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Snowflake divertor configuration studies for NSTX-Upgrade

Description: 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.
Date: November 12, 2011
Creator: Soukhanovskii, V A
Partner: UNT Libraries Government Documents Department

Electron Density Measurements in the National Spherical Torus Experiment Detached Divertor Region Using Stark Broadening of Deuterium Infrared Paschen Emission Lines

Description: Spatially resolved measurements of deuterium Balmer and Paschen line emission have been performed in the divertor region of the National Spherical Torus Experiment using a commercial 0.5 m Czerny-Turner spectrometer. While the Balmer emission lines, Balmer and Paschen continua in the ultraviolet and visible regions have been extensively used for tokamak divertor plasma temperature and density measurements, the diagnostic potential of infrared Paschen lines has been largely overlooked. We analyze Stark broadening of the lines corresponding to 2-n and 3-m transitions with principle quantum numbers n = 7-12 and m = 10-12 using recent Model Microfield Method calculations (C. Stehle and R. Hutcheon, Astron. Astrophys. Supl. Ser. 140, 93 (1999)). Densities in the range (5-50) x 10{sup 19} m{sup -3} are obtained in the recombining inner divertor plasma in 2-6 MW NBI H-mode discharges. The measured Paschen line profiles show good sensitivity to Stark effects, and low sensitivity to instrumental and Doppler broadening. The lines are situated in the near-infrared wavelength domain, where optical signal extraction schemes for harsh nuclear environments are practically realizable, and where a recombining divertor plasma is optically thin. These properties make them an attractive recombining divertor density diagnostic for a burning plasma experiment.
Date: April 27, 2007
Creator: Soukhanovskii, V A; Johnson, D W; Kaita, R & Roquemore, A L
Partner: UNT Libraries Government Documents Department

The role of parallel heat transport in the relation between upstream scrape-off layer widths and target heat flux width in H-mode plasmas of NSTX.

Description: The physics of parallel heat transport was tested in the Scrape-off Layer (SOL) plasma of the National Spherical Torus Experiment (NSTX) [M. Ono, et al., Nucl. Fusion 40, 557 (2000) and S. M. Kaye, et al., Nucl. Fusion 45, S168 (2005)] tokamak by comparing the upstream electron temperature (T{sub e}) and density (n{sub e}) profiles measured by the mid-plane reciprocating probe to the heat flux (q{sub {perpendicular}}) profile at the divertor plate measured by an infrared (IR) camera. It is found that electron conduction explains the near SOL width data reasonably well while the far SOL, which is in the sheath limited regime, requires an ion heat flux profile broader than the electron one to be consistent with the experimental data. The measured plasma parameters indicate that the SOL energy transport should be in the conduction-limited regime for R-R{sub sep} (radial distance from the separatrix location) < 2-3 cm. The SOL energy transport should transition to the sheath-limited regime for R-R{sub sep} > 2-3cm. The T{sub e}, n{sub e}, and q{sub {perpendicular}} profiles are better described by an offset exponential function instead of a simple exponential. The conventional relation between mid plane electron temperature decay length ({lambda}{sub Te}) and target heat flux decay length ({lambda}{sub q}) is {lambda}{sub Te} = 7/2{lambda}{sub q}, whereas the newly-derived relation, assuming offset exponential functional forms, implies {lambda}{sub Te} = (2-2.5){lambda}{sub q}. The measured values of {lambda}{sub Te}/{lambda}{sub q} differ from the new prediction by 25-30%. The measured {lambda}{sub q} values in the far SOL (R-R{sub sep} > 2-3cm) are 9-10cm, while the expected values are 2.7 < {lambda}{sub q} < 4.9 cm (for sheath-limited regime). We propose that the ion heat flux profile is substantially broader than the electron heat flux profile as an explanation for this discrepancy in the far SOL.
Date: January 5, 2009
Creator: Ahn, J W; Boedo, J A; Maingi, R & Soukhanovskii, V A
Partner: UNT Libraries Government Documents Department

Solenoid-free Plasma Start-up in NSTX using Transient CHI

Description: 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.
Date: January 5, 2009
Creator: Raman, R; Jarboe, T; Nelson, B; Mueller, D & Soukhanovskii, V A
Partner: UNT Libraries Government Documents Department

NSTX Report on FES Joint Facilities Research Milestone 2010

Description: Annual Target: Conduct experiments on major fusion facilities to improve understanding of the heat transport in the tokamak scrape-off layer (SOL) plasma, strengthening the basis for projecting divertor conditions in ITER. The divertor heat flux profiles and plasma characteristics in the tokamak scrape-off layer will be measured in multiple devices to investigate the underlying thermal transport processes. The unique characteristics of C-Mod, DIII-D, and NSTX will enable collection of data over a broad range of SOL and divertor parameters (e.g., collisionality ν*, beta β, parallel heat flux q||, and divertor geometry). Coordinated experiments using common analysis methods will generate a data set that will be compared with theory and simulation.
Date: March 24, 2011
Creator: Maingi, R.; Ahn, J.-W.; Gray, T. K.; McLean, A. G. & Soukhanovskii, V. A.
Partner: UNT Libraries Government Documents Department

NSTX Tangential Divertor Camera

Description: 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.
Date: July 16, 2004
Creator: Roquemore, A.L.; Biewer, Ted; Johnson, D.; Zweben, S.J.; Nishino, Nobuhiro & Soukhanovskii, V.A.
Partner: UNT Libraries Government Documents Department

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

Description: 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.
Date: August 5, 2003
Creator: Soukhanovskii, V.A.; Maingi, R.; Raman, R.; Kugel, H.; LeBlanc, B.; Roquemore, A.L. et al.
Partner: UNT Libraries Government Documents Department

Core Fueling and Edge Particle Flux Analysis in Ohmically and Auxiliary Heated NSTX Plasmas

Description: The Boundary Physics program of the National Spherical Torus Experiment (NSTX) is focusing on optimization of the edge power and particle flows in b * 25% L- and H-mode plasmas of t {approx} 0.8 s duration heated by up to 6 MW of high harmonic fast wave and up to 5 MW of neutral beam injection. Particle balance and core fueling efficiencies of low and high field side gas fueling of L-mode homic and NBI heated plasmas have been compared using an analytical zero dimensional particle balance model and measured ion and neutral fluxes. Gas fueling efficiencies are in the range of 0.05-0.20 and do not depend on discharge magnetic configuration, density or poloidal location of the injector. The particle balance modeling indicates that the addition of HFS fueling results in a reversal of the wall loading rate and higher wall inventories. Initial particle source estimates obtained from neutral pressure and spectroscopic measurements indicate that ion flux into the divertor greatly exceeds midplane ion flux from the main plasma, suggesting that the scrape-off cross-field transport plays a minor role in diverted plasmas. Present analysis provides the basis for detailed fluid modeling of core and edge particle flows and particle confinement properties of NSTX plasmas. This research was supported by the U.S. Department of Energy under contracts No. DE-AC02-76CH03073, DE-AC05-00OR22725, and W-7405-ENG-36.
Date: June 12, 2002
Creator: Soukhanovskii, V.A.; Maingi, R.; Raman, R.; Kugel, H.W.; LeBlanc, B.P.; Roquemore, L. et al.
Partner: UNT Libraries Government Documents Department

Observation And Modeling Of Inner Divertor Re-attachment In Discharges With Lithium Coatings in NSTX

Description: In the National Spherical Torus Experiment (NSTX), modifications to the inner divertor plasma regimes are observed in high triangularity, H-mode, NBI heated discharges due to lithium coatings evaporated on the plasma facing components. In particular, the drop in the recombination rate, the reduced neutral pressure and the reduced electron density (inferred from Stark broadening measurements of high−n deuterium Balmer lines) suggested that the inner divertor, which is usually detached in discharges without lithium, re-attached. Experimental results are compared to simulations obtained with a 1D partially ionized plasma transport model integrated in the non-local thermodynamic equilibrium radiation transport code CRETIN to understand how the reduced recycling affects the divertor parameters in NSTX discharges with lithium coatings.
Date: January 4, 2011
Creator: Scotti, Filippo; Soukhanovskii, V. A.; Adams, M. L.; Scott, H. A.; Kugel, H. W.; Kaita, R. et al.
Partner: UNT Libraries Government Documents Department

Physics Design of the National High-power Advanced Torus Experiment

Description: Moving beyond ITER toward a demonstration power reactor (Demo) will require the integration of stable high fusion gain in steady-state, advanced methods for dissipating very high divertor heat-fluxes, and adherence to strict limits on in-vessel tritium retention. While ITER will clearly address the issue of high fusion gain, and new and planned long-pulse experiments (EAST, JT60-SA, KSTAR, SST-1) will collectively address stable steady-state highperformance operation, none of these devices will adequately address the integrated heat-flux, tritium retention, and plasma performance requirements needed for extrapolation to Demo. Expressing power exhaust requirements in terms of Pheat/R, future ARIES reactors are projected to operate with 60-200MW/m, a Component Test Facility (CTF) or Fusion Development Facility (FDF) for nuclear component testing (NCT) with 40-50MW/m, and ITER 20-25MW/m. However, new and planned long-pulse experiments are currently projected to operate at values of Pheat/R no more than 16MW/m. Furthermore, none of the existing or planned experiments are capable of operating with very high temperature first-wall (Twall = 600-1000C) which may be critical for understanding and ultimately minimizing tritium retention with a reactor-relevant metallic first-wall. The considerable gap between present and near-term experiments and the performance needed for NCT and Demo motivates the development of the concept for a new experiment — the National High-power advanced-Torus eXperiment (NHTX) — whose mission is to study the integration of a fusion-relevant plasma-material interface with stable steady-state high-performance plasma operation.
Date: July 18, 2007
Creator: Menard, J E; Fu, G -Y; Gorelenkov, N; Kaye, S M; Kramer, G; Maingi, R et al.
Partner: UNT Libraries Government Documents Department

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

Description: 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.
Date: April 5, 2011
Creator: Lyons, B. C.; Zweben, S. J.; Gray, T. K.; Hosea, J.; Kaita, R.; Kugel, H. W. et al.
Partner: UNT Libraries Government Documents Department