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US blanket technology programs. [Directory of current research]

Description: Experimental research in US programs related to blanket technology is described through brief summaries of the objectives, facilities, recent experimental results and principal investigators for the Blanket Technology Program, TRIO-1 Experiment, TSTA, Fusion Hybrid Program and selected activities in the Fusion Materials and Fusion Safety Programs in neutronics research.
Date: January 1, 1985
Creator: Nygren, R.E.
Partner: UNT Libraries Government Documents Department

First Wall, Blanket, Shield Engineering Technology Program

Description: The First Wall/Blanket/Shield Engineering Technology Program sponsored by the Office of Fusion Energy of DOE has the overall objective of providing engineering data that will define performance parameters for nuclear systems in advanced fusion reactors. The program comprises testing and the development of computational tools in four areas: (1) thermomechanical and thermal-hydraulic performance of first-wall component facsimiles with emphasis on surface heat loads; (2) thermomechanical and thermal-hydraulic performance of blanket and shield component facsimiles with emphasis on bulk heating; (3) electromagnetic effects in first wall, blanket, and shield component facsimiles with emphasis on transient field penetration and eddy-current effects; (4) assembly, maintenance and repair with emphasis on remote-handling techniques. This paper will focus on elements 2 and 4 above and, in keeping with the conference participation from both fusion and fission programs, will emphasize potential interfaces between fusion technology and experience in the fission industry.
Date: January 1, 1982
Creator: Nygren, R.E.
Partner: UNT Libraries Government Documents Department

Qualitative Reliability Issues for In-Vessel Solid and Liquid Wall Fusion Designs

Description: This paper presents the results of a study of the qualitative aspects of plasma facing component (PFC) reliability for actively cooled solid wall and liquid wall concepts for magnetic fusion reactor vessels. These two designs have been analyzed for component failure modes. The most important results of that study are given here. A brief discussion of reliability growth in design is included to illustrate how solid wall designs have begun as workable designs and have evolved over time to become more optimized designs with better longevity. The increase in tolerable heat fluxes shows the improvement. Liquid walls could also have reliability growth if the designs had similar development efforts.
Date: October 1, 2001
Creator: Cadwallader, Lee Charles & Nygren, R. E.
Partner: UNT Libraries Government Documents Department

A comparison of stresses in armor joints with and without interlayers

Description: Reliable joining of armor to heat sinks for plasma facing components has been a persistent problem in fusion and a concern for the International Thermonuclear Experimental Reactor (ITER). Post-fabrication and operating stresses in heat sinks with a 1mm compliant layer (or no interlayer) between tungsten armor and a CuCrZr channel were analyzed with a 2-D finite element model with temperature dependent properties, generalized plane strain, and strain hardening.
Date: November 1, 1997
Creator: Nygren, R.E.
Partner: UNT Libraries Government Documents Department

In situ spectroscopic measurements of erosion behavior of TFTR-redeposited carbon materials under high-flux plasma bombardment in PISCES-A

Description: The chemical erosion behavior of graphite materials pre-exposed in Tokamak Fusion Test Reactor (TFTR) as the bumper limiter has been investigated spectroscopically under deuterium plasma bombardment in the PISCES-A facility. The deuterium plasma bombardment conditions are: ion bombarding energy of 300 eV; ion flux of 1.7 /times/ 10/sup 18/ ions s/sup /minus/1/ cm/sup /minus/2/; plasma density of 1.4 /times/ 10/sup 12/ cm/sup /minus/3/; electron temperature of 11 eV; and neutral pressure of 3 /times/ 10/sup /minus/4/ torr. The chemical erosion yield is measured with CD-band spectroscopy during the temperature ramp from 100 to 900/degree/C at an average rate of about 5 degrees/s. The materials used include virgin POCO AFX-5Q graphite, graphite tile pieces from the redeposition-dominated and erosion-dominated areas of the bumper limiter in TFTR. It has been found in common for these graphite materials that the chemical erosion yield maximizes at a temperature around 550/degree/C. However, graphite from the redeposited area has shown a somewhat higher maximum erosion yield and significantly steeper temperature dependence. In addition, the removability of the redeposited materials by helium plasma bombardment has been studied. The removal rate is found to be similar to the physical sputtering yield of carbon by helium. The surface morphology and surface composition has been analyzed with SEM and EMPA in parallel with these erosion behavior measurements. 38 refs., 5 figs., 1 tab.
Date: August 1, 1988
Creator: Hirooka, Y.; Pospieszczyk, A.; Conn, R.W.; Labombard, B.; Mills, B.; Nygren, R.E. et al.
Partner: UNT Libraries Government Documents Department

Initial report on calorimetry for the Tore Supra Outboard pump Limiter

Description: This report describes the instrumentation locations of the Tore Supra Phase III Outboard Limiter, including the locations and signal names of the flowmeters and thermocouples. Shot 11044 was evaluated in some detail. The heat loads in the fourteen cooling tubes that form the limiter head were calculated from the data and the results compared with the heat loads predicted using a 3-D model heat transfer calculation that calculates the distribution of power on the limiter based upon the power scrape-off length, the mag magnetic configuration and the shape of the limiter.
Date: January 1, 1994
Creator: Nygren, R. E.; Lutz, T. J. & Miller, J. D.
Partner: UNT Libraries Government Documents Department

Erosion and redeposition behavior of selected net-candidate materials under high-flux hydrogen, deuterium plasma bombardment in PISCES

Description: Plasma erosion and redeposition behavior of selected candidate materials for plasma-facing components in the NET-machine have been investigated using the PISCES-A facility. Materials studied include SiC-impregnated graphite, 2D graphite weaves with and without CVD- SiC coatings, and isotropic graphite. These specimens were exposed to continuous hydrogen or deuterium plasmas under the following conditions: electron temperature range from 5 to 35eV; plasma density range from 5 x 10/sup 11/ to 1 x 10/sup 12/ cm/sup -3/; flux range from 5 x 10/sup 17/ to 2 x 10/sup 18/ ions cm/sup -2/ s/sup -1/; fluence of the order from 10/sup 21/ to 10/sup 22/ ions/cm/sup 2/; bombarding energies of 50 and 100eV; target temperature range from 300 to 1000/degree/C. The erosion yield of SiC-impregnated graphite due to deuterium plasma bombardment is found to be a factor of 2 to 3 less than that of isotropic graphite materials. A further factor of 2-3 reduction in the erosion yield is observed in when redeposition associated with reionization of sputtered particle becomes significant. From postbombardment surface analysis with AES, the surface composition in terms of the Si/C of SiC-impregnated graphite ratio is found to increase from 0.15 to 0.7 after hydrogen plasma bombardment to a fluence around 4 x 10/sup 21/ ions/cm/sup 2/ at 350/degree/C. However, the final surface composition appears to remain unchanged up to 4 x 10/sup 22/ ions/cm/sup 2/, the highest fluence in the present study. Significant surface morphological modifications of SiC-impregnated graphite are observed after the high-fluence plasma exposure. Several structural problems such as coating-substrate adhesion have been pointed out for SiC-coated 2D graphite weave. 11 refs., 6 figs., 1 tab.
Date: June 1, 1988
Creator: Franconi, E.; Hirooka, Y.; Conn, R.W.; Leung, W.K.; LaBombard, B. & Nygren, R.E.
Partner: UNT Libraries Government Documents Department

Assessing braze quality in the actively cooled Tore Supra Phase III outboard pump limiter

Description: The quality of brazing of pyrolytic graphite armor brazed to copper tubes in Tore Supra`s Phase III Outboard Pump Limiter was assessed through pre-service qualification testing of individual copper/tile assemblies. The evaluation used non-destructive, hot water transient heating tests performed in the high-temperature, high-pressure flow loop at Sandia`s Plasma Materials Test Facility. Surface temperatures of tiles were monitored with an infrared camera as water at 120{degrees}C at about 2.07 MPa (300 psi) passed through a tube assembly initially at 30{degrees}C. For tiles with braze voids or cracks, the surface temperatures tagged behind those of adjacent well-bonded tiles. Temperature tags were correlated with flaw sizes observed during repairs based upon a detailed 2-D heat transfer analyses. {open_quotes}Bad{close_quotes} tiles, i.e., temperature tags of 10-20{degrees}C depending upon tile`s size, were easy to detect and, when removed, revealed braze voids of roughly 50% of the joint area. Eleven of the 14 tubes were rebrazed after bad tiles were detected and removed. Three tubes were rebrazed twice.
Date: December 31, 1994
Creator: Nygren, R.E.; Lutz, T.L.; Miller, J.D.; McGrath, R. & Dale, G.
Partner: UNT Libraries Government Documents Department

Deuterium pumping and erosion behavior of selected graphite materials under high flux plasma bombardment in PISCES

Description: Deuterium plasma recycling and chemical erosion behavior of selected graphite materials have been investigated using the PISCES-A facility. These materials include: Pyro-graphite; 2D-graphite weave; 4D-graphite weave; and POCO-graphite. Deuterium plasma bombardment conditions are: fluxes around 7 /times/ 10/sup 17/ ions s/sup /minus/1/cm/sup /minus/2/; exposure time in the range from 10 to 100 s; bombarding energy of 300 eV; and graphite temperatures between 20 and 120/degree/C. To reduce deuterium plasma recycling, several approaches have been investigated. Erosion due to high-fluence helium plasma conditioning significantly increases the surface porosity of POCO-graphite and 4D-graphite weave whereas little change for 2D-graphite weave and Pyro-graphite. The increased pore openings and refreshed in-pore surface sites are found to reduce the deuterium plasma recycling and chemical erosion rates at transient stages. The steady state recycling rates for these graphite materials can be also correlated to the surface porosity. Surface topographical modification by machined-grooves noticeably reduces the steady state deuterium recycling rate and the impurity emission from the surface. These surface topography effects are attributed to co-deposition of remitted deuterium, chemically sputtered hydrocarbon and physically sputtered carbon under deuterium plasma bombardment. The co-deposited film is found to have a characteristic surface morphology with dendritic microstructures. 18 ref., 4 figs., 1 tab.
Date: June 1, 1988
Creator: Hirooka, Y.; Conn, R.W.; Goebel, D.M.; LaBombard, B.; Lehmer, R.; Leung, W.K. et al.
Partner: UNT Libraries Government Documents Department

Brazing of the Tore Supra actively cooled Phase III Limiter

Description: The head of the water-cooled Tore Supra Phase 3 Limiter is a bank of 14 round OFHC copper tubes, curved to fit the plasma radius, onto which several hundred pyrolytic graphite (PG) tiles and a lesser number of carbon fiber composite tiles are brazed. The small allowable tolerances for fitting the tiles to the tubes and mating of compound curvatures made the brazing and fabrication extremely challenging. The paper describes the fabrication process with emphasis on the procedure for brazing. In the fixturing for vacuum furnace brazing, the tiles were each independently clamped to the tube with an elaborate set of window frame clamps. Braze quality was evaluated with transient heating tests. Some rebrazing was necessary.
Date: December 31, 1993
Creator: Nygren, R. E.; Walker, C. A.; Lutz, T. J.; Hosking, F. M. & McGrath, R. T.
Partner: UNT Libraries Government Documents Department

Presheath profiles in simulated tokamak edge plasmas

Description: The PISCES plasma surface interaction facility at UCLA generates plasmas with characteristics similar to those found in the edge plasmas of tokamaks. Steady state magnetized plasmas produced by this device are used to study plasma-wall interaction phenomena which are relevant to tokamak devices. We report here progress on some detailed investigations of the presheath region that extends from a wall surface into these /open quotes/simulated tokamak/close quotes/ edge plasma discharges along magnetic field lines.
Date: April 1, 1988
Creator: LaBombard, B.; Conn, R.W.; Hirooka, Y.; Lehmer, R.; Leung, W.K.; Nygren, R.E. et al.
Partner: UNT Libraries Government Documents Department

Particle exhaust of helium plasmas with actively cooled outboard pump limiter on Tore Supra

Description: The superconducting tokamak Tore Supra was designed for long-pulse (30-s) high input power operation. Here observations on the particle-handling characteristics of the actively cooled modular outboard pump limiter (OPL) are presented for helium discharges. The important experimental result was that a modest pumping speed (1 m{sup 3}/s) of the OPL turbomolecular pump (TMP) provided background helium exhaust. This result came about due to a well-conditioned vessel wall with helium discharges that caused no wall outgasing. The particle accountability in these helium discharges was excellent, and the well-conditioned wall did not play a significant role in the particle balance. The helium density control, 25% density drop with OPL exhaust efficiency of {approximately}1%, was possible with TMP although this may not be the case with reactive gases such as deuterium. The observed quadratic increase of the OPL neutral pressure with helium density was consistent with an improvement of the particle control with increasing plasma density.
Date: August 1, 1995
Creator: Uckan, T.; Mioduszewski, P.K.; Loarer, T.; Chatelier, M.; Guilhem, D.; Lutz, T. et al.
Partner: UNT Libraries Government Documents Department

Engineering testing requirements in FED/INTOR

Description: The FED/INTOR critical issues activity has addressed three key testing requirements that have the largest impact on the design, operation and cost of FED/INTOR. These are: (1) the total testing time (fluence) during the device lifetime, (2) the minimum number of back-to-back cycles, and (3) the neutron wall load (power density in the first wall/blanket). The testing program activities were structured into three tasks in order to define the benefits, and in some cases, costs and risks of these testing requirements. The three tasks were carried out with wide participation of experts from a number of organizations in the United States. Similar effort was performed by Japan, the European Community and the Soviet Union.
Date: October 1, 1982
Creator: Abdou, M. A.; Nygren, R. E.; Morgan, G. D.; Trachsel, C. A.; Wire, G.; Oppermann, E. et al.
Partner: UNT Libraries Government Documents Department

Summary of results from the TEXTOR helium self-pumping experiment

Description: Helium removal experiments were conducted in TEXTOR with a small helium self-pumping module located in a modified ALT-I limiter head. The module contained two heated nickel alloy trapping plates, a nickel deposition filament array, a Langmuir probe, flux probe, and thermocouples. The experiment examined plasma helium removal via trapping of helium ions in the deposited nickel surfaces. Such helium removal was successfully observed, with about 10% of the helium He/D plasma being removed in a {approximately}1 s period. The module was found to be compatible with overall tokamak operation with essentially no sputtered nickel entering the core plasma. The temperature rise on the ion-exposed inner trapping plate, during a plasma shot, is consistent with a nickel a local sheath potential of {approximately}3 kT{sub e}. Post-tokamak test examination of the trapping plates shows helium atom concentrations in the deposited nickel consistent with the observed helium removal, and shows very small D concentrations.
Date: March 1, 1992
Creator: Brooks, J. N.; Krauss, A.; Nygren, R. E.; Doyle, B. L.; Dippel, K. H. & Finken, K. H.
Partner: UNT Libraries Government Documents Department

Design Integration of Liquid Surface Divertors

Description: The US Enabling Technology Program in fusion is investigating the use of free flowing liquid surfaces facing the plasma. We have been studying the issues in integrating a liquid surface divertor into a configuration based upon an advanced tokamak, specifically the ARIES-RS configuration. The simplest form of such a divertor is to extend the flow of the liquid first wall into the divertor and thereby avoid introducing additional fluid streams. In this case, one can modify the flow above the divertor to enhance thermal mixing. For divertors with flowing liquid metals (or other electrically conductive fluids) MHD (magneto-hydrodynamics) effects are a major concern and can produce forces that redirect flow and suppress turbulence. An evaluation of Flibe (a molten salt) as a working fluid was done to assess a case in which the MHD forces could be largely neglected. Initial studies indicate that, for a tokamak with high power density, an integrated Flibe first wall and divertor does not seem workable. We have continued work with molten salts and replaced Flibe with Flinabe, a mixture of lithium and sodium fluorides, that has some potential because of its lower melting temperature. Sn and Sn-Li have also been considered, and the initial evaluations on heat removal with minimal plasma contamination show promise, although the complicated 3-D MHD flows cannot yet be fully modeled. Particle pumping in these design concepts is accomplished by conventional means (ports and pumps). However, trapping of hydrogen in these flowing liquids seems plausible and novel concepts for entrapping helium are also being studied.
Date: November 13, 2003
Creator: Nygren, R E; Cowgill, D F; Ulrickson, M A; Nelson, B E; Fogarty, P J; Rognlien, T D et al.
Partner: UNT Libraries Government Documents Department

The TEXTOR helium self-pumping experiment: Design, plans, and supporting ion-beam data on helium retention in nickel

Description: A proof-of-principle experiment to demonstrate helium self-pumping in a tokamak is being undertaken in TEXTOR. The experiment will use a helium self-pumping module installed in a modified ALT-I limiter head. The module consists of two, {approximately}25 {times} 25 cm{sup 2} heated nickel alloy trapping plates, a nickel deposition filament array, and associated diagnostics. Between plasma shots a coating of {approximately}50 {angstrom} nickel will be deposited on the two trapping plates. During a shot helium and hydrogen ions will impinge on the plates through a {approximately}3 cm wide entrance slot. The helium removal capability, due to trapping in the nickel, will be assessed for a variety of plasma conditions. In support of the tokamak experiment, the trapping of helium over a range of ion fluences and surface temperatures, and detrapping during subsequent exposure to hydrogen, were measured in ion beam experiments using evaporated nickel surfaces similar to that expected in TEXTOR. Also, the retention of H and He after exposure of a nickel surface to mixed He/H plasmas has bee measured. The results appear favorable, showing high helium trapping ({approximately}10--50% He/Ni) and little or no detrapping by hydrogen. The TEXTOR experiment is planned to begin in 1991. 12 refs., 2 figs., 2 tabs.
Date: January 1, 1990
Creator: Brooks, J. N.; Krauss, A.; Mattas, R. F.; Smith, D. L.; Nygren, R. E.; Doyle, B. L. et al.
Partner: UNT Libraries Government Documents Department

A Fusion Reactor Design with a Liquid First Wall and Divertor

Description: Within the magnetic fusion energy program in the US, a program called APEX is investigating the use of free flowing liquid surfaces to form the inner surface of the chamber around the plasma. As part of this work, the APEX Team has investigated several possible design implementations and developed a specific engineering concept for a fusion reactor with liquid walls. Our approach has been to utilize an already established design for a future fusion reactor, the ARIES-RS, for the basic chamber geometry and magnetic configuration and to replace the chamber technology in this design with liquid wall technology for a first wall and divertor and a blanket with adequate tritium breeding. This paper gives an overview of one design with a molten salt (a mixture of lithium, beryllium and sodium fluorides) forming the liquid surfaces and a ferritic steel for the structural material of the blanket. The design point is a reactor with 3840MW of fusion power of which 767MW is in the form of energetic particles (alpha power) and 3073MW is in the form of neutrons. The alpha plus auxiliary power total 909MW of which 430MW is radiated from the core mostly onto the first wall and the balance flows into the edge plasma and is distributed between the first wall and the divertor. In pursuing the application of liquid surfaces in APEX, the team has developed analytical tools that are significant achievements themselves and also pursued experiments on flowing liquids. This work is covered elsewhere, but the paper will also note several such areas to indicate the supporting science behind the design presented. Significant new work in modeling the plasma edge to understand the interaction of the plasma with the liquid walls is one example. Another is the incorporation of magneto-hydrodynamic (MHD) effects in fluid modeling and heat ...
Date: November 13, 2003
Creator: Nygren, R E; Rognlien, T D; Rensink, M E; Smolentsev, S S; Youssef, M E; Sawan, M Z et al.
Partner: UNT Libraries Government Documents Department

Physics Design Requirements for the National Spherical Torus Experiment Liquid Lithium Divertor

Description: Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on PFC's to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15-25% ne decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, ne/nGW~1), to enable ne scan capability (x2) in the H-mode, to test the ability to operate at significantly lower density for future ST-CTF reactor designs (e.g., ne/nGW = 0.25), and eventually to investigate high heat-flux power handling (10 MW/m2) with longpulse discharges (>1.5s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.
Date: September 26, 2008
Creator: Kugel, W.; Bell, M.; Berzak,L.; Brooks, A.; Ellis, R.; Gerhardt, S. et al.
Partner: UNT Libraries Government Documents Department

Helium-Cooled Refractory Alloys First Wall and Blanket Evaluation

Description: Under the APEX program the He-cooled system design task is to evaluate and recommend high power density refractory alloy first wall and blanket designs and to recommend and initiate tests to address critical issues. We completed the preliminary design of a helium-cooled, W-5Re alloy, lithium breeder design and the results are reported in this paper. Many areas of the design were assessed, including material selection, helium impurity control, and mechanical, nuclear and thermal hydraulics design, and waste disposal, tritium and safety design. System study results show that at a closed cycle gas turbine (CCGT) gross thermal efficiency of 57.5%, a superconducting coil tokamak reactor, with an aspect ratio of 4, and an output power of 2 GWe, can be projected to have a cost of electricity at 54.6 mill/kWh. Critical issues were identified and we plan to continue the design on some of the critical issues during the next phase of the APEX design study.
Date: August 1, 1999
Creator: Wong, C.P.C.; Nygren, R.E.; Baxi, C.B.; Fogarty, P.; Ghoniem, N.; Khater, H. et al.
Partner: UNT Libraries Government Documents Department

Spectroscopic studies of carbon containing molecules and their break-up in PISCES-A

Description: We have used the PISCES-A facility in order to study the behavior of carbon containing molecules in a representative plasma with parameters close to that of a tokamak boundary layer, CH{sub 4}, C{sub 2}H{sub 2}, C{sub 2}H{sub 4}, CO, and CO{sub 2} molecules were introduced through a slit aperture into a helium plasma and the radiation from these due to electronic excitation was spectrographically recorded. The imaging of the plasma onto the entrance slit of a 1.33m McPherson optical spectrometer was chosen in such a way that simultaneous information about spectral and spatial distribution of the emission could be obtained by an attached photographic camera and an optical multichannel analyser (OMA). The recorded spectra show that many features in previously obtained spectra from limiters originate -- beside from hydrocarbons -- from carbonoxides, which seem to play a major role in the transport of carbon and oxygen. It was also possible to calibrate the radiation intensity of several molecular bands versus the known molecular influx so that an absolute determination of these fluxes from the wall of a fusion device could be done. Measurements of the attenuation of the individual species were carried out, which describe the penetration of carbon, oxygen, and hydrogen atoms into a discharge by taking into account individual steps in the molecular breakup process. 36 refs., 35 figs.
Date: December 1, 1989
Creator: Pospieszczyk, A. (Association Euratom-Kernforschungsanlage Juelich (Germany, F.R.). Inst. fuer Plasmaphysik); Ra, Y.; Hirooka, Y.; Conn, R.W.; Goebel, D.M.; LaBombard, B. et al.
Partner: UNT Libraries Government Documents Department

PISCES and ALT-II: Juelich PSI papers

Description: This publication comprises papers from the PISCES and ALT-II Programs at UCLA which were presented at the International Plasma Surface Interactions Meeting held in Juelich, FRG, on May 2-6, 1988. A list of publications from the PISCES and ALT-II contained in this report are: Deuterium pumping and erosion behavior of selected graphite materials under high flux plasma bombardment in PISCES; Erosion and redeposition behavior of selected NET-candidate materials under high-flux hydrogen, deuterium plasma bombardment in PISCES; Presheath profiles in simulated tokamak edge plasmas; Boundary asymmetries and plasma flow to the ALT-II toroidal belt pump limiter; ALT-II toroidal belt pump limiter performance in TEXTOR; and An in-situ spectroscopic erosion yield measurement with applications to sputtering and surface morphology alterations.
Date: August 1, 1988
Creator: Conn, R.W.; Hirooka, Y.; LaBombard, B.; Moyer, R.; Goebel, D.M.; Leung, W.K. et al.
Partner: UNT Libraries Government Documents Department