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Measurements of sputtering yields for low-energy plasma ions

Description: Sputtering yields of various wall/limiter materials of fusion devices have been extensively measured in the relevant plasma environment for low-energy light ions (E < 300eV). As a plasma source we have used an energetic arc device (QED-1 machine) in which hydrogen, deuterium, helium and argon plasma can be generated with a density of up to 10/sup 14/ cm/sup -3/ and electron temperature up to 10eV. Target materials used were C (graphite), Ti, Mo, Ta, W, and Fe (stainless steel). In order to study the dependence of the sputtering yields on the incident energy of ions, the target samples were held at negative bias voltage up to 300V. The sputtering yields were determined by a weight-loss method and by spectral line intensity measurements. The data obtained in the present experiment agree well with those previously obtained at the higher energies (E greater than or equal to 200eV) by other authors using different schemes; the present data also extend to substantially lower energies (E approx. > 30eV) than hitherto.
Date: April 1, 1979
Creator: Nishi, M.; Yamada, M.; Suckewer, S. & Rosengaus, E.
Partner: UNT Libraries Government Documents Department

Martensitic transformation of CsFeS/sub 2/ driven by the singlet to Neel state transition

Description: CsFeS/sub 2/ is regarded as an example of the quasi one-dimensional alternating antiferromagnetic Heisenberg chain above about 70K. In accordance with this picture, an energy gap of 10MeV was observed for the singlet to triplet excitation at a zone center near the transition temperature by our neutron measurements. Large excitation width implies a strong coupling of excitons to phonons, and LA phonons along <001> direction become ill-defined for q greater than 0.3. At about 70K, the first order transition occurs, whereby the singlet ground state changes to a Neel state and the simultaneous structural transformation takes place, which is a martensitic transformation. Mechanism for such martensitic transformation is discussed based on the similarity of the magnetic excitation and phonon behaviors between the present compound and the ..gamma.. Mn alloys with Fe and Cu, which are itinerant electron magnetic systems. 10 refs., 3 figs.
Date: January 1, 1985
Creator: Ito, Y.; Nishi, M.; Passell, L.; Majkrzak, C.F. & Shirane, G.
Partner: UNT Libraries Government Documents Department

FENIX experimental results of large-scale CICC made of bronze-processed Nb{sub 3}Sn strands

Description: The Fusion ENgineering International eXperiments (FENIX) Test Facility recently has successfully complete the testing of a pair of Nb{sub 3}rSn cable-in-conduit conductors developed by the Japan Atomic Energy Research Institute. These conductors, made of bronze-processed strands, were designed to operate stably with 40-kA transport current at a magnetic field of 13 T. In addition to the measurements of major design parameters such as current-sharing temperature, FENIX provided several experiments specifically designed to provide results urgently needed by magnet designers. Performed experiments include measurements of ramp-rate limit, current-distribution, stability, and joint performance. This paper presents the design and results of these special experiments.
Date: October 13, 1994
Creator: Shen, S.S.; Felker, B.; Moller, J.M.; Parker, J.M.; Isono, T.; Yasukawa, Y. et al.
Partner: UNT Libraries Government Documents Department

Radiochemical reactions between tritium and humid air

Description: Radiochemical reactions between pure tritium (T{sub 2}) and moist air have been examined using real-time Raman spectroscopy. The reacting constituents were contained in a 1 cm{sup 3} quartz cell sealed by a quartz-to-metal seal leading to a valve. A near-stoichiometric mixture of T{sub 2} and O{sub 2} was introduced into the cell, and the time evolution of the composition was monitored at 297 K for twenty-nine days. The production of T{sub 2}O was observed in these experiments, for the first time unambiguously detected in Raman spectroscopy. T{sub 2}O exhibits a relatively weak vibrational band at {approximately}2,313 cm{sup {minus}1}. The radiochemical production of tritiated water did not occur in the expected 2:1 ratio, but rather with the O{sub 2} disappearing totally when the T{sub 2} was only slightly over halfway depleted. After the disappearance of O{sub 2}, the T{sub 2} partial pressure continued to decrease, but at a slower rate. The initial water in the moist-air mixture disappeared totally after about 15 hours, with no concomitant production of HT. A small quantity of CO{sub 2} was also detected, presumably produced by radiochemically driven reactions with stainless steel components.
Date: March 1, 1998
Creator: Sherman, R.H.; Taylor, D.J.; Honnell, K.G.; O`hira, S.; Kawamura, Y.; Nishi, M. et al.
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

Preliminary results of the partial array LCT coil tests

Description: The Large Coil Task (LCT) is a collaboration between the US, Euratom, Japan, and Switzerland for the production and testing of 2.5 x 3.5-m bore, superconducting 8-T magnets. The definitive tests in the design configuration, the six coils arrayed in a compact torus, will begin in 1985. Partial-array tests are being done in 1984. In January the initial cooldown of two coils was aborted because of helium-to-vacuum leaks that developed in certain seal welds when the coil temperatures were 170 to 180 K. In July three adjacent coils (designated JA, GD, CH) were cooled and in August two were energized to the limits of the test facility. An overview of the results are presented, including facility, cooldown (warmup has not yet begun), energization, dump, recovery from intentional normal zones, strain, and displacement, for operation up to 100% of design current but below full field and stress. These initial results are highly encouraging.
Date: September 10, 1984
Creator: Luton, J.N.; Cogswell, F.D.; Dresner, L.; Friesinger, G.M.; Gray, W.H.; Iwasa, Y. et al.
Partner: UNT Libraries Government Documents Department

ISX-A graphite limiter experiment

Description: Graphite limiters were installed and tested in the ISX-A tokamak as part of the ISX-A surface physics program and the TFTR materials research program. The puropse of the experiment was to compare plasma performance using graphite limiters as opposed to the standard ISX-A stainless steel limiters. Heaters were installed in the graphite limiters so that the effects of operation at elevated temperatures could be evaluated.
Date: January 1, 1979
Creator: Langley, R.A.; Colchin, R.J.; Isler, R.C.; Murakami, M.; Simpkins, J.E.; Cecchi, J.L. et al.
Partner: UNT Libraries Government Documents Department

Magnetic Excitation of CuGeO{sub 3} under Applied Pressure

Description: Magnetic excitations of the spin-Peierls compound CuGeO{sub 3} under applied pressure of 2 GPa have been studied. The dispersion along the chain direction up to zone boundary has been obtained. The spin-Peierls gap energy increases to 4.2 meV and the zone boundary energy decreases to 14.1 meV. The pressure dependence of dispersion relation can be interpreted by the increase of the next-nearest-neighbor intra-chain interaction under applied pressure causing the increase of both the spin-Peierls gap energy and transition temperature.
Date: July 31, 1997
Creator: Nishi, M.; Kakurai, K.; Fujii, Y.; Yethiraj, M.; Tennant, D. A.; Nagler, S. E. 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

Studies of tritiated co-deposited layers in TFTR

Description: Plasma facing components in TFTR contain an important record of plasma wall interactions in reactor grade DT plasmas. Tiles, flakes, wall coupons, a stainless steel shutter and dust samples have been retrieved from the TFTR vessel for analysis. Selected samples have been baked to release tritium and assay the tritium content. The in-vessel tritium inventory is estimated to be 0.56 g and is consistent with the in-vessel tritium inventory derived from the difference between tritium fueling and tritium exhaust. The distribution of tritium on the limiter and vessel wall showed complex patterns of co-deposition. Relatively high concentrations of tritium were found at the top and bottom of the bumper limiter, as predicted by earlier BBQ modeling.
Date: June 28, 2000
Creator: Skinner, C.H.; Gentile, C.A.; Ascione, G.; Carpe, A.; Causey, R.A.; Hayashi, T. et al.
Partner: UNT Libraries Government Documents Department

Studies of tritiated co-deposited layers in TFTR

Description: Plasma facing components in TFTR contain an important record of plasma wall interactions in reactor grade DT plasmas. Tiles, flakes, wall coupons and dust samples have been retrieved from the TFTR vessel for analysis. Selected samples have been baked to release tritium and assay the tritium content. The in-vessel tritium inventory is estimated to be 0.5 g and is consistent with the in-vessel tritium inventory derived from the difference between tritium fueling and tritium exhaust. Relatively high concentrations of tritium were found at the top and bottom of the bumper limiter, as predicted by earlier BBQ modeling. The distribution of tritium on the limiter and vessel wall showed complex patterns of co-deposition.
Date: May 30, 2000
Creator: SKINNER,C.H.; GENTILE,C.A.; ASCIONE,G.; CAUSEY,R.A.; HAYASKI,T.; HOGAN,J. et al.
Partner: UNT Libraries Government Documents Department

Visual tritium imaging of In-Vessel surfaces

Description: A 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: June 26, 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

Studies of tritiated co-deposited Layers in TFTR

Description: Plasma facing components in TFTR contain an important record of plasma wall interactions in reactor grade DT plasmas. Tiles, flakes, wall coupons, a stainless steel shutter and dust samples have been retrieved from the TFTR vessel for analysis. Selected samples have been baked to release tritium and assay the tritium content. The in-vessel tritium inventory is estimated to be 0.56 g and is consistent with the in-vessel tritium inventory derived from the difference between tritium fueling and tritium exhaust. The distribution of tritium on the limiter and vessel wall showed complex patterns of co-deposition. Relatively high concentrations of tritium were found at the top and bottom of the bumper limiter, as predicted by earlier BBQ modeling.
Date: May 22, 2000
Creator: Skinner, C. H.; Gentile, C. A.; Ascione, G.; Carpe, A.; Causey, R. A.; Hayashi, T. et al.
Partner: UNT Libraries Government Documents Department

Test data from the US-Demonstration Poloidal Coil experiment

Description: The US Demonstration Poloidal Field Coil (US-DPC) experiment took place successfully at the Japan Atomic Energy Research Institute (JAERI) in late 1990. The 8 MJ niobium-tin coil was leak tight; it performed very well in DC tests; it performed well in AC tests, achieving approximately 70% of its design goal. An unexpected ramp-rate barrier at high currents was identified. The barrier could not be explored in the regime of higher fields and slower ramp rates due to limitations of the background-field coils. This document presents the results of the experiment with as little editing as possible. The coil, conductor, and operating conditions are given. The intent is to present data in a form that can be used by magnet analysts and designers.
Date: January 1, 1992
Creator: Painter, T.A.; Steeves, M.M.; Takayasu, M.; Gung, C.; Hoenig, M.O. (Massachusetts Inst. of Tech., Cambridge, MA (United States). Plasma Fusion Center); Tsuji, H. et al.
Partner: UNT Libraries Government Documents Department

Test data from the US-Demonstration Poloidal Coil experiment

Description: The US Demonstration Poloidal Field Coil (US-DPC) experiment took place successfully at the Japan Atomic Energy Research Institute (JAERI) in late 1990. The 8 MJ niobium-tin coil was leak tight; it performed very well in DC tests; it performed well in AC tests, achieving approximately 70% of its design goal. An unexpected ramp-rate barrier at high currents was identified. The barrier could not be explored in the regime of higher fields and slower ramp rates due to limitations of the background-field coils. This document presents the results of the experiment with as little editing as possible. The coil, conductor, and operating conditions are given. The intent is to present data in a form that can be used by magnet analysts and designers.
Date: January 1, 1992
Creator: Painter, T. A.; Steeves, M. M.; Takayasu, M.; Gung, C.; Hoenig, M. O.; Tsuji, H. et al.
Partner: UNT Libraries Government Documents Department