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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

Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

Description: A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed.
Date: January 18, 2002
Creator: Lumia, M.E. & Gentile, C.A.
Partner: UNT Libraries Government Documents Department

Tritiated Dust Levitation by Beta Induced Static Charge

Description: Tritiated particles have been observed to spontaneously levitate under the influence of a static electric field. Tritium containing co-deposits were mechanically scraped from tiles that had been used in the Tokamak Fusion Test Reactor (TFTR) inner limiter during the deuterium-tritium campaign and were placed in a glass vial. On rubbing the plastic cap of the vial a remarkable ''fountain'' of particles was seen inside the vial. Particles from an unused tile or from a TFTR co-deposit formed during deuterium discharges did not exhibit this phenomenon. It appears that tritiated particles are more mobile than other particles and this should be considered in assessing tokamak accident scenarios and in occupational safety.
Date: June 4, 2003
Creator: Skinner, C.H.; Gentile, C.A.; Ciebiera, L. & Langish, S.
Partner: UNT Libraries Government Documents Department

The operation of the Tokamak Fusion Test Reactor Tritium Facility

Description: The TFTR tritium operations staff has successfully received, stored, handled, and processed over five hundred thousand curies of tritium for the purpose of supporting D-T (Deuterium-Tritium) operations at TFTR. Tritium operations personnel nominally provide continuous round the clock coverage (24 hours/day, 7 days/week) in shift complements consisting of I supervisor and 3 operators. Tritium Shift Supervisors and operators are required to have 5 years of operational experience in either the nuclear or chemical industry and to become certified for their positions. The certification program provides formal instruction, as well as on the job training. The certification process requires 4 to 6 months to complete, which includes an oral board lasting up to 4 hours at which time the candidate is tested on their knowledge of Tritium Technology and TFTR Tritium systems. Once an operator is certified, the training process continues with scheduled training weeks occurring once every 5 weeks. During D-T operations at TFTR the operators must evacuate the tritium area due to direct radiation from TFTR D-T pulses. During `` time operators maintain cognizance over tritium systems via a real time TV camera system. Operators are able to gain access to the Tritium area between TFTR D-T pulses, but have been excluded from die tritium area during D-T pulsing for periods up to 30 minutes. Tritium operators are responsible for delivering tritium gas to TFRR as well as processing plasma exhaust gases which lead to the deposition of tritium oxide on disposable molecular sieve beds (DMSB). Once a DMSB is loaded, the operations staff remove the expended DMSB, and replace it with a new DMSB container. The TFIR tritium system is operated via detailed procedures which require operator sign off for system manipulation. There are >300 procedures controlling the operation of the tritium systems.
Date: July 1, 1995
Creator: Gentile, C.A.; LaMarche, P.H. & Anderson, J.L.
Partner: UNT Libraries Government Documents Department

High Heat Flux Interactions and Tritium Removal from Plasma Facing Components by a Scanning Laser

Description: A new technique for studying high heat flux interactions with plasma facing components is presented. The beam from a continuous wave 300 W neodymium laser was focused to 80 W/mm2 and scanned at high speed over the surface of carbon tiles. These tiles were previously used in the TFTR [Tokamak Fusion Test Reactor] inner limiter and have a surface layer of amorphous hydrogenated carbon that was codeposited during plasma operations. Laser scanning released up to 84% of the codeposited tritium. The temperature rise of the codeposit on the tiles was significantly higher than that of the manufactured material. In one experiment, the codeposit surface temperature rose to 1,770 C while for the same conditions, the manufactured surface increased to only 1,080 C. The peak temperature did not follow the usual square-root dependence on heat pulse duration. Durations of order 100 ms resulted in brittle destruction and material loss from the surface, while a duration of approximately 10 ms showed minimal change. A digital microscope imaged the codeposit before, during, and after the interaction with the laser and revealed hot spots on a 100-micron scale. These results will be compared to analytic modeling and are relevant to the response of plasma facing components to disruptions and vertical displacement events (VDEs) in next-step magnetic fusion devices.
Date: January 28, 2002
Creator: Skinner, C.H.; Gentile, C.A. & Hassanein, A.
Partner: UNT Libraries Government Documents Department

Measurement of Tritium Surface Distribution on TFTR Bumper Limiter Tiles

Description: The tritium surface distribution on graphite tiles used in the Tokamak Fusion Test Reactor (TFTR) bumper limiter and exposed to TFTR deuterium-tritium (D-T) discharges from 1993 to 1997 was measured by the Tritium Imaging Plate Technique (TIPT). The TFTR bumper limiter shows both re-/co-deposition and erosion. The tritium images for all tiles measured are strongly correlated with erosion and deposition patterns, and long-term tritium retention was found in the re-/co-depositions and flakes. The CFC tiles located at erosion dominated areas clearly showed their woven structure in their tritium images owing to different erosion yields between fibers and matrix. Significantly high tritium retention was observed on all sides of the erosion tiles, indicating carbon transport via repetition of local erosion/deposition cycles.
Date: June 28, 2004
Creator: Sugiyama, K.; Tanabe, T.; Skinner, C.H. & Gentile, C.A.
Partner: UNT Libraries Government Documents Department

Tritium Removal from JET and TFTR Tiles by a Scanning Laser

Description: Fast and efficient tritium removal is needed for future D-T machines with carbon plasma-facing components. A novel method for tritium release has been demonstrated on co-deposited layers on tiles retrieved from the Tokamak Fusion Test Reactor (TFTR) and from the Joint European Torus (JET). A scanning continuous wave neodymium laser beam was focused to =100 W/mm2 and scanned at high speed over the co-deposits, heating them to temperatures =2000 C for about 10 ms in either air or argon atmospheres. Fiber optic coupling between the laser and scanner was implemented. Up to 87% of the co-deposited tritium was thermally desorbed from the JET and TFTR samples. This technique appears to be a promising in-situ method for tritium removal in a next-step D-T device as it avoids oxidation, the associated de-conditioning of the plasma-facing surfaces, and the expense of processing large quantities of tritium oxide.
Date: May 30, 2002
Creator: Skinner, C.H.; Bekris, N.; Coad, J.P.; Gentile, C.A. & Glugla, M.
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

Comparison and Evaluation of Various Tritium Decontamination Techniques and Processes

Description: In support of fusion energy development, various techniques and processes have been developed over the past two decades for the removal and decontamination of tritium from a variety of items, surfaces, and components. Tritium decontamination, by chemical, physical, mechanical, or a combination of these methods, is driven by two underlying motivational forces. The first of these motivational forces is safety. Safety is paramount to the established culture associated with fusion energy. The second of these motivational forces is cost. In all aspects, less tritium contamination equals lower operational and disposal costs. This paper will discuss and evaluate the various processes employed for tritium removal and decontamination.
Date: September 10, 2004
Creator: Gentile, C.A.; Langish, S.W.; Skinner, C.H. & Ciebiera, L.P.
Partner: UNT Libraries Government Documents Department

Plasma Fueling, Pumping, and Tritium Handling Considerations for FIRE

Description: Tritium pellet injection will be utilized on the Fusion Ignition Research Experiment (FIRE) for efficient tritium fueling and to optimize the density profile for high fusion power. Conventional pneumatic pellet injectors, coupled with a guidetube system to launch pellets into the plasma from the high, field side, low field side, and vertically, will be provided for fueling along with gas puffing for plasma edge density control. About 0.1 g of tritium must be injected during each 10-s pulse. The tritium and deuterium will be exhausted into the divertor. The double null divertor will have 16 cryogenic pumps located near the divertor chamber to provide the required high pumping speed of 200 torr-L/s.
Date: November 13, 1999
Creator: Fisher, P.W.; Foster, C.A.; Gentile, C.A.; Gouge, M.J. & Nelson, B.E.
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

The Development of a Hibachi Window for Electron Beam Transmission in a KrF Laser

Description: In support of Inertial Fusion Energy (IFE), a 150 {micro}m thick silicon (Si) wafer coated on one side with a 1.2 {micro}m nanocrystalline diamond foil is being fabricated as an electron beam transmission (hibachi) window for use in KrF lasers. The hibachi window separates the lasing medium from the electron beam source while allowing the electron beam to pass through. The hibachi window must be capable of withstanding the challenging environment presented in the lasing chamber, which include: fluorine gas, delta pressure >2 atm at 5 Hz, and a high heat flux due to the transmission of electrons passing through the foil. Tests at NRL/Electra and at PPPL have shown that a device employing these novel components in the stated configuration provide for a robust hibachi window with structural integrity.
Date: November 7, 2003
Creator: Gentile, C.A.; Parsells, R.; Butler, J.E.; Sethian, J.D.; Ciebiera, L.; Hegeler, F. 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

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 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

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

Thermal Response of Tritiated Co-deposits from JET and TFTR to Transient Heat Pulses

Description: High heat flux interactions with plasma-facing components have been studied at microscopic scales. The beam from a continuous wave neodymium laser was scanned at high speed over the surface of graphite and carbon fiber composite tiles that had been retrieved from TFTR (Tokamak Fusion Test Reactor) and JET (Joint European Torus) after D-T plasma operations. The tiles have a surface layer of amorphous hydrogenated carbon that was co-deposited during plasma operations, and laser scanning has released more than 80% of the co-deposited tritium. The temperature rise of the co-deposit was much higher than that of the manufactured material and showed an extended time history. The peak temperature varied dramatically (e.g., 1,436 C compared to >2,300 C), indicating strong variations in the thermal conductivity to the substrate. A digital microscope imaged the co-deposit before, during, and after the interaction with the laser and revealed 100-micron scale hot spots during the interaction. Heat pulse durations of order 100 ms resulted in brittle destruction and material loss from the surface, whilst a duration of =10 ms showed minimal changes to the co-deposit. These results show that reliable predictions for the response of deposition areas to off-normal events such as ELMs (edge-localized modes) and disruptions in next-step devices need to be based on experiments with tokamak generated co-deposits.
Date: May 30, 2002
Creator: Skinner, C.H.; Bekrisl, N.; Coad, J.P.; Gentile, C.A.; Hassanein, A.; Reiswig, R. 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

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: 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

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