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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

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