The lowest ν*ped, and thus largest spectrum broadening, is achieved with low pedestal density via lithium wall conditioning and when Zeff in the pedestal is significantly reduced via large edge rotation shear from external 3D fields or a large ELM. Kinetic neoclassical transport calculations (XGC0) confirm that Zeff is reduced when edge rotation braking leads to a more negative Er that shifts the impurity density profiles inward relative to the main ion density. These calculations also describe the role kinetic neoclassical and anomalous transport effects play in the decoupling of energy, momentum and particle transport at the bifurcation to EPH-mode. This work was sponsored by the U.S. Department of Energy.

  4. Alignment of the Thomson scattering diagnostic on NSTX

    International Nuclear Information System (INIS)

    LeBlanc, B P; Diallo, A

    2013-01-01

    The Thomson scattering diagnostic can provide profile measurement of the electron temperature, T e , and density, n e , in plasmas. Proper laser beam path and optics arrangement permits profiles T e (R) and n e (R) measurement along the major radius R. Keeping proper alignment between the laser beam path and the collection optics is necessary for an accurate determination of the electron density. As time progresses the relative position of the collection optics field of view with respect to the laser beam path will invariably shift. This can be kept to a minimum by proper attention to the physical arrangement of the collection and laser-beam delivery optics. A system has been in place to monitor the relative position between laser beam and collection optics. Variation of the alignment can be detected before it begins to affect the quality of the profile data. This paper discusses details of the instrumentation and techniques used to maintain alignment during NSTX multi-month experimental campaigns

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

    Energy Technology Data Exchange (ETDEWEB)

    Laugier, F

    2000-10-19

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

  6. Te(R,t) Measurements using Electron Bernstein Wave Thermal Emission on NSTX

    International Nuclear Information System (INIS)

    Diem, S.J.; Taylor, G.; Efthimion, P.C.; LeBlanc, B.P.; Carter, M.; Caughman, J.; Wilgen, J.B.; Harvey, R.W.; Preinhaelter, J.; Urban, J.

    2006-01-01

    The National Spherical Torus Experiment (NSTX) routinely studies overdense plasmas with n e of (1-5) x 10 19 m -3 and total magnetic field of e measurement. A significant upgrade to the previous NSTX EBW emission diagnostic to measure thermal EBW emission via the oblique B-X-O mode conversion process has been completed. The new EBW diagnostic consists of two remotely steerable, quad-ridged horn antennas, each of which is coupled to a dual channel radiometer. Fundamental (8-18 GHz) and second and third harmonic (18-40 GHz) thermal EBW emission and polarization measurements can be obtained simultaneously.

  7. Impact of the Wall Conditioning Program on Plasma Performance in NSTX

    International Nuclear Information System (INIS)

    H.W. Kuge; V. Soukhanovskii; M. Bell; , W. Blanchard; D. Gates; B. LeBlanc; R. Maingi; D. Mueller; H.K. Na; S. Paul; C.H. Skinner; D. Stutman; and W.R. Wampler

    2002-01-01

    High performance operating regimes have been achieved on NSTX (National Spherical Torus Experiment) through impurity control and wall-conditioning techniques. These techniques include HeGDC-aided boronization using deuterated trimethylboron, inter-discharge HeGDC, 350 C PFC bake-out followed by D2 and HeGDC, and experiments to test fueling discharges with either a He-trimethylboron mixture or pure trimethylboron. The impact of this impurity and density control program on recent advances in NSTX plasma performance is discussed

  8. Vessel Eddy Current Measurement for the National Spherical Torus Experiment (NSTX)

    International Nuclear Information System (INIS)

    Gates, D.A.; Menard, J.; Marsala, R.

    2004-01-01

    A simple analog circuit that measures the NSTX axisymmetric eddy current distribution has been designed and constructed. It is based on simple circuit model of the NSTX vacuum vessel that was calibrated using a special axisymmetric eddy current code which was written so that accuracy was maintained in the vicinity of the current filaments. The measurement and the model have been benchmarked against data from numerous vacuum shots and they are in excellent agreement. This is an important measurement that helps give more accurate equilibrium reconstructions

  9. Momentum Transport Studies in High E x B Shear Plasmas in NSTX

    International Nuclear Information System (INIS)

    Solomon, W.M.; Kaye, S.M.; Bell, S.M.; LeBlanc, B.P.; Menard, B.P.; Rewoldt, B.P.; Wang, W.; Levinton, F.M.; Yuh, H.; Sabbagh, S.A.

    2008-01-01

    Experiments have been conducted on NSTX to study both steady state and perturbative momentum transport. These studies are unique in their parameter space under investigation, where the low aspect ratio of NSTX results in rapid plasma rotation with E x B shearing rates high enough to suppress low-k turbulence. In some cases, the ratio of momentum to energy confinement time is found to exceed five. Momentum pinch velocities of order 10-40 m/s are inferred from the measured angular momentum flux evolution after non-resonant magnetic perturbations are applied to brake the plasma

  10. Using LGI experiments to achieve better understanding of pedestal-edge coupling in NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Zhehui [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-02-23

    PowerPoint presentation. Latest advances in granule or dust injection technologies, fast and high-resolution imaging, together with micro-/nano-structured material fabrication, provide new opportunities to examine plasma-material interaction (PMI) in magnetic fusion environment. Some of our previous work in these areas is summarized. The upcoming LGI experiments in NSTX-U will shed new light on granular matter transport in the pedestal-edge region. In addition to particle control, these results can also be used for code validation and achieving better understanding of pedestal-edge coupling in fusion plasmas in both NSTX-U and others.

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

    Science.gov (United States)

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

    2017-10-01

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

  12. Design analysis of the ITER divertor

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  13. Design analysis of the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

  14. On the refuelling of large divertor experiments

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  15. Nuclear graphite ageing and turnaround

    International Nuclear Information System (INIS)

    Marsden, B.J.; Hall, G.N.; Smart, J.

    2001-01-01

    Graphite moderated reactors are being operated in many countries including, the UK, Russia, Lithuania, Ukraine and Japan. Many of these reactors will operate well into the next century. New designs of High Temperature Graphite Moderated Reactors (HTRS) are being built in China and Japan. The design life of these graphite-moderated reactors is governed by the ageing of the graphite core due to fast neutron damage, and also, in the case of carbon dioxide cooled reactors by the rate of oxidation of the graphite. Nuclear graphites are polycrystalline in nature and it is the irradiation-induced damage to the individual graphite crystals that determines the material property changes with age. The life of a graphite component in a nuclear reactor can be related to the graphite irradiation induced dimensional changes. Graphites typically shrink with age, until a point is reached where the shrinkage stops and the graphite starts to swell. This change from shrinkage to swelling is known as ''turnaround''. It is well known that pre-oxidising graphite specimens caused ''turnaround'' to be delayed, thus extending the life of the graphite, and hence the life of the reactor. However, there was no satisfactory explanation of this behaviour. This paper presents a numerical crystal based model of dimensional change in graphite, which explains the delay in ''turnaround'' in the pre-oxidised specimens irradiated in a fast neutron flux, in terms of crystal accommodation and orientation and change in compliance due to radiolytic oxidation. (author)

  16. Differentiating the role of lithium and oxygen in retaining deuterium on lithiated graphite plasma-facing components

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, C. N. [Fusion Safety Program, Idaho National Laboratory, P.O. Box 1625-7113, Idaho Falls, Idaho 83415 (United States); School of Nuclear Engineering, Purdue University, 400 Central Drive, West Lafayette, Indiana 47907 (United States); Allain, J. P. [School of Nuclear Engineering, Purdue University, 400 Central Drive, West Lafayette, Indiana 47907 (United States); Department of Nuclear, Plasma and Radiological Engineering, University of Illinois at Urbana-Champaign, Illinois 61801 (United States); Luitjohan, K. E. [School of Nuclear Engineering, Purdue University, 400 Central Drive, West Lafayette, Indiana 47907 (United States); Krstic, P. S. [Institute for Advanced Computational Science, Stony Brook University, New York 11794 (United States); Department of Physics and Astronomy, University of Tennessee, Knoxville, Tennessee 37996 (United States); TheoretiK, Knoxville, Tennessee 379XX (United States); Dadras, J. [Department of Physics and Astronomy, University of Tennessee, Knoxville, Tennessee 37996 (United States); Department of Chemistry and Biochemistry, University of California Los Angeles, Los Angeles, California 90095 (United States); Skinner, C. H. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

    2014-05-15

    Laboratory experiments have been used to investigate the fundamental interactions responsible for deuterium retention in lithiated graphite. Oxygen was found to be present and play a key role in experiments that simulated NSTX lithium conditioning, where the atomic surface concentration can increase to >40% when deuterium retention chemistry is observed. Quantum-classical molecular dynamic simulations elucidated this oxygen-deuterium effect and showed that oxygen retains significantly more deuterium than lithium in a simulated matrix with 20% lithium, 20% oxygen, and 60% carbon. Simulations further show that deuterium retention is even higher when lithium is removed from the matrix. Experiments artificially increased the oxygen content in graphite to ∼16% and then bombarded with deuterium. X-ray photoelectron spectroscopy showed depletion of the oxygen and no enhanced deuterium retention, thus demonstrating that lithium is essential in retaining the oxygen that thereby retains deuterium.

  17. Recompressed exfoliated graphite articles

    Science.gov (United States)

    Zhamu, Aruna; Shi, Jinjun; Guo, Jiusheng; Jang, Bor Z

    2013-08-06

    This invention provides an electrically conductive, less anisotropic, recompressed exfoliated graphite article comprising a mixture of (a) expanded or exfoliated graphite flakes; and (b) particles of non-expandable graphite or carbon, wherein the non-expandable graphite or carbon particles are in the amount of between about 3% and about 70% by weight based on the total weight of the particles and the expanded graphite flakes combined; wherein the mixture is compressed to form the article having an apparent bulk density of from about 0.1 g/cm.sup.3 to about 2.0 g/cm.sup.3. The article exhibits a thickness-direction conductivity typically greater than 50 S/cm, more typically greater than 100 S/cm, and most typically greater than 200 S/cm. The article, when used in a thin foil or sheet form, can be a useful component in a sheet molding compound plate used as a fuel cell separator or flow field plate. The article may also be used as a current collector for a battery, supercapacitor, or any other electrochemical cell.

  18. Bromine intercalated graphite for lightweight composite conductors

    KAUST Repository

    Amassian, Aram; Patole, Archana

    2017-01-01

    A method of fabricating a bromine-graphite/metal composite includes intercalating bromine within layers of graphite via liquid-phase bromination to create brominated-graphite and consolidating the brominated-graphite with a metal nanopowder via a

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  20. Operation method for thermonuclear device and divertor for it

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  1. Physical study of experimental fusion breeder FEB divertor

    International Nuclear Information System (INIS)

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

    1999-10-01

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

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

    International Nuclear Information System (INIS)

    Sadakov, S.; Bondarchuk, E.; Doinikov, N.; Kitaev, B.; Kozhukhovskaya, N.; Maximiva, I.; Hirai, T.; Mertens, P.; Neubauer, O.; Obidenko, T.

    2006-01-01

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

  3. Cesium diffusion in graphite

    International Nuclear Information System (INIS)

    Evans, R.B. III; Davis, W. Jr.; Sutton, A.L. Jr.

    1980-05-01

    Experiments on diffusion of 137 Cs in five types of graphite were performed. The document provides a completion of the report that was started and includes a presentation of all of the diffusion data, previously unpublished. Except for data on mass transfer of 137 Cs in the Hawker-Siddeley graphite, analyses of experimental results were initiated but not completed. The mass transfer process of cesium in HS-1-1 graphite at 600 to 1000 0 C in a helium atmosphere is essentially pure diffusion wherein values of (E/epsilon) and ΔE of the equation D/epsilon = (D/epsilon) 0 exp [-ΔE/RT] are about 4 x 10 -2 cm 2 /s and 30 kcal/mole, respectively

  4. Irradiation Creep in Graphite

    Energy Technology Data Exchange (ETDEWEB)

    Ubic, Rick; Butt, Darryl; Windes, William

    2014-03-13

    An understanding of the underlying mechanisms of irradiation creep in graphite material is required to correctly interpret experimental data, explain micromechanical modeling results, and predict whole-core behavior. This project will focus on experimental microscopic data to demonstrate the mechanism of irradiation creep. High-resolution transmission electron microscopy should be able to image both the dislocations in graphite and the irradiation-induced interstitial clusters that pin those dislocations. The team will first prepare and characterize nanoscale samples of virgin nuclear graphite in a transmission electron microscope. Additional samples will be irradiated to varying degrees at the Advanced Test Reactor (ATR) facility and similarly characterized. Researchers will record microstructures and crystal defects and suggest a mechanism for irradiation creep based on the results. In addition, the purchase of a tensile holder for a transmission electron microscope will allow, for the first time, in situ observation of creep behavior on the microstructure and crystallographic defects.

  5. Suppression of turbulent transport in NSTX internal transport barriers

    Science.gov (United States)

    Yuh, Howard

    2008-11-01

    Electron transport will be important for ITER where fusion alphas and high-energy beam ions will primarily heat electrons. In the NSTX, internal transport barriers (ITBs) are observed in reversed (negative) shear discharges where diffusivities for electron and ion thermal channels and momentum are reduced. While neutral beam heating can produce ITBs in both electron and ion channels, High Harmonic Fast Wave (HHFW) heating can produce electron thermal ITBs under reversed magnetic shear conditions without momentum input. Interestingly, the location of the electron ITB does not necessarily match that of the ion ITB: the electron ITB correlates well with the minimum in the magnetic shear determined by Motional Stark Effect (MSE) [1] constrained equilibria, whereas the ion ITB better correlates with the maximum ExB shearing rate. Measured electron temperature gradients can exceed critical linear thresholds for ETG instability calculated by linear gyrokinetic codes in the ITB confinement region. The high-k microwave scattering diagnostic [2] shows reduced local density fluctuations at wavenumbers characteristic of electron turbulence for discharges with strongly negative magnetic shear versus weakly negative or positive magnetic shear. Fluctuation reductions are found to be spatially and temporally correlated with the local magnetic shear. These results are consistent with non-linear gyrokinetic simulations predictions showing the reduction of electron transport in negative magnetic shear conditions despite being linearly unstable [3]. Electron transport improvement via negative magnetic shear rather than ExB shear highlights the importance of current profile control in ITER and future devices. [1] F.M. Levinton, H. Yuh et al., PoP 14, 056119 [2] D.R. Smith, E. Mazzucato et al., RSI 75, 3840 [3] Jenko, F. and Dorland, W., PRL 89 225001

  6. High spatial sampling global mode structure measurements via multichannel reflectometry in NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Crocker, N A; Peebles, W A; Kubota, S; Zhang, J [Department of Physics and Astronomy, University of California-Los Angeles, Los Angeles, CA 90095-7099 (United States); Bell, R E; Fredrickson, E D; Gorelenkov, N N; LeBlanc, B P; Menard, J E; Podesta, M [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); Sabbagh, S A [Department of Applied Physics and Applied Mathematics, Columbia University, New York, NY 10027 (United States); Tritz, K [Johns Hopkins University, Baltimore, MD 21218 (United States); Yuh, H [Nova Photonics, Princeton, NJ 08540 (United States)

    2011-10-15

    Global modes-including kinks and tearing modes (f <{approx} 50 kHz), toroidicity-induced Alfven eigenmodes (TAE; f {approx} 50-250 kHz) and global and compressional Alfven eigenmodes (GAE and CAE; f >{approx} 400 kHz)-play critical roles in many aspects of plasma performance. Their investigation on NSTX is aided by an array of fixed-frequency quadrature reflectometers used to determine their radial density perturbation structure. The array has been recently upgraded to 16 channels spanning 30-75 GHz (n{sub cutoff} = (1.1-6.9) x 10{sup 19} m{sup -3} in O-mode), improving spatial sampling and access to the core of H-mode plasmas. The upgrade has yielded significant new results that advance the understanding of global modes in NSTX. The GAE and CAE structures have been measured for the first time in the core of an NSTX high-power (6 MW) beam-heated H-mode plasma. The CAE structure is strongly core-localized, which has important implications for electron thermal transport. The TAE structure has been measured with greatly improved spatial sampling, and measurements of the TAE phase, the first in NSTX, show strong radial variation near the midplane, indicating radial propagation caused by non-ideal MHD effects. Finally, the tearing mode structure measurements provide unambiguous evidence of coupling to an external kink.

  7. Predictions and observations of global beta-induced Alfven-acoustic modes in JET and NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Gorelenkov, N N [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States); Berk, H L [Institute for Fusion Studies, University of Texas, Austin, TX 78712 (United States); Crocker, N A [Institute of Plasma and Fusion Research, University of California, Los Angeles, CA 90095-1354 (United States); Fredrickson, E D [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States); Kaye, S [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States); Kubota, S [Institute of Plasma and Fusion Research, University of California, Los Angeles, CA 90095-1354 (United States); Park, H [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States); Peebles, W [Institute of Plasma and Fusion Research, University of California, Los Angeles, CA 90095-1354 (United States); Sabbagh, S A [Department of Applied Physics, Columbia University, New York, NY 10027-6902 (United States); Sharapov, S E [Euroatom/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Stutmat, D [Department of Physics and Astronomy, Johns Hopkins University, Baltimore, MD 21218 (United States); Tritz, K [Department of Physics and Astronomy, Johns Hopkins University, Baltimore, MD 21218 (United States); Levinton, F M [Nova Photonics, One Oak Place, Princeton, NJ 08540 (United States); Yuh, H [Nova Photonics, One Oak Place, Princeton, NJ 08540 (United States)

    2007-12-15

    In this paper we report on observations and interpretations of a new class of global MHD eigenmode solutions arising in gaps in the low frequency Alfven-acoustic continuum below the geodesic acoustic mode frequency. These modes have been just reported (Gorelenkov et al 2007 Phys. Lett. 370 70-7) where preliminary comparisons indicate qualitative agreement between theory and experiment. Here we show a more quantitative comparison emphasizing recent NSTX experiments on the observations of the global eigenmodes, referred to as beta-induced Alfven-acoustic eigenmodes (BAAEs), which exist near the extrema of the Alfven-acoustic continuum. In accordance to the linear dispersion relations, the frequency of these modes may shift as the safety factor, q, profile relaxes. We show that BAAEs can be responsible for observations in JET plasmas at relatively low beta <2% as well as in NSTX plasmas at relatively high beta >20%. In NSTX plasma observed magnetic activity has the same properties as predicted by theory for the mode structure and the frequency. Found numerically in NOVA simulations BAAEs are used to explain the observed properties of relatively low frequency experimental signals seen in NSTX and JET tokamaks.

  8. Intercomparison of graphite irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Hering, H; Perio, P; Seguin, M [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    While fast neutrons only are effective in damaging graphite, results of irradiations are more or less universally expressed in terms of thermal neutron fluxes. This paper attempts to correlate irradiations made in different reactors, i.e., in fluxes of different spectral compositions. Those attempts are based on comparison of 1) bulk length change and volume expansion, and 2) crystalline properties (e.g., lattice parameter C, magnetic susceptibility, stored energy, etc.). The methods used by various authors for determining the lattice constants of irradiated graphite are discussed. (author)

  9. Disruption characteristics in PDX with limiter and divertor discharges

    International Nuclear Information System (INIS)

    Couture, P.; McGuire, K.

    1986-09-01

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

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

    Directory of Open Access Journals (Sweden)

    T.D. Rognlien

    2017-08-01

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

  11. Graphite-based photovoltaic cells

    Science.gov (United States)

    Lagally, Max; Liu, Feng

    2010-12-28

    The present invention uses lithographically patterned graphite stacks as the basic building elements of an efficient and economical photovoltaic cell. The basic design of the graphite-based photovoltaic cells includes a plurality of spatially separated graphite stacks, each comprising a plurality of vertically stacked, semiconducting graphene sheets (carbon nanoribbons) bridging electrically conductive contacts.

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  13. Resonant island divertor experiments on text

    International Nuclear Information System (INIS)

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

    1988-09-01

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

  14. Evaluation of helium cooling for fusion divertors

    International Nuclear Information System (INIS)

    Baxi, C.B.

    1993-09-01

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

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

    International Nuclear Information System (INIS)

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

    1974-10-01

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

  16. Two-dimensional divertor modeling and scaling laws

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    1993-03-01

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

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

    International Nuclear Information System (INIS)

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

    1998-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

  20. Divertor IR thermography on Alcator C-Moda)

    Science.gov (United States)

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

    2010-10-01

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