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Fusion Engineering, 1999. 18th Symposium on

Date 25-29 Oct. 1999

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  • 18th IEEE/NPSS Symposium on Fusion Engineering. Symposium Proceedings (Cat. No.99CH37050)

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    Freely Available from IEEE
  • Author index

    Page(s): 0_2 - 0_4
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    Freely Available from IEEE
  • CHF comparison of an attached-fin hypervapotron and porous-coated channels

    Page(s): 388 - 391
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    Critical heat flux experiments were undertaken on two medium-scale, bare copper, mock-ups of the ITER vertical target. One was an attached-fin hypervapotron made of CuNiBe and manufactured by Boeing Company, and the other contained two round porous-coated channels made of CuCrZr and manufactured by the Efremov Institute. All the experiments were performed in the EB-1200 electron beam facility at Sandia and used ITER relevant flow conditions of 100°C inlet subcooling, 4 MPa, 10 m/s water. Both the flat ITER nominal, and the highly peaked, ITER transient, heat flux profiles were used, For both mock-ups, we maintained a near constant exit subcooling by fixing the total beam power and decreasing the heated length from 360 mm to 150 mm and 140 mm, respectively. Results are compared to finite element model predictions and to data obtained on swirl tubes and helical wire mock-ups tested under similar conditions. Local CHF was detected at 21.3 MW/m2 for the hypervapotron and 24.5 MW/m2 for the porous coated -up mock when subjected to the flat heat flux profile. New Global CHF/burnout data are presented for the attached-fin hypervapotron for better comparison to global CHF values obtained on swirl tapes and helical wires View full abstract»

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  • Electron streaming from central core region in inertial-electrostatic confinement fusion

    Page(s): 213 - 216
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    In a steady operation of inertial electrostatic confinement fusion (IECF), the electrons produced within a wire cathode should escape out and reach the anode-on the other hand the ions accelerated towards the cathode should enter the cathode eventually. To verify the guess, we developed a 3-D simulation code and studied formation of a potential distribution, electron trajectories and ion trajectories. The simulation results support electron streaming from the central region through the center area of opening of the cathode, which explains the distinctive discharge called a star mode commonly observed in the IECF experiments View full abstract»

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  • Fast matching of load changes in the ion cyclotron resonance frequency range

    Page(s): 395 - 398
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    In the ion cyclotron resonance frequency (ICRF) range, it is necessary to match the antenna load to the impedance at which the generator delivers its maximum power. This is typically done by adding impedances (mechanical tuners i.e. pieces of transmission lines) at appropriate locations in parallel to the feeder line of the antenna. Large and fast load changes can result from variations in the plasma such as H-mode transitions and ELMs (edge localised modes). For optimum power delivery, those load changes have to be matched dynamically. The required variation of the impedance is performed by a change of electrical length of the tuners. A new type of tuners, fast ferrite tuners (FFT) developed by AFT, relies on the change of magnetisation of ferrites to achieve this. Since there are no mechanically moving parts, the change can be fast. Acceptance tests of a system, developed for General Atomics, using these tuners were performed successfully and first experience was gained on a proof of principle test under plasma conditions View full abstract»

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  • Preliminary radiological assessment of the Fusion Ignition Research Experiment (FIRE)

    Page(s): 475 - 478
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    Detailed activation analysis has been performed for FIRE. The machine is assumed to have an operation schedule of 3000 D-T pulses with a pulse burn of 10 seconds and 2 hours between pulses. At shutdown, the decay heat induced in the first wall is less than 0.1% of the nuclear heating generated in the first wall during operation. The ratio between the shutdown decay heat and nuclear heating generated in the vacuum vessel during operation is on the order of 0.01%. At the end of the machine life, all components would qualify for disposal as Class C low level waste. The biological dose rates behind the vacuum vessel and the divertor remain high during the first year following shutdown. The biological dose rates behind the outboard magnet are acceptable only at locations where the vacuum vessel is more than 40 cm thick. Using a 30 cm thick POLY/CAST shield drops the dose rates on the top of the shield to acceptable levels within a week from shutdown View full abstract»

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  • ITER dome fabrication processes

    Page(s): 377 - 380
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    The 3/4-ton International Thermonuclear Experimental Reactor (ITER) divertor dome component is fabricated from three dissimilar metals-stainless steel, copper, and tungsten. It is actively cooled to withstand heat loads of 5 MW/m2 during nominal operation, transient beat loads of 15 MW/m2, for 1-2 seconds, and volumetric heating of 0.5-5 MW/m3. The divertor dome is comprised of two subcomponents: a 316L cast stainless steel dome shield block with internal coolant passages and a CuCrZr plasma-facing component (PFC) with internal coolant passages and a 15-mm-thick layer of diffusion bonded W-brush armor. This paper describes the processes developed for fabricating the stainless steel dome shield block and CuCrZr plasma-facing component (PFC) and the resulting mechanical properties of the copper and stainless steel joined by the hot-isostatic-pressing assisted diffusion bonding techniques View full abstract»

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  • Schemes and optimization of gas flowing into the ion source and the neutralizer of the DIII-D neutral beam systems

    Page(s): 507 - 510
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    Performance comparisons of a DIII-D neutral beam ion source operated with two different schemes of supplying neutral gas to the arc chamber were performed. Superior performance was achieved when gas was puffed into both the arc chamber and the neutralizer with the gas flows optimized as compared to supplying gas through the neutralizer alone. To form a neutral beam, ions extracted from the arc chamber and accelerated are passed through a neutralizing cell of gas. Neutral gas is commonly puffed into the neutralizing cell to supplement the residual neutral gas from the arc chamber to obtain maximum neutralization efficiency. However, maximizing neutralization efficiency does not necessarily provide the maximum available neutral beam power, since high levels of neutral gas can increase beam loss through collisions and cause larger beam divergence View full abstract»

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  • D-3He fusion in an inertial electrostatic confinement device

    Page(s): 35 - 37
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    Advanced fusion fuels, D and 3He, have been successfully fused in an inertial electrostatic confinement device at the University of Wisconsin. It is thought that this is the first known fusion of helium-3 with deuterium on a steady state basis. The detection of 14.7 MeV protons has confirmed the reaction of D-3He fusion, and has produced a continuous, charged particle flux in excess of 1.4×105 protons/s. Using the same device with D-D fuel a neutron rate of 2.2×107 was achieved. Operating parameters that affect the reaction rate are discussed View full abstract»

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  • Safety and environmental considerations in the design of the Fusion Ignition Research Experiment

    Page(s): 479 - 483
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    The Fusion Ignition Research Experiment (FIRE) is being designed as a next step option in the U.S. Fusion program. Based upon a review of the DOE Fusion Safety Standard, inventory thresholds for DOE facility hazard classification, radiological release limits to meet the no-evacuation objective, key safety functions, and the potential safety concerns that could threaten the safety functions have been identified. Safety analysis calculations of these events have been performed and the resulting thermal hydraulic behavior and radiological consequences are discussed. We conclude with an overall assessment of the safety of FIRE at this stage in the design process and discuss our future plans View full abstract»

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  • Development of high heat flux components in JAERI

    Page(s): 381 - 384
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    Recent progress on the development of new divertor high heat flux components in JAERI is presented in this paper. Large-scale divertor mock-ups, 90 cm long×40 cm wide with carbon fiber composite armors, were fabricated and tested. The mock-ups withstood a heat load of 5 MW/m2 for more than 3000 cycles, and 20 MW/m2 for 1000 cycles. A small mock-up made of reduced activation ferritic steel without armors was also fabricated and tested. The mock-up endured a heat load of 5 MW/m2 for 10,000 cycles View full abstract»

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  • A neural network approach for the detection of the locking position in RFX

    Page(s): 575 - 578
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    Recently in RFX, where wall locked modes were always present, a new technique has demonstrated the possibility to induce a continuous rotation of the modes with respect to the wall. In this technique the nonlinear coupling of the m=0 and m=1 modes has been used to decouple the modes themselves. In the present experiments the mode rotation is induced with a preprogrammed waveform of a toroidal magnetic field rotating ripple. A feedback system able to create a continuous rotation with variable and increasing speed is now under implementation. A neural network (NN) has been developed to identify the locked mode position. In the paper different NNs are presented, discussed and compared View full abstract»

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  • Remote handling of JET in-torus components-a practical experience

    Page(s): 151 - 154
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    This paper summarises the experiences gained from the extensive handling of JET components inside the torus. The problems involved with handling components not designed to be remotely handled and the methods used to overcome them are described and discussed with specific examples from recent JET remote operations. The method employed for remotely producing structural TIG welds is explained. The problems of dextrous manipulation in an inverted attitude are discussed and the methods of amelioration are described View full abstract»

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  • Automated calculation of DIII-D neutral beam availability

    Page(s): 511 - 514
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    The neutral beam systems for the DIII-D tokamak are an extremely reliable source of auxiliary plasma heating, capable of supplying up to 20 MW of injected power, from eight separate beam sources into each tokamak discharge. The high availability of these systems for tokamak operations is sustained by careful monitoring of performance and following up on failures. One of the metrics for this performance is the requested injected power profile as compared to the power profile delivered for a particular pulse. Calculating this was a relatively straightforward task, however innovations such as the ability to modulate the beams and more recently the ability to substitute an idle beam for one which has failed during a plasma discharge, have made the task very complex View full abstract»

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  • Plasma fueling, pumping, and tritium handling considerations for FIRE

    Page(s): 463 - 466
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    Tritium pellet injection will be utilized on the Fusion Ignition Research Experiment (FIRE) for efficient tritium fueling and to optimize the density profile for high fusion power. Conventional pneumatic pellet injectors, coupled with a guidetube system to launch pellets into the plasma from the high field side, low field side, and vertically, will be provided for fueling along with gas puffing for plasma edge density control. About 0.1 g of tritium must be injected during each 10-s pulse. The tritium and deuterium will be exhausted into the divertor. The double null divertor will have 16 cryogenic pumps located near the divertor chamber to provide the required high pumping speed of 200 torr-L/s View full abstract»

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  • Design of the vacuum liner for the National Compact Stellarator Experiment (NCSX)

    Page(s): 235 - 238
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    The National Compact Stellarator Experiment (NCSX) requires an inner vacuum vessel (liner) which will serve as a structural component, withstanding full atmospheric pressure and magnetic disruption loads. It must also be capable of bake out to 350 °C. Without imparting excessive thermal loading on saddle coils which reside on an outer shell surrounding the liner. NCSX will be sited in the Princeton Beta Experiment (PBX) test cell. Many of the existing site assets including the test cell, TF and PF coils, power supplies, neutral beam heating systems, and site utilities will be utilized to minimize the cost of the project. The conceptual design features a stellarator core that is pre-fabricated and dropped into place on the PBX platform. The existing TF and PF coils are then reassembled around the stellarator core. Trade studies have been conducted to explore different configurations for the liner and the saddle coil structure. These included a design which used a low conductance non vacuum liner assembled within an outer vessel which surrounded the NCSX device, and served as the primary vacuum containment. The plasma has three periods and is very convoluted, changing in cross section from nearly circular to bean shaped with a very concave contour on the inner surface. The contour repeats twice every period, that is the shape is identical after 60 degrees toroidally, but with the shape inverted 180 degrees. The liner is required to follow the plasma surface very closely and, as result, is complex in shape. A trade off study has been made to determine the best fabrication technique to produce such a liner. The study included casting, pressing, explosive forming, isostatic pressing, and brake bending from flat developed patterns. The decision on a fabrication method relied heavily on previous experience by other experiments such as HSX, W7AS, and W7-X which dealt with similar, convoluted shapes. The method picked was the flat pattern technique utilized by Wendelstein W7-AS. The liner is designed to be built in three sections corresponding to the plasma field periods, and final assembled around an inner shell core. After welding of the liner sections, the outer shell segments will assembled around the liner. Saddle coils will then installed around the shell insulation must be installed between the liner and shell to protect the shell during liner bake out, and to minimize heat transfer from the cooler shell during operation. Because the radial build up is limited and the shell is designed to operate at cryogenic (LN2) temperatures the choice of insulation materials is limited. A trade off study has been performed and extensive thermal analyses have been done to determine an insulation design which can operate in the extreme temperature range from 77 K to 623 K. The engineering design is being developed by a team from Oak Ridge National Laboratory and Princeton Plasma Physics Laboratory. A Physics Validation Review of NCSX is planned for September, 1999. The Conceptual Design effort will then commence in earnest, culminating in a conceptual design review in March, 7000. First plasma is planned for September, 2004 View full abstract»

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  • Innovative, ultra-low cost fabrication methods

    Page(s): 77 - 80
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    Two fabrication processes are being developed that may significantly enhance fusion as a competitive electrical power source. Currently, the capital cost for experimental, developmental, and commercial fusion power cores is too high. Innovative ultralow cost fabrication methods being developed by Boeing and its vendors may significantly lower the cost of normally-conducting toroidal field (TF) coils. Laser or plasma-arc forming is proposed for the large copper centerpost. This process uses a laser or plasma are to “additively machine” the centerpost with all necessary features including integral coolant passages. The 2.7×106 kg aluminum TF return shell can be fabricated using a spray casting process that atomizes and sprays molten aluminum onto a preform structure and around embedded coolant tubes. A scoping cost study indicated the capital cost of the TF centerpost and return shell leg will be reduced by $330 M (greater than an order of magnitude) which will reduce the projected cost of electricity (COE) by approximately 10% View full abstract»

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  • Engineering overview of the Fusion Ignition Research Experiment (FIRE)

    Page(s): 163 - 167
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    FIRE is an option for the next step in the US magnetic fusion energy program. The primary engineering requirement is to produce a pulse length of greater than 10 s with a plasma current of about 6.6 MA. The FIRE tokamak has a major radius of 2 m, and a minor radius of 0.525 m. Liquid nitrogen cooled copper coils are used for both the toroidal field (TF) and poloidal field (PF) coils. The overall engineering features of FIRE are outlined in this paper View full abstract»

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  • Operation and control of JT-60U ECRF system

    Page(s): 403 - 406
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    This paper describes the operation and control for the 110 GHz ECRF system for JT-60U, which has started operation since March in 1999. This system is composed of a 1 MW gyrotron, its high voltage power supply, a transmission line about 60 m in length and a quasi-optical antenna using a steerable mirror. Key issues of the ECRF system are to drive the gyrotron stably and to control the local RF deposition profile in a plasma View full abstract»

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  • The influence of divertor target plates on the magnetic field of the dynamic ergodic divertor of TEXTOR-94

    Page(s): 115 - 118
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    The dynamic ergodic divertor DED is under construction as a novel tool for TEXTOR-94 to control transport at the plasma edge and possibly also plasma rotation. The DED will be operated with DC and AC currents at frequencies between 50 Hz and 10 kHz. The efficiency of the magnetic field in the plasma region especially at frequencies above 5 kHz is influenced by the complete structural arrangement in the vicinity of the DED coils. The main influence is given by the graphite tiles. For the design of the coil arrangement and the layout of the power supply the knowledge of the shielding effect especially of the divertor tiles on the magnetic field strength is important. Furthermore the knowledge of induced currents in the tiles as function of the tile dimensions and the resulting temperatures is necessary in order to optimize their dimensions and to calculate the forces for the layout of the mechanical design View full abstract»

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  • Evaluation of welding deformation on ITER vacuum vessel

    Page(s): 245 - 248
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    The evaluation of welding deformation on the vacuum vessel sector model of the Interaction Thermonuclear Experimental Reactor (ITER) has been performed using a finite element method analysis. The welding data of a simple plate test is translated to thermal shrinkage and applied to the model as an equivalent welding deformation force. The calculation results are compared with the results of full-scale mock-up test simulating on-site welding. As a result, the error of this analysis was obtained to be 10-25% in shrinkage and factor 5 in cross sectional deformation View full abstract»

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  • Facility and site needs for the FIRE project

    Page(s): 488 - 491
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    This paper describes concepts that have been developed for the layout of the FIRE buildings and facilities, and identifies major site requirements. The FIRE test cell, approximately 30×30 meters, is part of a structurally integrated tokamak and hot cell building. The tokamak centerline is set 4.2 m above the floor. Overhead space is sufficient to allow a bridge crane to remove the tie rod structure from the assembled tokamak. FIRE will be designed so that the thick outer wall of the vacuum vessel and the TF coils provide enough shielding so that the exterior of the machine remains accessible for hands-on maintenance. However, during operation and during maintenance activities that require the removal of a port plug additional shielding is required. A movable shield roof, which operates above the tokamak but below the crane, is provided to close the shield boundary during operations. The FIRE facility will have a tritium inventory large enough to require that it must be licensed by the NRC. A non-site-specific plot plan has been developed and is described. The tokamak-hot cell building is placed in the center of the site and arbitrarily oriented so crane rails run north-south, and the hot cell is to the north of the tokamak. Major electrical facilities are located to the west, while vacuum pumping, cooling, and tritium processing facilities are located to the east. Office space is provided for a laboratory staff assumed to be 800. All buildings and facilities are developed assuming a “green-field” site, and costs are estimated accordingly. Site requirements and alternative site selection strategies are described. Savings could be obtained if an existing facility, particularly one where nuclear licensing is already established, could be utilized View full abstract»

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  • PF and TF power systems for the Fusion Ignition Research Experiment (FIRE)

    Page(s): 455 - 458
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    The primary goal of the FIRE preconceptual design is to affordably conduct physics experiments in the fusion ignition regime with self-heated plasmas. The device could also permit advanced tokamak experiments in deuterium lasting several minutes. Tradeoff studies considering MVA and flattop duration have identified TF and PF magnet power system designs expected to have low cost, consistent both with the mission to enter the ignition regime and the possibility to conduct long pulse deuterium experiments. In addition, a performance-extending upgrade path for the power system has been identified which could be followed later, if experimental results justify further increasing the maximum toroidal field and plasma current or lengthening the duration View full abstract»

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  • Management of combined manual and remote handling shutdowns at JET

    Page(s): 257 - 260
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    The Tile Carrier Transfer Facility (TCTF) was used during the remote tile exchange for the robotic transfer of all equipment in and out of the torus using the short boom in coordination with the mascot and articulated boom. The TCTF has been up rated to support limited duration manned access to the torus where remote handling is impractical View full abstract»

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  • Development of a closed loop simulator for poloidal field control in DIII-D

    Page(s): 531 - 534
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    The design of a model-based simulator of the DIII-D poloidal field system is presented. The simulator is automatically configured to match a particular DIII-D discharge circuit. The simulator can be run in a data input mode, in which prior acquired DIII-D shot data is input to the simulator, or in a stand-alone predictive mode, in which the model operates in closed loop with the plasma control system. The simulator is used to design and validate a multi-input-multi-output controller which has been implemented on DIII-D to control plasma shape. Preliminary experimental controller results are presented View full abstract»

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