76 Matching Results

Search Results

Advanced search parameters have been applied.

Initial-temperature profiles on the PDX inner toroidal limiter

Description: The temperature profiles resulting from plasma operation on the PDX vertical, large area, inner toroidal limiter have been measured during both ohmic and neutral-beam-heated discharges using a scanning infrared camera. An asymmetric double-peaked temperature profile is seen after neutral-beam-heated discharges. Disruptions in ohmically heated discharges are found to be preceded by a single-peaked deposition and succeeded by an initially symmetric double-peaked deposition. The results were compared with the Schmidt model for scrape-off at a toroidal limiter and it was found that the measured double-peaked temperature profiles yielded scrape-off lengths consistent with previous measurements.
Date: September 1, 1983
Creator: Ulrickson, M. & Kugel, H.W.
Partner: UNT Libraries Government Documents Department

Design of the PDX Tokamak wall armor and inner limiter system

Description: The inner wall protective plates for the PDX Tokamak are designed to absorb 8 MW of neutral deuterium beam power at maximum power densities of 3 kW/cm/sup 2/ for pulse lengths of 0.5 sec. Preliminary studies indicate that the design could survive several pulses of 1 sec duration. The design consists of a tile and mounting plate structure. The mounting plates are water cooled to allow short duty cycles and beam calorimetry. The temperature and flow of the coolant is measured to obtain the injected power. A thermocouple array on the tiles provides beam position and power density profiles. Several material combinations for the tiles were subjected to thermal tests using both electron and neutral beams, and titanium carbide coated graphite was selected as the tile material. The heat transfer coefficient of the tile backing plate structure was measured to determine the maximum pulse rate allowable. The design of the armor system allows the structure to be used as a neutral beam power diagnostic and as an inner plasma limiter. The electrical and cooling systems external to the vacuum vessel are discussed.
Date: August 1, 1981
Creator: Kugel, H.W. & Ulrickson, M.
Partner: UNT Libraries Government Documents Department

Method and apparatus for measuring the momentum, energy, power, and power density profile of intense particle beams

Description: A method and apparatus for determining the power, momentum, energy, and power density profile for high momentum mass flow. Small probe projectiles of appropriate size, shape and composition are propelled through an intense particle beam at equal intervals along an axis perpendicular to the beam direction. Probe projectiles are deflected by collisions with beam particles. The net beam-induced deflection of each projectile is measured after it passes through the intense particle beam into an array of suitable detectors.
Date: December 31, 1991
Creator: Gammel, G.M. & Kugel, H.W.
Partner: UNT Libraries Government Documents Department

Neutral beam species measurements using in situ Rutherford backscatter spectrometry

Description: This work describes a new in situ method for measuring the neutral particle fractions in high power deuterium neutral beams, used to heat magnetically confined fusion plasmas. Deuterium beams, of variable energies, pulse lengths, and powers up to 47 keV, 100 msec, 1.6 MW, were Rutherford backscattered at 135/sup 0/ from TiC inner neutral beam armor of the PDX, and detected using an electrostatic analyzer with microchannel plates. Complete energy scans were made every 20 msec and data were obtained simultaneously from five different positions across the beam profile. The neutral particle fractions were measured to be D/sup 0/(E):D/sup 0/(E/2):D/sup 0/(E/3)=53:32:15. The corresponding neutral power fractions were P/sup 0/(E):P/sup 0/(E/2):P/sup 0/(E/3)=72:21:7, and the associated ionic fractions at the output of the ion source were D/sub 1//sup +/(E):D/sub 2//sup +/(E):D/sub 3//sup +/(E)=74:20:6. The measured neutral particle fractions were relatively constant over more than 70% of the beam power distribution. A decrease in the yield of the full energy component in the outer regions of the beam was observed. Other possible experimental configurations and geometries are discussed.
Date: December 1, 1984
Creator: Kugel, H.W.; Kaita, R.; Gammel, G. & Williams, M.D.
Partner: UNT Libraries Government Documents Department

Technique for measuring cooling patterns in ion source grids by infrared scanning

Description: Many plasma sources designed for neutral beam injection heating of plasmas now employ copper beam acceleration grids which are water-cooled by small capillary tubes fed from one or more headers. To prevent thermally-induced warpage of these grids it is essential that one be able to detect inhomogeneities in the cooling. Due to the very strong thermal coupling between adjacent cooling lines and the concomitant rapid equilibration times, it is not practical to make such measurements in a direct manner with a contact thermometer. We have developed a technique whereby we send a burst of hot water through an initially cool grid, followed by a burst of cool water, and record the transient thermal behavior usng an infrared television camera. This technique, which would be useful for any system with cooling paths that are strongly coupled thermally, has been applied to a number of sources built for the PLT and PDX tokamaks, and has proven highly effective in locating cooling deficiencies and blocked capillary tubes.
Date: February 1, 1980
Creator: Grisham, L.R.; Eubank, H.P. & Kugel, H.W.
Partner: UNT Libraries Government Documents Department

Diagnostic method for measuring plasma-induced voltages on the PBX-M (Princeton Beta Experiment-Modified) stabilizing shell

Description: The Princeton Beta Experiment-Modified (PBX-M) has a close-fitting conducting, passive plate, stabilizing shell which nearly surrounds highly indented, bean-shaped plasmas. The proximity of this electrically isolated shell to a large fraction of the plasma surface allows measurements similar to previous work on other tokamaks using floating probes and limiters. Measurements were performed to characterize the plasma-induced voltages on the PBX-M passive plate stabilizing shell during high-{beta} plasmas. Voltage differences were measured between the respective passive plate toroidal and poloidal gaps, the respective passive plates and the vessel, and an outer poloidal graphite limiter and its passive plate. The calibration and qualification testing procedures are discussed. The initial measurements found that the largest voltages were observed at plasma start-up and at the plasma current disruption and exhibited characteristics depending on operating conditions. The highest voltages observed have been at disruption and were less than 2 kV. 9 refs., 5 figs.
Date: July 1, 1990
Creator: Kugel, H.W.; Okabayashi, M. & Schweitzer, S.
Partner: UNT Libraries Government Documents Department

Infrared Camera Diagnostic for Heat Flux Measurements on NSTX

Description: 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.
Date: March 25, 2003
Creator: Mastrovito, D.; Maingi, R.; Kugel, H. W. & Roquemore, A. L.
Partner: UNT Libraries Government Documents Department

NCSX Plasma Heating Methods

Description: The NCSX (National Compact Stellarator Experiment) has been designed to accommodate a variety of heating systems, including ohmic heating, neutral-beam injection, and radio-frequency. Neutral beams will provide one of the primary heating methods for NCSX. In addition to plasma heating, beams are also expected to provide a means for external control over the level of toroidal plasma rotation velocity and its profile. The plan is to provide 3 MW of 50 keV balanced neutral-beam tangential injection with pulse lengths of 500 msec for initial experiments, and to be upgradeable to pulse lengths of 1.5 sec. Subsequent upgrades will add 3 MW of neutral-beam injection. This Chapter discusses the NCSX neutral-beam injection requirements and design issues, and shows how these are provided by the candidate PBX-M (Princeton Beta Experiment-Modification) neutral-beam injection system. In addition, estimations are given for beam-heating efficiencies, scaling of heating efficiency with machine size an d magnetic field level, parameter studies of the optimum beam-injection tangency radius and toroidal injection location, and loss patterns of beam ions on the vacuum chamber wall to assist placement of wall armor and for minimizing the generation of impurities by the energetic beam ions. Finally, subsequent upgrades could add an additional 6 MW of radio-frequency heating by mode-conversion ion-Bernstein wave (MCIBW) heating, and if desired as possible future upgrades, the design also will accommodate high-harmonic fast-wave and electron-cyclotron heating. The initial MCIBW heating technique and the design of the radio-frequency system lend themselves to current drive, so that if current drive became desirable for any reason only minor modifications to the heating system described here would be needed. The radio-frequency system will also be capable of localized ion heating (bulk or tail), and possibly ion-Bernstein-wave-generated sheared flows.
Date: February 28, 2003
Creator: Kugel, H.W.; Spong, D.; Majeski, R. & Zarnstorff, M.
Partner: UNT Libraries Government Documents Department

Neutral Beam Injection Requirements and Design Issues for the National Compact Stellarator Experiment

Description: The National Compact Stellarator Experiment (NCSX) will require 6 MW of 50 keV neutral beam injection (NBI) with initial pulse lengths of 500 msec and upgradeable to pulse lengths of 1.5 sec. This paper discusses the NCSX NBI requirements and design issues, and shows how these are provided by the candidate PBX-M [Princeton Beta Experiment-Modification] NBI system.
Date: February 11, 2002
Creator: Kugel, H. W.; Neilson, H.; Reiersen, W. & Zarnstorff, M.
Partner: UNT Libraries Government Documents Department

Performance of the PDX neutral beam wall armor

Description: The PDX wall armor was designed to function as an inner wall thermal armor, a neutral beam diagnostic, and a large area inner toroidal plasma limiter. In this paper we discuss its thermal performance as wall armor during two years of PDX neutral beam heating experiments. During this period it provided sufficient inner wall protection to permit perpendicular heating injections into normal and disruptive plasmas as well as injections in the absence of plasma involving special experiments, calibrations, and tests important for the optimization and development of the PDX neutral beam injection system. Many of the design constraints and performance issues encountered in this work are relevant to the design of larger fusion devices.
Date: February 1, 1985
Creator: Kugel, H.W.; Eubank, H.P.; Kozub, T.A. & Williams, M.D.
Partner: UNT Libraries Government Documents Department

Neutral Probe Beam q-profile measurements in PDX and PBX-M

Description: Using the Fast Ion Diagnostic Experiment (FIDE) technique, a Neutral Probe Beam (NPB) can be aimed to inject tangentially to a magnetic surface. The resultant ion orbit shifts, due to conservation of canonical toroidal angular momentum, can be measured with a multi-sightline charge-exchange analyzer to yield direct measurements of radial magnetic flux profiles, current density profiles, the radial position of the magnetic axis, flux surface inner and outer edges, q-profiles, and central-q time dependencies. An extensive error analysis was performed on previous PDX q-measurements in circular plasmas and the resulting estimated contributions of various systematic effects are discussed. Preliminary results of fast ion orbit shift measurements at early times in indented PBX-M plasmas are given. Methods for increasing the absolute experimental precision of similar measurements in progress on PBX-M are discussed. 4 refs., 3 figs.
Date: June 1, 1988
Creator: Kugel, H.W.; Gammel, G.M.; Kaita, R.; Reusch, M.F. & Roberts, D.W.
Partner: UNT Libraries Government Documents Department

Performance of the PBX-M passive plate stabilization system

Description: The PBX-M passive plate stabilization system provides significant stabilization of long-wavelength external kink modes, the slowing of vertical instability growth rates, and the amelioration of disruption characteristics. The passive plate stabilization system has allowed the use of LHCD and IBW to induce current density and pressure profile modifications, and m = 1 divertor biasing for modifying edge plasma transport. Improvements in the passive plate system insulators and support structures have provided reliable operation. Impurity influxes with the close-fitting passive plates are low. Solid target boronization is applied routinely to reduce conditioning time and maintain clean conditions.
Date: February 1, 1994
Creator: Kugel, H. W.; Bell, R. & Bernabei, S.
Partner: UNT Libraries Government Documents Department

Plasma Facing Surface Composition During NSTX Li Experiments

Description: Lithium conditioned plasma facing surfaces have lowered recycling and enhanced plasma performance on many fusion devices. However, the nature of the plasma-lithium surface interaction has been obscured by the difficulty of in-tokamak surface analysis. We report laboratory studies of the chemical composition of lithium surfaces exposed to typical residual gases found in tokamaks. Solid lithium and a molybdenum alloy (TZM) coated with lithium has been examined using x-ray photoelectron spectroscopy, temperature programmed desorption, and Auger electron spectroscopy both in ultrahigh vacuum conditions and after exposure to trace gases. Lithium surfaces near room temperature were oxidized after exposure to 1-2 Langmuirs of oxygen or water vapor. The oxidation rate by carbon monoxide was four times less. Lithiated PFC surfaces in tokamaks will be oxidized in about 100 s depending on the tokamak vacuum conditions.
Date: July 20, 2012
Creator: Skinner, C. H.; Sullenberger, R.; Koel, B. E.; Jaworski, M. A. & Kugel, H. W.
Partner: UNT Libraries Government Documents Department

Development of Lithium Deposition Techniques for TFTR

Description: The ability to increase the quantity of lithium deposition into TFTR beyond that of the Pellet Injector while minimizing perturbations to the plasma provides interesting experimental and operational options. Two additional lithium deposition tools were developed for possible application during the 1996 Experimental Schedule: a solid lithium target probe for real-time deposition, and a lithium effusion oven for deposition between discharges. The lithium effusion oven was operated in TFTR to deposit lithium on the Inner Limiter in the absence of plasma. This resulted in the third highest power TFTR discharge.
Date: October 1, 1997
Creator: Gorman, J.; Johnson, D.; Kugel, H.W.; Labik, G.; Lemunyan, G. & al, et
Partner: UNT Libraries Government Documents Department

Remote Metrology, Mapping, and Motion Sensing of Plasma Facing Components Using FM Coherent Laser Radar

Description: Metrology inside a D/T burning fusion reactor must necessarily be conducted remotely since the in-vessel environment would be highly radioactive due to neutron activation of the torus walls. A technique based on frequency modulated coherent laser radar (FM CLR) for such remote metrology is described. Since the FM CLR relies on frequency shift to measure distances, the results are largely insensitive to surface reflectance characteristics. Results of measurements in TFTR and NSTX fusion devices using a prototype FM CLR unit, capable of remotely measuring distances (range) up to 22 m with better than 0.1-mm precision, are provided. These results illustrate that the FM CLR can be used for precision remote metrology as well as viewing. It is also shown that by conducting Doppler corrected range measurements using the CLR, the motion of objects can be tracked. Thus, the FM CLR has the potential to remotely measure the motion of plasma facing components (PFCs) during plasma disruptions.
Date: September 11, 2000
Creator: Menon, M.M.; Barry, R.E.; Slotwinsky, A.; Kugel, H.W. & Skinner, C.H.
Partner: UNT Libraries Government Documents Department

Overview of impurity control and wall conditioning in NSTX

Description: The National Spherical Torus Experiment (NSTX) started plasma operations in 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 condition. In this paper, the authors describe the internal geometry, and initial plasma discharge, impurity control, wall conditioning, erosion, and deposition results.
Date: May 23, 2000
Creator: Kugel, H.W.; Maingi, R.; Wampler, W.; Berry, R.E. & al, et
Partner: UNT Libraries Government Documents Department

Neutron Activation Cool-down of the Tokamak Fusion Test Reactor

Description: Tokamak Fusion Test Reactor (TFTR) final operations and post-shutdown neutron activation measurements were made. Ionization chambers were used to follow TFTR activation during operations and after shutdown. Gamma-ray energy spectroscopy measurements were performed to characterize TFTR activation at accessible vessel-bays and on sample hardware removed from structures at various distances from the vessel. The results demonstrate long-lived activations from common, commercially available materials used in the fabrication and field engineering of TFTR. The measurements allow characterization of residual TFTR neutron activation, the projection of residual activation decay, and benchmarking of low activation issues simulations.
Date: June 10, 1998
Creator: Ascione, G.; Kugel, H.W.; Kumar, A. & Tilson, Jr, C.
Partner: UNT Libraries Government Documents Department

NSTX High Temperature Sensor Systems

Description: 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.
Date: November 1, 1999
Creator: B.McCormack; Kugel, H.W.; Goranson, P.; Kaita, R. & al, et
Partner: UNT Libraries Government Documents Department

Fast Neutral Pressure Gauges in NSTX

Description: 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.
Date: April 26, 2004
Creator: Raman, R.; Kugel, H.W.; Gernhardt, R.; Provost, T.; Jarboe, T.R. & Soukhanovskii, V.
Partner: UNT Libraries Government Documents Department

Boronization on NSTX using Deuterated Trimethylboron

Description: 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.
Date: January 28, 2002
Creator: Blanchard, W.R.; Gernhardt, R.C.; Kugel, H.W. & LaMarche, P.H.
Partner: UNT Libraries Government Documents Department

Neutral beam interlock system on TFTR using infrared pyrometry

Description: Although the region of the TFTR vacuum vessel wall which is susceptible to damage by neutral beam strike is armored with a mosaic of TiC-clad POCO graphite titles, at power deposition levels above 2.5 kW/cm/sup 2/ the armor surface temperature exceeds 1200/sup 0/C within 250 ms and itself becomes susceptible to damage. In order to protect the wall armor, a neutral beam interlock system based on infrared pyrometry measurement of the armor surface temperature was installed on TFTR. For each beamline, a three-fiber-optic telescope views three areas of approx.30 cm diameter centered on the armor hot spots for the three ion sources. Each signal is fiber-optic coupled to a remote 900 nm pyrometer which feeds analog signals to the neutral beam interrupt circuits. The pyrometer interlock system is designed to interrupt each of the twelve ion sources independently within 10 ms of the temperature exceeding a threshold settable in the range of 500 to 2300/sup 0/C. A description of the pyrometer interlock system and its performance will be presented.
Date: June 1, 1986
Creator: Medley, S.S.; Kugel, H.W.; Kozub, T.A.; Lowrance, J.L.; Mastrocola, V.; Renda, G. et al.
Partner: UNT Libraries Government Documents Department

Initial operation and performance of the PDX neutral-beam injection system

Description: In 1981, the joint ORNL/PPPL PDX neutral beam heating project succeeded in reliably injecting 7.2 MW of D/sup 0/ into the PDX plasma, at nearly perpendicular angles, and achieved ion temperatures up to 6.5 keV. The expeditious achievement of this result was due to the thorough conditioning and qualification of the PDX neutral beam ion sources at ORNL prior to delivery coupled with several field design changes and improvements in the injection system made at PPPL as a result of neutral beam operating experience with the PLT tokamak. It has been found that the operation of high power neutral beam injection systems in a tokamak-neutral beam environment requires procedures and performance different from those required for development operation on test stands. In this paper, we review the installatin of the PDX neutral beam injection system, and its operation and performance during the initial high power plasma heating experiments with the PDX tokamak.
Date: January 19, 1982
Creator: Kugel, H.W.; Eubank, H.P.; Kozub, T.A.; Rossmassler, J.E.; Schilling, G.; van Halle, A. et al.
Partner: UNT Libraries Government Documents Department