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1. Inferring the scrape-off layer heat flux width in a divertor with a low degree of axisymmetry

Author

Marsden, C., Zhang, X., Moscheni, M., Gray, T. K., Vekshina, E., Rengle, A., Scarabosio, A., Sertoli, M., Romanelli, M., and team, the ST40

Subjects
Physics - Plasma Physics
Abstract

Plasma facing components (PFCs) in the next generation of tokamak devices will operate in challenging environments, with heat loads predicted to exceed 10 MWm$^{-2}$. The magnitude of these heat loads is set by the width of the channel, the "scrape-off layer" (SOL), into which heat is exhausted, and can be characterised by an e-folding length scale for the decay of heat flux across the channel. It is expected this channel will narrow as tokamaks move towards reactor relevant conditions. Understanding the processes involved in setting the SOL heat flux width is imperative to be able to predict the heat loads PFCs must handle in future devices. Measurements of the SOL width are performed on the high-field spherical tokamak, ST40, using a newly commissioned infrared thermography system. With its high on-axis toroidal magnetic field ($\geq$1.5 T) ST40 is uniquely positioned to investigate the influence of toroidal field on the heat flux width in spherical tokamaks, whilst also extending measurements of the SOL width in spherical tokamaks to increased poloidal field ($\geq$0.3 T). Due to the divertor on ST40 having a low degree of axisymmetry, it is necessary for a set of radial measurements of the heat flux to be taken across the divertor, made possible using an automated toolchain that fully incorporates its 3D geometry. These radial profiles are combined with the magnetic topology of the plasma to infer the width of the SOL, with both single and double exponential profiles of heat flux observed. A reduction in the heat flux is observed toroidally across part of the divertor, with preliminary investigations indicating that partial shadowing occurs, resulting from the separation between magnetic field lines and trailing edges upstream of the observed region becoming comparable to the ion gyro-radius., Comment: 7 pages, 5 figures, Submitted to the Journal of Nuclear Materials and Energy Special Issue : Proceedings of 26th International Conference on Plasma Surface Interactions in Controlled Fusion Devices

Published
2024

2. Experimental observations of bifurcated power decay lengths in the near Scrape-Off Layer of tokamak plasmas

Author

Zhang, X, Marsden, C, Moscheni, M, Maartensson, E, Rengle, A, Robinson, M, OGorman, T, Lowe, H F, Vekshina, E, Janhunen, S, Buxton, P F, Sertoli, M, Romanelli, M, Scarabosio, A, Gray, T K, Lopez, N A, and Team, the ST40

Subjects
Physics - Plasma Physics
Abstract

The scrape-off layer parallel heat flux decay lengths measured at ST40, a high field, low aspect ratio spherical tokamak, have been observed to bifurcate into two groups. The wide group matches closely with the scale of ion poloidal Larmour radius and follows existing H-mode scalings, while the narrow group falls up to 10 times below scalings, on the scale of ion total Larmour radius. The onset of the narrow scrape-off layer width is observed to be associated with suppressed magnetic fluctuations, suggesting reduced electromagnetic turbulence levels in the SOL., Comment: 6 pages, 4 figures

Published
2024

3. Overview of the SPARC tokamak

Author

Massachusetts Institute of Technology. Plasma Science and Fusion Center, Massachusetts Institute of Technology. Department of Nuclear Science and Engineering, Creeely, A. J., Greenwald, Martin J., Ballinger, Sean B, Brunner, D., Canik, J., Doody, Jeffrey, Fülöp, T., Garnier, D.T., Granetz, Robert S., Gray, T. K., Holland, C., Howard, N.T., Hughes, J.W., Irby, James Henderson, Izzo, Viviana A., Kramer, G. J., Kuang, Adam QingYang, Labombard, Brian, Lin, Y., Lipschultz, B., Logan, N. C., Lore, J. D., Marmar, Earl S., Montes, Kevin J., Mumgaard, R. T., Paz-Soldan, C., Rea, C., Reinke, M. L., Rodriguez Fernandez, Pablo, Särkimäki, K., Sciortino, Francesco, Scott, S. D., Snicker, A., Snyder, P. B., Sorbom, B. N., Sweeney, R., Tinguely, Roy Alexander, Tolman, Elizabeth Ann, Umansky, M., Vallhagen, O., Varje, J., Whyte, Dennis G, Wright, J.C., Wukitch, Stephen James, The SPARC Team, Zhu, Jiazhou, Massachusetts Institute of Technology. Plasma Science and Fusion Center, Massachusetts Institute of Technology. Department of Nuclear Science and Engineering, Creeely, A. J., Greenwald, Martin J., Ballinger, Sean B, Brunner, D., Canik, J., Doody, Jeffrey, Fülöp, T., Garnier, D.T., Granetz, Robert S., Gray, T. K., Holland, C., Howard, N.T., Hughes, J.W., Irby, James Henderson, Izzo, Viviana A., Kramer, G. J., Kuang, Adam QingYang, Labombard, Brian, Lin, Y., Lipschultz, B., Logan, N. C., Lore, J. D., Marmar, Earl S., Montes, Kevin J., Mumgaard, R. T., Paz-Soldan, C., Rea, C., Reinke, M. L., Rodriguez Fernandez, Pablo, Särkimäki, K., Sciortino, Francesco, Scott, S. D., Snicker, A., Snyder, P. B., Sorbom, B. N., Sweeney, R., Tinguely, Roy Alexander, Tolman, Elizabeth Ann, Umansky, M., Vallhagen, O., Varje, J., Whyte, Dennis G, Wright, J.C., Wukitch, Stephen James, The SPARC Team, and Zhu, Jiazhou

Abstract

© 2020 The Author(s). The SPARC tokamak is a critical next step towards commercial fusion energy. SPARC is designed as a high-field (T), compact (m, m), superconducting, D-T tokamak with the goal of producing fusion gain 2$]]> from a magnetically confined fusion plasma for the first time. Currently under design, SPARC will continue the high-field path of the Alcator series of tokamaks, utilizing new magnets based on rare earth barium copper oxide high-temperature superconductors to achieve high performance in a compact device. The goal of 2$]]> is achievable with conservative physics assumptions () and, with the nominal assumption of, SPARC is projected to attain and MW. SPARC will therefore constitute a unique platform for burning plasma physics research with high density (), high temperature (keV) and high power density () relevant to fusion power plants. SPARC's place in the path to commercial fusion energy, its parameters and the current status of SPARC design work are presented. This work also describes the basis for global performance projections and summarizes some of the physics analysis that is presented in greater detail in the companion articles of this collection.

Published
2022

7. Overview of the SPARC tokamak

Author

Creely, A. J., Greenwald, M. J., Ballinger, S. B., Brunner, D., Canik, J., Doody, J., Fulop, T., Garnier, D. T., Granetz, R., Gray, T. K., Holland, C., Howard, N. T., Hughes, J. W., Irby, J. H., Izzo, V. A., Kramer, G. J., Kuang, A. Q., LaBombard, B., Lin, Yijun, Lipschultz, B., Logan, N. C., Lore, J. D., Marmar, E. S., Montes, K., Mumgaard, R. T., Paz-Soldan, C., Rea, C., Reinke, M. L., Rodriguez-Fernandez, P., Sarkimaki, K., Sciortino, F., Scott, S. D., Snicker, A., Snyder, P. B., Sorbom, B. N., Sweeney, R., Tinguely, R. A., Tolman, E. A., Umansky, M., Vallhagen, O., Varje, J., Whyte, D. G., Wright, J. C., Wukitch, S. J., Zhu, J., SPARC Team, Department of Applied Physics, Aalto-yliopisto, and Aalto University

Subjects
Tokamak, DATABASE, fusion plasma, Fusion plasma, Plasma confinement, 01 natural sciences, 010305 fluids & plasmas, law.invention, PHYSICS, PLASMA-FACING COMPONENTS, DESIGN, law, 0103 physical sciences, 010306 general physics, PROGRESS, DEMO, Physics, plasma confinement, TUNGSTEN, CONSTRUCTION, Fusion power, Condensed Matter Physics, H-MODE CONFINEMENT, plasma devices, Systems engineering, CHAPTER 2
Abstract

The SPARC tokamak is a critical next step towards commercial fusion energy. SPARC is designed as a high-field ($B_0 = 12.2$T), compact ($R_0 = 1.85$m,$a = 0.57$m), superconducting, D-T tokamak with the goal of producing fusion gain$Q>2$from a magnetically confined fusion plasma for the first time. Currently under design, SPARC will continue the high-field path of the Alcator series of tokamaks, utilizing new magnets based on rare earth barium copper oxide high-temperature superconductors to achieve high performance in a compact device. The goal of$Q>2$is achievable with conservative physics assumptions ($H_{98,y2} = 0.7$) and, with the nominal assumption of$H_{98,y2} = 1$, SPARC is projected to attain$Q \approx 11$and$P_{\textrm {fusion}} \approx 140$MW. SPARC will therefore constitute a unique platform for burning plasma physics research with high density ($\langle n_{e} \rangle \approx 3 \times 10^{20}\ \textrm {m}^{-3}$), high temperature ($\langle T_e \rangle \approx 7$keV) and high power density ($P_{\textrm {fusion}}/V_{\textrm {plasma}} \approx 7\ \textrm {MW}\,\textrm {m}^{-3}$) relevant to fusion power plants. SPARC's place in the path to commercial fusion energy, its parameters and the current status of SPARC design work are presented. This work also describes the basis for global performance projections and summarizes some of the physics analysis that is presented in greater detail in the companion articles of this collection.

Published
2020

17. High Heat Flux Testing of Castellated Graphite Plasma- Facing Components

Author

Gray, T. K., Youchison, D. L., Ellis, R. E., Jaworski, M. A., Khodak, A., Looby, T., Reinke, M. L., Smalley, G., and Wolfe, D. E.

Abstract

AbstractAs part of the recovery project of the National Spherical Tokamak Experiment–Upgrade (NSTX-U), the divertor plasma-facing components (PFCs) were redesigned to handle significantly higher heat fluxes and longer pulse lengths than NSTX. The design process resulted in a castellated, graphite PFC tile. To verify the thermal performance of this design, dedicated electron beam, high heat flux (HHF) testing was carried out on a de-optimized mock-up PFC target. These tests demonstrated that the tile design is itself robust to large, localized thermal gradients. No mechanical damage to the mock-up was observed during HHF testing, though the actual PFC tile mechanical tie-down was not tested. Rather, when the surface temperature exceeded the sublimation temperature of graphite, carbon blooms from the mock-up tile surface were observed. This resulted in 1 to 2 mm of surface material ablating from the mock-up after repeated, highly localized electron beam exposures.

Published
2021
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19. Design and characterization of a prototype divertor viewing infrared video bolometer for NSTX-U

Author

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

Abstract

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

Published
2016

23. High performance discharges in the Lithium Tokamak eXperiment with liquid lithium wallsa)

Author

Schmitt, J. C., primary, Bell, R. E., additional, Boyle, D. P., additional, Esposti, B., additional, Kaita, R., additional, Kozub, T., additional, LeBlanc, B. P., additional, Lucia, M., additional, Maingi, R., additional, Majeski, R., additional, Merino, E., additional, Punjabi-Vinoth, S., additional, Tchilingurian, G., additional, Capece, A., additional, Koel, B., additional, Roszell, J., additional, Biewer, T. M., additional, Gray, T. K., additional, Kubota, S., additional, Beiersdorfer, P., additional, Widmann, K., additional, and Tritz, K., additional

Published
2015
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24. 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

Author

Perkins, R. J., primary, Hosea, J. C., additional, Jaworski, M. A., additional, Ahn, J.-W., additional, Diallo, A., additional, Bell, R. E., additional, Bertelli, N., additional, Gerhardt, S., additional, Gray, T. K., additional, Kramer, G. J., additional, LeBlanc, B. P., additional, McLean, A., additional, Phillips, C. K., additional, Podestà, M., additional, Roquemore, L., additional, Sabbagh, S., additional, Taylor, G., additional, and Wilson, J. R., additional

Published
2015
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27. Towards identifying the mechanisms underlying field-aligned edge-loss of HHFW power on NSTX

Author

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

Published
2014
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28. Characterization of divertor footprints and the pedestal plasmas in the presence of appliedn= 3 fields for the attached and detached conditions in NSTX

Author

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

Published
2013
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30. Particle control and plasma performance in the Lithium Tokamak eXperiment

Author

Majeski, R., primary, Abrams, T., additional, Boyle, D., additional, Granstedt, E., additional, Hare, J., additional, Jacobson, C. M., additional, Kaita, R., additional, Kozub, T., additional, LeBlanc, B., additional, Lundberg, D. P., additional, Lucia, M., additional, Merino, E., additional, Schmitt, J., additional, Stotler, D., additional, Biewer, T. M., additional, Canik, J. M., additional, Gray, T. K., additional, Maingi, R., additional, McLean, A. G., additional, Kubota, S., additional, Peebles, W. A., additional, Beiersdorfer, P., additional, Clementson, J. H. T., additional, and Tritz, K., additional

Published
2013
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32. SPIRAL field mapping on NSTX for comparison to divertor RF heat deposition.

Author

Hosea, J. C., Perkins, R., Jaworski, M. A., Kramer, G. J., Ahn, J.-W., Bertelli, N., Gerhardt, S., Gray, T. K., LeBlanc, B. P., Maingi, R., Phillips, C. K., Roquemore, L., Ryan, P. M., Sabbagh, S., Taylor, G., Tritz, K., and Wilson, J. R.

Subjects
PHASE transitions, FUSION (Phase transformation), MAGNETIC fields, FUSION reactor divertors, FUSION reactor materials, FUSION reactor limiters, LANGMUIR probes, STATIC equilibrium (Physics)
Abstract

Field-aligned losses of HHFW power in the SOL of NSTX have been studied with IR cameras and probes, but the interpretation of the data depends somewhat on the magnetic equilibrium reconstruction. Both EFIT02 and LRDFIT04 magnetic equilibria have been used with the SPIRAL code to provide field mappings in the scrape off layer (SOL) on NSTX from the midplane SOL in front of the HHFW antenna to the divertor regions, where the heat deposition spirals are measured. The field-line mapping spiral produced at the divertor plate with LRDFIT04 matches the HHFWproduced heat deposition best, in general. An independent method for comparing the field-line strike patterns on the outer divertor for the two equilibria is provided by measuring Langmuir probe characteristics in the vicinity of the outer vessel strike radius (OVSR) and observing the effect on floating potential, saturation current, and zero-probe-voltage current (IV=0) with the crossing of the OVSR over the probe. Interestingly, these comparisons also reveal that LRDFIT04 gives the more accurate location of the predicted OVSR, and confirm that the RF power flow in the SOL is essentially along the magnetic field lines. Also, the probe characteristics and IV=0 data indicate that current flows under the OVSR in the divertor tiles in most cases studied. [ABSTRACT FROM AUTHOR]

Published
2014
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33. Towards Identifying The Mechanisms Underlying Field-Aligned Edge-Loss Of HHFW Power On NSTX.

Author

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

Subjects
PLASMA gases, RADIOFREQUENCY heating, CURRENT-drive heating, FUSION reactor divertors, MAGNETIC fields, LOW temperature plasmas, PLASMA boundary layers
Abstract

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 their importance in ITER. [ABSTRACT FROM AUTHOR]

Published
2014
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35. High-Harmonic Fast-Wave Power Flow along Magnetic Field Lines in the Scrape-Off Layer of NSTX

Author

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

Published
2012
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36. Analysis of a multi-machine database on divertor heat fluxes

Author

Makowski, M. A., primary, Elder, D., additional, Gray, T. K., additional, LaBombard, B., additional, Lasnier, C. J., additional, Leonard, A. W., additional, Maingi, R., additional, Osborne, T. H., additional, Stangeby, P. C., additional, Terry, J. L., additional, and Watkins, J., additional

Published
2012
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38. Continuous Improvement of H-Mode Discharge Performance with Progressively Increasing Lithium Coatings in the National Spherical Torus Experiment

Author

Maingi, R., primary, Kaye, S. M., additional, Skinner, C. H., additional, Boyle, D. P., additional, Canik, J. M., additional, Bell, M. G., additional, Bell, R. E., additional, Gray, T. K., additional, Jaworski, M. A., additional, Kaita, R., additional, Kugel, H. W., additional, LeBlanc, B. P., additional, Mansfield, D. K., additional, Osborne, T. H., additional, Sabbagh, S. A., additional, and Soukhanovskii, V. A., additional

Published
2011
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44. The Effect of ELMs on HHFW Heating of NBI Generated H-modes.

Author

Hosea, J. C., Ahn, J.-W., Bell, R. E., Gerhardt, S., Gray, T. K., Green, D. L., Jaeger, E. F., LeBlanc, B. P., Maingi, R., McLean, A., Phillips, C. K., Roquemore, L., Ryan, P. M., Sabbagh, S., Taylor, G., Tritz, K., Wilgen, J., and Wilson, J. R.

Subjects
ELECTRIC heating, ELECTRICAL harmonics, ENERGY storage, RADIO wave propagation, ANTENNAS (Electronics), MAGNETIC fields, INFRARED radiation
Abstract

ELMs reduce the stored energy achieved with HHFW heating compared with the ELM free case as also occurs for NBI heating alone. This reduction can be attributed both to direct ELM ejection of stored energy and to an increase in edge density with ELMs that exceeds the onset density for perpendicular wave propagation near the antenna [1,2], and leads to significantly more edge RF power deposition. This latter effect causes a more intense RF 'hot' zone in the lower divertor scrape off region due to an increase in edge RF power propagating to the divertor from the antenna region along the magnetic field lines. Fast IR measurements of the direct ELM heat deposition at the lower divertor shows it to be peaked in the vicinity of the outer strike radius and to fall off strongly as the 'hot' zone is approached, indicating little direct ELM effect on the 'hot' zone heating. Physics studies of the 'hot' zone have begun with sweeping the RF 'hot' zone spiral over Rogowski instrumented divertor region tiles to show that the tile currents track the location of the zone. The relation of the results here to the AORSA RF code simulations including the SOL region is pointed out. [ABSTRACT FROM AUTHOR]

Published
2011
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45. Characterization of divertor footprints and the pedestal plasmas in the presence of applied n = 3 fields for the attached and detached conditions in NSTX.

Author

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

Subjects
DIVERTERS (Electronics), PLASMA devices, ELECTROMAGNETIC fields, AXIAL flow, QUANTUM perturbations, ELECTRON temperature measurement
Abstract

Recent progress in the study of 3D field effects on the divertor and pedestal plasmas is reported with the use of a new set of diagnostics. A wide angle visible camera provides 2D data of lower divertor surface covering almost the full range of radius (r) and toroidal angle (Φ), a significant advantage over the conventional 1D radial profile in examining non-axisymmetric effects of 3D fields on the divertor footprints. The spatial distribution of connection lengths (Lc) calculated by vacuum field line tracing in the presence of 3D fields (n = 3) agrees with the footprint pattern observed in the 2D wide angle camera images. The full (r, Φ) image data with high temporal resolution revealed that the spatial structure of modified divertor footprints is maintained even during the edge-localized modes (ELMs) triggered by applied n = 3 fields, when the ELM size is sufficiently small, i.e. the ELMs are ‘phase locked’ to the imposed perturbation field structure. This phase-lock is lost during the ELM rise time for ELMs with large energy loss, e.g. ΔWELM/WMHD > 4–5%. Divertor gas puff was used to create detached divertor condition and the effect of 3D fields on the detachment was investigated. The divertor remains partially detached with the 3D field application when a sufficient amount of gas is injected into the divertor region, which is accompanied by a noticeable drop of pedestal electron temperature (Te). However, with a lower gas puff, the divertor plasma re-attaches, when 3D fields were applied to the detached plasma, and the pedestal Te rises back up. There observed no other change in the pedestal profile associated with the re-attachment, indicating that this is likely to be dominated by a change in the electron thermal transport processes. A TRANSP analysis shows that the drop of pedestal electron heat diffusivity (χe) is responsible for this change but the source of this reduction is yet unclear. [ABSTRACT FROM AUTHOR]

Published
2014
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46. Liquid lithium divertor characteristics and plasma-material interactions in NSTX high-performance plasmas.

Author

Jaworski, M. A., Abrams, T., Allain, J. P., Bell, M. G., Bell, R. E., Diallo, A., Gray, T. K., Gerhardt, S. P., Kaita, R., Kugel, H. W., LeBlanc, B. P., Maingi, R., McLean, A. G., Menard, J., Nygren, R., Ono, M., Podesta, M., Roquemore, A. L., Sabbagh, S. A., and Scotti, F.

Subjects
LITHIUM, PLASMA physics, FUSION reactors, TOKAMAKS, MOLYBDENUM, PROBLEM solving
Abstract

Liquid metal plasma-facing components (PFCs) have been proposed as a means of solving several problems facing the creation of economically viable fusion power reactors. To date, few demonstrations exist of this approach in a diverted tokamak and we here provide an overview of such work on the National Spherical Torus Experiment (NSTX). The Liquid Lithium Divertor (LLD) was installed and operated for the 2010 run campaign using evaporated coatings as the filling method. The LLD consisted of a copper-backed structure with a porous molybdenum front face. Nominal Li filling levels by the end of the run campaign exceeded the porosity void fraction by 150%. Despite a nominal liquid level exceeding the capillary structure and peak current densities into the PFCs exceeding 100 kA m-2, no macroscopic ejection events were observed. In addition, no substrate line emission was observed after achieving lithium-melt temperatures indicating the lithium wicks and provides a protective coating on the molybdenum porous layer. Impurity emission from the divertor suggests that the plasma is interacting with oxygen-contaminated lithium whether diverted on the LLD or not. A database of LLD discharges is analysed to consider whether there is a net effect on the discharges over the range of total deposited lithium in the machine. Examination of H-97L indicates that performance was constant throughout the run, consistent with the hypothesis that it is the quality of the surface layers of the lithium that impact performance. The accumulation of impurities suggests a fully flowing liquid lithium system to obtain a steady-state PFC on timescales relevant to NSTX. [ABSTRACT FROM AUTHOR]

Published
2013
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47. Fast-wave power flow along SOL field lines in NSTX and the associated power deposition profile across the SOL in front of the antenna.

Author

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

Subjects
PLASMA boundary layers, MAGNETIC fields, THEORY of wave motion, PREDICTION models, FUSION reactor divertors, STANDING waves
Abstract

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. [ABSTRACT FROM AUTHOR]

Published
2013
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49. The effect of progressively increasing lithium coatings on plasma discharge characteristics, transport, edge profiles and ELM stability in the National Spherical Torus Experiment.

Author

Maingi, R., Boyle, D. P., Canik, J. M., Kaye, S. M., Skinner, C. H., Allain, J. P., Bell, M. G., Bell, R. E., Gerhardt, S. P., Gray, T. K., Jaworski, M. A., Kaita, R., Kugel, H. W., LeBlanc, B. P., Manickam, J., Mansfield, D. K., Menard, J. E., Osborne, T. H., Raman, R., and Roquemore, A. L.

Subjects
PLASMA gases, LITHIUM compounds, SURFACE coatings, WASTE recycling, IONIZED gases, SURFACES (Technology)
Abstract

Lithium wall coatings have been shown to reduce recycling, suppress edge-localized modes (ELMs), and improve energy confinement in the National Spherical Torus Experiment (NSTX). Here we document the effect of gradually increasing lithium wall coatings on the discharge characteristics, with the reference ELMy discharges obtained in boronized, i.e. non-lithiated conditions. We observed a continuous but not quite monotonic reduction in recycling and improvement in energy confinement, a gradual alteration of edge plasma profiles, and slowly increasing periods of ELM quiescence. The measured edge plasma profiles during the lithium-coating scan were simulated with the SOLPS code, which quantified the reduction in divertor recycling coefficient from∼98% to∼90%. The reduction in recycling and fuelling, coupled with a drop in the edge particle transport rate, reduced the average edge density profile gradient, and shifted it radially inwards from the separatrix location. In contrast, the edge electron temperature (Te) profile was unaffected in the H-mode pedestal steep gradient region within the last 5% of normalized poloidal flux, ψN; however, the Te gradient became steeper at the top of the H-mode pedestal for 0.8 < ψN < 0.94 with lithium coatings. The peak pressure gradients were comparable during ELMy and ELM-free phases, but were shifted away from the separatrix in the ELM-free discharges, which is stabilizing to the current-driven instabilities thought to be responsible for ELMs in NSTX. [ABSTRACT FROM AUTHOR]

Published
2012
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