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Design of the ISX-B differentially pumped bellows

Description: The structural design and operational aspects of the ISX-B differentially pumped bellows assembly are discussed. The ISX-B vacuum vessel consists of two monolithic halves which are rectangular in cross section and joined together by short bellows which provide electrical resistance around the vesel in the toroidal direction. The bellows are unable to withstand atmospheric pressure and must be mounted within an outer, thick-wall vessel with the space between bellows and wall being pumped down to a partial vacuum.
Date: January 1, 1977
Creator: Goranson, P.L.
Partner: UNT Libraries Government Documents Department

Design of the ICRH antenna for TPX

Description: A 6-MW ion cyclotron (IC) system for the Tokamak Physics Experiment (TPX) is in the preliminary design phase. In conjunction with the 3-MW Lower Hybrid system and the 8-MW neutral beam system, the IC system will provide heating and current-drive capabilities to explore advanced tokamak physics and long-pulse (1000 s) operation. The IC launcher consists of six nickel-plated current straps arranged toroidally in pairs behind three water-cooled Faraday shields. The Faraday shields can be independently mid remotely detached by cutting water lines at the back of the launcher and removing bolts at the front to free each shield. The antenna can be located at the +2 cm flux line and retracted 10 cm. Faraday shields are usually copper- or nickel-plated stainless steel or inconel. Titanium is the preferred material to minimize activation without greatly decreasing electrical resistivity and therefore increasing disruption loads. The IC antenna research and development programs have provided data that confirm the feasibility of B{sub 4}C-coated nickel-plated titanium alloy in the TPX environment.
Date: January 1, 1996
Creator: Fogelman, C.H.; Goranson, P.L. & Swain, D.W.
Partner: UNT Libraries Government Documents Department

Heating and current drive systems for TPX

Description: The heating and current drive (H and CD) system proposed for the TPX tokamak will consist of ion cyclotron, neutral beam, and lower hybrid systems. It will have 17.5 MW of installed H and CD power initially, and can be upgraded to 45 MW. It will be used to explore advanced confinement and fully current-driven plasma regimes with pulse lengths of up to 1,000 s.
Date: May 24, 1994
Creator: Swain, D.; Goranson, P.; Halle, A. von; Bernabei, S. & Greenough, N.
Partner: UNT Libraries Government Documents Department

A modified lower hybrid coupler for TPX

Description: Efforts have concentrated on redesigning the configuration of the Lower Hybrid coupler for TPX tokamak. Several concerns motivated this redesign: reduce the effect of thermal incompatibility between coupler and rf-window material, reduce weight, reduce the risk of wind failure and address the problem of replaceability, increase the reliability by reducing the number connections and finally, reduce the total cost. The result is a highly compact, light and easily serviceable coupler which incorporates some of the simplicity of the multifunction coupler but preserves the spectral flexibility of a conventional coupler.
Date: July 1, 1995
Creator: Bernabei, S.; Greenough, N.; Goranson, P. & Swain, D.
Partner: UNT Libraries Government Documents Department

Mechanical design issues associated with mounting, maintenance, and handling of an ITER divertor

Description: Several designs that address plasma-facing plate configurations and thermal-hydraulic design issues have been developed for the ITER divertor. Design criteria growing out of physics requirements, physical constraints, and remote handling requirements impose severe mechanical requirements on the support structure and its attachments. These pose a challenge to the mechanical design of a divertor, which must be addressed before a functional divertor is practical -- that is, one that can be remotely handled, aligned, and maintained; that functions reliably under thermal loading and disruptions; and that gives the required life in the nuclear environment predicted for ITER. This paper discusses the design criteria for the divertor mounting structure and identifies the mechanical design issues that need to be addressed. Achieving the criteria may require the development of new components and innovative configurations, specifically a new class of remote fasteners and electrically resistant material for mounts. The possible design of such components and an R D program to develop them are described, and issues specific to the high-aspect-ratio design (HARD) configuration are summarized. Analysis and experiments that will resolve these issues and concerns and lead to a final ITER design are identified. 2 refs., 2 figs.
Date: January 1, 1991
Creator: Goranson, P.L.; Fogarty, P.J. & Jones, G.H.
Partner: UNT Libraries Government Documents Department

Mechanical testing and development of the helical field coil joint for the Advanced Toroidal Facility

Description: The helical field (HF) coil set for the Advanced Toroidal Facility (ATF) is an M = 12, l = 2, constant-ratio torsatron winding consisting of 2 coils, each with 14 turns of heavy copper conductor. The coils are divided into 24 identical segments to facilitate fabrication and minimize the assembly schedule. The segments are connected across through-bolted lap joints that must carry up to 124,000 A per turn for 5 s or 62,500 A steady-state. In addition, the joints must carry the high magnetic and thermal loads induced in the conductor and still fit within the basic 140- by 30-mm copper envelope. Extensive testing and development were undertaken to verify and refine the basic joint design. Tests included assembly force and clamping force for various types of misalignment; joint resistance as a function of clamping force; clamp bolt relaxation due to thermal cycling; fatigue testing of full-size, multiturn joint prototypes; and low-cycle fatigue and tensile tests of annealed CDA102 copper. The required performance parameters and actual test results, as well as the final joint configuration, are presented. 2 refs., 9 figs., 4 tabs.
Date: January 1, 1985
Creator: Nelson, B.E.; Bryan, W.E.; Goranson, P.L. & Warwick, J.E.
Partner: UNT Libraries Government Documents Department

Design of an ion cyclotron resonance heating system for the Compact Ignition Tokamak

Description: The Compact Ignition Tokamak (CIT) requires 10-20 MW of ion cyclotron resonance heating (ICRH) power to raise the plasma temperature to ignition. The initial ICRH system will provide 10 MW of power to the plasma, utilizing a total of six rf power units feeding six current straps in three ports. The systems may be expanded to 20 MW with additional rf power units, antennas, and ports. Plasma heating will be achieved through coupling to the fundamental ion cyclotron resonance of a /sup 3/He minority species (also the second harmonic of tritium). The proposed antenna is a resonant double loop (RDL) structure with vacuum, shorted stubs at each end for tuning and impedance matching. The antennas are of modular, compact construction for installation and removal through the midplane port. Remote maintainability and the reactorlike operating environment have a major impact on the design of the launcher for this machine. 6 refs., 7 figs., 5 tabs.
Date: January 1, 1987
Creator: Yugo, J.J.; Goranson, P.L.; Swain, D.W.; Baity, F.W. & Vesey, R.
Partner: UNT Libraries Government Documents Department

Lower hybrid system design for the Tokamak physics experiment

Description: The lower hybrid (LH) launcher configuration has been redesigned to integrate the functions of the vertical four-way power splitter and the front waveguide array (front array). This permits 256 waveguide channels to be fed by only 64 waveguides at the vacuum window interface. The resulting configuration is a more compact coupler, which incorporates the simplicity of a multijunction coupler while preserving the spectral flexibility of a conventional lower hybrid launcher. Other spin-offs of the redesign are reduction in thermal incompatibility between the front array and vacuum windows, improved maintainability, in situ vacuum window replacement, a reduced number of radio frequency (rf) connections, and a weight reduction of 7300 kg. There should be a significant cost reduction as well. Issues associated with the launcher design and fabrication have been addressed by a research and development program that includes brazing of the front array and testing of the power splitter configuration to confirm that phase errors due to reflections in the shorted splitter legs will not significantly impact the rf spectrum. The Conceptual Design Review requires that radiation levels at the torus radial port mounting flange and outer surface of the toroidal field coils should be sufficiently low to permit hands-on maintenance. Low activation materials and neutron shielding are incorporated in the launcher design to meet these requirements. The launcher is configured to couple 3 MW of steady state LH heating/LH current drive power at 3.7 GHz to the Tokamak Physics Experiment plasma.
Date: December 31, 1995
Creator: Goranson, P.L.; Conner, D.L.; Swain, D.W.; Yugo, J.J.; Bernabei, S. & Greenough, N.
Partner: UNT Libraries Government Documents Department

NSTX High Temperature Sensor Systems

Description: The design of the more than 300 in-vessel sensor systems for the National Spherical Torus Experiment (NSTX) has encountered several challenging fusion reactor diagnostic issues involving high temperatures and space constraints. This has resulted in unique miniature, high temperature in-vessel sensor systems mounted in small spaces behind plasma facing armor tiles, and they are prototypical of possible high power reactor first-wall applications. In the Center Stack, Divertor, Passive Plate, and vessel wall regions, the small magnetic sensors, large magnetic sensors, flux loops, Rogowski Coils, thermocouples, and Langmuir Probes are qualified for 600 degrees C operation. This rating will accommodate both peak rear-face graphite tile temperatures during operations and the 350 degrees C bake-out conditions. Similar sensor systems including flux loops, on other vacuum vessel regions are qualified for 350 degrees C operation. Cabling from the sensors embedded in the graphite tiles follows narrow routes to exit the vessel. The detailed sensor design and installation methods of these diagnostic systems developed for high-powered ST operation are discussed.
Date: November 1, 1999
Creator: B.McCormack; Kugel, H.W.; Goranson, P.; Kaita, R. & al, et
Partner: UNT Libraries Government Documents Department

Evaluation of Demo 1C composite flywheel rotor burst test and containment design

Description: Laboratory-Directed funds were provided in FY 1995 for research to develop flywheel containment specifications and to consider concepts that could satisfy these specifications and produce a prototype small, lightweight, inexpensive, mobile flywheel containment. Research activities have included an analytical and pictorial review of the Demo 1C flywheel failure test, which provided significant insight about radial and axial failure modes; calculations of the thickness of ultra-conservative pressure vessel containment; entertainment of advanced containment concepts using lightweight materials and armor literature; consideration of fabrication assembly procedures; and participation in a Flywheel Energy Storage Workshop during which additional flywheel failure experiences were discussed. Based on these activities, calculations, and results, a list of conclusions concerning flywheel containment and its relation to the flywheel are presented followed by recommendations for further research.
Date: July 1, 1998
Creator: Kass, M.D.; McKeever, J.W.; Akerman, M.A.; Goranson, P.L.; Litherland, P.S. & O`Kain, D.U.
Partner: UNT Libraries Government Documents Department

ICRF antenna designs for CIT and Alcator C-Mod

Description: An ion cyclotron range of frequencies (ICRF) launcher for the Compact Ignition Tokamak (CIT) has been designed. This launcher incorporates four current straps in a 2 /times/ 2 configuration. The current straps consist of end-fed loops that are grounded in the middle. An antenna similar in geometry, size, and feed configuration to a single strap of the CIT launcher will be built for use on Alcator C-Mod. The design must provide maximum power levels of 4 MW/port for CIT and 2 MW/port for C-Mod, pulse lengths of 5--10 s for CIT and 1 s for C-Mod, and power densities up to 2 kW/cm/sup 2/. The design uses a Faraday shield consisting of Inconel rods with mechanically attached graphite tiles; the shield and the current strap are cooled by radiating to a gas-cooled backplane. A feed configuration compatible with the end-fed antenna design has been developed and features tunability in three bands in the range 65--130 MHz. It uses an external resonant loop with integral tuning elements. It has been designed to maximize power handling capabilities, minimize space requirements, and facilitate remote handling. 1 ref., 6 figs.
Date: January 1, 1989
Creator: Goulding, R.H.; Baity, F.W.; Goranson, P.L.; Hoffman, D.J.; Ryan, P.M.; Traylor, D.J. et al.
Partner: UNT Libraries Government Documents Department

Conceptual design of the tokamak radiation shielding for the Tokamak Physics Experiment (TPX)

Description: The tokamak radiation shielding includes the neutron and gamma shielding around the torus and penetrations required to (1) limit activation of components outside the shield to levels that permit hands-on maintenance and (2) limit the nuclear heating of the superconducting coils and cold structure. The primary design drivers are space, the 350{degree}C bakeout temperature, and cost; therefore, different shield materials were used for different shield components and locations. The shielding is divided into three areas: (1) torus shielding around the vacuum vessel, (2) duct shielding around the vacuum pumping ducts and vertical diagnostic ducts, and (3) penetration shielding in and around the radial ports. The major shield components include water between the walls of the vacuum vessel, lead monosilicate/boron carbide tiles that are attached to the exterior of the vacuum vessel, shield plugs that rill the openings of the large radial ports, and polyethylene/lead/boron shield blocks for duct shielding. A description of the shielding configuration and the performance and operational requirements will be discussed.
Date: November 1, 1993
Creator: Cole, M. J.; Nelson, B. E.; Jones, G. H.; Goranson, P. L.; Gohar, Y. & Liew, S. L.
Partner: UNT Libraries Government Documents Department

Methodology for first wall design

Description: An analytic parametric scoping tool has been developed for application to first wall (FW) design problems. Both thermal and disruption force effects are considered. For the high heat flux and high disruption load conditions expected in the International Thermonuclear Experimental Reactor (ITER) device, Vanadium alloy and dispersion-strengthened copper offer the best stress margins using a somewhat flattened plasma-facing configuration. Ferritic steels also appear to have an acceptable stress margin, whereas the conventional stainless steel 316 does not appear feasible. If a full semicircle shape FW is required, only the Vanadium and ferritic steel alloy have acceptable solutions.
Date: November 1, 1993
Creator: Galambos, J. D.; Conner, D. L.; Goranson, P. L.; Lousteau, D. C.; Williamson, D. E.; Nelson, B. E. et al.
Partner: UNT Libraries Government Documents Department

NCSX Vacuum Vessel Fabrication

Description: The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in conjunction with the Oak Ridge National Laboratory (ORNL). The goal of this experiment is to develop a device which has the steady state properties of a traditional stellarator along with the high performance characteristics of a tokamak. A key element of this device is its highly shaped Inconel 625 vacuum vessel. This paper describes the manufacturing of the vessel. The vessel is being fabricated by Major Tool and Machine, Inc. (MTM) in three identical 120º vessel segments, corresponding to the three NCSX field periods, in order to accommodate assembly of the device. The port extensions are welded on, leak checked, cut off within 1" of the vessel surface at MTM and then reattached at PPPL, to accommodate assembly of the close-fitting modular coils that surround the vessel. The 120º vessel segments are formed by welding two 60º segments together. Each 60º segment is fabricated by welding ten press-formed panels together over a collapsible welding fixture which is needed to precisely position the panels. The vessel is joined at assembly by welding via custom machined 8" (20.3 cm) wide spacer "spool pieces." The vessel must have a total leak rate less than 5 X 10-6 t-l/s, magnetic permeability less than 1.02μ, and its contours must be within 0.188" (4.76 mm). It is scheduled for completion in January 2006.
Date: October 7, 2005
Creator: Viola, M. E.; Brown, T.; Heitzenroeder, P.; Malinowski, F.; Reiersen, W.; Sutton, L. et al.
Partner: UNT Libraries Government Documents Department

Progress In NCSX and QPS Design and Construction

Description: The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge National Laboratory (ORNL). The stellarator core is designed to produce a compact 3-D plasma that combines stellarator and tokamak physics advantages. The engineering challenges of NCSX stem from its complex geometry. From the project's start in April, 2003 to September, 2004, the fabrication specifications for the project's two long-lead components, the modular coil winding forms and the vacuum vessel, were developed. An industrial manufacturing R&D program refined the processes for their fabrication as well as production cost and schedule estimates. The project passed a series of reviews and established its performance baseline with the Department of Energy. In September 2004, fabrication was approved and contracts for these components were awarded. The suppliers have completed the engineering and tooling preparations and are in production. Meanwhile, the project completed preparations for winding the coils at PPPL by installing a coil manufacturing facility and developing all necessary processes through R&D. The main activities for the next two years will be component manufacture, coil winding, and sub-assembly of the vacuum vessel and coil subsets. Machine sector sub-assembly, machine assembly, and testing will follow, leading to First Plasma in July 2009.
Date: October 20, 2005
Creator: Reiersen, W.; Heitzenroeder, P.; Neilson, G. H.; Nelson, B.; Zarnstorff, M.; Brown, T. et al.
Partner: UNT Libraries Government Documents Department

Component Manufacturing Development for the National Compact Stellarator Experiment (NCSX)

Description: NCSX [National Compact Stellarator Experiment] is the first of a new class of stellarators called compact stellarators which hold the promise of retaining the steady state feature of the stellarator but at a much lower aspect ratio and using a quasi-axisymmetric magnetic field to obtain tokamak-like performance. Although much of NCSX is conventional in design and construction, the vacuum vessel and modular coils provide significant engineering challenges due to their complex shapes, need for high dimensional accuracy, and the high current density required in the modular coils due space constraints. Consequently, a three-phase development program has been undertaken. In the first phase, laboratory/industrial studies were performed during the development of the conceptual design to permit advances in manufacturing technology to be incorporated into NCSX's plans. In the second phase, full-scale prototype modular coil winding forms, compacted cable conductors, and 20 degree sectors of the vacuum vessel were fabricated in industry. In parallel, the NCSX project team undertook R&D studies that focused on the windings. The third (production) phase began in September 2004. First plasma is scheduled for January 2008.
Date: October 28, 2004
Creator: Heitzenroeder, P.J.; Brown, T.G.; Chrzanowski, J.H.; Cole, M.J.; Goranson, P.L.; Neilson, G.H. et al.
Partner: UNT Libraries Government Documents Department

Diagnostics for the National Compact Stellarator Experiment

Description: The status of planning of the National Compact Stellarator Experiment (NCSX) diagnostics is presented, with the emphasis on resolution of diagnostics access issues and on diagnostics required for the early phases of operation.
Date: September 16, 2003
Creator: Stratton, B.C.; Johnson, D.; Feder, R.; Fredrickson, E.; Neilson, H.; Takahashi, H. et al.
Partner: UNT Libraries Government Documents Department

NCSX Construction Progress and Research Plans

Description: Stellarators use 3D plasma and magnetic field shaping to produce a steady-state disruption-free magnetic confinement configuration. Compact stellarators have additional attractive properties — quasi-symmetric magnetic fields and low aspect ratio. The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge National Laboratory (ORNL) to test the physics of a high-beta compact stellarator with a lowripple, tokamak-like magnetic configuration. The engineering challenges of NCSX stem from its complex geometry requirements. These issues are addressed in the construction project through manufacturing R&D and system engineering. As a result, the fabrication of the coil winding forms and vacuum vessel are proceeding in industry without significant technical issues, and preparations for winding the coils at PPPL are in place. Design integration, analysis, and dimensional control are functions provided by system engineering to ensure that the finished product will satisfy the physics requirements, especially accurate realization of the specified coil geometries. After completion of construction in 2009, a research program to test the expected physics benefits will start.
Date: September 26, 2005
Creator: Neilson, G. H.; Heitzenroeder, P.; Lyon, J.; Nelson, B.; Reiersen, W.; Zarnstorff, M. et al.
Partner: UNT Libraries Government Documents Department

Thermal-hydraulic design issues and analysis for the ITER (International Thermonuclear Experimental Reactor) divertor

Description: Critical Heat Flux (CHF), also called burnout, is one of the major design limits for water-cooled divertors in tokamaks. Another important design issue is the correct thermal modeling of the divertor plate geometry where heat is applied to only one side of the plate and highly subcooled flow boiling in internal passages is used for heat removal. This paper discusses analytical techniques developed to address these design issues, and the experimental evidence gathered in support of the approach. Typical water-cooled divertor designs for the International Thermonuclear Experimental Reactor (ITER) are analyzed, and design margins estimated. Peaking of the heat flux at the tube-water boundary is shown to be an important issue, and design concerns which could lead to imposing large design safety margins are identified. The use of flow enhancement techniques such as internal twisted tapes and fins are discussed, and some estimates of the gains in the design margin are presented. Finally, unresolved issues and concerns regarding hydraulic design of divertors are summarized, and some experiments which could help the ITER final design process identified. 23 refs., 10 figs.
Date: January 1, 1990
Creator: Koski, J.A.; Watson, R.D. (Sandia National Labs., Albuquerque, NM (USA)); Hassanien, A.M. (Argonne National Lab., IL (USA)); Goranson, P.L. (Oak Ridge National Lab., TN (USA). Fusion Engineering Design Center) & Salmonson, J.C. (EG and G Energy Measurements Group, Inc., Albuquerque, NM (USA). Special Projects)
Partner: UNT Libraries Government Documents Department

US solid breeder blanket design for ITER

Description: The US blanket design activity has focused on the developments and the analyses of a solid breeder blanket concept for ITER. The main function of this blanket is to produce the necessary tritium required for the ITER operation and the test program. Safety, power reactor relevance, low tritium inventory, and design flexibility are the main reasons for the blanket selection. The blanket is designed to operate satisfactorily in the physics and the technology phases of ITER without the need for hardware changes. Mechanical simplicity, predictability, performance, minimum cost, and minimum R D requirements are the other criteria used to guide the design process. The design aspects of the blanket are summarized in this paper. 2 refs., 7 figs., 3 tabs.
Date: September 1, 1990
Creator: Gohar, Y.; Attaya, H.; Billone, M.; Lin, C.; Johnson, C.; Majumdar, S. et al.
Partner: UNT Libraries Government Documents Department

Engineering Accomplishments in the Construction of NCSX

Description: The National Compact Stellarator Experiment (NCSX) was designed to test a compact, quasiaxisymmetric stellarator configuration. Flexibility and accurate realization of its complex 3D geometry were key requirements affecting the design and construction. While the project was terminated before completing construction, there were significant engineering accomplishments in design, fabrication, and assembly. The design of the stellarator core device was completed. All of the modular coils, toroidal field coils, and vacuum vessel sectors were fabricated. Critical assembly steps were demonstrated. Engineering advances were made in the application of CAD modeling, structural analysis, and accurate fabrication of complex-shaped components and subassemblies. The engineering accomplishments of the project are summarized
Date: September 1, 2008
Creator: Neilson, G. H.; Heitzenroeder, P.J.; Nelson, B.E.; Reiersen, W.T.; Brooks, A.; Brown, T.G. et al.
Partner: UNT Libraries Government Documents Department