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Nuclear aspects of tokamak fusion test reactor (TFTR) diagnostics and instrumentation

Description: There are five principal aspects of the nuclear radiation from the high temperature plasmas of TFTR on its plasma diagnostic equipment. i) Important information about the plasma properties to be obtained from measurement of the neutrons, or other fusion reaction products. ii) Experimental studies to give design data for future tokamak devices and their instrumentation. iii) Transient noise or damage effects on the array of detectors for the collection of physics data about the plasma. iv) The effect of tritium on detectors that necessarily are in vacuum, directly connected to the tokamak vacuum vessel. v) Damage of diagnostic components mounted close to the vacuum vessel. Each of these topics will be addressed after a brief description of the TFTR tokamak and its radiation environment.
Date: January 1, 1982
Creator: Young, K.M.
Partner: UNT Libraries Government Documents Department

Alpha-particle diagnostics

Description: This paper will focus on the state of development of diagnostics which are expected to provide the information needed for {alpha}- physics studies in the future. Conventional measurement of detailed temporal and spatial profiles of background plasma properties in DT will be essential for such aspects as determining heating effectiveness, shaping of the plasma profiles and effects of MHD, but will not be addressed here. This paper will address (1) the measurement of the neutron source, and hence {alpha}-particle birth profile, (2) measurement of the escaping {alpha}-particles and (3) measurement of the confined {alpha}-particles over their full energy range. There will also be a brief discussion of (4) the concerns about instabilities being generated by {alpha}-particles and the methods necessary for measuring these effects. 51 refs., 10 figs.
Date: January 1, 1991
Creator: Young, K.M.
Partner: UNT Libraries Government Documents Department

Requirements for ITER diagnostics

Description: The development and design of plasma diagnostics for the International Thermonuclear Experimental Reactor (ITER) present a formidable challenge for experimental plasma physicists. The large plasma size, the high central density and temperature and the very high thermal wall loadings provide new challenges for present measurement techniques and lead to a search for new methods. But the physics and control requirements for the long burn phase of the discharge, combined with very limited access to the plasma, constrained by the requirement for radiation shielding of the coils and sharing of access ports with heating and current drive power, remote manipulation, fueling and turn blanket modules, make for very difficult design choices. An initial attempt at these choices has been made by an international team of diagnostic physicists, gathering together in a series of three workshops during the ITER Conceptual Design Activity. This paper is based on that report and provides a summary of its most important points. To provide a background against which to place the diagnostic requirements and design concepts, the ITER device, its most important plasma properties and the proposed experimental program will be described. The specifications for the measurement of the plasma parameters and the proposed diagnostics for these measurements will then be addressed, followed by some examples of the design concepts that have been proposed. As a result of these design studies, it was clear that there were many uncertainties associated with these concepts, particularly because of the nuclear radiation environment, so that a Research and Development Program for diagnostic hardware was established. It will also be briefly summarized.
Date: January 1, 1991
Creator: Young, K.M.
Partner: UNT Libraries Government Documents Department

Observations of Flaking of Co-deposited Layers in TFTR

Description: Flaking of co-deposited layers in the Tokamak Fusion Test Reactor (TFTR) has been observed after the termination of plasma operations. This unexpected flaking affects approximately 15% of the tiles and appears on isotropic graphite tiles but not on carbon fiber composite tiles. Samples of tiles, flakes and dust were recently collected from the inside of the vacuum vessel and will be analyzed to better characterize the behavior of tritium on plasma facing components in DT fusion devices.
Date: November 1, 1999
Creator: Gentile, C.A.; Skinner, C.H. & Young, K.M.
Partner: UNT Libraries Government Documents Department

TPX diagnostics for tokamak operation, plasma control and machine protection

Description: The diagnostics for TPX are at an early design phase, with emphasis on the diagnostic access interface with the major tokamak components. Account has to be taken of the very severe environment for diagnostic components located inside the vacuum vessel. The placement of subcontracts for the design and fabrication of the diagnostic systems is in process.
Date: August 1, 1995
Creator: Edmonds, P.H.; Medley, S.S. & Young, K.M.
Partner: UNT Libraries Government Documents Department

Ion cyclotron and spin-flip emissions from fusion products in tokamaks

Description: Power emission by fusion products of tokamak plasmas in their ion cyclotron range of frequencies (ICRF) and at their spin-flip resonance frequency is calculated for some specific model fusion product velocity-space distribution functions. The background plasma of say deuterium (D) is assumed to be in equilibrium with a Maxwellian distribution both for the electrons and ions. The fusion product velocity distributions analyzed here are: (1) A monoenergetic velocity space ring distribution. (2) A monoenergetic velocity space spherical shell distribution. (3) An anisotropic Maxwellian distribution with T [perpendicular] [ne] T[parallel]and with appreciable drift velocity along the confining magnetic field. Single dressed'' test particle spontaneous emission calculations are presented first and the radiation temperature for ion cyclotron emission (ICE) is analyzed both for black-body emission and nonequilibrium conditions. Thresholds for instability and overstability conditions are then examined and quasilinear and nonlinear theories of the electromagnetic ion cyclotron modes are discussed. Distinctions between kinetic or causal instabilities'' and hydrodynamic instabilities'' are drawn and some numerical estimates are presented for typical tokamak parameters. Semiquantitative remarks are offered on wave accessibility, mode conversion, and parametric decay instabilities as possible for spatially localized ICE. Calculations are carried out both for k[parallel] = 0 for k[parallel] [ne] 0. The effects of the temperature anisotropy and large drift velocities in the parallel direction are also examined. Finally, proton spin-flip resonance emission and absorption calculations are also presented both for thermal equilibrium conditions and for an inverted'' population of states.
Date: February 1, 1993
Creator: Arunasalam, V.; Greene, G.J. & Young, K.M.
Partner: UNT Libraries Government Documents Department

Ion cyclotron and spin-flip emissions from fusion products in tokamaks

Description: Power emission by fusion products of tokamak plasmas in their ion cyclotron range of frequencies (ICRF) and at their spin-flip resonance frequency is calculated for some specific model fusion product velocity-space distribution functions. The background plasma of say deuterium (D) is assumed to be in equilibrium with a Maxwellian distribution both for the electrons and ions. The fusion product velocity distributions analyzed here are: (1) A monoenergetic velocity space ring distribution. (2) A monoenergetic velocity space spherical shell distribution. (3) An anisotropic Maxwellian distribution with T {perpendicular} {ne} T{parallel}and with appreciable drift velocity along the confining magnetic field. Single ``dressed`` test particle spontaneous emission calculations are presented first and the radiation temperature for ion cyclotron emission (ICE) is analyzed both for black-body emission and nonequilibrium conditions. Thresholds for instability and overstability conditions are then examined and quasilinear and nonlinear theories of the electromagnetic ion cyclotron modes are discussed. Distinctions between ``kinetic or causal instabilities`` and ``hydrodynamic instabilities`` are drawn and some numerical estimates are presented for typical tokamak parameters. Semiquantitative remarks are offered on wave accessibility, mode conversion, and parametric decay instabilities as possible for spatially localized ICE. Calculations are carried out both for k{parallel} = 0 for k{parallel} {ne} 0. The effects of the temperature anisotropy and large drift velocities in the parallel direction are also examined. Finally, proton spin-flip resonance emission and absorption calculations are also presented both for thermal equilibrium conditions and for an ``inverted`` population of states.
Date: February 1, 1993
Creator: Arunasalam, V.; Greene, G. J. & Young, K. M.
Partner: UNT Libraries Government Documents Department

Alpha-physics and measurement requirements for ITER

Description: This paper reviews alpha particle physics issues in ITER and their implications for alpha particle measurements. A comparison is made between alpha heating in ITER and NBI and ICRH heating systems in present tokamaks, and alpha particle issues in ITER are discussed in three physics areas: `single particle` alpha effects, `collective` alpha effects, and RF interactions with alpha particles. 29 refs., 4 figs., 4 tabs.
Date: December 31, 1995
Creator: Zweben, S.J.; Young, K.M.; Putvinski, S.; Petrov, M.P.; Sadler, G. & Tobita, K.
Partner: UNT Libraries Government Documents Department

Tritium Decontamination of TFTR D-T Graphite Tiles Employing Ultra Violet Light and a Nd:YAG Laser

Description: The use of an ultra violet (UV) light source (wavelength = 172 nm) and a Nd:YAG Laser for the decontamination of the Tokamak Fusion Test Reactor (TFTR) deuterium-tritium (D-T) tiles will be investigated at the Princeton Plasma Physics Laboratory (PPPL). The development of this form of tritium decontamination may be useful for future D-T burning fusion devices which employ carbon plasma-facing components on the first wall. Carbon tiles retain hydrogen isotopes, and the in-situ tritium decontamination of carbon can be extremely important in maintaining resident in-vessel tritium inventory to a minimum. A test chamber has been designed and fabricated at PPPL. The chamber has the ability to be maintained under vacuum, be baked to 200 *C, and provides sample ports for gas analyses. Tiles from TFTR that have been exposed to D-T plasmas will be placed within the chamber and exposed to either an UV light source or the ND:YAG Laser. The experiment will determine the effectiveness of these two techniques for the removal of tritium. In addition, exposure rates and scan times for the UV light source and/or Nd:YAG Laser will be determined for tritium removal optimization from D-T tiles.
Date: October 1, 1999
Creator: Gentile, C.A.; Skinner, C.H.; Young, K.M.; Ciebiera, L. & al, et
Partner: UNT Libraries Government Documents Department

Characterization of the TFTR plasma edge by Langmuir-calorimeter probes

Description: Two combination Langmuir-calorimeter probes were operated on the TFTR midplane to measure plasma properties of the scrape-off layer. Two different caloriemter elements were used: 0.5 x 3 x 8 mm Ta plates or 6.4 mm diameter graphite rods fitted with thermocouples. Separate graphite rods served as Langmuir elements. This paper presents Langmuir probe measurements of the radial profiles of edge density and electron temperature, and calorimeter measurements of heat flux. The variation of these quantities in certain operational regimes are presented, including dependence on ohmic and neutral beam heating, compression and free expansion, plasma major and minor radii, plasma current, and line integral density with and without Cr gettering.
Date: March 1, 1986
Creator: Kilpatrick, S.J.; Manos, D.M.; Budny, R.V.; Stangeby, P.C.; Ritter, R.S. & Young, K.M.
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

Absolute calibration of TFTR neutron detectors for D-T plasma operation

Description: The two most sensitive TFTR fission-chamber detectors were absolutely calibrated in situ by a D-T neutron generator ({approximately}5 {times} 10{sup 7} n/s) rotated once around the torus in each direction, with data taken at about 45 positions. The combined uncertainty for determining fusion neutron rates, including the uncertainty in the total neutron generator output ({plus_minus}9%), counting statistics, the effect of coil coolant, detector stability, cross-calibration to the current mode or log Campbell mode and to other fission chambers, and plasma position variation, is about {plus_minus}13%. The NE-451 (ZnS) scintillators and {sup 4}He proportional counters that view the plasma in up to 10 collimated sightlines were calibrated by scanning. the neutron generator radially and toroidally in the horizontal midplane across the flight tubes of 7 cm diameter. Spatial integration of the detector responses using the calibrated signal per unit chord-integrated neutron emission gives the global neutron source strength with an overall uncertainty of {plus_minus}14% for the scintillators and {plus_minus}15% for the {sup 4}He counters.
Date: March 1, 1995
Creator: Jassby, D.L.; Johnson, L.C.; Roquemore, A.L.; Strachan, J.D.; Johnson, D.W.; Medley, S.S. et al.
Partner: UNT Libraries Government Documents Department

Tritium Removal from Codeposits on Carbon Tiles by a Scanning Laser

Description: A novel method for tritium release has been demonstrated on codeposited layers on graphite and carbon-fiber-composite tiles from the Tokamak Fusion Test Reactor (TFTR). A scanning continuous wave Nd laser beam heated the codeposits to a temperature of 1200-2300 degrees C for 10 to 200 milliseconds in an argon atmosphere. The temperature rise of the codeposit was significantly higher than that of the manufactured tile material (e.g., 1770 degrees C cf. 1080 degrees C). A major fraction of tritium was thermally desorbed with minimal change to the surface appearance at a laser intensity of 8 kW/cm(superscript ''2''), peak temperatures above 1230 degrees C and heating duration 10-20 milliseconds. In two experiments, 46% and 84% of the total tritium was released during the laser scan. The application of this method for tritium removal from a tokamak reactor appears promising and has significant advantages over oxidative techniques.
Date: September 28, 2001
Creator: Skinner, C.H.; Gentile, C.A.; Carpe, A.; Guttadora, G.; Langish, S.; Young, K.M. et al.
Partner: UNT Libraries Government Documents Department

Long Term Tritium Trapping in TFTR and JET

Description: Tritium retention in TFTR [Tokamak Fusion Test Reactor] and JET [Joint European Torus] shows striking similarities and contrasts. In TFTR, 5 g of tritium were injected into circular plasmas over a 3.5 year period, mostly by neutral-beam injection. In JET, 35 g were injected into divertor plasmas over a 6 month campaign, mostly by gas puffing. In TFTR, the bumper limiter provided a large source of eroded carbon and a major part of tritium was co-deposited on the limiter and vessel wall. Only a small area of the co-deposit flaked off. In JET, the wall is a net erosion area, and co-deposition occurs principally in shadowed parts of the inner divertor, with heavy flaking. In both machines, the initial tritium retention, after a change from deuterium [D] to tritium [T] gas puffing, is high and is due to isotope exchange with deuterium on plasma-facing surfaces (dynamic inventory). The contribution of co-deposition is lower but cumulative, and is revealed by including periods of D fueling that reversed the T/D isotope exchange. Ion beam analysis of flakes from TFTR showed an atomic D/C ratio of 0.13 on the plasma facing surface, 0.25 on the back surface and 0.11 in the bulk. Data from a JET divertor tile showed a larger D/C ratio with 46% C, 30% D, 20% H and 4% O. Deuterium, tritium, and beryllium profiles have been measured and show a thin less than 50 micron co-deposited layer. Flakes retrieved from the JET vacuum vessel exhibited a high tritium release rate of 2e10 Bq/month/g. BBQ modeling of the effect of lithium on retention in TFTR showed overlapping lithium and tritium implantation and a 1.3x increase in local T retention.
Date: July 24, 2001
Creator: Skinner, C.H.; Gentile, C.A.; Young, K.M.; Coad, J.P.; Hogan, J.T.; Penzhorn, R.-D. et al.
Partner: UNT Libraries Government Documents Department

A Visual Detection System for Determining Tritium Surface Deposition Employing Phosphor Coated Materials

Description: A method for visually observing tritium deposition on the surface of the Tokamak Fusion Test Reactor (TFTR) deuterium-tritium (D-T) tiles is being investigated at the Princeton Plasma Physics Laboratory. A green phosphor (P31, zinc sulfide: copper) similar to that used in oscilloscope screens with a wavelength peak of 530 nm was positioned on the surface of a TFTR D-T tile. The approximately 600 gram tile, which contains approximately 1.5 Ci of tritium located on the top approximately 1-50 microns of the surface, was placed in a two liter lexan chamber at Standard Temperature and Pressure (STP). The phosphor plates and phosphor powder were placed on the surface of the tile which resulted in visible light being observed, the consequence of tritium betas interacting with the phosphor. This technique provides a method of visually observing varying concentrations of tritium on the surface of D-T carbon tiles, and may be employed (in a calibrated system) to obtain quantitative data.
Date: October 1, 1999
Creator: Gentile, C.A.; Skinner, C.H.; Young, K.M.; Zweben, S.J. & al, et
Partner: UNT Libraries Government Documents Department

In-situ Tritium Measurements of the Tokamak Fusion Test Reactor Bumper Limiter Tiles Post D-T Operations

Description: The Princeton Plasma Physics Laboratory (PPPL) Engineering and Research Staff in collaboration with members of the Japan Atomic Energy Research Institute (JAERI), Tritium Engineering Laboratory have commenced in-situ tritium measurements of the TFTR bumper limiter. The Tokamak Fusion Test Reactor (TFTR) operated with tritium from 1993 to 1997. During this time {approximately} 53,000 Ci of tritium was injected into the TFTR vacuum vessel. After the cessation of TFTR plasma operations in April 1997 an aggressive tritium cleanup campaign lasting {approximately} 3 months was initiated. The TFTR vacuum vessel was subjected to a regimen of glow discharge cleaning (GDC) and dry nitrogen and ''moist air'' purges. Currently {approximately} 7,500 Ci of tritium remains in the vacuum vessel largely contained in the limiter tiles. The TFTR limiter is composed of 1,920 carbon tiles with an average weight of {approximately} 600 grams each. The location and distribution of tritium on the TFTR carbon tiles are of considerable interest. Future magnetically confined fusion devices employing carbon as a limiter material may be considerably constrained due to potentially large tritium inventories being tenaciously held on the surface of the tiles. In-situ tritium measurements were conducted in TFTR bay L during August and November 1998. During the bay L measurement campaign open wall ion chambers and ultra thin thermoluminscent dosimeters (TLD) affixed to a boom and end effector were deployed into the vacuum vessel. The detectors were designed to make contact with the surface of the bumper limiter tile and to provide either real time (ion chamber) or passive (TLD) indication of the surface tritium concentration. The open wall ion chambers were positioned onto the surface of the tile in a manner which employed the surface of the tile as one of the walls of the chamber. The ion chambers, which are (electrically) gamma insensitive, were ...
Date: September 1, 1999
Creator: Gentile, C.A.; Skinner, C.H.; Young, K.M.; Nishi, M.; Langish, S. & al, et
Partner: UNT Libraries Government Documents Department

Tritium Removal by Laser Heating and Its Application to Tokamaks

Description: A novel laser heating technique has recently been applied to removing tritium from carbon tiles that had been exposed to deuterium-tritium (DT) plasmas in the Tokamak Test Fusion Reactor (TFTR). A continuous wave neodymium laser, of power up to 300 watts, was used to heat the surface of the tiles. The beam was focused to an intensity, typically 8 kW/cm{sup 2}, and rapidly scanned over the tile surface by galvanometer-driven scanning mirrors. Under the laser irradiation, the surface temperature increased dramatically, and temperatures up to 2,300 degrees C were recorded by an optical pyrometer. Tritium was released and circulated in a closed-loop system to an ionization chamber that measured the tritium concentration. Most of the tritium (up to 84%) could be released by the laser scan. This technique appears promising for tritium removal in a next-step DT device as it avoids oxidation, the associated deconditioning of the plasma facing surfaces, and the expense of processing large quantities of tritium oxide. Some engineering aspects of the implementation of this method in a next-step fusion device will be discussed.
Date: November 16, 2001
Creator: Skinner, C.H.; Gentile, C.A.; Guttadora, G.; Carpe, A.; Langish, S.; Young, K.M. et al.
Partner: UNT Libraries Government Documents Department

Tritium Removal by Laser Heating and Its Application to Tokamaks

Description: A novel laser heating technique has recently been applied to removing tritium from carbon tiles that had been exposed to deuterium-tritium (DT) plasmas in the Tokamak Test Fusion Reactor (TFTR). A continuous wave neodymium laser, of power up to 300 watts, was used to heat the surface of the tiles. The beam was focused to an intensity, typically 8 kW/cm{sup 2}, and rapidly scanned over the tile surface by galvanometer-driven scanning mirrors. Under the laser irradiation, the surface temperature increased dramatically, and temperatures up to 2,300 degrees C were recorded by an optical pyrometer. Tritium was released and circulated in a closed-loop system to an ionization chamber that measured the tritium concentration. Most of the tritium (up to 84%) could be released by the laser scan. This technique appears promising for tritium removal in a next-step DT device as it avoids oxidation, the associated deconditioning of the plasma facing surfaces, and the expense of processing large quantities of tritium oxide. Some engineering aspects of the implementation of this method in a next-step fusion device will be discussed.
Date: November 16, 2001
Creator: Skinner, C.H.; Gentile, C.A.; Guttadora, G.; Carpe, A.; Langish, S.; Young, K.M. et al.
Partner: UNT Libraries Government Documents Department

Tritium Removal by Laser Heating and Its Application to Tokamaks

Description: A novel laser heating technique has recently been applied to removing tritium from carbon tiles that had been exposed to deuterium-tritium (DT) plasmas in the Tokamak Test Fusion Reactor (TFTR). A continuous wave neodymium laser, of power up to 300 watts, was used to heat the surface of the tiles. The beam was focused to an intensity, typically 8 kW/cm{sup 2}, and rapidly scanned over the tile surface by galvanometer-driven scanning mirrors. Under the laser irradiation, the surface temperature increased dramatically, and temperatures up to 2,300 degrees C were recorded by an optical pyrometer. Tritium was released and circulated in a closed-loop system to an ionization chamber that measured the tritium concentration. Most of the tritium (up to 84%) could be released by the laser scan. This technique appears promising for tritium removal in a next-step DT device as it avoids oxidation, the associated deconditioning of the plasma facing surfaces, and the expense of processing large quantities of tritium oxide. Some engineering aspects of the implementation of this method in a next-step fusion device will be discussed.
Date: November 16, 2001
Creator: Skinner, C.H.; Gentile, C.A.; Guttadora, G.; Carpe, A.; Langish, S.; Young, K.M. et al.
Partner: UNT Libraries Government Documents Department

Tritium Removal by Laser Heating and Its Application to Tokamaks

Description: A novel laser heating technique has recently been applied to removing tritium from carbon tiles that had been exposed to deuterium-tritium (DT) plasmas in the Tokamak Test Fusion Reactor (TFTR). A continuous wave neodymium laser, of power up to 300 watts, was used to heat the surface of the tiles. The beam was focused to an intensity, typically 8 kW/cm{sup 2}, and rapidly scanned over the tile surface by galvanometer-driven scanning mirrors. Under the laser irradiation, the surface temperature increased dramatically, and temperatures up to 2,300 degrees C were recorded by an optical pyrometer. Tritium was released and circulated in a closed-loop system to an ionization chamber that measured the tritium concentration. Most of the tritium (up to 84%) could be released by the laser scan. This technique appears promising for tritium removal in a next-step DT device as it avoids oxidation, the associated deconditioning of the plasma facing surfaces, and the expense of processing large quantities of tritium oxide. Some engineering aspects of the implementation of this method in a next-step fusion device will be discussed.
Date: November 16, 2001
Creator: Skinner, C.H.; Gentile, C.A.; Guttadora, G.; Carpe, A.; Langish, S.; Young, K.M. et al.
Partner: UNT Libraries Government Documents Department

Periscope-camera system for visible and infrared imaging diagnostics on TFTR

Description: An optical diagnostic consisting of a periscope which relays images of the torus interior to an array of cameras is used on the Tokamak Fusion Test Reactor (TFTR) to view plasma discharge phenomena and inspect vacuum vessel internal structures in both visible and near-infrared wavelength regions. Three periscopes view through 20-cm-diameter fused-silica windows which are spaced around the torus midplane to provide a viewing coverage of approximately 75% of the vacuum vessel internal surface area. The periscopes have f/8 optics and motor-driven controls for focusing, magnification selection (5/sup 0/, 20/sup 0/, and 60/sup 0/ field of view), elevation and azimuth setting, mast rotation, filter selection, iris aperture, and viewing port selection. The four viewing ports on each periscope are equipped with multiple imaging devices which include: (1) an inspection eyepiece, (2) standard (RCA TC2900) and fast (RETICON) framing rate television cameras, (3) a PtSi CCD infrared imaging camera, (4) a 35 mm Nikon F3 still camera, or (5) a 16 mm Locam II movie camera with variable framing up to 500 fps. Operation of the periscope-camera system is controlled either locally or remotely through a computer-CAMAC interface. A description of the equipment and examples of its application are presented.
Date: May 1, 1985
Creator: Medley, S.S.; Dimock, D.L.; Hayes, S.; Long, D.; Lowrence, J.L.; Mastrocola, V. et al.
Partner: UNT Libraries Government Documents Department

TFTR initial operations

Description: The Tokamak Fusion Test Reactor (TFTR) has operated since December 1982 with ohmically heated plasmas. Routine operation with feedback control of plasma current, position, and density has been obtained for plasmas with I/sub p/ approx. = 800 kA, a = 68 cm, R = 250 cm, and B/sub t/ = 27 kG. A maximum plasma current of 1 MA was achieved with q approx. = 2.5. Energy confinement times of approx. 150 msec were measured for hydrogen and deuterium plasmas with anti n/sub e/ approx. = 2 x 10/sup 13/ cm/sup -3/, T/sub e/ (0) approx. = 1.5 keV, T/sub i/ (0) approx. = 1.5 keV, and Z/sub eff/ approx. = 3. The preliminary results suggest a size-cubed scaling from PLT and are consistent with Alcator C scaling where tau approx. nR/sup 2/a. Initial measurements of plasma disruption characteristics indicate current decay rates of approx. 800 kA in 8 ms which is within the TFTR design requirement of 3 MA in 3 ms.
Date: November 1, 1983
Creator: Young, K.M.; Bell, M.; Blanchard, W.R.; Bretz, N.; Cecchi, J.; Coonrod, J. et al.
Partner: UNT Libraries Government Documents Department

Satellite spectra for helium-like titanium. Part II

Description: K/sup ..cap alpha../ x-ray spectra of helium-like titanium, Ti XXI, from Tokamak Fusion Test Reactor (TFTR) plasmas have been observed with a high resolution crystal spectrometer and have been used as a diagnostic of central plasma parameters. The data allow detailed comparison with recent theoretical predictions for the Ti XXI helium-like lines and the associated satellite spectrum in the wavelength range from 2.6000 to 2.6400 A. Improved values for the excitation rate coefficients of the Ti XXI resonance line, the intercombination lines and the forbidden line, and new theoretical results on the wavelengths and transition probabilities for beryllium-like satellites due to transitions of the type 1s/sup 2/ 2lnl' - 1s2p2l'' nl'' with n = 2-4 have been calculated.
Date: August 1, 1985
Creator: Bitter, M.; Hill, K.W.; Zarnstorff, M.; von Goeler, S.; Hulse, R.; Johnson, L.C. et al.
Partner: UNT Libraries Government Documents Department

Visual tritium imaging of in-vessel surfaces

Description: An imaging detector has been developed for the purpose of providing a non-destructive, real time method of determining tritium concentrations on the surface of internal TFTR vacuum vessel components. The detector employs a green phosphor screen (P31, zinc sulfide: copper) with a wave length peak of 530 nm, a charge-coupled device (CCD) camera linked to a computer, and a detection chamber for inserting components recovered from the vacuum vessel. This detector is capable of determining tritium concentrations on the surfaces. The detector provides a method of imaging tritium deposition on the surfaces in a fairly rapid fashion.
Date: May 22, 2000
Creator: Gentile, C.A.; Zweben, S.J.; Skinner, C.H.; Young, K.M.; Langish, S.W.; Nishi, M.F. et al.
Partner: UNT Libraries Government Documents Department