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Proceedings of the 4th International Workshop on Tritium Effects in Plasma Facing Components

Description: The 4th International Workshop on Tritium Effects in Plasma Facing Components was held in Santa Fe, New Mexico on May 14-15, 1998. This workshop occurs every two years, and has previously been held in Livermore/California, Nagoya/Japan, and the JRC-Ispra Site in Italy. The purpose of the workshop is to gather researchers involved in the topic of tritium migration, retention, and recycling in materials used to line magnetic fusion reactor walls and provide a forum for presentation and discussions in this area. This document provides an overall summary of the workshop, the workshop agenda, a summary of the presentations, and a list of attendees.
Date: February 1, 1999
Creator: Causey, R. A.
Partner: UNT Libraries Government Documents Department

Silver transport in H-451 graphite

Description: Diffusion profiles have been measured for silver in hollow and solid cylinders in the temperature range 490 to 800/sup 0/C. The profiles show two components of diffusion. There is a high concentration, relatively immobile zone near the surface and a low concentration, rapidly moving component deeper into the sample. The rapid component is well fitted by the diffusion equation when irreversible trapping is assumed. The diffusion coefficients determined by least-square analysis of the profiles can be represented by D (m/sup 2//sec) = 17.4 exp (-22,100/T). Excessive scatter was seen in the trapping coefficients determined from the profiles. This is believed to be due to varying amounts of impurities in the different samples. A limited number of desorption measurements were performed for silver from H-451 graphite. Evidence of Elovich-type behavior was noted.
Date: January 1, 1981
Creator: Causey, R.A. & Wichner, R.P.
Partner: UNT Libraries Government Documents Department

Gas retention in irradiated beryllium

Description: Helium (an inert gas) with low solubility in beryllium is trapped in irradiated beryllium at low temperatures (<100{degree}C) while the tritium generated may have some mobility and be released. The subject of tritium retention in irradiated beryllium within fusion reactor blankets is of considerable interest in their conceptual design. Results from experiments on three sets of irradiated beryllium specimens are examined in this paper. The beryllium specimens were irradiated at abut 75{degree}C in capsules to protect them from the cooling water. One set of samples was irradiated to {approximately}3 {times} 10{sup 22} n/cm{sup 2} (E > 1 MeV). In these samples the calculated helium generated was {approximately} 14,000 appm. They are described in terms of swelling, annealing, microstructure, and helium bubble behavior (size, density and mobility). A second sample was irradiated to {approximately}5 {times} 10{sup 22} n/cm{sup 2} (E > 1 MeV). In that one the calculated helium and tritium generated were {approximately}24,000 appm He and {approximately}3720 appm, and tritium content was examined in a dissolution experiment. Most of the tritium was released as gas to the glovebox indicating the generated tritium was retained in the helium bubbles. In a third set of experiments a specimen was examined by annealing at a succession of temperatures to more than 600{degree}C for tritium release. In the temperature range of 300--500{degree}C little release (0.01--0.4%) occurred, but there was a massive release at just over 600{degree}C. Theories of swelling appear to adequately describe bubble behavior with breakaway release occurring at high helium contents and at large bubble diameters. 8 refs., 6 figs.
Date: June 1, 1990
Creator: Beeston, J.M.; Miller, L.G.; Longhurst, G.R. (EG and G Idaho, Inc., Idaho Falls, ID (USA)) & Causey, R.A. (Sandia National Labs., Albuquerque, NM (USA))
Partner: UNT Libraries Government Documents Department

Erosion and redeposition experiments in the PISCES facility

Description: The modification of surfaces during exposure to plasma bombardment is a critical issue in the development of limiter and wall materials for fusion confinement experiments. Controlled studies of the erosion and redeposition of materials during high flux and fluence plasma exposure are now possible in the PISCES facility. PISCES is a continuously operating plasma device which has achieved hydrogen plasma densities of over 10/sup 13/ cm/sup -3/ and electron temperatures of 5 to 24 eV over large areas. Ion fluxes of 10/sup 17/ to 10/sup 19/ cm/sup -2/ sec/sup -1/ and fluences of up to 10/sup 23/ cm/sup -2/ have been used to bombard biased samples inserted into the plasma. The plasma parameters can be selected to produce simple sputtering, or redeposition by the ionization and recycling of the sputtered target materials. Collaborative studies on the performance of Cu and Cu-Li alloys (with ANL), stainless steel (with SNLL), and graphite (with IPP at Garching, and SNLL) have been undertaken. Surface topography modification is always observed after a sufficient fluence is achieved. The net erosion rate is significantly lower during redeposition than one would expect from classical sputtering yields. The transport and deposition of different materials by the plasma to the samples during redeposition conditions results in greatly modified surface composition and morphology. Chemical sputtering of graphite during low energy, high flux (>10/sup 18/ cm/sup -2/ sec/sup -1/) plasma bombardment is observed. Chemically formed hydrocarbons are relatively easily redeposited compared to sputtered carbon. The performance of these materials, the surface morphology evolution, and the characteristics of the redeposited materials are discussed.
Date: May 1, 1986
Creator: Goebel, D.M.; Hirooka, Y.; Conn, R.W.; Leung, W.K.; Campbell, G.A.; Bohdansky, J. et al.
Partner: UNT Libraries Government Documents Department

Materials surface modification by plasma bombardment under simultaneous erosion and redeposition conditions

Description: The first in-depth investigation of surface modification of materials by continuous, high-flux argon plasma bombardment under simultaneous erosion and redeposition conditions have been carried out for copper and 304 stainless steel using the PISCES facility. The plasma bombardment conditions are: incident ion flux range from 10/sup 17/ to 10/sup 19/ ions sec/sup -1/cm/sup -2/, total ion fluence is controlled between 10/sup 19/ and 10/sup 22/ ions cm/sup -2/, electron temperature range from 5 to 15 eV, and plasma density range from 10/sup 11/ to 10/sup 13/cm/sup -3/. The incident ion energy is 100 eV. The sample temperature is between 300 and 700K. Under redeposition dominated conditions, the material erosion rate due to the plasma bombardment is significantly smaller (by a factor up to 10) than that can be expected from the classical ion beam sputtering yield data. It is found that surface morphologies of redeposited materials strongly depend on the plasma bombardment condition. The effect of impurities on surface morphology is elucidated in detail. First-order modelings are implemented to interpret the reduced erosion rate and the surface evolution. Also, fusion related surface properties of redeposited materials such as hydrogen reemission and plasma driven permeation have been characterized.
Date: July 1, 1986
Creator: Hirooka, Y.; Goebel, D.M.; Conn, R.W.; Campbell, G.A.; Leung, W.K.; Wilson, K.L. et al.
Partner: UNT Libraries Government Documents Department

Recent Advances on Hydrogenic Retention in ITER&#x27;s Plasma-Facing Materials: BE, C, W.

Description: Management of tritium inventory remains one of the grand challenges in the development of fusion energy and the choice of plasma-facing materials is a key factor for in-vessel tritium retention. The Atomic and Molecular Data Unit of the International Atomic Energy Agency organized a Coordinated Research Project (CRP) on the overall topic of tritium inventory in fusion reactors during the period 2001-2006. This dealt with hydrogenic retention in ITER&#x27;s plasma-facing materials, Be, C, W, and in compounds (mixed materials) of these elements as well as tritium removal techniques. The results of the CRP are summarized in this article together with recommendations for ITER. Basic parameters of diffusivity, solubility and trapping in Be, C and W are reviewed. For Be, the development of open porosity can account for transient hydrogenic pumping but long term retention will be dominated by codeposition. Codeposition is also the dominant retention mechanism for carbon and remains a serious concern for both Be and C containing layers. Hydrogenic trapping in unirradiated tungsten is low but will increase with ion and neutron damage. Mixed materials will be formed in a tokamak and these can also retain significant amounts of hydrogen isotopes. Oxidative and photon-based techniques for detritiation of plasma-facing components are described.
Date: March 29, 2008
Creator: Skinner, C H; Alimov, Kh; Bekris, N; Causey, R A; Clark, R.E.H.; Coad, J P 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