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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

Dynamics of an Isolated Blob in the Presence of the X-Point

Description: The interplay of X-point shearing and axial plasma redistribution along a moving flux tube is discussed. Blobs limited to the main scrape-off-layer and the blobs entirely confined in the divertor region are identified. A strong effect of the radial tilt of the divertor plate on ''divertor'' blobs is found.
Date: October 10, 2005
Creator: Cohen, R H & Ryutov, D D
Partner: UNT Libraries Government Documents Department

Experimental divertor similarity database parameters

Description: A set of experimentally-determined dimensionless parameters is proposed for characterizing the regime of divertor operation. The objective is to be able to compare as unambiguously as possible the operation of different divertors and to understand what physical similarities and differences they represent. Examples from Alcator C- Mod are given. 4 refs., 5 figs,. 3 tabs.
Date: December 1, 1995
Creator: Hutchinson, I.H.; LaBombard, B. & Lipschultz, B.
Partner: UNT Libraries Government Documents Department

Response of NSTX Liquid Lithium divertor to High Heat Loads

Description: Samples of the NSTX Liquid Lithium Divertor (LLD) with and without an evaporative Li coating were directly exposed to a neutral beam ex-situ at a power of ~1.5 MW/m2 for 1-3 seconds. Measurements of front face and bulk sample temperature were obtained. Predictions of temperature evolution were derived from a 1D heat flux model. No macroscopic damage occurred when the "bare" sample was exposed to the beam but microscopic changes to the surface were observed. The Li-coated sample developed a lithium hydroxide (LiOH) coating, which did not change even when the front face temperature exceeded the pure Li melting point. These results are consistent with the lack of damage to the LLD surface and imply that heating alone may not expose pure liquid Li if the melting point of surface impurities is not exceeded. This suggests that flow and heat are needed for future PFCs requiring a liquid Li surface. __________________________________________________
Date: July 18, 2012
Creator: Abrams, Tyler; Kallman, J; Kaitaa, R; Foley, E L; Grayd, T K; Kugel, H et al.
Partner: UNT Libraries Government Documents Department

EDGE2D Simulations of JET{sup 13}C Migration Experiments

Description: Material migration has received renewed interest due to tritium retention associated with carbon transport to remote vessel locations. Those results influence the desirability of carbon usage on ITER. Subsequently, additional experiments have been performed, including tracer experiments attempting to identify material migration from specific locations. In this paper, EDGE2D models a well-diagnosed JET{sup 13}C tracer migration experiment. The role of SOL flows upon the migration patterns is identified.
Date: June 16, 2004
Creator: Strachan, J.D.; Coad, J.P.; Corrigan, G.; Matthews, G.F. & Spence, J.
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

Rapidly Moving Divertor Plates In A Tokamak

Description: 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.
Date: May 16, 2011
Creator: Zweben, S.
Partner: UNT Libraries Government Documents Department

Moving Divertor Plates in a Tokamak

Description: 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.
Date: February 12, 2009
Creator: S.J. Zweben, H. Zhang
Partner: UNT Libraries Government Documents Department

Strike Point Control for the National Spherical Torus Experiment (NSTX)

Description: This paper presents the first control algorithm for the inner and outer strike point position for a Spherical Torus (ST) fusion experiment and the performance analysis of the controller. A liquid lithium divertor (LLD) will be installed on NSTX which is believed to provide better pumping than lithium coatings on carbon PFCs. The shape of the plasma dictates the pumping rate of the lithium by channeling the plasma to LLD, where strike point location is the most important shape parameter. Simulations show that the density reduction depends on the proximity of strike point to LLD. Experiments were performed to study the dynamics of the strike point, design a new controller to change the location of the strike point to desired location and stabilize it. The most effective PF coils in changing inner and outer strike points were identified using equilibrium code. The PF coil inputs were changed in a step fashion between various set points and the step response of the strike point position was obtained. From the analysis of the step responses, PID controllers for the strike points were obtained and the controller was tuned experimentally for better performance. The strike controller was extended to include the outer-strike point on the inner plate to accommodate the desired low outer-strike points for the experiment with the aim of achieving "snowflake" divertor configuration in NSTX.
Date: July 9, 2010
Creator: Kolemen, E.; Gates, D. A.; Rowley, C. W.; Kasdin, N. J.; Kallman, J.; Gerhardt, S. et al.
Partner: UNT Libraries Government Documents Department

A Midsize Tokamak As Fast Track To Burning Plasmas

Description: This paper presents a midsize tokamak as a fast track to the investigation of burning plasmas. It is shown that it could reach large values of energy gain (≥10) with only a modest improvement in confinement over the scaling that was used for designing the International Thermonuclear Experimental Reactor (ITER). This could be achieved by operating in a low plasma recycling regime that experiments indicate can lead to improved plasma confinement. The possibility of reaching the necessary conditions of low recycling using a more efficient magnetic divertor than those of present tokamaks is discussed.
Date: July 14, 2010
Creator: Mazzucato, E.
Partner: UNT Libraries Government Documents Department

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

Description: 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
Date: August 9, 2004
Creator: Paul, S.F.; Maingi, R.; Soukhanovskii, V.; Kaye, S.M.; Kugel, H. & Team, the NSTX Research
Partner: UNT Libraries Government Documents Department

Diverted Tokamak Carbon Screening: Scaling with Machine Size and Consequences to Core Contamination

Description: Plasma impurity content depends upon the impurity sources, fueling efficiency, and confinement. In JET [Joint European Torus], carbon is the primary impurity, and its fueling efficiency has been studied using methane gas injection and modeled with the SOL [scrape-off layer] codes: DIVIMP and EDGE2D. In this paper, EDGE2D modeling of similar AUG [ASDEX-Upgrade] experiments and projections to ITER are described. The parameters have been identified which govern the size scaling of carbon screening. Size scaling is complex. For carbon injected from the main chamber, the important factors include: the SOL temperature, the magnitude of the thermal force at the divertor entrance, and the parallel distance to the divertor. For carbon injected at the strike points, the intersection of the carbon ionization region with the region of strong thermal force determines the carbon fueling efficiency ITER projects to have much better carbon screening than JET. The ITER SOL is hotter so that main chamber carbon is ionized further from the separatrix making the calculated carbon fueling efficiency lower. Also, the carbon originating near the strike point has less chance of escaping the divertor by factors of about 100. The carbon sputtering is projected to be larger by similar factors, making the projected ITER core contamination similar to JET. However, that result is based upon the assumption that the wall materials have similar composition and behavior as observed on JET. A general result is that the core contamination at fixed total sputtering rate and core impurity confinement increases when the fraction of carbon ionized in the main chamber SOL increases, and decreases for larger machine size and higher density operation.
Date: January 23, 2004
Creator: Strachan, J.D.; Corrigan, G.; Kallenbach, A.; Matthews, G.F.; Meister, H.; Neu, R. et al.
Partner: UNT Libraries Government Documents Department

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

Description: This report describes the work of the Divertor Task Force of the Massachusetts Institute of Technology Plasma Fusion Center, particularly the Task Force`s founding meeting, original research and development needs, organization, and achievements of its first year. The Task Force`s goal is to obtain an increasingly complete physics understanding of existing divertor plasmas, to build analytical and numerical models of the scrape-off-layer divertor plasmas, and to extrapolate them to find design solutions for the high power divertors of ignited tokamak plasmas such as those of ITER and other high performance future tokamaks. 67 refs., 2 figs.
Date: October 1, 1995
Partner: UNT Libraries Government Documents Department

Fast pressure measurements of the local island divertor on the compact helical system

Description: Development of an effective divertor is critical for the viability of the stellarator (helical system) concept. In the local island divertor (LID) concept particle and heat fluxes are channeled to the back of the LID head by the magnetic field structure of an externally produced m = 1, n = I island that is outside the last closed flux surface. The leading edge of the LID head is protected from the outward heat flux from the plasma because it is located inside the 1/1 island and the particles that strike the target plates on the back of the LID head in the throat of the LID pump module are then pumped efficiently. A set of 16 coils was used to create a 1/1 island in the Compact Helical System (CHS). The current (I{sub LID}) in the LID coils was chosen to position either the 0-point or the X-point of the external 1/1 magnetic island at the location of the LID head. The principal diagnostic in this study was an ASDEX-style ionization gauge that allowed fast ({approx}1-ms) measurements of the neutral pressure buildup behind the divertor head in the LID module.
Date: August 1, 1997
Creator: Lyon, J.F.; Klepper, C.C. & England, A.C.
Partner: UNT Libraries Government Documents Department

Comparison of impurities and time-dependent behavior for the ITER divertor

Description: This a the second part-of an ongoing project to model the divertor plasma for ITER. The UEDGE 2-D edge transport code is used to study the effect of impurities and tilted divertor plates to make a radiative divertor that can prevent excessive heat loads and adequately pump helium produced by fusion reactions in the core. The impurities are modeled using individual charge states with the local concentrations being determined by transport or as a fixed fraction of the hydrogenic ion density. For the multi-species model, helium, beryllium, carbon, and neon impurities are considered separately, together with the majority hydrogenic species, and a comparison is made of impurity spatial distribution and the power radiated at low impurity levels. At moderate to high impurity levels, typically only time-dependent solutions are found which are studied here for neon using both impurity models.
Date: February 25, 1997
Creator: Rensink, M.E.; Rognlien, T.D. & Hua, D.D.
Partner: UNT Libraries Government Documents Department

Divertor characterization experiments

Description: 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.
Date: June 18, 1996
Creator: Porter, G.D.; Allen, S.; Fenstermacher, M.; Hill, D.; Brown, M.; Jong, R.A, et al.
Partner: UNT Libraries Government Documents Department

Recent Progress in the NSTX/NSTX-U Lithium Program and Prospects for Reactor-Relevant Liquid-Lithium Based Divertor Development

Description: Developing a reactor compatible divertor has been identified as a particularly challenging technology problem for magnetic confinement fusion. While tungsten has been identified as the most attractive solid divertor material, the NSTX/NSTX-U lithium (Li) program is investigating the viability of liquid lithium (LL) as a potential reactor compatible divertor plasma facing component (PFC) . In the near term, operation in NSTX-U is projected to provide reactor-like divertor heat loads < 40 MW/m^2 for 5 s. During the most recent NSTX campaign, ~ 0.85 kg of Li was evaporated onto the NSTX PFCs where a ~50% reduction in heat load on the Liquid Lithium Divertor (LLD) was observed, attributable to enhanced divertor bolometric radiation. This reduced divertor heat flux through radiation observed in the NSTX LLD experiment is consistent with the results from other lithium experiments and calculations. These results motivate an LL-based closed radiative divertor concept proposed here for NSTX-U and fusion reactors. With an LL coating, the Li is evaporated from the divertor strike point surface due to the intense heat. The evaporated Li is readily ionized by the plasma due to its low ionization energies, and the ionized Li ions can radiate strongly, resulting in a significant reduction in the divertor heat flux. Due to the rapid plasma transport in divertor plasma, the radiation values can be significantly enhanced up to ~ 11 MJ/cc of LL. This radiative process has the desired function of spreading the focused divertor heat load to the entire divertor chamber facilitating the divertor heat removal. The LL divertor surface can also provide a "sacrificial" surface to protect the substrate solid material from transient high heat flux such as the ones caused by the ELMs. The closed radiative LLD concept has the advantages of providing some degree of partition in terms of plasma disruption ...
Date: October 27, 2012
Creator: M. Ono, et al.
Partner: UNT Libraries Government Documents Department

ORNL stellarator divertor studies

Description: Oak Ridge National Laboratory (ORNL) is studying various aspects of divertors in different stellarators. We are looking at a local island divertor (LID) on the CHS helical system, and potential designs of divertors that use islands for modular stellarators such as W7-AS and the modular helias-like heliac MHH chosen for the US Stellarator Power Plant Study. Although the helical CHS configuration is quite different from the modular W 7-AS configuration both rely on the island structure outside the last closed flux surface (LCFS) to aid in funneling escaping plasma flux and to protect the leading edges of the divertor plates. However, the way that the islands are used is quite different.
Date: September 1, 1995
Creator: Rome, J.A.; Lee, D.K.; England, A.C. & Lyon, J.F.
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

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

Description: 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.
Date: January 25, 2010
Creator: 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
Partner: UNT Libraries Government Documents Department

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

Description: While intermittent filamentary structures, also known as blobs, are routinely seen in the low-field-side scrape-off layer of the National Spherical Torus Experiment (NSTX) (Ono et al 2000 Nucl. Fusion 40 557), fine structured filaments are also seen on the lower divertor target plates of NSTX. These filaments, not associated with edge localized modes, correspond to the interaction of the turbulent blobs seen near the midplane with the divertor plasma facing components. The fluctuation level of the neutral lithium light observed at the divertor, and the skewness and kurtosis of its probability distribution function, is similar to that of midplane blobs seen in Dα; e.g. increasing with increasing radii outside the outer strike point (OSP) (separatrix). In addition, their toroidal and radial movement agrees with the typical movement of midplane blobs. Furthermore, with the appropriate magnetic topology, i.e. mapping between the portion of the target plates being observed into the field of view of the midplane gas puff imaging diagnostic, very good correlation is observed between the blobs and the divertor filaments. The correlation between divertor plate filaments and midplane blobs is lost close to the OSP. This latter observation is consistent with the existence of ‘magnetic shear disconnection’ due to the lower X-point, as proposed by Cohen and Ryutov (1997 Nucl. Fusion 37 621).
Date: May 19, 2010
Creator: Maqueda, R. J. & Stotler, D. P.
Partner: UNT Libraries Government Documents Department

Design of a fusion engineering test facility

Description: The fusion Engineering Test Facility (ETF) is being designed to provide for engineering testing capability in a program leading to the demonstration of fusion as a viable energy option. It will combine power-reactor-type components and subsystems into an integrated tokamak system and provide a test bed to test blanket modules in a fusion environment. Because of the uncertainties in impurity control two basic designs are being developed: a design with a bundle divertor (Design 1) and one with a poloidal divertor (Design 2). The two designs are similar where possible, the latter having somewhat larger toroidal field (TF) coils to accommodate removal of the larger torus sectors required for the single-null poloidal divertor. Both designs have a major radius of 5.4 m, a minor radius of 1.3 m, and a D-shaped plasma with an elongation of 1.6. Ten TF coils are incorporated in both designs, producing a toroidal field of 5.5 T on-axis. The ohmic heating and equilibrium field (EF) coils supply sufficient volt-seconds to produce a flat-top burn of 100 s and a duty cycle of 135 s, including a start of 12 s, a burn termination of 10 s, and a pumpdown of 13 s. The total fusion power during burn is 750 MW, giving a neutron wall loading of 1.5 MW/m/sup 2/. In Design 1 of the poloidal field (PF) coils except the fast-response EF coils are located outside the FT coils and are superconducting. The fast-response coils are located inside the TF coil bore near the torus and are normal conducting so that they can be easily replaced.In Design 2 all of the PF coils are located outside the TF coils and are superconducting. Ignition is achieved with 60 MW of neutral beam injection at 150 keV. Five megawatts of radio frequency heating (electron cyclotron resonance ...
Date: January 1, 1980
Creator: Sager, P.H.
Partner: UNT Libraries Government Documents Department