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Fusion Engineering, 2007. SOFE 2007. 2007 IEEE 22nd Symposium on

Date 17-21 June 2007

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  • Proceedings of the 22nd IEEE/NPSS Symposium on Fusion Engineering - SOFE 07

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  • Table of contents

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  • ITER Neutronics Analysis for the Design of Diagnostics and Port Plugs Using ATTILA Discrete Ordinates Software

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    A collaboration is underway between Princeton Plasma Physics Lab and UCLA to develop skill in the use of ATTILA, to benchmark ATTILA against MCNP and to develop Solid Works CAD models for neutronics analysis of the diagnostic ports. MCNP along with the cross section library FENDL 2.1 is the accepted standard tool for neutronics analysis of ITER against which results from ATTILA are being compared. The MCNP community has established a set of benchmark results for a standardized 40 degree CAD model of ITER. These benchmark results create a framework for the acceptance of new applications like A TTILA by the ITER central neutronics, quality assurance and nuclear safety groups. Analysis of the benchmark model with ATTILA also leads to the setting of discrete ordinates solution parameters and model mesh refinement that will help to accelerate the analysis of future diagnostic port design iterations. Flux and heating results in the Divertor, Blanket Shield Modules and Equatorial Port Shielding from the ATTILA benchmarking show good correlation with MCNP results. TF heating results were in error by up to 50% due to poor mesh refinement and boundary condition issues in that area. Detailed models of the Upper and Equatorial ports, port plugs and diagnostics are under development. The detailed port study models will be 40 degree ITER models to preserve the shape of the neutron source loading. These models will include the inner and outer vacuum vessel, inner-wall shielding, blanket shield modules, divertor and cryostat. Models of the OH, TF and PF coils will not be included to save on element count. View full abstract»

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  • High Pressure Supersonic Gas Jet Fueling on NSTX

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    A supersonic gas injector (SGI) has been developed for fueling and diagnostic applications on NSTX. The SGI is comprised of a small de Laval converging-diverging graphite nozzle, a commercial piezoelectric gas valve, and a diagnostic package, all mounted on a movable probe at a low field side midplane port location. The nozzle operated in a pulsed regime at room temperature, reservoir deuterium pressure up to 2500 Torr (50 PSIA), flow rate up to 65 Torr 1 /s (4.55e2f particles/s), and a measured Mach number of about 4. In initial experiments the SGI was used for fueling of ohmic and 2 -6 MW NBI-heated L-and H-mode plasmas. Reliable H-mode access was obtained with SGI fueling, with a fueling efficiency in the range 0.1 -0.3. Good progress was also made toward a controlled density SGI-fueled H-mode plasma scenario with the flow rate of the uncontrolled high field side (HFS) gas injector reduced by up to 20. These experiments motivated a number of SGI upgrades: 1) the maximum plenum pressure has been increased to 5000 Torr (100 PSIA), 2) the plenum pressure volume has been doubled, 3) the gas delivery system has been changed to allow for injection of various gases, 4) a multi-pulse capability has been implemented. As a result of the upgrades, the maximum flow rate increased to about 130 Torr 1 /s. Laboratory gas jet characterization tests indicated a Mach number of about 4 with H2 and I) , and 4-6 with He and N2. Plasma experiments demonstrated the high-pressure gas jet fueling compatibility with H-mode plasmas, high fueling efficiency (0.1 -0.3), and high SOL penetration. View full abstract»

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  • Pellet Dropper Device for ELM Control on DIII-D

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    On several experimental tokamaks, pellet injection has been found to trigger edge localized modes (ELMs) in H-mode plasmas. This can provide a technique for ELM amelioration by reducing the ELM size with small high-frequency pellets. The key for success appears to be small pellets that penetrate just beyond the separatix, enough to trigger an ELM, but not enough to strongly fuel the plasma. To provide a source of small pellets, a pellet dropper device has been developed at the Oak Ridge National Laboratory and installed on the DIII-D tokamak. The pellet dropper consists of a batch extruder with an exit nozzle to provide a filament of solid deuterium (nominal 1-mm diameter), from which pellets are punched/dropped at rates of up to ap50 Hz and at speeds of <10 m/s. The pellets are propelled directly downward and through a vertical injection port on DIII-D. In this paper, the design and the initial test results are presented, and the installation on DIII-D is described. View full abstract»

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  • Metrology for the NCSX Project

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    The National Compact Stellerator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge National Laboratory (ORNL). The complex geometry and tight fabrication tolerances of the NCSX's non-planar coils and vacuum vessel necessitate the use of computerized, CAD-based metrology systems capable of very accurate and reasonably quick measurements. To date, multi-link, portable coordinate measuring machines (pCMM) are used in the fabrication of the non-planar coils. Characterization of the CNC machined coil winding form and subsequent positioning of the conductor centroid (to within +/-0.5 mm) are accomplished via multiple sets of detailed measurements. A laser tracker is used for all phases of work on the vacuum vessel including positioning magnetic diagnostics and vessel ports prior to welding. Future tasks requiring metrology include positioning of the magnet systems and assembly of the three vacuum vessel sub-assemblies onto the final machine configuration. This paper describes the hardware and software used for metrology, as well as the methodology for achieving the required dimensional control and will present an overview of the measurement results to date. View full abstract»

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  • NCSX Component Fabrication Challenges

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    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 complex geometry and tight fabrication tolerances of NCSX create some unique engineering and assembly challenges. This paper will describe a few of the challenges of the machine's Modular Coils and vacuum vessel field period assembly and how they are being solved. Coil assembly began in November 2005 and to date 3 Modular Coils have been completed. One vacuum vessel 120deg section has been delivered and field period assembly work began in May 2006. Machine sector sub-assembly, machine assembly, and testing will follow, leading to First Plasma in 2011. View full abstract»

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  • Assembly Status of the W7-X Stellarator

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    Wendelstein 7-X is a superconducting stellarator being constructed in Greifswald, Germany. The assembly of W7-X is now well underway with the delivery of the main components for the construction of the first module. The spiral shape of the stellarator vacuum vessel means the 70 coils must be threaded individually over the vacuum vessel. This has now been achieved using the foreseen 5 axis coil handling devices. The distance between the coils is fixed by a series of radial and lateral support elements, central support elements and planar support elements, the assembly technology has been tested by an extensive test programme, the first of these radial elements have been installed and the tight assembly tolerances for assembly maintained. W7-X is equipped with 299 ports for diagnostics and supply of in vessel components; they connect the ICRH, ECRH and neutral injection systems to the plasma vessel. All ports must be positioned accurately at various angles around the vacuum vessel, ranging in weight from 100 kg up to 900 kg. For the ports this requires close integration of the manufacturing, metrology and assembly technologies. The powered ramp and bridge structure are under construction and the welding technology for welding and leak testing the port/plasma vessel weld is qualified. Detailed studies for the in-vessel components including a test campaign using the actual plasma vessel sectors in planned for later this year. The adjustability of the divertor and baffle plus the required flexibility of water connections is under investigation. The bus bar system links all coils to the power supply system is in manufacture and assembly tests. Space constraints within the machine cryostat require compact super conducting joints which, have been designed and tested with a resistance in the range of 1nOhm are now in manufacture. The paper will give an overview of the assembly status of W7-X; outline the assembly problems on the various components and the foreseen solutions - in order to bring the machine into operation. View full abstract»

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  • Development of Water Hydraulic Remote Handling System for Divertor Maintenance of ITER

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    In hostile industrial environments where human access could be a health risk, a reliable and flexible teleoperation system is an eminent need. ITER is such an example where a dexterous teleoperation system is required for remote handling tasks in a nuclear environment. The compactness of space, high load capacity and reliability makes hydraulic manipulator an obvious choice. However, possible oil leakage from traditional hydraulic systems and the characteristics of water (fire and environmentally safe, chemically neutral, not activated, not affected by radiation) makes the use of water hydraulics the only choice. This paper describes the development of teleoperation system for ITER consisting of a water hydraulic manipulator as a slave, a commercial haptic device as a master, a human machine interface to assist the operator and a graphical system providing a virtual 3D view of the environment. View full abstract»

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  • Manufacture of Inter-Coil Support-Elements of the W7-X Magnet System

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    The magnet system of W7-X consists of 50 non planar and 20 planar, super conducting coils. Each of these coils has to be fixed on the central-coil support-structure. This is a massive inner ring with extensions for attaching the coils. Additionally the coils have to support each other which is realized by special inter-coil support-elements (ICSE) (Damiani et al., 2005). These elements were designed to resist the high loads of maximum 1500 KN. In order to accelerate the "hand made", already qualified manufacture process of these ICSE an automatic production procedure was developed. This procedure starts with a laser scan measurement, followed by the evaluation of the measured data and the calculation of the dimension which is necessary for the manufacture and creating a design file including a drawing. The outcome is an iges-format file which will be sent directly to the milling/drilling machine. View full abstract»

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  • Hydrogen isotopes permeation evaluation in the advanced material for nuclear fusion blanket use

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    Behavior of hydrogen isotopes such as permeation, diffusion, and dissolution infusion blanket materials were investigated from a viewpoint of the material development and a material design for a fusion reactor. Experiments were conducted with Reduced Activation Ferritic Steel (F82H) and various kinds of SiC materials at elevated temperature. To evaluate permeability of SiC materials and RAFM, deuterium gas permeability was measured using newly designed device. Permeation of hydrogen through RAFM was not significantly different from that of Austenitic Steel, however the temperature dependence of the permeability diffusivity, and solubility showed marked discontinuous change around 850 degree C. Change in crystal structure from bcc to fee is a suspected cause. The measurement of permeability of deuterium gas in Hexoloy and CVD SiC samples were attempted at the temperature above 800 degree C The permeability and deuterium diffusivity of Hexoloy is 2 orders of magnitude smaller than that of CVD SiC. On the pressure dependence of permeability, both linear and square-root dependence were seen. View full abstract»

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  • Basic research of the solid electrolyte as a sensor for the purity control of the liquid metal LiPb

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    Liquid metal Pb-17Li is expected to be an advanced breeding material of a nuclear fusion reactor blanket. The purpose of this study is to evaluate possibilities of continuous monitoring of the concentration of oxygen and hydrogen in LiPb using solid electrolyte cells. In the experiment g 8 gen partial pressure were evaluated to be 10-47 Pa and 663 MPa, respectively. The results showed a good agreement with the theoretical values obtained by thermodynamic calculation. View full abstract»

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  • Improvements of the DIII-D Cryosystem from Analysis of Failure Modes and Their Effects

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    The cryogenic system at General Atomics' DIII-D Tokamak fusion facility provides cryopumping for the fusion vessel and neutral beams, cooling for the superconducting magnets for electron cyclotron heating, and cooling for the deuterium pellet injector. The operation of the fusion facility requires a reliable cryosystem, since upsets can result in significant downtime. A failure modes and effects analysis of the cryosystem was performed to identify and evaluate potential failures. This paper describes the methodology used for identifying the potential failures, based on likelihood of occurrence and consequence to operations. This paper further describes the analysis developed to understand these failure modes, and the improvements and testing to reduce the likelihood and the consequences of these failures. This failure analysis focused on maintaining the helium liquefier operation by providing reliable backup power and cooling for the helium compressors, and resulted in several recommendations which improved the liquifier's availability. View full abstract»

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  • Intensity Distribution of D-3He Fusion Reaction Rate in an IEC Device

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    An inertial electrostatic confinement (IEC) fusion device can produce copious amount of neutrons and protons from D-D and D-3He fusion reactions using D2 and 3He fuels. In IEC researches aiming at drastically enhanced neutron/proton yields, understanding the intensity distribution of fusion reactions is one of the most intensive interests. In order to make clear the spatial distribution of D-3He fusion reaction rate in an IEC device, we analyzed the experimentally observed proton count rates as function of collimation geometry by use of most likelihood-expectation maximization (ML-EM) method. Requirements of the measurement system were studied for reconstructing the D-3He reaction distribution, especially on the feedthrough surface that has been neglected so far, and an upgrade measurement system was developed and introduced in this study. From the experimental results, we found that more than 99 % of the D-3He fusion reactions occur on the cathode and feedthrough surfaces. View full abstract»

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  • Design and Performance of NSTX Movable GDC Probe

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    The NSTX GDC system has been improved by replacing one of the two fixed anodes with a movable GDC probe (MGP) anode that can be inserted 1.2 m to about midway between the inner and outer vessel walls. The purpose was to provide more spatially uniform HeGDC for improving discharge stability and reliability. The MGP has been used reliably between every discharge during the last two NSTX experimental campaigns. It has also been used to apply HeGDC assisted boronization, and more recently, HeGDC assisted lithiumization. The MGP has contributed to improved NSTX performance during long pulse and H-mode discharges, and enabled a faster discharge repetition rate. View full abstract»

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  • Divertor Heat Loads from Thermal and Alpha Particles in a Compact Stellarator Reactor

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    Divertor heat load distributions due to thermal and alpha particles have been assessed in an NCSX-based compact stellarator reactor. A divertor plate system is envisaged, with 4 plates per field period and covering 7% of the plasma surface area. The field-line tracing technique is employed; for thermal flux, the conventional approach is used, while for alphas, their characteristic exit pattern from the plasma and subsequent gyro- orbits are approximated. For the ARIES-CS reference design point (R=7.75 m, A=4.5, B=5.7 T, beta=6.4% and Pnet=1000 MW), combined peak heat loads in the 5 -18 MW/m2 range on the plates have been obtained, assuming a 75% radiation fraction both in the core and at the edge, and a 5% alpha loss fraction. The alpha heat flux could be a dominant determining factor. Further optimization study is warranted to lower all peak heat loads to satisfy the accepted limit of les10 MW/m2. View full abstract»

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  • Phase Lag Infra-red Thermal Examination (PLITE); A New Non-destructive Test Process

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    The International Organization of ITER (International Thermonuclear Experimental Reactor) specifies a requirement of 3 mm in diameter for the largest permissible flaw in the joint of the beryllium (Be) armor tiles and the underlying heat sink made of a copper-chrome-zirconium (CuCrZr) alloy for the first wall (FW). We investigated the sensitivity of a new non-destructive process of detecting these flaws using a method in which we mapped the phase lag of the temperatures on the surface of a sample during thermal cycling with a sinusoidally varying water temperature. A method with hot-cold water test that we had pioneered during the 1990's for the development of a water-cooled mid-plane modular limiter for Tore Supra had worked well with the high conductivity armor made of pyrolytic graphite brazed to copper tubes. The paper describes the experimental system, test samples and some experimental results. View full abstract»

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  • Design of an experimental facility to study convection in liquid lithium

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    Liquid lithium has been proposed as a possible material for both the first wall and the divertor/limiter of a fusion device. Recent experiments on the CDX-U device show that lithium can absorb a surface heat flux of greater than 40(MW/m2) with negligible evaporation. Observation of a focused electron beam hitting solid lithium in the CDX-U lithium tray saw melting of a large section and induced flows. It is believed that these flows redistributed the incident power flux. This paper presents the design of an experiment which will diagnose the flows induced by an intense heat flux onto a lithium pool and measure the maximum heat flux lithium can absorb with applied magnetic fields. A simplified analytical treatment of the expected fluid flow magnitude with increasing magnetic field and surface thermal gradient is shown. Experimental results of the system electron beam source are also shown. These results are the first step in the creation of an experimental facility to study the heat transfer capabilities of free-surface liquid lithium at the University of Illinois. View full abstract»

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  • Engineered Surfaces for the Lithium Tokamak Experiment (LTX)

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    Reactor studies have identified liquid lithium walls as a promising solution to magnetic fusion energy (MFE) first wall problems. The difficulty of translating thick (0.1-1 mm) liquid metals into a full-wall solution has led to the pursuit of the "thin-film" approach (100-10,000 nm) for near-term applications such as the Lithium Tokamak experiment (LTX). However, thin lithium films can become saturated with hydrogen and form LiH, which is not attractive as a plasma facing component. A "thick" lithium film approach would enable hundreds of discharges without the formation of LiH. During this investigation, an engineered surface comprised of a porous refractory metal in which lithium is embedded is being developed to enable the evaluation of a thick lithium film approach for plasma facing components (PFCs). Innovative vacuum plasma spray forming techniques are being used to produce the porous refractory metal surface. Initial resistive heating tests have demonstrated the excellent wetting characteristics of the plasma spray formed porous deposits with liquid lithium. This paper will discuss the development of the engineered surfaces including resistive heating experiments of the porous surfaces with liquid lithium and the status of the effort to coat the full size LTX shell. View full abstract»

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  • NSTX OH Coil Design Improvements

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    The National Spherical Torus Experiment (NSTX) has been operating successfully since February of 1999. A unique element of NSTX is the center solenoid or OH coil that from the start has been a design challenged by the low aspect ratio/geometry of the device. To achieve this low aspect ratio the OH coil's outer diameter is constrained to a narrow profile creating the need for creative design solutions concerning cooling connections, lead orientation, and insulation schemes. The original design has succeeded overall, but NSTX run time has been lost due to coil reliability issues. It was decided in the last year that it would be prudent to fabricate a new OH coil and have it available as an upgrade to the experiment. The experience of operating and maintaining the OH coil has provided the basis for an improved OH coil design. A collaboration was arranged with ASIPP in China to fabricate a spare coil for NSTX. The new OH coil will incorporate both design improvements intended to increase reliability as well as upgrades that will provide flexibility during future operation by allowing for an expanded operational profile. This paper summarizes and reviews these design and reliability improvements. View full abstract»

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  • A Conceptual Design for the Magnets in an IFE Magnetic Intervention Chamber

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    A conceptual design for a magnetic intervention system is presented in support of a 2 GW IFE direct drive fusion power reactor. The system is designed employing a cusp field to deflect ions generated by an IFE implosion away from the first wall of the reactor core and into specifically designed ion dumps. The magnetic coil system will employ liquid helium cooled 5083 Aluminum alloy casing on a Rutherford NbTi cable. The cables are configured as four double pancakes with a 5083 Aluminum alloy case for structural support. The conceptual design and corresponding preliminary load and field calculations will be presented. View full abstract»

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  • Design Optimization of the ITER TF Coil Structure for Manufacturing and Assembly

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    This paper presents some of the design features of the ITER toroidal field coil structure that are being modified to minimize anticipated problems with manufacturing, and assembly, while satisfying their functional design criteria. View full abstract»

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  • ITER Magnet Design Criteria and their Impact on Manufacturing and Assembly

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    The ITER structural design criteria have played an important role in the present design of the ITER magnet system. The performed analyses and structural assessment has given the ITER International Team valuable insight in the mechanical behaviour of the magnet system. A high reliability of its design is obtained to ensure the required operating life and to justify the financial investment. View full abstract»

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  • The Superconducting Bus-Bar System for Wendelstein 7-X - A Challenge to Engineering

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    Forschungszentrum Julich is contributing to the Wendelstein 7-X project among others with the superconducting bus-bar system. The scope is the design, qualification, manufacturing, and assembly of the superconducting bus bars and its appropriate supports. An overall concept of the bus-bar manufacturing was developed with the goal to optimize the process, to simplify the system assembly, and to provide easy transportation. A suitable insulation set up fulfilling all specifications was developed. For the qualification of the insulation and the fabrication process different samples have been fabricated and tested. The bus-bar support system has to take considerable loads due to cool down shrinkage, electromagnetic forces, and displacements imposed by the magnet system. General considerations led to the choice of a system of standardized components which allows assembly of specific adjustable supports for each position. The complexity of the system geometry and its loading required an iterative approach of support design and stress analysis. Finally, an assembly sequence for the bus-bar system was found matching the technological requirements of the bus itself with all geometric and other constraints of the assembly facility. View full abstract»

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