WorldWideScience

Sample records for high-power-density fusion devices

  1. Exploring novel high power density concepts for attractive fusion systems

    Energy Technology Data Exchange (ETDEWEB)

    Abdou, M.A. [California State Univ., Los Angeles, CA (United States). Dept. of Mechanical Engineering; APEX Team

    1999-05-01

    The advanced power extraction study is aimed at exploring innovative concepts for fusion power technology (FPT) that can tremendously enhance the potential of fusion as an attractive and competitive energy source. Specifically, the study is exploring new and `revolutionary` concepts that can provide the capability to efficiently extract heat from systems with high neutron and surface heat loads while satisfying all the FPT functional requirements and maximizing reliability, maintainability, safety, and environmental requirements. The primary criteria for measuring performance of the new concepts are: (1) high power density capability with a peak neutron wall load (NWL) of {proportional_to}10 MW m{sup -2} and surface heat flux of {proportional_to}2 MW m{sup -2}; (2) high power conversion efficiency, {proportional_to}40% net; and (3) clear potential to achieve high availability; specifically low failure rate, large design margin, and short downtime for maintenance. A requirement that MTBF{>=}43 MTTR was derived as a necessary condition to achieve the required first wall/blanket availability, where MTBF is the mean time between failures and MTTR is the mean time to recover. Highlights of innovative and promising new concepts that may satisfy these criteria are provided. (orig.) 40 refs.

  2. New directions in fusion machines: report on the MFAC Panel X on high power density options

    International Nuclear Information System (INIS)

    Linford, R.K.

    1985-01-01

    The high cost of fusion is motivating a shift in research interest toward smaller, lower-cost systems. Panel X of the Magnetic Fusion Advisory Committee (MFAC) was charged to assess the potential benefits and problems associated with small, high-power-density approaches to fusion. The Panel identified figures of merit which are useful in evaluating various approaches to reduce the development costs and capital costs of fusion systems. As a result of their deliberations, the Panel recommended that ''...increased emphasis should be given to improving the mass power density of fusion systems, aiming at a minimum target of 100 kWe/tonne'', and that ''Increased emphasis should be given to concepts that offer the potential to reduce substantially the cost of development steps in physics and technology.''

  3. New directions in fusion machines: report on the MFAC Panel X on high power density options

    Energy Technology Data Exchange (ETDEWEB)

    Linford, R.K.

    1985-01-01

    The high cost of fusion is motivating a shift in research interest toward smaller, lower-cost systems. Panel X of the Magnetic Fusion Advisory Committee (MFAC) was charged to assess the potential benefits and problems associated with small, high-power-density approaches to fusion. The Panel identified figures of merit which are useful in evaluating various approaches to reduce the development costs and capital costs of fusion systems. As a result of their deliberations, the Panel recommended that ''...increased emphasis should be given to improving the mass power density of fusion systems, aiming at a minimum target of 100 kWe/tonne'', and that ''Increased emphasis should be given to concepts that offer the potential to reduce substantially the cost of development steps in physics and technology.''

  4. Engineering and physics of high-power-density, compact, reversed-field-pinch fusion reactors

    International Nuclear Information System (INIS)

    Najmabadi, F.; Conn, R.W.; Krakowski, R.A.; Schultz, K.R.; Steiner, D.

    1989-01-01

    The technical feasibility and key developmental issues of compact, high-power-density Reversed-Field-Pinch (RFP) reactors are the primary results of the TITAN RFP reactor study. Two design approaches emerged, TITAN-I and TITAN-II, both of which are steady-state, DT-burning, circa 1000 MWe power reactors. The TITAN designs are physically compact and have a high neutron wall loading of 18 MW m 2 . Detailed analyses indicate that: a) each design is technically feasible; b) attractive features of compact RFP reactors can be realized without sacrificing the safety and environmental potential of fusion; and c) major features of this particular embodiment of the RFP reactor are retained in a design window of neutron wall loading ranging from 10 to 20 MW/m 2 . A major product of the TITAN study is the identification and quantification of major engineering and physics requirements for this class of RFP reactors. These findings are the focus of this paper. (author). 26 refs.; 4 figs.; 1 tab

  5. New directions in fusion machines: Report on the MFAC panel X on high power density options

    International Nuclear Information System (INIS)

    Linford, R.K.

    1986-01-01

    The high cost of fusion is motivating a shift in research interest toward smaller, lower-cost systems. Panel X of the Magnetic Fusion Advisory Committee (MFAC) was charged to assess the potential benefits and problems associated with small, highpower-density approaches to fusion. The Panel identified figures of merit which are useful in evaluating various approaches to reduce the development costs and capital costs of fusion systems. As a result of their deliberations, the Panel recommended that ''...increased emphasis should be given to improving the mass power density of fusion systems, aiming at a minimum target of 100 kWe/tonne'', and that ''Increased emphasis should be given to concepts that offer the potential to reduce4 substantially the cost of development steps in physics and technology.''

  6. Simulation study of a high power density rectenna array for biomedical implantable devices

    Science.gov (United States)

    Day, John; Yoon, Hargsoon; Kim, Jaehwan; Choi, Sang H.; Song, Kyo D.

    2016-04-01

    The integration of wireless power transmission devices using microwaves into the biomedical field is close to a practical reality. Implanted biomedical devices need a long lasting power source or continuous power supply. Recent development of high efficiency rectenna technology enables continuous power supply to these implanted devices. Due to the size limit of most of medical devices, it is imperative to minimize the rectenna as well. The research reported in this paper reviews the effects of close packing the rectenna elements which show the potential of directly empowering the implanted devices, especially within a confined area. The rectenna array is tested in the X band frequency range.

  7. The reversed-field pinch as a poloidal-field-dominated, compact, high-power-density fusion system

    International Nuclear Information System (INIS)

    Krakowski, R.A.

    1988-01-01

    This paper discusses the feasibility of reversed-field pinch devices as future thermonuclear reactors. Safety, cost, ion temperatures, Lawson numbers, and power densities are reviewed for these types of devices. 12 refs., 2 figs., 1 tab

  8. Fusion devices

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1977-01-01

    Three types of thermonuclear fusion devices currently under development are reviewed for an electric utilities management audience. Overall design features of laser fusion, tokamak, and magnetic mirror type reactors are described and illustrated. Thrusts and trends in current research on these devices that promise to improve performance are briefly reviewed. Twenty photographs and drawings are included

  9. Electronic DC transformer with high power density

    NARCIS (Netherlands)

    Pavlovský, M.

    2006-01-01

    This thesis is concerned with the possibilities of increasing the power density of high-power dc-dc converters with galvanic isolation. Three cornerstones for reaching high power densities are identified as: size reduction of passive components, reduction of losses particularly in active components

  10. A High Power Density Integrated Charger for Electric Vehicles with Active Ripple Compensation

    OpenAIRE

    Pan, Liwen; Zhang, Chengning

    2015-01-01

    This paper suggests a high power density on-board integrated charger with active ripple compensation circuit for electric vehicles. To obtain a high power density and high efficiency, silicon carbide devices are reported to meet the requirement of high-switching-frequency operation. An integrated bidirectional converter is proposed to function as AC/DC battery charger and to transfer energy between battery pack and motor drive of the traction system. In addition, the conventional H-bridge cir...

  11. Cold nuclear fusion device

    International Nuclear Information System (INIS)

    Ogino, Shinji.

    1991-01-01

    Selection of cathode material is a key to the attainment of cold nuclear fusion. However, there are only few reports on the cathode material at present and an effective development has been demanded. The device comprises an anode and a cathode and an electrolytic bath having metal salts dissolved therein and containing heavy water in a glass container. The anode is made of gold or platinum and the cathode is made of metals of V, Sr, Y, Nb, Hf or Ta, and a voltage of 3-25V is applied by way of a DC power source between them. The metal comprising V, Sr, Y, Nb, Hf or Ta absorbs deuterium formed by electrolysis of heavy water effectively to cause nuclear fusion reaction at substantially the same frequency and energy efficiency as palladium and titanium. Accordingly, a cold nuclear fusion device having high nuclear fusion generation frequency can be obtained. (N.H.)

  12. ADX: a high field, high power density, Advanced Divertor test eXperiment

    Science.gov (United States)

    Vieira, R.; Labombard, B.; Marmar, E.; Irby, J.; Shiraiwa, S.; Terry, J.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; ADX Team

    2014-10-01

    The MIT PSFC and collaborators are proposing an advanced divertor experiment (ADX) - a tokamak specifically designed to address critical gaps in the world fusion research program on the pathway to FNSF/DEMO. This high field (6.5 tesla, 1.5 MA), high power density (P/S ~ 1.5 MW/m2) facility would utilize Alcator magnet technology to test innovative divertor concepts for next-step DT fusion devices (FNSF, DEMO) at reactor-level boundary plasma pressures and parallel heat flux densities while producing high performance core plasma conditions. The experimental platform would also test advanced lower hybrid current drive (LHCD) and ion-cyclotron range of frequency (ICRF) actuators and wave physics at the plasma densities and magnetic field strengths of a DEMO, with the unique ability to deploy launcher structures both on the low-magnetic-field side and the high-field side - a location where energetic plasma-material interactions can be controlled and wave physics is most favorable for efficient current drive, heating and flow drive. This innovative experiment would perform plasma science and technology R&D necessary to inform the conceptual development and accelerate the readiness-for-deployment of FNSF/DEMO - in a timely manner, on a cost-effective research platform. Supported by DE-FC02-99ER54512.

  13. Ceramics for fusion devices

    International Nuclear Information System (INIS)

    Clinard, F.W. Jr.

    1984-01-01

    Ceramics are required for a number of applications in fusion devices, among the most critical of which are magnetic coil insulators, windows for RF heating systems, and structural uses. Radiation effects dominate consideration of candidate materials, although good pre-irradiation properties are a requisite. Materials and components can be optimized by careful control of chemical and microstructural content, and application of brittle material design and testing techniques. Future directions for research and development should include further extension of the data base in the areas of electrical, structural, and thermal properties; establishment of a fission neutron/fusion neutron correlation including transmutation gas effects; and development of new materials tailored to meet the specific needs of fusion reactors

  14. ADX: a high field, high power density, advanced divertor and RF tokamak

    Science.gov (United States)

    LaBombard, B.; Marmar, E.; Irby, J.; Terry, J. L.; Vieira, R.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; Baek, S.; Beck, W.; Bonoli, P.; Brunner, D.; Doody, J.; Ellis, R.; Ernst, D.; Fiore, C.; Freidberg, J. P.; Golfinopoulos, T.; Granetz, R.; Greenwald, M.; Hartwig, Z. S.; Hubbard, A.; Hughes, J. W.; Hutchinson, I. H.; Kessel, C.; Kotschenreuther, M.; Leccacorvi, R.; Lin, Y.; Lipschultz, B.; Mahajan, S.; Minervini, J.; Mumgaard, R.; Nygren, R.; Parker, R.; Poli, F.; Porkolab, M.; Reinke, M. L.; Rice, J.; Rognlien, T.; Rowan, W.; Shiraiwa, S.; Terry, D.; Theiler, C.; Titus, P.; Umansky, M.; Valanju, P.; Walk, J.; White, A.; Wilson, J. R.; Wright, G.; Zweben, S. J.

    2015-05-01

    The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)—a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (⩾6.5 T, 1.5 MA), high power density facility (P/S ˜ 1.5 MW m-2) will test innovative divertor ideas, including an ‘X-point target divertor’ concept, at the required performance parameters—reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region—while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side—the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination—advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions—will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept

  15. High Power Density Power Electronic Converters for Large Wind Turbines

    DEFF Research Database (Denmark)

    Senturk, Osman Selcuk

    . For these VSCs, high power density is required due to limited turbine nacelle space. Also, high reliability is required since maintenance cost of these remotely located wind turbines is quite high and these turbines operate under harsh operating conditions. In order to select a high power density and reliability......In large wind turbines (in MW and multi-MW ranges), which are extensively utilized in wind power plants, full-scale medium voltage (MV) multi-level (ML) voltage source converters (VSCs) are being more preferably employed nowadays for interfacing these wind turbines with electricity grids...... VSC solution for wind turbines, first, the VSC topology and the switch technology to be employed should be specified such that the highest possible power density and reliability are to be attained. Then, this qualitative approach should be complemented with the power density and reliability...

  16. Flexible asymmetric supercapacitors with high energy and high power density in aqueous electrolytes

    Science.gov (United States)

    Cheng, Yingwen; Zhang, Hongbo; Lu, Songtao; Varanasi, Chakrapani V.; Liu, Jie

    2013-01-01

    Supercapacitors with both high energy and high power densities are critical for many practical applications. In this paper, we discuss the design and demonstrate the fabrication of flexible asymmetric supercapacitors based on nanocomposite electrodes of MnO2, activated carbon, carbon nanotubes and graphene. The combined unique properties of each of these components enable highly flexible and mechanically strong films that can serve as electrodes directly without using any current collectors or binders. Using these flexible electrodes and a roll-up approach, asymmetric supercapacitors with 2 V working voltage were successfully fabricated. The fabricated device showed excellent rate capability, with 78% of the original capacitance retained when the scan rate was increased from 2 mV s-1 to 500 mV s-1. Owing to the unique composite structure, these supercapacitors were able to deliver high energy density (24 W h kg-1) under high power density (7.8 kW kg-1) conditions. These features could enable supercapacitor based energy storage systems to be very attractive for a variety of critical applications, such as the power sources in hybrid electric vehicles and the back-up powers for wind and solar energy, where both high energy density and high power density are required.Supercapacitors with both high energy and high power densities are critical for many practical applications. In this paper, we discuss the design and demonstrate the fabrication of flexible asymmetric supercapacitors based on nanocomposite electrodes of MnO2, activated carbon, carbon nanotubes and graphene. The combined unique properties of each of these components enable highly flexible and mechanically strong films that can serve as electrodes directly without using any current collectors or binders. Using these flexible electrodes and a roll-up approach, asymmetric supercapacitors with 2 V working voltage were successfully fabricated. The fabricated device showed excellent rate capability, with 78% of

  17. Fusion engineering device design description

    Energy Technology Data Exchange (ETDEWEB)

    Flanagan, C.A.; Steiner, D.; Smith, G.E.

    1981-12-01

    The US Magnetic Fusion Engineering Act of 1980 calls for the operation of a Fusion Engineering Device (FED) by 1990. It is the intent of the Act that the FED, in combination with other testing facilities, will establish the engineering feasibility of magnetic fusion energy. During 1981, the Fusion Engineering Design Center (FEDC), under the guidance of a Technical Management Board (TMB), developed a baseline design for the FED. This design is summarized herein.

  18. Fusion Engineering Device design description

    International Nuclear Information System (INIS)

    Flanagan, C.A.; Steiner, D.; Smith, G.E.

    1981-12-01

    The US Magnetic Fusion Engineering Act of 1980 calls for the operation of a Fusion Engineering Device (FED) by 1990. It is the intent of the Act that the FED, in combination with other testing facilities, will establish the engineering feasibility of magnetic fusion energy. During 1981, the Fusion Engineering Design Center (FEDC), under the guidance of a Technical Management Board (TMB), developed a baseline design for the FED. This design is summarized herein

  19. Fusion engineering device design description

    International Nuclear Information System (INIS)

    Flanagan, C.A.; Steiner, D.; Smith, G.E.

    1981-12-01

    The US Magnetic Fusion Engineering Act of 1980 calls for the operation of a Fusion Engineering Device (FED) by 1990. It is the intent of the Act that the FED, in combination with other testing facilities, will establish the engineering feasibility of magnetic fusion energy. During 1981, the Fusion Engineering Design Center (FEDC), under the guidance of a Technical Management Board (TMB), developed a baseline design for the FED. This design is summarized herein

  20. Beam dancer fusion device

    International Nuclear Information System (INIS)

    Maier, H.B.

    1984-01-01

    To accomplish fusion of two or more fusion fuel elements numerous minute spots of energy or laser light are directed to a micro target area, there to be moved or danced about by a precision mechanical controlling apparatus at the source of the laser light or electromagnetic energy beams, so that merging and coinciding patterns of light or energy beams can occur around the area of the fuel atoms or ions. The projecting of these merging patterns may be considered as target searching techniques to locate responsive clusters of fuel elements and to compress such elements into a condition in which fusion may occur. Computerized programming may be used

  1. Generation of high-power-density atmospheric pressure plasma with liquid electrodes

    International Nuclear Information System (INIS)

    Dong Lifang; Mao Zhiguo; Yin Zengqian; Ran Junxia

    2004-01-01

    We present a method for generating atmospheric pressure plasma using a dielectric barrier discharge reactor with two liquid electrodes. Four distinct kinds of discharge, including stochastic filaments, regular square pattern, glow-like discharge, and Turing stripe pattern, are observed in argon with a flow rate of 9 slm. The electrical and optical characteristics of the device are investigated. Results show that high-power-density atmospheric pressure plasma with high duty ratio in space and time can be obtained. The influence of wall charges on discharge power and duty ratio has been discussed

  2. A High Power Density Integrated Charger for Electric Vehicles with Active Ripple Compensation

    Directory of Open Access Journals (Sweden)

    Liwen Pan

    2015-01-01

    Full Text Available This paper suggests a high power density on-board integrated charger with active ripple compensation circuit for electric vehicles. To obtain a high power density and high efficiency, silicon carbide devices are reported to meet the requirement of high-switching-frequency operation. An integrated bidirectional converter is proposed to function as AC/DC battery charger and to transfer energy between battery pack and motor drive of the traction system. In addition, the conventional H-bridge circuit suffers from ripple power pulsating at second-order line frequency, and a scheme of active ripple compensation circuit has been explored to solve this second-order ripple problem, in which a pair of power switches shared traction mode, a ripple energy storage capacitor, and an energy transfer inductor. Simulation results in MATLAB/Simulink validated the eligibility of the proposed topology. The integrated charger can work as a 70 kW motor drive circuit or a converter with an active ripple compensation circuit for 3 kW charging the battery. The impact of the proposed topology and control strategy on the integrated charger power losses, efficiency, power density, and thermal performance has also been analysed and simulated.

  3. Toroidal nuclear fusion device

    International Nuclear Information System (INIS)

    Ito, Yutaka; Kasahara, Tatsuo; Takizawa, Teruhiro.

    1975-01-01

    Object: To design a device so as to be formed into a large-size and to arrange ports, through which neutral particles enter, in inclined fashion. Structure: Toroidal coils are wound about vacuum vessels which are divided into plural number. In the outer periphery of the vacuum vessels, ports are disposed inclined in the peripheral direction of the vacuum vessels and communicated with the vacuum vessels, and wall surfaces opposed to the ports of the toroidal coils adjacent at least the inclined sides of the ports are inclined substantially simularly to the port wall surfaces. (Kamimura, M.)

  4. APEX and ALPS, high power density technology programs in the U.S

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Berk, S.; Abdou, M.; Mattas, R.

    1999-02-01

    In fiscal year (FY) 1998 two new fusion technology programs were initiated in the US, with the goal of making marked progress in the scientific understanding of technologies and materials required to withstand high plasma heat flux and neutron wall loads. APEX is exploring new and revolutionary concepts that can provide the capability to extract heat efficiently from a system with high neutron and surface heat loads while satisfying all the fusion power technology requirements and achieving maximum reliability, maintainability, safety, and environmental acceptability. ALPS program is evaluating advanced concepts including liquid surface limiters and divertors on the basis of such factors as their compatibility with fusion plasma, high power density handling capabilities, engineering feasibility, lifetime, safety and R and D requirements. The APEX and ALPS are three-year programs to specify requirements and evaluate criteria for revolutionary approaches in first wall, blanket and high heat flux component applications. Conceptual design and analysis of candidate concepts are being performed with the goal of selecting the most promising first wall, blanket and high heat flux component designs that will provide the technical basis for the initiation of a significant R and D effort beginning in FY2001. These programs are also considering opportunities for international collaborations

  5. High power density yeast catalyzed microbial fuel cells

    Science.gov (United States)

    Ganguli, Rahul

    Microbial fuel cells leverage whole cell biocatalysis to convert the energy stored in energy-rich renewable biomolecules such as sugar, directly to electrical energy at high efficiencies. Advantages of the process include ambient temperature operation, operation in natural streams such as wastewater without the need to clean electrodes, minimal balance-of-plant requirements compared to conventional fuel cells, and environmentally friendly operation. These make the technology very attractive as portable power sources and waste-to-energy converters. The principal problem facing the technology is the low power densities compared to other conventional portable power sources such as batteries and traditional fuel cells. In this work we examined the yeast catalyzed microbial fuel cell and developed methods to increase the power density from such fuel cells. A combination of cyclic voltammetry and optical absorption measurements were used to establish significant adsorption of electron mediators by the microbes. Mediator adsorption was demonstrated to be an important limitation in achieving high power densities in yeast-catalyzed microbial fuel cells. Specifically, the power densities are low for the length of time mediator adsorption continues to occur. Once the mediator adsorption stops, the power densities increase. Rotating disk chronoamperometry was used to extract reaction rate information, and a simple kinetic expression was developed for the current observed in the anodic half-cell. Since the rate expression showed that the current was directly related to microbe concentration close to the electrode, methods to increase cell mass attached to the anode was investigated. Electrically biased electrodes were demonstrated to develop biofilm-like layers of the Baker's yeast with a high concentration of cells directly connected to the electrode. The increased cell mass did increase the power density 2 times compared to a non biofilm fuel cell, but the power density

  6. Study of In-Cylinder Reactions of High Power-Density Direct Injection Diesel Engines

    National Research Council Canada - National Science Library

    Jansons, M

    2004-01-01

    Direct-injection (DI) Diesel or compression-ignition (CI) engine combustion process is investigated when new design and operational strategies are employed in order to achieve a high power-density (HPD) engine...

  7. Efficient, High Power Density Hydrocarbon-Fueled Solid Oxide Stack System, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — Precision Combustion, Inc. (PCI) proposes to develop and demonstrate an innovative high power density design for direct internal reforming of regolith off-gases...

  8. Efficient, high power density hydrocarbon-fueled solid oxide stack system, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Precision Combustion, Inc. (PCI) proposes to develop and demonstrate an innovative high power density design for direct internal reforming of regolith off-gases...

  9. Nuclear fusion power supply device

    International Nuclear Information System (INIS)

    Nakagawa, Satoshi.

    1975-01-01

    Object: To use a hybrid power supply device, which comprises a thyristor power supply and a diode power supply, to decrease cost of a nuclear fusion power supply device. Structure: The device comprises a thyristor power supply connected through a closing unit and a diode power supply connected in parallel through a breaker, input of each power supply being applied with an output voltage of a flywheel AC generator. When a current transformer is excited, a disconnecting switch is turned on to close the diode power supply and a current of the current transformer is increased by an automatic voltage regulator to a set value within a predetermined period of time. Next, the current is cut off by a breaker, and when the breaker is in on position, the disconnecting switch is opened to turn on the closing unit. Thus, when a plasma electric current reaches a predetermined value, the breaker is turned on, and the current of the current transformer is controlled by the thyristor power supply. (Kamimura, M.)

  10. Management of water leaks on Tore Supra actively cooled fusion device

    International Nuclear Information System (INIS)

    Hatchressian, J.C.; Gargiulo, L.; Samaille, F.; Soler, B.

    2005-01-01

    Up to now, Tore Supra is the only fusion device fully equipped with actively cooled Plasma Facing Components (PFCs). In case of abnormal events during a plasma discharge, the PFCs could be submitted to a transient high power density (run away electrons) or to a continuous phenomena as local thermal flux induced by trapped suprathermal electrons or ions). It could lead to a degradation of the PFC integrity and in the worst case to a water leak occurrence. Such water leak has important consequence on the tokamak operation that concerns PFCs themselves, monitoring equipment located in the vacuum vessel or connected to the ports as RF antennas, diagnostics or pumping systems. Following successive water leak events (the most important water leak, that occurred in September 2002, is described in the paper), a large feedback experience has been gained on Tore supra since more than 15 years that could be useful to actively cooled next devices as W7X and ITER. (authors)

  11. High-Power-Density, High-Energy-Density Fluorinated Graphene for Primary Lithium Batteries

    Directory of Open Access Journals (Sweden)

    Guiming Zhong

    2018-03-01

    Full Text Available Li/CFx is one of the highest-energy-density primary batteries; however, poor rate capability hinders its practical applications in high-power devices. Here we report a preparation of fluorinated graphene (GFx with superior performance through a direct gas fluorination method. We find that the so-called “semi-ionic” C-F bond content in all C-F bonds presents a more critical impact on rate performance of the GFx in comparison with sp2 C content in the GFx, morphology, structure, and specific surface area of the materials. The rate capability remains excellent before the semi-ionic C-F bond proportion in the GFx decreases. Thus, by optimizing semi-ionic C-F content in our GFx, we obtain the optimal x of 0.8, with which the GF0.8 exhibits a very high energy density of 1,073 Wh kg−1 and an excellent power density of 21,460 W kg−1 at a high current density of 10 A g−1. More importantly, our approach opens a new avenue to obtain fluorinated carbon with high energy densities without compromising high power densities.

  12. High-temperature and high-power-density nanostructured thermoelectric generator for automotive waste heat recovery

    International Nuclear Information System (INIS)

    Zhang, Yanliang; Cleary, Martin; Wang, Xiaowei; Kempf, Nicholas; Schoensee, Luke; Yang, Jian; Joshi, Giri; Meda, Lakshmikanth

    2015-01-01

    Highlights: • A thermoelectric generator (TEG) is fabricated using nanostructured half-Heusler materials. • The TE unicouple devices produce superior power density above 5 W/cm"2. • A TEG system with over 1 kW power output is demonstrated by recovering automotive waste heat. - Abstract: Given increasing energy use as well as decreasing fossil fuel sources worldwide, it is no surprise that interest in promoting energy efficiency through waste heat recovery is also increasing. Thermoelectric generators (TEGs) are one of the most promising pathways for waste heat recovery. Despite recent thermoelectric efficiency improvement in nanostructured materials, a variety of challenges have nevertheless resulted in few demonstrations of these materials for large-scale waste heat recovery. Here we demonstrate a high-performance TEG by combining high-efficiency nanostructured bulk materials with a novel direct metal brazing process to increase the device operating temperature. A unicouple device generates a high power density of 5.26 W cm"−"2 with a 500 °C temperature difference between hot and cold sides. A 1 kW TEG system is experimentally demonstrated by recovering the exhaust waste heat from an automotive diesel engine. The TEG system operated with a 2.1% heat-to-electricity efficiency under the average temperature difference of 339 °C between the TEG hot- and cold-side surfaces at a 550 °C exhaust temperature. The high-performance TEG reported here open up opportunities to use TEGs for energy harvesting and power generation applications.

  13. Electromagnetic computations for fusion devices

    International Nuclear Information System (INIS)

    Turner, L.R.

    1989-09-01

    Among the difficulties in making nuclear fusion a useful energy source, two important ones are producing the magnetic fields needed to drive and confine the plasma, and controlling the eddy currents induced in electrically conducting components by changing fields. All over the world, researchers are developing electromagnetic codes and employing them to compute electromagnetic effects. Ferromagnetic components of a fusion reactor introduce field distortions. Eddy currents are induced in the vacuum vessel, blanket and other torus components of a tokamak when the plasma current disrupts. These eddy currents lead to large forces, and 3-D codes are being developed to study the currents and forces. 35 refs., 6 figs

  14. Maximum neutron yeidls in experimental fusion devices

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1979-02-01

    The optimal performances of 12 types of fusion devices are compared with regard to neutron production rate, neutrons per pulse, and fusion energy multiplication, Q/sub p/ (converted to the equivalent value in D-T operation). The record values in all categories are held by the beam-injected tokamak plasma, followed by other beam-target systems. The achieved values of Q/sub p/ for nearly all laboratory plasma fusion devices (magnetically or inertially confined) are found to roughly satisfy a common empirical scaling, Q/sub p/ approx. 10 -6 E/sub in//sup 3/2/, where E/sub in/ is the energy (in kilojoules) injected into the plasma during one or two energy confinement times, or the total energy delivered to the target for inertially confined systems. Fusion energy break-even (Q/sub p/ = 1) in any system apparently requires E/sub in/ approx. 10,000 kJ

  15. Power Requirements Determined for High-Power-Density Electric Motors for Electric Aircraft Propulsion

    Science.gov (United States)

    Johnson, Dexter; Brown, Gerald V.

    2005-01-01

    Future advanced aircraft fueled by hydrogen are being developed to use electric drive systems instead of gas turbine engines for propulsion. Current conventional electric motor power densities cannot match those of today s gas turbine aircraft engines. However, if significant technological advances could be made in high-power-density motor development, the benefits of an electric propulsion system, such as the reduction of harmful emissions, could be realized.

  16. A 380 V High Efficiency and High Power Density Switched-Capacitor Power Converter using Wide Band Gap Semiconductors

    DEFF Research Database (Denmark)

    Fan, Lin; Knott, Arnold; Jørgensen, Ivan Harald Holger

    2018-01-01

    . This paper presents such a high voltage low power switched-capacitor DC-DC converter with an input voltage upto 380 V (compatible with rectified European mains) and an output power experimentally validated up to 21.3 W. The wideband gap semiconductor devices of GaN switches and SiC diodes are combined...... to compose the proposed power stage. Their switching and loss characteristics are analyzed with transient waveforms and thermal images. Different isolated driving circuits are compared and a compact isolated halfbridge driving circuit is proposed. The full-load efficiencies of 98.3% and 97.6% are achieved......State-of-the-art switched-capacitor DC-DC power converters mainly focus on low voltage and/or high power applications. However, at high voltage and low power levels, new designs are anticipated to emerge and a power converter that has both high efficiency and high power density is highly desirable...

  17. Fusion Engineering Device. Volume II. Design description

    International Nuclear Information System (INIS)

    1981-10-01

    This volume summarizes the design of the FED. It includes a description of the major systems and subsystems, the supporting plasma design analysis, a projected device cost and associated construction schedule, and a description of the facilities to house and support the device. This effort represents the culmination of the FY81 studies conducted at the Fusion Engineering Design Center (FEDC). Unique in these design activities has been the collaborative involvement of the Design Center personnel and numerous resource physicists from the fusion community who have made significant contributions in the physics design analysis as well as the physics support of the engineering design of the major FED systems and components

  18. Open-ended fusion devices and reactors

    International Nuclear Information System (INIS)

    Kawabe, T.; Nariai, H.

    1983-01-01

    Conceptual design studies on fusion reactors based upon open-ended confinement schemes, such as the tandem mirror and rf plugged cusp, have been carried out in Japan. These studies may be classified into two categories: near-term devices (Fusion Engineering Test Facility), and long-term fusion power recators. In the first category, a two-component cusp neutron source was proposed. In the second category, the GAMMA-R, a tandem-mirror power reactor, and the RFC-R, an axisymetric mirror and cusp, reactor studies are being conducted at the University of Tsukuba and the Institute of Plasma Physics. Mirror Fusion Engineering Facility parameters and a schematic are shown. The GAMMA-R central-cell design schematic is also shown

  19. Magnetic systems for fusion devices

    International Nuclear Information System (INIS)

    Henning, C.D.

    1985-02-01

    Mirror experiments have led the way in applying superconductivity to fusion research because of unique requirements for high and steady magnetic fields. The first significant applications were Baseball II at LLNL and IMP at ORNL. More recently, the MFTF-B yin-yang coil was successfully tested and the entire tandem configuration is nearing completion. Tokamak magnets have also enjoyed recent success with the large coil project tests at ORNL, preceded by single coil tests in Japan and Germany. In the USSR, the T-7 Tokamak has been operational for many years and the T-15 Tokamak is under construction, with the TF coils nearing completion. Also the Tore Supra is being built in France

  20. Tokamak devices: towards controlled fusion

    International Nuclear Information System (INIS)

    Trocheris, M.

    1975-01-01

    The Tokamak family is from Soviet Union. These devices were exclusively studied at the Kurchatov Institute in Moscow for more than ten years. The first occidental Tokamak started in 1970 at Princeton. The TFR (Tokamak Fontenay-aux-Roses) was built to be superior to the Russian T4. Tokamak future is now represented by the JET (Joint European Tokamak) [fr

  1. Plasma surface interactions in controlled fusion devices

    International Nuclear Information System (INIS)

    Ghendrih, Ph.; Becoulet, M.; Costanzo, L.

    2000-07-01

    This report brings together all the contributions of EURATOM/CEA association to the 14. international conference on plasma surface interactions in controlled fusion devices. 24 papers are presented and they deal mainly with the ergodic divertor and the first wall of Tore-supra tokamak

  2. Data-Acquisition Systems for Fusion Devices

    NARCIS (Netherlands)

    van Haren, P. C.; Oomens, N. A.

    1993-01-01

    During the last two decades, computerized data acquisition systems (DASs) have been applied at magnetic confinement fusion devices. Present-day data acquisition is done by means of distributed computer systems and transient recorders in CAMAC systems. The development of DASs has been technology

  3. Plasma surface interactions in controlled fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Ghendrih, Ph.; Becoulet, M.; Costanzo, L. [and others

    2000-07-01

    This report brings together all the contributions of EURATOM/CEA association to the 14. international conference on plasma surface interactions in controlled fusion devices. 24 papers are presented and they deal mainly with the ergodic divertor and the first wall of Tore-supra tokamak.

  4. Analysis and Design Considerations of a High-Power Density, Dual Air Gap, Axial-Field Brushless, Permanent Magnet Motor.

    Science.gov (United States)

    Cho, Chahee Peter

    1995-01-01

    Until recently, brush dc motors have been the dominant drive system because they provide easily controlled motor speed over a wide range, rapid acceleration and deceleration, convenient control of position, and lower product cost. Despite these capabilities, the brush dc motor configuration does not satisfy the design requirements for the U.S. Navy's underwater propulsion applications. Technical advances in rare-earth permanent magnet materials, in high-power semiconductor transistor technology, and in various rotor position-sensing devices have made using brushless permanent magnet motors a viable alternative. This research investigates brushless permanent magnet motor technology, studying the merits of dual-air gap, axial -field, brushless, permanent magnet motor configuration in terms of power density, efficiency, and noise/vibration levels. Because the design objectives for underwater motor applications include high-power density, high-performance, and low-noise/vibration, the traditional, simplified equivalent circuit analysis methods to assist in meeting these goals were inadequate. This study presents the development and verification of detailed finite element analysis (FEA) models and lumped parameter circuit models that can calculate back electromotive force waveforms, inductance, cogging torque, energized torque, and eddy current power losses. It is the first thorough quantification of dual air-gap, axial -field, brushless, permanent magnet motor parameters and performance characteristics. The new methodology introduced in this research not only facilitates the design process of an axial field, brushless, permanent magnet motor but reinforces the idea that the high-power density, high-efficiency, and low-noise/vibration motor is attainable.

  5. Philosophy and physics of predemonstration fusion devices

    International Nuclear Information System (INIS)

    Clarke, J.F.

    1976-01-01

    A PDFD will operate in the 1980's and must provide the plasma and plasma support technology information necessary to warrant design, construction, and operation of succeeding experimental power reactors and then the demonstration plant. The PDFD must be prototypical of economic fusion devices to justify its cost. Therefore, development of the fusion core will be the focus of the PDFD. The physics performance, power production objectives, and characteristics of the PDFD, and their relationship to the research and development needs to achieve them are outlined. The design criteria for a PDFD which satisfied these constraints will be established

  6. Engineering science research issues in high power density transmission dynamics for aerospace applications. [rotorcraft geared rotors

    Science.gov (United States)

    Singh, Rajendra; Houser, Donald R.

    1993-01-01

    This paper discusses analytical and experimental approaches that will be needed to understand dynamic, vibro-acoustic and design characteristics of high power density rotorcraft transmissions. Complexities associated with mathematical modeling of such systems will be discussed. An overview of research work planned during the next several years will be presented, with emphasis on engineering science issues such as gear contact mechanics, multi-mesh drive dynamics, parameter uncertainties, vibration transmission through bearings, and vibro-acoustic characteristics of geared rotor systems and housing-mount structures. A few examples of work in progress are cited.

  7. Development of superconducting equipment for fusion device

    International Nuclear Information System (INIS)

    Konno, Masayuki; Ueda, Toshio; Hiue, Hisaaki; Ohgushi, Kouzou

    1993-01-01

    At Fuji Electric Co., Ltd., the development of superconductivity was started from 1960, and superconducting equipment for fusion device has been developed for ten years. The superconducting equipment, which is developed for fusion by Fuji Electric Co., Ltd., are able to be grouped in three categories which are current lead, superconducting coil and superconducting bus-line. The current lead is an electrical feeder between a superconducting coil and an electrical power supply. The rated current of developed current lead is 30kA at continuous use and 100kA at short time use respectively. The advanced disk type coil is developed for the toroidal field coil and some coils are developed for critical current measurement. Superconductor is applied to the superconducting bus-line between the superconducting coils and the current leads, and the bus-line is being developed for the Large Helical Device. This report describes an abstract of these equipment. (author)

  8. Arcing phenomena in fusion devices workshop

    International Nuclear Information System (INIS)

    Clausing, R.E.

    1979-01-01

    The workshop on arcing phenomena in fusion devices was organized (1) to review the pesent status of our understanding of arcing as it relates to confinement devices, (2) to determine what informaion is needed to suppress arcing and (3) to define both laboratory and in-situ experiments which can ultimately lead to reduction of impurities in the plasma caused by arcing. The workshop was attended by experts in the area of vacuum arc electrode phenomena and ion source technology, materials scientists, and both theoreticians and experimentalists engaged in assessing the importance of unipolar arcing in today's tokamaks. Abstracts for papers presented at the workshop are included

  9. High Efficiency Hybrid Energy Storage Utilizing High Power Density Ultracapacitors For Long Duration Balloon Flights, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — FastCAP proposes to develop an ultra-high power density and high frequency ultracapacitor capable of surviving over the wide temperature range of -60C to 130C and...

  10. A High Power Density Single-Phase PWM Rectifier With Active Ripple Energy Storage

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Ruxi [Virginia Polytechnic Institute and State University (Virginia Tech); Wang, Fei [ORNL; Boroyevich, Dushan [Virginia Polytechnic Institute and State University (Virginia Tech); Burgos, Rolando [ABB; Lai, Rixin [General Electric; Ning, Puqi [ORNL; Rajashekara, Kaushik [Rolls Royce

    2011-01-01

    It is well known that single-phase pulse width modulation rectifiers have second-order harmonic currents and corresponding ripple voltages on the dc bus. The low-frequency harmonic current is normally filtered using a bulk capacitor in the bus, which results in low power density. However, pursuing high power density in converter design is a very important goal in the aerospace applications. This paper studies methods for reducing the energy storage capacitor for single-phase rectifiers. The minimum ripple energy storage requirement is derived independently of a specific topology. Based on theminimum ripple energy requirement, the feasibility of the active capacitor s reduction schemes is verified. Then, we propose a bidirectional buck boost converter as the ripple energy storage circuit, which can effectively reduce the energy storage capacitance. The analysis and design are validated by simulation and experimental results.

  11. Three-dimensional Core Design of a Super Fast Reactor with a High Power Density

    International Nuclear Information System (INIS)

    Cao, Liangzhi; Oka, Yoshiaki; Ishiwatari, Yuki; Ikejiri, Satoshi; Ju, Haitao

    2010-01-01

    The SuperCritical Water-cooled Reactor (SCWR) pursues high power density to reduce its capital cost. The fast spectrum SCWR, called a super fast reactor, can be designed with a higher power density than thermal spectrum SCWR. The mechanism of increasing the average power density of the super fast reactor is studied theoretically and numerically. Some key parameters affecting the average power density, including fuel pin outer diameter, fuel pitch, power peaking factor, and the fraction of seed assemblies, are analyzed and optimized to achieve a more compact core. Based on those sensitivity analyses, a compact super fast reactor is successfully designed with an average power density of 294.8 W/cm 3 . The core characteristics are analyzed by using three-dimensional neutronics/thermal-hydraulics coupling method. Numerical results show that all of the design criteria and goals are satisfied

  12. ICRF Traveling Wave launcher for fusion devices

    International Nuclear Information System (INIS)

    Ragona, R

    2017-01-01

    Ion Cyclotron Resonance Heating and Current Drive is a method that has the ability to heat directly the ions in the Deuterium-Tritrium fuel to the high temperature needed for the fusion reaction to works. The capability of efficiently couple the Radio Frequency power to the plasma plays a big role in the overall performance of a fusion device. A Traveling Wave Antenna in a resonant ring configuration is a good candidate for an Ion Cyclotron Resonance Heating and Current Drive system. It has the capability to increase the coupled power with respect to present designs and to have a highly selective power spectrum that can be peaked around the maximally absorbed wave. It is also insensitive to the loading variations due to fluctuation of the plasma edge increasing the reliability and the efficiency of the system. It works as a low power density launcher due to the possible large number of current carrying elements. (paper)

  13. Neutral particle kinetics in fusion devices

    International Nuclear Information System (INIS)

    Tendler, M.; Heifetz, D.

    1986-05-01

    The theory of neutral particle kinetics treats the transport of mass, momentum, and energy in a plasma due to neutral particles which themselves are unaffected by magnetic fields. This transport affects the global power and particle balances in fusion devices, as well as profile control and plasma confinement quality, particle and energy fluxes onto device components, performance of pumping systems, and the design of diagnostics and the interpretation of their measurements. This paper reviews the development of analytic, numerical, and Monte Carlo methods of solving the time-independent Boltzmann equation describing neutral kinetics. These models for neutral particle behavior typically use adaptations of techniques developed originally for computing neutron transport, due to the analogy between the two phenomena, where charge-exchange corresponds to scattering and ionization to absorption. Progress in the field depends on developing multidimensional analytic methods, and obtaining experimental data for the physical processes of wall reflection, the neutral/plasma interaction, and for processes in fusion devices which are directly related to neutral transport, such as H/sub α/ emission rates, plenum pressures, and charge-exchange emission spectra

  14. Neutral particle kinetics in fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Tendler, M.; Heifetz, D.

    1986-05-01

    The theory of neutral particle kinetics treats the transport of mass, momentum, and energy in a plasma due to neutral particles which themselves are unaffected by magnetic fields. This transport affects the global power and particle balances in fusion devices, as well as profile control and plasma confinement quality, particle and energy fluxes onto device components, performance of pumping systems, and the design of diagnostics and the interpretation of their measurements. This paper reviews the development of analytic, numerical, and Monte Carlo methods of solving the time-independent Boltzmann equation describing neutral kinetics. These models for neutral particle behavior typically use adaptations of techniques developed originally for computing neutron transport, due to the analogy between the two phenomena, where charge-exchange corresponds to scattering and ionization to absorption. Progress in the field depends on developing multidimensional analytic methods, and obtaining experimental data for the physical processes of wall reflection, the neutral/plasma interaction, and for processes in fusion devices which are directly related to neutral transport, such as H/sub ..cap alpha../ emission rates, plenum pressures, and charge-exchange emission spectra.

  15. Pressure measurements in magnetic-fusion devices

    International Nuclear Information System (INIS)

    Dylla, H.F.

    1981-11-01

    Accurate pressure measurements are important in magnetic fusion devices for: (1) plasma diagnostic measurements of particle balance and ion temperature; (2) discharge cleaning optimization; (3) vacuum system performance; and (4) tritium accountability. This paper reviews the application, required accuracy, and suitable instrumentation for these measurements. Demonstrated uses of ionization-type and capacitance-diaphragm gauges for various pressure and gas-flow measurements in tokamaks are presented, with specific reference to the effects of magnetic fields on gauge performance and the problems associated with gauge calibration

  16. Plasma Surface interaction in Controlled fusion devices

    International Nuclear Information System (INIS)

    1990-05-01

    The subjects presented in the 9th conference on plasma surface interaction in controlled fusion devices were: the modifications of power scrape-off-length and power deposition during various configurations in Tore Supra plasmas; the effects observed in ergodic divertor experiments in Tore-Supra; the diffuse connexion induced by the ergodic divertor and the topology of the heat load patterns on the plasma facing components in Tore-Supra; the study of the influence of air exposure on graphite implanted by low energy high density deuterium plasma

  17. Protector in a nuclear fusion device

    International Nuclear Information System (INIS)

    Furukawa, Masayuki; Yamane, Katsumi; Niwa, Sadahiko; Ogata, Fumio; Masuda, Jun-ichi.

    1975-01-01

    Object: To block an abnormal voltage, which shifts from plasma to coil or power supply by means of action of mutual induction, by a circuit utilizing non-linear impedance elements. Structure: The nuclear fusion device includes a current transformer coil, a vertical field coil and a plasma circuit, with a non-linear impedance element disposed in parallel with at least the current transformer coil, said impedance element being disposed in parallel with a short-circuiting switch, relative to the abnormal voltage moving from the plasma by means of action of mutual induction. (Kamimura, M.)

  18. Pressure measurements in magnetic-fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Dylla, H.F.

    1981-11-01

    Accurate pressure measurements are important in magnetic fusion devices for: (1) plasma diagnostic measurements of particle balance and ion temperature; (2) discharge cleaning optimization; (3) vacuum system performance; and (4) tritium accountability. This paper reviews the application, required accuracy, and suitable instrumentation for these measurements. Demonstrated uses of ionization-type and capacitance-diaphragm gauges for various pressure and gas-flow measurements in tokamaks are presented, with specific reference to the effects of magnetic fields on gauge performance and the problems associated with gauge calibration.

  19. Eddy current analysis in fusion devices

    International Nuclear Information System (INIS)

    Turner, L.R.

    1988-06-01

    In magnetic fusion devices, particularly tokamaks and reversed field pinch (RFP) experiments, time-varying magnetic fields are in intimate contact with electrically conducting components of the device. Induced currents, fields, forces, and torques result. This note reviews the analysis of eddy current effects in the following systems: Interaction of a tokamak plasma with the eddy currents in the first wall, blanket, and shield (FWBS) systems; Eddy currents in a complex but two-dimensional vacuum vessel, as in TFTR, JET, and JT-60; Eddy currents in the FWBS system of a tokamak reactor, such as NET, FER, or ITER; and Eddy currents in a RFP shell. The cited studies are chosen to be illustrative, rather than exhaustive. 42 refs

  20. NASA Glenn Research Center Program in High Power Density Motors for Aeropropulsion

    Science.gov (United States)

    Brown, Gerald V.; Kascak, Albert F.; Ebihara, Ben; Johnson, Dexter; Choi, Benjamin; Siebert, Mark; Buccieri, Carl

    2005-01-01

    Electric drive of transport-sized aircraft propulsors, with electric power generated by fuel cells or turbo-generators, will require electric motors with much higher power density than conventional room-temperature machines. Cryogenic cooling of the motor windings by the liquid hydrogen fuel offers a possible solution, enabling motors with higher power density than turbine engines. Some context on weights of various systems, which is required to assess the problem, is presented. This context includes a survey of turbine engine weights over a considerable size range, a correlation of gear box weights and some examples of conventional and advanced electric motor weights. The NASA Glenn Research Center program for high power density motors is outlined and some technical results to date are presented. These results include current densities of 5,000 A per square centimeter current density achieved in cryogenic coils, finite element predictions compared to measurements of torque production in a switched reluctance motor, and initial tests of a cryogenic switched reluctance motor.

  1. A Cryogenic High-Power-Density Bearingless Motor for Future Electric Propulsion

    Science.gov (United States)

    Choi, Benjamin; Siebert, Mark

    2008-01-01

    The NASA Glenn Research Center (GRC) is developing a high-power-density switched-reluctance cryogenic motor for all-electric and pollution-free flight. However, cryogenic operation at higher rotational speeds markedly shortens the life of mechanical rolling element bearings. Thus, to demonstrate the practical feasibility of using this motor for future flights, a non-contact rotor-bearing system is a crucial technology to circumvent poor bearing life that ordinarily accompanies cryogenic operation. In this paper, a bearingless motor control technology for a 12-8 (12 poles in the stator and 8 poles in the rotor) switched-reluctance motor operating in liquid nitrogen (boiling point, 77 K (-196 C or -321 F)) was presented. We pushed previous disciplinary limits of electromagnetic controller technique by extending the state-of-the-art bearingless motor operating at liquid nitrogen for high-specific-power applications. The motor was levitated even in its nonlinear region of magnetic saturation, which is believed to be a world first for the motor type. Also we used only motoring coils to generate motoring torque and levitation force, which is an important feature for developing a high specific power motor.

  2. Dispersion interferometer for controlled fusion devices

    International Nuclear Information System (INIS)

    Drachev, V.P.; Krasnikov, Yu.I.; Bagryansky, P.A.

    1992-01-01

    A common feature in interferometry is the presence of two independent optical channels. Since wave phase in a medium depends on the geometrical path, polarization and radiation frequency, respectively, one can distinguish three types of interferometric schemes when the channels are geometrically separated, or separation occurs in polarizations or radiation frequencies. We have developed a measurement scheme based on a dispersion interferometer (DI) for plasma diagnostics in the experiments on controlled fusion. DI optical channels have the same geometrical path and are separated in radiation frequency. Use of a common optical path causes the main advantage of the DI technique - low sensitivity to vibrations of optical elements. The use of the DI technique for diagnostics of a laser spark in air and of arc discharges has shown its essential advantages as compared to classical interferometers. Interest in the DI technique from the viewpoint of its application in controlled fusion devices is determined also generated by the possibility of developing a compact multichannel interferometer not requiring a vibration isolation structure. (author) 14 refs., 3 figs

  3. Optimized design of a high-power-density PM-assisted synchronous reluctance machine with ferrite magnets for electric vehicles

    Directory of Open Access Journals (Sweden)

    Liu Xiping

    2017-06-01

    Full Text Available This paper proposes a permanent magnet (PM-assisted synchronous reluctance machine (PMASynRM using ferrite magnets with the same power density as rareearth PM synchronous motors employed in Toyota Prius 2010. A suitable rotor structure for high torque density and high power density is discussed with respect to the demagnetization of ferrite magnets, mechanical strength and torque ripple. Some electromagnetic characteristics including torque, output power, loss and efficiency are calculated by 2-D finite element analysis (FEA. The analysis results show that a high power density and high efficiency of PMASynRM are obtained by using ferrite magnets.

  4. Data acquisition systems for fusion devices

    International Nuclear Information System (INIS)

    Van Haren, P.C.; Oomens, N.A.

    1993-01-01

    During the last two decades, computerized data acquisition systems (DASs) have been applied at magnetic confinement fusion devices. Present-day data acquisition is done by means of distributed computer systems and transient recorders in CAMAC systems. The development of DASs has been technology driven; the emphasis has been on the development of computer hardware and system software. For future DASs, challenging problems are to be solved: The DASs have to be better optimized with respect to the needs of the users. Existing bottlenecks, such as CAMAC-computer coupling or pulse file merging, need to be eliminated. Continuous or long-pulse operation will require the introduction of event abstraction in DAS design. 59 refs., 4 figs., 1 tab

  5. Vacuum vessel for a nuclear fusion device

    International Nuclear Information System (INIS)

    Watanabe, Takashi; Sato, Hiroshi; Owada, Koro.

    1976-01-01

    Object: To provide a reinforcing member on a bellows portion to reduce a stress at the bellows portion thereby increasing the strength of a vessel. Structure: A vacuum vessel for a nuclear fusion device has a bellows portion and a wall thick portion. A support extended toward the bellows portion is secured inside of a toroidal section in order to reduce the stress at the bellows portion. An insulator is interposed between the support and the bellows portion and is retained on the support by a bolt. Since the stress may be reduced by the support, the wall thick of the bellows portion may be decreased to sufficiently secure the low electric resistance value. (Yoshihara, H.)

  6. High-power density miniscale power generation and energy harvesting systems

    Energy Technology Data Exchange (ETDEWEB)

    Lyshevski, Sergey Edward [Department of Electrical and Microelectronics Engineering, Rochester Institute of Technology, Rochester, NY 14623-5603 (United States)

    2011-01-15

    This paper reports design, analysis, evaluations and characterization of miniscale self-sustained power generation systems. Our ultimate objective is to guarantee highly-efficient mechanical-to-electrical energy conversion, ensure premier wind- or hydro-energy harvesting capabilities, enable electric machinery and power electronics solutions, stabilize output voltage, etc. By performing the advanced scalable power generation system design, we enable miniscale energy sources and energy harvesting technologies. The proposed systems integrate: (1) turbine which rotates a radial- or axial-topology permanent-magnet synchronous generator at variable angular velocity depending on flow rate, speed and load, and, (2) power electronic module with controllable rectifier, soft-switching converter and energy storage stages. These scalable energy systems can be utilized as miniscale auxiliary and self-sustained power units in various applications, such as, aerospace, automotive, biotechnology, biomedical, and marine. The proposed systems uniquely suit various submersible and harsh environment applications. Due to operation in dynamic rapidly-changing envelopes (variable speed, load changes, etc.), sound solutions are researched, proposed and verified. We focus on enabling system organizations utilizing advanced developments for various components, such as generators, converters, and energy storage. Basic, applied and experimental findings are reported. The prototypes of integrated power generation systems were tested, characterized and evaluated. It is documented that high-power density, high efficiency, robustness and other enabling capabilities are achieved. The results and solutions are scalable from micro ({proportional_to}100 {mu}W) to medium ({proportional_to}100 kW) and heavy-duty (sub-megawatt) auxiliary and power systems. (author)

  7. High-power density miniscale power generation and energy harvesting systems

    International Nuclear Information System (INIS)

    Lyshevski, Sergey Edward

    2011-01-01

    This paper reports design, analysis, evaluations and characterization of miniscale self-sustained power generation systems. Our ultimate objective is to guarantee highly-efficient mechanical-to-electrical energy conversion, ensure premier wind- or hydro-energy harvesting capabilities, enable electric machinery and power electronics solutions, stabilize output voltage, etc. By performing the advanced scalable power generation system design, we enable miniscale energy sources and energy harvesting technologies. The proposed systems integrate: (1) turbine which rotates a radial- or axial-topology permanent-magnet synchronous generator at variable angular velocity depending on flow rate, speed and load, and, (2) power electronic module with controllable rectifier, soft-switching converter and energy storage stages. These scalable energy systems can be utilized as miniscale auxiliary and self-sustained power units in various applications, such as, aerospace, automotive, biotechnology, biomedical, and marine. The proposed systems uniquely suit various submersible and harsh environment applications. Due to operation in dynamic rapidly-changing envelopes (variable speed, load changes, etc.), sound solutions are researched, proposed and verified. We focus on enabling system organizations utilizing advanced developments for various components, such as generators, converters, and energy storage. Basic, applied and experimental findings are reported. The prototypes of integrated power generation systems were tested, characterized and evaluated. It is documented that high-power density, high efficiency, robustness and other enabling capabilities are achieved. The results and solutions are scalable from micro (∼100 μW) to medium (∼100 kW) and heavy-duty (sub-megawatt) auxiliary and power systems.

  8. Coil supporting device in a nuclear fusion device

    International Nuclear Information System (INIS)

    Takano, Hirohisa; Sasaki, Katsutoki.

    1976-01-01

    Object: To slide a vacuum vessel in the nuclear fusion device and a coil within the vacuum vessel and to mount the coil within the vacuum vessel in a manner that it may not be moved by an electromagnetic force, thereby preventing stress from being produced in the coil. Structure: A coil supporting plate mounted at upper and lower parts prevents damage to an insulation of the coil, said coil being held in a U-shaped groove, and can be moved integral with the coil by the action of a roller bearing with a plurality of needle-like rollers arranged in parallel. The coil supporting plate has a plurality of projections disposed on the lower surface thereof, and flat springs are placed in the projections one over another so that the spring action exerted in the lower plate causes the coil to be resiliently bias in a direction of an electromagnetic force applied thereto and to support the coil. (Yoshino, Y.)

  9. Neutronic analysis of fusion tokamak devices by PHITS

    International Nuclear Information System (INIS)

    Sukegawa, Atsuhiko M.; Takiyoshi, Kouji; Amano, Toshio; Kawasaki, Hiromitsu; Okuno, Koichi

    2011-01-01

    A complete 3D neutronic analysis by PHITS (Particle and Heavy Ion Transport code System) has been performed for fusion tokamak devices such as JT-60U device and JT-60 Superconducting tokamak device (JT-60 Super Advanced). The mono-energetic neutrons (E n =2.45 MeV) of the DD fusion devices are used for the neutron source in the analysis. The visual neutron flux distribution for the estimation of the port streaming and the dose rate around the fusion tokamak devices has been calculated by the PHITS. The PHITS analysis makes it clear that the effect of the port streaming of superconducting fusion tokamak device with the cryostat is crucial and the calculated neutron spectrum results by PHITS agree with the MCNP-4C2 results. (author)

  10. Durability of Low Platinum Fuel Cells Operating at High Power Density

    Energy Technology Data Exchange (ETDEWEB)

    Polevaya, Olga [Nuvera Fuel Cells Inc.; Blanchet, Scott [Nuvera Fuel Cells Inc.; Ahluwalia, Rajesh [Argonne National Lab; Borup, Rod [Los-Alamos National Lab; Mukundan, Rangachary [Los-Alamos National Lab

    2014-03-19

    Understanding and improving the durability of cost-competitive fuel cell stacks is imperative to successful deployment of the technology. Stacks will need to operate well beyond today’s state-of-the-art rated power density with very low platinum loading in order to achieve the cost targets set forth by DOE ($15/kW) and ultimately be competitive with incumbent technologies. An accelerated cost-reduction path presented by Nuvera focused on substantially increasing power density to address non-PGM material costs as well as platinum. The study developed a practical understanding of the degradation mechanisms impacting durability of fuel cells with low platinum loading (≤0.2mg/cm2) operating at high power density (≥1.0W/cm2) and worked out approaches for improving the durability of low-loaded, high-power stack designs. Of specific interest is the impact of combining low platinum loading with high power density operation, as this offers the best chance of achieving long-term cost targets. A design-of-experiments approach was utilized to reveal and quantify the sensitivity of durability-critical material properties to high current density at two levels of platinum loading (the more conventional 0.45 mgPt.cm–1 and the much lower 0.2 mgPt.cm–2) across several cell architectures. We studied the relevance of selected component accelerated stress tests (AST) to fuel cell operation in power producing mode. New stress tests (NST) were designed to investigate the sensitivity to the addition of electrical current on the ASTs, along with combined humidity and load cycles and, eventually, relate to the combined city/highway drive cycle. Changes in the cathode electrochemical surface area (ECSA) and average oxygen partial pressure on the catalyst layer with aging under AST and NST protocols were compared based on the number of completed cycles. Studies showed elevated sensitivity of Pt growth to the potential limits and the initial particle size distribution. The ECSA loss

  11. Magnetic field coil in nuclear fusion device

    International Nuclear Information System (INIS)

    Yamaguchi, Mitsugi; Takano, Hirohisa.

    1975-01-01

    Object: To provide an electrical-insulatively stabilized magnetic field coil in nuclear fusion device, restraining an increase in voltage when plasma current is rapidly changed. Structure: A magnetic field coil comprises coils arranged coaxial with respective vacuum vessels, said coils being wound in positive and reverse polarities so as to form a vertical magnetic field within the plasma. The coils of the positive polarity are arranged along the vacuum vessel inside of an axis vertical in section of the annular plasma and are arranged symmetrically up and down of a horizontal axis. On the other hand, the coils of the reverse polarity are arranged along the vacuum vessel outside of a vertical axis and arranged symmetrically up and down of the horizontal axis. These positive and reverse polarity coils are alternately connected in series, and lead portions of the coils are connected to a power source by means of connecting wires. In this case, lead positions of the coils are arranged in one direction, and the connecting wires are disposed in closely contact relation to offset magnetic fields formed by the connecting wires each other. (Kawakami, Y.)

  12. Use of high current density superconducting coils in fusion devices

    International Nuclear Information System (INIS)

    Green, M.A.

    1979-11-01

    Superconducting magnets will play an important role in fusion research in years to come. The magnets which are currently proposed for fusion research use the concept of cryostability to insure stable operation of the superconducting coils. This paper proposes the use of adiabatically stable high current density superconducting coils in some types of fusion devices. The advantages of this approach are much lower system cold mass, enhanced cryogenic safety, increased access to the plasma and lower cost

  13. Fusion Engineering Device. Volume 1. Mission and program summary

    International Nuclear Information System (INIS)

    1981-10-01

    This volume presents, in summary form, a recommended approach to implementing the Magnetic Fusion Energy Engineering Act of 1980. These recommendations constitute the findings of the FED Technical Management Board (TMB). The TMB and the affiliated technical managers gave particular scrutiny to elucidating the role of FED in fusion development and to defining the device mission

  14. Development of Optimized Core Design and Analysis Methods for High Power Density BWRs

    Science.gov (United States)

    Shirvan, Koroush

    Increasing the economic competitiveness of nuclear energy is vital to its future. Improving the economics of BWRs is the main goal of this work, focusing on designing cores with higher power density, to reduce the BWR capital cost. Generally, the core power density in BWRs is limited by the thermal Critical Power of its assemblies, below which heat removal can be accomplished with low fuel and cladding temperatures. The present study investigates both increases in the heat transfer area between ~he fuel and coolant and changes in operating parameters to achieve higher power levels while meeting the appropriate thermal as well as materials and neutronic constraints. A scoping study is conducted under the constraints of using fuel with cylindrical geometry, traditional materials and enrichments below 5% to enhance its licensability. The reactor vessel diameter is limited to the largest proposed thus far. The BWR with High power Density (BWR-HD) is found to have a power level of 5000 MWth, equivalent to 26% uprated ABWR, resulting into 20% cheaper O&M and Capital costs. This is achieved by utilizing the same number of assemblies, but with wider 16x16 assemblies and 50% shorter active fuel than that of the ABWR. The fuel rod diameter and pitch are reduced to just over 45% of the ABWR values. Traditional cruciform form control rods are used, which restricts the assembly span to less than 1.2 times the current GE14 design due to limitation on shutdown margin. Thus, it is possible to increase the power density and specific power by 65%, while maintaining the nominal ABWR Minimum Critical Power Ratio (MCPR) margin. The plant systems outside the vessel are assumed to be the same as the ABWR-Il design, utilizing a combination of active and passive safety systems. Safety analyses applied a void reactivity coefficient calculated by SIMULA TE-3 for an equilibrium cycle core that showed a 15% less negative coefficient for the BWR-HD compared to the ABWR. The feedwater

  15. Health physics appraisal guidelines for fusion/confinement devices

    International Nuclear Information System (INIS)

    Neeson, P.M.

    1987-01-01

    Several types of fusion/confinement devices have been developed for a variety of research applications. The health physics considerations for these devices can vary, depending on a number of parameters. This paper presents guidelines for health physics appraisal of such devices, which can be tailored to apply to specific systems. The guidelines can also be useful for establishing ongoing health physics programs for safe operation of the devices

  16. Process and device for energy production from thermonuclear fusion reactions

    International Nuclear Information System (INIS)

    Bussard, R.W.; Coppi, Bruno.

    1977-01-01

    An energy generating system is described using a fusion reaction. It includes several contrivances for confining a plasma in an area, a protective device around a significant part of each of these confinement contrivances, an appliance for introducing a fusion reaction fuel in each of the confinements so that the plasma may be formed. Each confinement can be separated from the protective device so that it may be replaced by another. The system is connected to the confinements, to the protective devices or to both. It enables the thermal energy to be extracted and transformed into another form, electric, mechanical or both [fr

  17. Particle and impurity control in toroidal fusion devices

    International Nuclear Information System (INIS)

    Wootton, A.J.

    1986-01-01

    A review of working particle and impurity control techniques used in and proposed for magnetic fusion devices is presented. The requirements of both present-day machines and envisaged fusion reactors are considered. The various techniques which have been proposed are characterized by whether they affect sources, sinks, or fluxes; in many cases a particular method or device can appear in more than one category. Examples are drawn from published results. The solutions proposed for the large devices which will be operating during the next 5 years are discussed

  18. Improved thermal/MHD design of self-cooled blankets for high-power-density fusion reactors

    International Nuclear Information System (INIS)

    Sedehi, S.; Lund, K.O.

    1986-01-01

    In this work, an improved self-cooled blanket design is conceived that seeks to minimize the induced current and pressure loss, while maintaining effective cooling and power output. Standard solutions for fully developed MHD flows in rectangular ducts are utilized to describe the magnetic pressure drop in rectangular ducts in terms of the duct aspects ratio. A newly available analytical result for developing and fully developed temperatures is utilized in determining the maximum wall temperature and outlet temperature. Based on results from rectangular ducts, improved annular-type duct designs are proposed and evaluated. The results from the rectangular duct analysis indicate reduced pressure drop and increased thermal performance for large aspect ratio (ratio of duct width in the toroidal B-field direction to width normal to B-field). An infinite aspect ratio occurs for the annular duct design and it is shown that this configuration has superior characteristics as a self-cooled blanket design concept

  19. High power density dc/dc converter: Selection of converter topology

    Science.gov (United States)

    Divan, Deepakraj M.

    1990-01-01

    The work involved in the identification and selection of a suitable converter topology is described. Three new dc/dc converter topologies are proposed: Phase-Shifted Single Active Bridge DC/DC Converter; Single Phase Dual Active Bridges DC/DC Converter; and Three Phase Dual Active Bridges DC/DC Converter (Topology C). The salient features of these topologies are: (1) All are minimal in structure, i.e., each consists of an input and output bridge, input and output filter and a transformer, all components essential for a high power dc/dc conversion process; (2) All devices of both the bridges can operate under near zero-voltage conditions, making possible a reduction of device switching losses and hence, an increase in switching frequency; (3) All circuits operate at a constant frequency, thus simplifying the task of the magnetic and filter elements; (4) Since, the leakage inductance of the transformer is used as the main current transfer element, problems associated with the diode reverse recovery are eliminated. Also, this mode of operation allows easy paralleling of multiple modules for extending the power capacity of the system; (5) All circuits are least sensitive to parasitic impedances, infact the parasitics are efficently utilized; and (6) The soft switching transitions, result in low electromagnetic interference. A detailed analysis of each topology was carried out. Based on the analysis, the various device and component ratings for each topology operating at an optimum point, and under the given specifications, are tabulated and discussed.

  20. ADX: A high Power Density, Advanced RF-Driven Divertor Test Tokamak for PMI studies

    Science.gov (United States)

    Whyte, Dennis; ADX Team

    2015-11-01

    The MIT PSFC and collaborators are proposing an advanced divertor experiment, ADX; a divertor test tokamak dedicated to address critical gaps in plasma-material interactions (PMI) science, and the world fusion research program, on the pathway to FNSF/DEMO. Basic ADX design features are motivated and discussed. In order to assess the widest range of advanced divertor concepts, a large fraction (>50%) of the toroidal field volume is purpose-built with innovative magnetic topology control and flexibility for assessing different surfaces, including liquids. ADX features high B-field (>6 Tesla) and high global power density (P/S ~ 1.5 MW/m2) in order to access the full range of parallel heat flux and divertor plasma pressures foreseen for reactors, while simultaneously assessing the effect of highly dissipative divertors on core plasma/pedestal. Various options for efficiently achieving high field are being assessed including the use of Alcator technology (cryogenic cooled copper) and high-temperature superconductors. The experimental platform would also explore advanced lower hybrid current drive and ion-cyclotron range of frequency actuators located at the high-field side; a location which is predicted to greatly reduce the PMI effects on the launcher while minimally perturbing the core plasma. The synergistic effects of high-field launchers with high total B on current and flow drive can thus be studied in reactor-relevant boundary plasmas.

  1. Three equipment concepts for the Fusion Engineering Device

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Masson, L.S.; Watts, K.D.; Grant, N.R.; Kuban, D.P.

    1982-01-01

    Maintenance equipment which is needed to remotely handle fusion device components is being conceptually developed for the Fusion Engineering Design Center. This will test the assumption that these equipment needs can be satisfied by present technology. In addition, the development of equipment conceptual designs will allow for cost estimates which have a much higher degree of certainty. Accurate equipment costs will be useful for assessments which trade off gains in availability as a function of increased investments in maintenance equipment

  2. Database for fusion devices and associated fuel systems

    International Nuclear Information System (INIS)

    Woolgar, P.W.

    1983-03-01

    A computerized database storage and retrieval system has been set up for fusion devices and the associated fusion fuel systems which should be a useful tool for the CFFTP program and other users. The features of the Wang 'Alliance' system are discussed for this application, as well as some of the limitations of the system. Recommendations are made on the operation, upkeep and further development that should take place to implement and maintain the system

  3. Helical-type device and laser fusion. Rivals for tokamak-type device at n-fusion development in Japan

    International Nuclear Information System (INIS)

    Anon.

    1994-01-01

    Under the current policy on the research and development of nuclear fusion in Japan, as enunciated by the Atomic Energy Commission of Japan, the type of a prototype fusion reactor will be chosen after 2020 from tokamak, helical or some other type including the inertial confinement fusion using lasers. A prototype fusion reactor is the next step following the tokamak type International Thermonuclear Experimental Reactor (ITER). With the prototype reactor, the feasibility as a power plant will be examined. At present the main research and development of nuclear fusion in Japan are on tokamak type, which have been promoted by Japan Atomic Energy Research Institute (JAERI). As for the other types of nuclear fusion, researches have been carried out on the helical type in Kyoto University and National Institute for Fusion Science (NIFS), the mirror type in Tsukuba University, the tokamak type using superconductive coils in Kyushu University, and the laser fusion in Osaka University. The features and the present state of research and development of the Large Helical Device and the laser fusion which is one step away from the break-even condition are reported. (K.I.)

  4. Institute for Fusion Research and Large Helical Device program

    International Nuclear Information System (INIS)

    Iiyoshi, Atsuo

    1989-01-01

    In the research on nuclear fusion, the final objective is to materialize nuclear fusion reactors, and for the purpose, it is necessary to cause nuclear combustion by making the plasma of higher than 100 million deg and confine it for a certain time. So far in various universities, the researches on diversified fusion processes have been advanced, but in February, 1986, the Science Council issued the report 'Nuclear fusion research in universities hereafter'. As the next large scale device, an external conductor system helical device was decided, and it is desirable to found the organization for joint utilization by national universities to promote the project. The researches on the other processes are continued by utilizing the existing facilitie. The reason of selecting a helical device is the data base of the researches carried out so far can be utilized sufficiently, it is sufficiently novel even after 10 years from now, and many researchers can be collected. The place of the research is Toki City, Gifu Prefecture, where the Institute of Plasma Physics, Nagoya University, is to be moved. The basic concept of the superconducting helical device project, the trend of nuclear fusion development in the world, the physical research using a helical system and so on are reported. (Kako, I.)

  5. Coil for a nuclear fusion device

    International Nuclear Information System (INIS)

    Kadotani, Kenzo.

    1975-01-01

    Object: To provide a thin nuclear fusion coil having good thermal insulation and insulating properties in which mica and glass materials are wound round conductors subjected to varnish treatment and hardened, which is then sealed into a metallic case along with negative gases of more than two atmospheric pressures. Structure: A plurality of conductors impregnated with varnish are hardened by a rare insulating layer, after which it is coated with a layer of mica not impregnated with varnish and a layer of glass substance and is then received into a metallic case and filled under pressure with negative gases at a pressure more than two atmospheric pressures. (Kamimura, M.)

  6. Extremely low recycling and high power density handling in CDX-U lithium experiments

    International Nuclear Information System (INIS)

    Kaita, R.; Majeski, R.; Doerner, R.; Gray, T.; Kugel, H.; Lynch, T.; Maingi, R.; Mansfield, D.; Soukhanovskii, V.; Spaleta, J.; Timberlake, J.; Zakharov, L.

    2007-01-01

    The mission of the Current Drive eXperiment-Upgrade (CDX-U) spherical tokamak is to investigate lithium as a plasma-facing component (PFC). The latest CDX-U experiments used a combination of a toroidal liquid lithium limiter and lithium wall coatings applied between plasma shots. Recycling coefficients for these plasmas were deduced to be 30% or below, and are the lowest ever observed in magnetically-confined plasmas. The corresponding energy confinement times showed nearly a factor of six improvement over discharges without lithium PFC's. An electron beam (e-beam) for evaporating lithium from the toroidal limiter was one of the techniques used to create lithium wall coatings in CDX-U. The evaporation was not localized to the e-beam spot, but occurred only after the entire volume of lithium in toroidal limiter was liquefied. This demonstration of the ability of lithium to handle high heat loads can have significant consequences for PFC's in future burning plasma devices

  7. Initial trade and design studies for the fusion engineering device

    International Nuclear Information System (INIS)

    Flanagan, C.A.; Steiner, D.; Smith, G.E.

    1981-06-01

    The Magnetic Fusion Energy Engineering Act of 1980 calls for the operation of a Fusion Engineering Device (FED) by 1990. It is the intent of the Act that the FED, in combination with other testing facilities, will establish the engineering feasibility of magnetic fusion energy. The Fusion Engineering Design Center (FEDC), under the guidance of a Technical Management Board (TMB), initiated a program of trade and design studies in October 1980 to support the selection of the FED concept. This document presents the results of these initial trade and design studies. Based on these results, a baseline configuration has been identified and the Design Center effort for the remainder of the fiscal year will be devoted to the development of a self-consistent FED design description

  8. Local wall power loading variations in thermonuclear fusion devices

    International Nuclear Information System (INIS)

    Carroll, M.C.; Miley, G.H.

    1989-01-01

    A 2 1/2-dimensional geometric model is presented that allows calculation of power loadings at various points on the first wall of a thermonuclear fusion device. Given average wall power loadings for brems-strahlung, cyclotron radiation charged particles, and neutrons, which are determined from various plasma-physics computation models, local wall heat loads are calculated by partitioning the plasma volume and surface into cells and superimposing the heating effects of the individual cells on selected first-wall differential areas. Heat loads from the entire plasma are thus determined as a function of position on the first-wall surface. Significant differences in local power loadings were found for most fusion designs, and it was therefore concluded that the effect of local power loading variations must be taken into account when calculating temperatures and heat transfer rates in fusion device first walls

  9. High Power Density Motors

    Science.gov (United States)

    Kascak, Daniel J.

    2004-01-01

    With the growing concerns of global warming, the need for pollution-free vehicles is ever increasing. Pollution-free flight is one of NASA's goals for the 21" Century. , One method of approaching that goal is hydrogen-fueled aircraft that use fuel cells or turbo- generators to develop electric power that can drive electric motors that turn the aircraft's propulsive fans or propellers. Hydrogen fuel would likely be carried as a liquid, stored in tanks at its boiling point of 20.5 K (-422.5 F). Conventional electric motors, however, are far too heavy (for a given horsepower) to use on aircraft. Fortunately the liquid hydrogen fuel can provide essentially free refrigeration that can be used to cool the windings of motors before the hydrogen is used for fuel. Either High Temperature Superconductors (HTS) or high purity metals such as copper or aluminum may be used in the motor windings. Superconductors have essentially zero electrical resistance to steady current. The electrical resistance of high purity aluminum or copper near liquid hydrogen temperature can be l/lOO* or less of the room temperature resistance. These conductors could provide higher motor efficiency than normal room-temperature motors achieve. But much more importantly, these conductors can carry ten to a hundred times more current than copper conductors do in normal motors operating at room temperature. This is a consequence of the low electrical resistance and of good heat transfer coefficients in boiling LH2. Thus the conductors can produce higher magnetic field strengths and consequently higher motor torque and power. Designs, analysis and actual cryogenic motor tests show that such cryogenic motors could produce three or more times as much power per unit weight as turbine engines can, whereas conventional motors produce only 1/5 as much power per weight as turbine engines. This summer work has been done with Litz wire to maximize the current density. The current is limited by the amount of heat it generates. By increasing the heat transfer out of the wire, the wires can carry a larger current and therefore produce more force. This was done by increasing the surface area of the wire to allow more coolant to flow over it. Litz wire was used because it can carry high frequency current. It also can be deformed into configurations that would increase the surface area. The best configuration was determined by heat transfer and force plots that were generated using Maxwell 2D. Future work will be done by testing and measuring the thrust force produced by the wires in a magnetic field.

  10. Interfacing between concrete and steel construction and fusion research devices

    International Nuclear Information System (INIS)

    Willoughby, E.

    1981-01-01

    In 1976 Giffels Associates, Inc. an architect/engineer organization, was retained by the United States Department of Energy to provide Title I and Title II design services and Title III construction inspection services for the Tokamak Fusion Test Reactor now being installed at the Princeton Plasma Physics Laboratory in Princeton, New Jersey. Construction of the complex required to house and serve the reactor itself, designed by others, now commencing. During building construction several problems occurred with respect to the interface between the building design, construction and the fusion device (reactor). A brief description of some of these problems and related factors is presented, which may be of benefit to those persons active in continuing fusion research and experimental work

  11. Progress of research and development of nuclear fusion and development of large nuclear fusion device technology

    International Nuclear Information System (INIS)

    1994-01-01

    In the last several years, the results of tokamak experiments were conspicuous, and the progress of plasma confinement performance, transport mechanism, divertors and impurities, helium transport and exhaust, electric current drive, magnetic field ripple effect and high speed particle transport and DT experiment are reported. The other confinement methods than tokamak, the related theories and reactor technology are described. The conceptual design of ITER was carried out by the cooperation of Japan, USA, EC and the former USSR. The projects of developing nuclear fusion in various countries, the design and the required research and development of ITER, the reconstruction and the required research and development of JT-60, JET and TFTR, the design and the required research and development of large helical device, the state of research and development of laser nuclear fusion and inversion magnetic field pinch nuclear fusion, the activities and roles of industrial circles in large nuclear fusion device technology, and the long term perspective of the technical development of nuclear fusion are described. (K.I.)

  12. Numerical Experiments Providing New Insights into Plasma Focus Fusion Devices

    Directory of Open Access Journals (Sweden)

    Sing Lee

    2010-04-01

    Full Text Available Recent extensive and systematic numerical experiments have uncovered new insights into plasma focus fusion devices including the following: (1 a plasma current limitation effect, as device static inductance is reduced towards very small values; (2 scaling laws of neutron yield and soft x-ray yield as functions of storage energies and currents; (3 a global scaling law for neutron yield as a function of storage energy combining experimental and numerical data showing that scaling deterioration has probably been interpreted as neutron ‘saturation’; and (4 a fundamental cause of neutron ‘saturation’. The ground-breaking insights thus gained may completely change the directions of plasma focus fusion research.

  13. Final Technical Report for "Nuclear Technologies for Near Term Fusion Devices"

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, Paul P.H. [Univ. of Wisconsin, Madison, WI (United States); Sawan, Mohamed E. [Univ. of Wisconsin, Madison, WI (United States); Davis, Andrew [Univ. of Wisconsin, Madison, WI (United States); Bohm, Tim D. [Univ. of Wisconsin, Madison, WI (United States)

    2017-09-05

    Over approximately 18 years, this project evolved to focus on a number of related topics, all tied to the nuclear analysis of fusion energy systems. For the earliest years, the University of Wisconsin (UW)’s effort was in support of the Advanced Power Extraction (APEX) study to investigate high power density first wall and blanket systems. A variety of design concepts were studied before this study gave way to a design effort for a US Test Blanket Module (TBM) to be installed in ITER. Simultaneous to this TBM project, nuclear analysis supported the conceptual design of a number of fusion nuclear science facilities that might fill a role in the path to fusion energy. Beginning in approximately 2005, this project added a component focused on the development of novel radiation transport software capability in support of the above nuclear analysis needs. Specifically, a clear need was identified to support neutron and photon transport on the complex geometries associated with Computer-Aided Design (CAD). Following the initial development of the Direct Accelerated Geoemtry Monte Carlo (DAGMC) capability, additional features were added, including unstructured mesh tallies and multi-physics analysis such as the Rigorous 2-Step (R2S) methodology for Shutdown Dose Rate (SDR) prediction. Throughout the project, there were also smaller tasks in support of the fusion materials community and for the testing of changes to the nuclear data that is fundamental to this kind of nuclear analysis.

  14. Impurity studies in fusion devices using laser-fluorescence-spectroscopy

    International Nuclear Information System (INIS)

    Husinsky, W.R.

    1980-08-01

    Resonance fluorescence excitation of neutral atoms using tunable radiation from dye lasers offers a number of unique advantages for impurity studies in fusion devices. Using this technique, it is possible to perform local, time-resolved measurements of the densities and velocity distributions of metallic impurities in fusion devices without disturbing the plasma. Velocities are measured by monitoring the fluorescence intensity while tuning narrow bandwidth laser radiation through the Doppler - broadened absorbtion spectrum of the transition. The knowledge of the velocity distribution of neutral impurities is particularly useful for the determination of impurity introduction mechanisms. The laser fluorescence technique will be described in terms of its application to metallic impurities in fusion devices and related laboratory experiments. Particular attention will be given to recent results from the ISX-B tokamak using pulsed dye lasers where detection sensitivities for neutral Fe of 10 6 atoms/cm 3 with a velocity resolution of 600 m/sec (0.1 eV) have been achieved. Techniques for exciting plasma particles (H,D) will also be discussed

  15. NOx, Soot, and Fuel Consumption Predictions under Transient Operating Cycle for Common Rail High Power Density Diesel Engines

    Directory of Open Access Journals (Sweden)

    N. H. Walke

    2016-01-01

    Full Text Available Diesel engine is presently facing the challenge of controlling NOx and soot emissions on transient cycles, to meet stricter emission norms and to control emissions during field operations. Development of a simulation tool for NOx and soot emissions prediction on transient operating cycles has become the most important objective, which can significantly reduce the experimentation time and cost required for tuning these emissions. Hence, in this work, a 0D comprehensive predictive model has been formulated with selection and coupling of appropriate combustion and emissions models to engine cycle models. Selected combustion and emissions models are further modified to improve their prediction accuracy in the full operating zone. Responses of the combustion and emissions models have been validated for load and “start of injection” changes. Model predicted transient fuel consumption, air handling system parameters, and NOx and soot emissions are in good agreement with measured data on a turbocharged high power density common rail engine for the “nonroad transient cycle” (NRTC. It can be concluded that 0D models can be used for prediction of transient emissions on modern engines. How the formulated approach can also be extended to transient emissions prediction for other applications and fuels is also discussed.

  16. Safety considerations in the design of the fusion engineering device

    International Nuclear Information System (INIS)

    Barrett, R.J.

    1983-01-01

    Safety considerations play a significant role in the design of a near-term Fusion Engineering Device (FED). For the safety of the general public and the plant workers, the radiation environment caused by the reacting plasma and the potential release of tritium fuel are the dominant considerations. The U.S. Department of Energy (DOE) regulations and guidelines for radiation protection have been reviewed and are being applied to the device design. Direct radiation protection is provided by the device shield and the reactor building walls. Radiation from the activated device components and the tritium fuel is to be controlled with shielding, contamination control, and ventilation. The potential release of tritium from the plant has influenced the selection of reactor building and plant designs and specifications. The safety of the plant workers is affected primarily by the radiation from the activated device components and from plasma chamber debris. The highly activated device components make it necessary to design many of the maintenance activities in the reactor building for totally remote operation. The hot cell facility has evolved as a totally remote maintenance facility due to the high radiation levels of the device components. Safety considerations have had substantial impacts on the design of FED. Several examples of safety-related design impacts are discussed in the paper. Feasible solutions have been identified for all outstanding safety-related items, and additional optimization of these solutions is anticipated in future design studies

  17. Development of new low activation aluminum alloys for fusion devices

    International Nuclear Information System (INIS)

    Kamada, Kohji; Kakihana, Hidetake.

    1985-01-01

    As the materials for the R facility (a tokamak nuclear fusion device in the R project intended for D-T burning) in the Institute of Plasma Physics, Nagoya University, Al-4 % Mg-0.2 % Bi (5083 improved type) and Al-4 % Mg-1 % Li, aimed at low radioactivability, high electric resistance and high strength, have been developed. The results of the nuclear properties evaluation with 14 MeV neutrons and of the measurements of electric resistance and mechanical properties were satisfactory. The possibility of producing large Al-4 % Mg-1 % Li plate (1 m x 2 m x 25 mm) in the existing factory was confirmed, with the properties retained. The electric resistances were higher than those in the conventional aluminum alloys, and still with feasibility for the further improvement. General properties of the fusion aluminum alloys and the 26 Al formation in (n, 2n) reaction were studied. (Mori, K.)

  18. Overview of the Fusion Engineering Device (FED) design

    International Nuclear Information System (INIS)

    Steiner, D.; Flanagan, C.A.

    1981-01-01

    The device has a major radius of 5.0 m with a plasma minor radius of 1.3 m elongated by 1.6. Capability is provided for operating the toroidal field coils up to 10 T, but the bulk of the operations are designed for 8 T. At 8-T conditions the fusion power is approx. 180 MW (neutron wall loading approx. 0.4 MW/m 2 ) and a plasma Q of approx. 5 is expected. At 10-T conditions, which are expected to be limited to about 10% of the total operations, the fusion power is approx. 450 MW (approx. 1.0 MW/m 2 ) and ignition is expected

  19. Overview of the fusion engineering device (FED) design

    International Nuclear Information System (INIS)

    Steiner, D.; Flanagan, C.A.

    1981-10-01

    The device has a major radius of 5.0 m with a plasma minor radius of 1.3 m elongated by 1.6. Capability is provided for operating the toroidal field coils up to 10 T, but the bulk of the operations are designed for 8 T. At 8-T conditions, the fusion power is approx. 180 MW (neutron wall loading approx. 0.4 MW/m 2 ) and a plasma Q of approx. 5 is expected. At 10-T conditions, which are expected to be limited to about 10% of the total operations, the fusion power is approx. 450 MW (approx. 1.0 MW/m 2 ) and ignition is expected

  20. Railgun pellet injection system for fusion experimental devices

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, M. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan). Adv. Tech. Dev. Dept.; Oda, Y. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan). Adv. Tech. Dev. Dept.; Azuma, K. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan). Adv. Tech. Dev. Dept.; Satake, K. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan). Adv. Tech. Dev. Dept.; Kasai, S. [Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun 319-11 (Japan); Hasegawa, K. [Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun 319-11 (Japan)

    1995-11-01

    A railgun pellet injection system has been developed for fusion experimental devices. Using a low electric energy railgun system, hydrogen pellet acceleration tests have been conducted to investigate the application of the electromagnetic railgun system for high speed pellet injection into fusion plasmas. In the system, the pellet is pre-accelerated before railgun acceleration. A laser beam is used to induce plasma armature. The ignited plasma armature is accelerated by an electromagnetic force that accelerates the pellet. Under the same operational conditions, the energy conversion coefficient for the dummy pellets was around 0.4%, while that for the hydrogen pellets was around 0.12%. The highest hydrogen pellet velocity was 1.4 km s{sup -1} using a 1 m long railgun. Based on the findings, it is estimated that the hydrogen pellet has the potential to be accelerated to 5 km s{sup -1} using a 3 m long railgun. (orig.).

  1. Railgun pellet injection system for fusion experimental devices

    International Nuclear Information System (INIS)

    Onozuka, M.; Hasegawa, K.

    1995-01-01

    A railgun pellet injection system has been developed for fusion experimental devices. Using a low electric energy railgun system, hydrogen pellet acceleration tests have been conducted to investigate the application of the electromagnetic railgun system for high speed pellet injection into fusion plasmas. In the system, the pellet is pre-accelerated before railgun acceleration. A laser beam is used to induce plasma armature. The ignited plasma armature is accelerated by an electromagnetic force that accelerates the pellet. Under the same operational conditions, the energy conversion coefficient for the dummy pellets was around 0.4%, while that for the hydrogen pellets was around 0.12%. The highest hydrogen pellet velocity was 1.4 km s -1 using a 1 m long railgun. Based on the findings, it is estimated that the hydrogen pellet has the potential to be accelerated to 5 km s -1 using a 3 m long railgun. (orig.)

  2. Superconducting (radiation hardened) magnets for mirror fusion devices

    International Nuclear Information System (INIS)

    Henning, C.D.; Dalder, E.N.C.; Miller, J.R.; Perkins, J.R.

    1983-01-01

    Superconducting magnets for mirror fusion have evolved considerably since the Baseball II magnet in 1970. Recently, the Mirror Fusion Test Facility (MFTF-B) yin-yang has been tested to a full field of 7.7 T with radial dimensions representative of a full scale reactor. Now the emphasis has turned to the manufacture of very high field solenoids (choke coils) that are placed between the tandem mirror central cell and the yin-yang anchor-plug set. For MFTF-B the choke coil field reaches 12 T, while in future devices like the MFTF-Upgrade, Fusion Power Demonstration and Mirror Advanced Reactor Study (MARS) reactor the fields are doubled. Besides developing high fields, the magnets must be radiation hardened. Otherwise, thick neutron shields increase the magnet size to an unacceptable weight and cost. Neutron fluences in superconducting magnets must be increased by an order of magnitude or more. Insulators must withstand 10 10 to 10 11 rads, while magnet stability must be retained after the copper has been exposed to fluence above 10 19 neutrons/cm 2

  3. Development and application of charcoal sorbents for cryopumping fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Sedgley, D.W. (Grumman Corp., Bethpage, NY (USA). Space Systems Div.)

    1989-06-01

    Progress has been made in defining the capabilities of charcoal as the most promising absorbent to be used in cryopumps for fusion power application. The capabilities of alternative methods of cryopumping helium have been examined in a literature survey and by test, and the results are described here. Considerations include pumping speed, capacity to accumulate pumped gas, ease of reconditioning, use of alternative materials and tolerance to the fusion environment. Vacuum pumps for future fusion devices must handle large quantities of helium/hydrogen isotopes and other impurities. Cryopumps or turbomolecular pumps have demonstrated the capability on a small scale, and each has an important advantage: TMPs do not accumulate gases; cryopumps can separate helium from other effluents. This paper includes a review of a method for selecting charcoals for helium cryopumping, testing of a continuously operating cryopump system, and definition of a design that is based on the requirements of the Next European Torus. Tritium limits are satisfied. The pump design incorporates the charcoal sorbent system that has been recently developed and is based on a reasonable extrapolation of current state-of-the-art. Evaluation of alternative methods of separating helium and other gases led to selection of a movable barrier as the preferred solution. (orig.).

  4. Mirror Fusion Test Facility: an intermediate device to a mirror fusion reactor

    International Nuclear Information System (INIS)

    Karpenko, V.N.

    1983-01-01

    The Mirror Fusion Test Facility (MFTF-B) now under construction at Lawrence Livermore National Laboratory represents more than an order-of-magnitude step from earlier magnetic-mirror experiments toward a future mirror fusion reactor. In fact, when the device begins operating in 1986, the Lawson criteria of ntau = 10 14 cm -3 .s will almost be achieved for D-T equivalent operation, thus signifying scientific breakeven. Major steps have been taken to develop MFTF-B technologies for tandem mirrors. Steady-state, high-field, superconducting magnets at reactor-revelant scales are used in the machine. The 30-s beam pulses, ECRH, and ICRH will also introduce steady-state technologies in those systems

  5. Electrical insulation and conduction coating for fusion experimental devices

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, Masanori; Tsujimura, Seiji; Toyoda, Masahiko; Inoue, Masahiko [Mitsubishi Heavy Industries, Ltd., Yokohama (Japan); Abe, Tetsuya; Murakami, Yoshio [Japan Atomic Energy Research Inst., Naka (Japan)

    1996-01-01

    The development of electrical insulation and conduction coating methods that can be applied to large components of fusion experimental devices has been investigated. A thermal spraying method is used to coat the insulation or conduction materials on the structural components because of its applicability for large surfaces. The insulation material chosen was Al{sub 2}O{sub 3}, while Cr{sub 3}C{sub 2}-NiCr and WC-NiCr were chosen as conduction materials. These materials were coated on stainless steel substrates to examine the basic characteristics of the coated layers, such as their adhesive strength to the substrate, thermal shock resistance, electrical resistance, dielectric breakdown voltage, and thermal conductivity. It was found that they have sufficient electrical insulation and conduction properties, respectively. In addition, the sliding tests of the coated layers showed adequate frictional properties. The spraying method was tested on a 100- x 1000-mm surface and found to be applicable for large surfaces of experimental fusion devices. 9 refs., 6 figs., 15 tabs.

  6. Electrical insulation and conduction coating for fusion experimental devices

    International Nuclear Information System (INIS)

    Onozuka, Masanori; Tsujimura, Seiji; Toyoda, Masahiko; Inoue, Masahiko; Abe, Tetsuya; Murakami, Yoshio

    1996-01-01

    The development of electrical insulation and conduction coating methods that can be applied to large components of fusion experimental devices has been investigated. A thermal spraying method is used to coat the insulation or conduction materials on the structural components because of its applicability for large surfaces. The insulation material chosen was Al 2 O 3 , while Cr 3 C 2 -NiCr and WC-NiCr were chosen as conduction materials. These materials were coated on stainless steel substrates to examine the basic characteristics of the coated layers, such as their adhesive strength to the substrate, thermal shock resistance, electrical resistance, dielectric breakdown voltage, and thermal conductivity. It was found that they have sufficient electrical insulation and conduction properties, respectively. In addition, the sliding tests of the coated layers showed adequate frictional properties. The spraying method was tested on a 100- x 1000-mm surface and found to be applicable for large surfaces of experimental fusion devices. 9 refs., 6 figs., 15 tabs

  7. Tritium inventory and recovery in next-step fusion devices

    International Nuclear Information System (INIS)

    Causey, R.A.; Brooks, J.N.; Federici, G.

    2002-01-01

    Future fusion devices will use tritium and deuterium fuel. Because tritium is both radioactive and expensive, it is absolutely necessary that there be an understanding of the tritium retention characteristics of the materials used in these devices as well as how to recover the tritium. There are three materials that are strong candidates for plasma-facing-material use in next-step fusion devices. These are beryllium, tungsten, and carbon. While beryllium has the disadvantage of high sputtering and low melting point (which limits its power handling capabilities in divertor areas), it has the advantages of being a low-Z material with a good thermal conductivity and the ability to get oxygen from the plasma. Due to beryllium's very low solubility for hydrogen, implantation of beryllium with deuterium and tritium results in a saturated layer in the very near-surface with limited inventory (J. Nucl. Mater. 273 (1999) 1). Unfortunately, there are nuclear reactions generated by neutrons that will breed tritium and helium in the material bulk (J. Nucl. Mater. 179 (1991) 329). This process will lead to a substantial tritium inventory in the bulk of the beryllium after long-term neutron exposure (i.e. well beyond the operation life time of a next-step reactor like ITER). Tungsten is a high-Z material that will be used in the divertor region of next-step devices (e.g. ITER) and possibly as a first wall material in later devices. The divertor is the preferred location for tungsten use because net erosion is very low there due to low sputtering and high redeposition. While experiments are still continuing on tritium retention in tungsten, present data suggest that relatively low tritium inventories will result with this material (J. Nucl. Mater. 290-293 (2001) 505). For tritium inventories, carbon is the problem material. Neutron damage to the graphite can result in substantial bulk tritium retention (J. Nucl. Mater. 191-194 (1992) 368), and codeposition of the sputtered carbon

  8. Reducing the tritium inventory in waste produced by fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Pamela, J., E-mail: jerome.pamela@cea.fr [CEA, Agence ITER-France, F-13108 Saint-Paul-lez-Durance (France); Decanis, C. [CEA, DEN, Centre de Cadarache, F-13108 Saint-Paul-lez-Durance (France); Canas, D. [CEA, DEN/DADN, Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Liger, K.; Gaune, F. [CEA, DEN, Centre de Cadarache, F-13108 Saint-Paul-lez-Durance (France)

    2015-04-15

    Highlights: • Fusion devices including ITER will generate tritiated waste, some of which will need to be detritiated before disposal. • Interim storage is the reference solution offering an answer for all types of tritiated radwaste. • Incineration is very attractive for VLLW and possibly SL-LILW soft housekeeping waste, since it offers higher tritium and waste volume reduction than the alternative thermal treatment technique. • For metallic waste, further R&D efforts should be made to optimize tritium release management and minimize the need for interim storage. - Abstract: The specific issues raised by tritiated waste resulting from fusion machines are described. Of the several categories of tritium contaminated waste produced during the entire lifespan of a fusion facility, i.e. operating phase and dismantling phase, only two categories are considered here: metal components and solid combustible waste, especially soft housekeeping materials. Some of these are expected to contain a high level of tritium, and may therefore need to be processed using a detritiation technique before disposal or interim storage. The reference solution for tritiated waste management in France is a 50-year temporary storage for tritium decay, with options for reducing the tritium content as alternatives or complement. An overview of the strategic issues related to tritium reduction techniques is proposed for each radiological category of waste for both metallic and soft housekeeping waste. For this latter category, several options of detritiation techniques by thermal treatment like heating up or incineration are described. A comparison has been made between these various technical options based on several criteria: environment, safety, technical feasibility and costs. For soft housekeeping waste, incineration is very attractive for VLLW and possibly SL-LILW. For metallic waste, further R&D efforts should be conducted.

  9. OVERVIEW OF NEUTRON MEASUREMENTS IN JET FUSION DEVICE.

    Science.gov (United States)

    Batistoni, P; Villari, R; Obryk, B; Packer, L W; Stamatelatos, I E; Popovichev, S; Colangeli, A; Colling, B; Fonnesu, N; Loreti, S; Klix, A; Klosowski, M; Malik, K; Naish, J; Pillon, M; Vasilopoulou, T; De Felice, P; Pimpinella, M; Quintieri, L

    2017-10-05

    The design and operation of ITER experimental fusion reactor requires the development of neutron measurement techniques and numerical tools to derive the fusion power and the radiation field in the device and in the surrounding areas. Nuclear analyses provide essential input to the conceptual design, optimisation, engineering and safety case in ITER and power plant studies. The required radiation transport calculations are extremely challenging because of the large physical extent of the reactor plant, the complexity of the geometry, and the combination of deep penetration and streaming paths. This article reports the experimental activities which are carried-out at JET to validate the neutronics measurements methods and numerical tools used in ITER and power plant design. A new deuterium-tritium campaign is proposed in 2019 at JET: the unique 14 MeV neutron yields produced will be exploited as much as possible to validate measurement techniques, codes, procedures and data currently used in ITER design thus reducing the related uncertainties and the associated risks in the machine operation. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  10. Interrupter and hybrid-switch testing for fusion devices

    International Nuclear Information System (INIS)

    Parsons, W.M.; Warren, R.W.; Honig, E.M.; Lindsay, J.D.G.; Bellamo, P.; Cassel, R.L.

    1979-01-01

    This paper discusses recent and ongoing switch testing for fusion devices. The first part describes testing for the TFTR ohmic-heating circuit. In this set of tests, which simulated the stresses produced during a plasma initiation pulse, circuit breakers were required to interrupt a current of 24 kA with an associated recovery voltage of 25 kV. Two interrupter systems were tested for over 1000 operations each, and both appear to satisfy TFTR requirements. The second part discusses hybrid-switch development for superconducting coil protection. These switching systems must be capable of carrying large currents on a continuous basis as well as performing interruption duties. The third part presents preliminary results on an early-counterpulse technique applied to vacuum interrupters. Implementation of this technique has resulted in large increases in interruptible current as well as a marked reduction in contact erosion

  11. Speckle interferometry application for erosion measurements in fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Gauthier, E.; Roupillard, R. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee

    2003-07-01

    In order to measure erosion/redeposition in fusion devices, a new diagnostic based on speckle interferometry is investigated. First experiments performed on carbon fibre composite (CFC) materials have shown that this technique is able to measure a modification of the surface in the range of 1 {mu}m. Further experiments have been performed on different materials using a second wavelength in order to carry out 3-dimensional measurements of the surface and to increase the dynamic range of the depth measurement. A diagnostic, based on two-wavelength TV-holography to measure in situ erosion/redeposition during long duration discharges on the CIEL limiter in Tore Supra, is under development at CEA Cadarache. (authors)

  12. Fusion Engineering Device (FED) first wall/shield design

    International Nuclear Information System (INIS)

    Sager, P.H.; Fuller, G.; Cramer, B.; Davisson, J.; Haines, J.; Kirchner, J.

    1981-01-01

    The torus of the Fusion Engineering Device (FED) is comprised of the bulk shield and its associated spool lstructure and support system, the first wall water-cooled panel and armor systems, and the pumped limiter. The bulk shielding is provided by ten shield sectors that are installed in the spool structure in such a way as to permit extraction of the sectors through the openings between adjacent toroidal field coils with a direct radial movement. The first wall armor is installed on the inboard and top interior walls of these sectors, and the water-cooled panels are installed on the outboard interior walls and the pumped limiter in the bottom of the sectors. The overall design of the first wall and shield system is described in this paper

  13. Erosion and redeposition at the vessel walls in fusion devices

    International Nuclear Information System (INIS)

    Naujoks, D.; Behrisch, R.

    1995-01-01

    The plasma induced erosion and redeposition at the vessel walls in today's fusion devices have been investigated both with the computer simulation code ERO, and in experiments. Well prepared carbon probes with implanted and evaporated markers in the surface layers have been exposed in the scrape-off layer (SOL) of several tokamaks such as JET, TEXTOR and ASDEX-Upgrade. The main plasma parameters (electron density and temperature, impurity concentration in the SOL) are simultaneously determined. After exposure to single plasma discharges, erosion and redeposition of the marker material were measured by surface layer analysis with MeV ion beam techniques. The experimental results were compared with the results from the ERO code. The measured erosion/redeposition could be described with ERO, which takes into account the impurity concentration in the SOL, the dynamical change of the surface composition (causing a modification of the sputtering yield during the exposure) and ExB drift effects. ((orig.))

  14. Numerical modelling of electromagnetic loads on fusion device structures

    International Nuclear Information System (INIS)

    Bettini, Paolo; Palumbo, Maurizio Furno; Specogna, Ruben

    2014-01-01

    In magnetic confinement fusion devices, during abnormal operations (disruptions) the plasma begins to move rapidly towards the vessel wall in a vertical displacement event (VDE), producing plasma current asymmetries, vessel eddy currents and open field line halo currents, each of which can exert potentially damaging forces upon the vessel and in-vessel components. This paper presents a methodology to estimate electromagnetic loads, on three-dimensional conductive structures surrounding the plasma, which arise from the interaction of halo-currents associated to VDEs with a magnetic field of the order of some Tesla needed for plasma confinement. Lorentz forces, calculated by complementary formulations, are used as constraining loads in a linear static structural analysis carried out on a detailed model of the mechanical structures of a representative machine

  15. Alternative divertor target concepts for next step fusion devices

    Science.gov (United States)

    Mazul, I. V.

    2016-12-01

    The operational conditions of a divertor target in the next steps of fusion devices are more severe in comparison with ITER. The current divertor designs and technologies have a limited application concerning these conditions, and so new design concepts/technologies are required. The main reasons which practically prevent the use of the traditional motionless solid divertor target are analyzed. We describe several alternative divertor target concepts in this paper. The comparative analysis of these concepts (including the advantages and the drawbacks) is made and the prospects for their practical implementation are prioritized. The concept of the swept divertor target with a liquid metal interlayer between the moving armour and motionless heat-sink is presented in more detail. The critical issues of this design are listed and outlined, and the possible experiments are presented.

  16. Numerical modelling of electromagnetic loads on fusion device structures

    Science.gov (United States)

    Bettini, Paolo; Furno Palumbo, Maurizio; Specogna, Ruben

    2014-03-01

    In magnetic confinement fusion devices, during abnormal operations (disruptions) the plasma begins to move rapidly towards the vessel wall in a vertical displacement event (VDE), producing plasma current asymmetries, vessel eddy currents and open field line halo currents, each of which can exert potentially damaging forces upon the vessel and in-vessel components. This paper presents a methodology to estimate electromagnetic loads, on three-dimensional conductive structures surrounding the plasma, which arise from the interaction of halo-currents associated to VDEs with a magnetic field of the order of some Tesla needed for plasma confinement. Lorentz forces, calculated by complementary formulations, are used as constraining loads in a linear static structural analysis carried out on a detailed model of the mechanical structures of a representative machine.

  17. Conceptual design report for a Fusion Engineering Device sector-handling machine and movable manipulator system

    International Nuclear Information System (INIS)

    Watts, K.D.; Masson, L.S.; McPherson, R.S.

    1982-10-01

    Design requirements, trade studies, design descriptions, conceptual designs, and cost estimates have been completed for the Fusion Engineering Device sector handling machine, movable manipulator system, subcomponent handling machine, and limiter blade handling machine. This information will be used by the Fusion Engineering Design Center to begin to determine the cost and magnitude of the effort required to perform remote maintenance on the Fusion Engineering Device. The designs presented are by no means optimum, and the costs estimates are rough-order-of-magnitude

  18. Safety analysis and evaluation of the next fusion device

    International Nuclear Information System (INIS)

    Kobayashi, Shigetada; Honda, Tsutomu; Ohmura, Hiroshi; Kawai, Masayoshi; Shimizu, Takeshi; Yamaoka, Mitsuaki; Nakahara, Katsuhiko; Seki, Yasushi.

    1988-12-01

    As a part of safety evaluation, a probabilistic risk assessment (PRA) has been attempted for the Next Fusion Device system. Among the various events related to safety, a number of representative events have been selected for assessment, from the events in normal operation state, repair and maintenance state and accidental state. In the first chapter, in order to conduct the probabilistic risk assessment of the whole Fusion Experimental Reactor (FER), the data base required for the analysis was investigated in 1.1, the results on the failure mode and effects analysis (FMEA), accident sequence, radioactive inventory leakage flow path, event tree analysis (ETA) and fault tree analysis (FTA) were summarized in 1.2 to 1.5, respectively. Based on these results, accident initiating events were evaluated in 1.6, and overall risk was assessed in 1.7 and the tasks for the future were summarized in 1.8. It is important to analyze and evaluate various events during normal operations, repair and maintenance and accidents. However, due to the large uncertainties in the modeling of phenomena or the data base, there are many events for which realistic analyses are difficult. Three such events were selected and studied in chapter two. In 2.1, the temperature rise in the reactor structure after the Loss-of-Coolant-Accident caused by the decay heat under various heat removal conditions were investigated. In 2.2, the radiation dose of personnel during repair and maintenance period caused by the release of activated dust were estimated. Lastly, in 2.3 tritium behavior in the stainless steel first wall and graphite armour were studied. (author)

  19. Conceptual radiation shielding design of superconducting tokamak fusion device by PHITS

    International Nuclear Information System (INIS)

    Sukegawa, Atsuhiko M.; Kawasaki, Hiromitsu; Okuno, Koichi

    2010-01-01

    A complete 3D neutron and photon transport analysis by Monte Carlo transport code system PHITS (Particle and Heavy Ion Transport code System) have been performed for superconducting tokamak fusion device such as JT-60 Super Advanced (JT-60SA). It is possible to make use of PHITS in the port streaming analysis around the devices for the tokamak fusion device, the duct streaming analysis in the building where the device is installed, and the sky shine analysis for the site boundary. The neutron transport analysis by PHITS makes it clear that the shielding performance of the superconducting tokamak fusion device with the cryostat is improved by the graphical results. From the standpoint of the port streaming and the duct streaming, it is necessary to calculate by 3D Monte Carlo code such as PHITS for the neutronics analysis of superconducting tokamak fusion device. (author)

  20. Protective coatings for in-vessel fusion devices

    International Nuclear Information System (INIS)

    Brossa, F.

    1984-01-01

    Coatings of Al/Si, SAP (Sintered Aluminium Powder), Al 2 O 3 , TiC (low-Z material) and Ta have been developed for in-vessel component protection. Anodic oxidation, vapor depositions, reactive sputtering, chemical vapor deposition (CVD) and plasma spray have been the coating formation methods studied. AISI 316, 310, 304, Inconel 600 and Mo were adopted as base materials. the coatings were characterized in terms of composition, structure and connection with the supporting material. The behavior of coatings under H + , D + and He + irradiation in the energy range 100 eV-8 keV was tested and compared to the solid massive samples. TiC and Ta coatings were tested with thermal shock under power density pulses of 1 kW/cm 2 generated by an electron beam gun. Temperature-dependence of the erosion of TiC by vacuum arcs in a magnetic field was also studied. TiC coatings have low sputtering values, good resistance to arcing and a high chemical stability. TiC and Ta, CVD and plasma spray coatings are thermal-shock resistant. High thermal loads produce cracks but no spalling. Destruction occurred only after melting of the base material. The plasma spray coating method seems to be most appropriate for developing remote handling applications in fusion devices. (orig.)

  1. Turbomolecular pumping systems for nuclear fusion devices in JAERI

    International Nuclear Information System (INIS)

    Ohga, Tokumichi; Arai, Takashi

    1978-01-01

    The turbomolecular pumping systems for the nuclear fusion devices JFT-2, JFT-2a and the injector test stands ITS-1, 2 and 3 in the Japan Atomic Energy Research Institute are mainly reported. For these vacuum systems, many requirements exist, such as oil free, large exhausting speed up to high pressure region (10 -3 Torr), compactness and easy operation and maintenance, etc., for the special usage. The outline of the systems and components, and the functions and the operational characteristics of the turbomolecular pumps are introduced. Concerning to the vacuum systems for JFT-2 and JFT-2a, the main system flow charts, the key specifications, the exhausting characteristic curves in case of starting from the atmospheric pressure for both JFT-2 and JFT-2a, and the conductance for hydrogen gas in the high vacuum side of JFT-2a are explained. As for the vacuum system for ITS-2, the main specification, the system flow chart, the main components, the functions, the conductance for hydrogen gas, the pumping characteristic curve, the starting characteristic of the turbomolecular pump, the exhausting speed for hydrogen gas and an example of mass spectrum are shown. The vacuum pressure obtained is almost 10 -5 -- 10 -6 torr for the three pumping systems. (Nakai, Y.)

  2. "Fan-Tip-Drive" High-Power-Density, Permanent Magnet Electric Motor and Test Rig Designed for a Nonpolluting Aircraft Propulsion Program

    Science.gov (United States)

    Brown, Gerald V.; Kascak, Albert F.

    2004-01-01

    A scaled blade-tip-drive test rig was designed at the NASA Glenn Research Center. The rig is a scaled version of a direct-current brushless motor that would be located in the shroud of a thrust fan. This geometry is very attractive since the allowable speed of the armature is approximately the speed of the blade tips (Mach 1 or 1100 ft/s). The magnetic pressure generated in the motor acts over a large area and, thus, produces a large force or torque. This large force multiplied by the large velocity results in a high-power-density motor.

  3. AxiaLIF system: minimally invasive device for presacral lumbar interbody spinal fusion.

    Science.gov (United States)

    Rapp, Steven M; Miller, Larry E; Block, Jon E

    2011-01-01

    Lumbar fusion is commonly performed to alleviate chronic low back and leg pain secondary to disc degeneration, spondylolisthesis with or without concomitant lumbar spinal stenosis, or chronic lumbar instability. However, the risk of iatrogenic injury during traditional anterior, posterior, and transforaminal open fusion surgery is significant. The axial lumbar interbody fusion (AxiaLIF) system is a minimally invasive fusion device that accesses the lumbar (L4-S1) intervertebral disc spaces via a reproducible presacral approach that avoids critical neurovascular and musculoligamentous structures. Since the AxiaLIF system received marketing clearance from the US Food and Drug Administration in 2004, clinical studies of this device have reported high fusion rates without implant subsidence, significant improvements in pain and function, and low complication rates. This paper describes the design and approach of this lumbar fusion system, details the indications for use, and summarizes the clinical experience with the AxiaLIF system to date.

  4. High power densities from high-temperature material interactions. [in thermionic energy conversion and metallic fluid heat pipes

    Science.gov (United States)

    Morris, J. F.

    1981-01-01

    Thermionic energy conversion (TEC) and metallic-fluid heat pipes (MFHPs), offering unique advantages in terrestrial and space energy processing by virtue of operating on working-fluid vaporization/condensation cycles that accept great thermal power densities at high temperatures, share complex materials problems. Simplified equations are presented that verify and solve such problems, suggesting the possibility of cost-effective applications in the near term for TEC and MFHP devices. Among the problems discussed are: the limitation of alkali-metal corrosion, protection against hot external gases, external and internal vaporization, interfacial reactions and diffusion, expansion coefficient matching, and creep deformation.

  5. Achieving High-Energy-High-Power Density in a Flexible Quasi-Solid-State Sodium Ion Capacitor.

    Science.gov (United States)

    Li, Hongsen; Peng, Lele; Zhu, Yue; Zhang, Xiaogang; Yu, Guihua

    2016-09-14

    Simultaneous integration of high-energy output with high-power delivery is a major challenge for electrochemical energy storage systems, limiting dual fine attributes on a device. We introduce a quasi-solid-state sodium ion capacitor (NIC) based on a battery type urchin-like Na2Ti3O7 anode and a capacitor type peanut shell derived carbon cathode, using a sodium ion conducting gel polymer as electrolyte, achieving high-energy-high-power characteristics in solid state. Energy densities can reach 111.2 Wh kg(-1) at power density of 800 W kg(-1), and 33.2 Wh kg(-1) at power density of 11200 W kg(-1), which are among the best reported state-of-the-art NICs. The designed device also exhibits long-term cycling stability over 3000 cycles with capacity retention ∼86%. Furthermore, we demonstrate the assembly of a highly flexible quasi-solid-state NIC and it shows no obvious capacity loss under different bending conditions.

  6. Managing fusion high-level waste-A strategy for burning the long-lived products in fusion devices

    International Nuclear Information System (INIS)

    El-Guebaly, L.A.

    2006-01-01

    Fusion devices appear to be a viable option for burning their own high-level waste (HLW). We propose a novel strategy to eliminate (or minimize) the HLW generated by fusion systems. The main source of the fusion HLW includes the structural and recycled materials, refractory metals, and liquid breeders. The basic idea involves recycling and reprocessing the waste, separating the long-lived radionuclides from the bulk low-level waste, and irradiating the limited amount of HLW in a specially designed module to transmute the long-lived products into short-lived radioisotopes or preferably, stable elements. The potential performance of the new concept seems promising. Our analysis indicated moderate to excellent transmutation rates could be achieved in advanced fusion designs. Successive irradiation should burn the majority of the HLW. The figures of merit for the concept relate to the HLW burn-up fraction, neutron economy, and impact on tritium breeding. Hopefully, the added design requirements could be accommodated easily in fusion power plants and the cost of the proposed system would be much less than disposal in a deep geological HLW repository. Overall, this innovative approach offers benefits to fusion systems and helps earn public acceptance for fusion as a HLW-free source of clean nuclear energy

  7. Simulation, measurement, and emulation of photovoltaic modules using high frequency and high power density power electronic circuits

    Science.gov (United States)

    Erkaya, Yunus

    The number of solar photovoltaic (PV) installations is growing exponentially, and to improve the energy yield and the efficiency of PV systems, it is necessary to have correct methods for simulation, measurement, and emulation. PV systems can be simulated using PV models for different configurations and technologies of PV modules. Additionally, different environmental conditions of solar irradiance, temperature, and partial shading can be incorporated in the model to accurately simulate PV systems for any given condition. The electrical measurement of PV systems both prior to and after making electrical connections is important for attaining high efficiency and reliability. Measuring PV modules using a current-voltage (I-V) curve tracer allows the installer to know whether the PV modules are 100% operational. The installed modules can be properly matched to maximize performance. Once installed, the whole system needs to be characterized similarly to detect mismatches, partial shading, or installation damage before energizing the system. This will prevent any reliability issues from the onset and ensure the system efficiency will remain high. A capacitive load is implemented in making I-V curve measurements with the goal of minimizing the curve tracer volume and cost. Additionally, the increase of measurement resolution and accuracy is possible via the use of accurate voltage and current measurement methods and accurate PV models to translate the curves to standard testing conditions. A move from mechanical relays to solid-state MOSFETs improved system reliability while significantly reducing device volume and costs. Finally, emulating PV modules is necessary for testing electrical components of a PV system. PV emulation simplifies and standardizes the tests allowing for different irradiance, temperature and partial shading levels to be easily tested. Proper emulation of PV modules requires an accurate and mathematically simple PV model that incorporates all known

  8. Nanofluidic crystal: a facile, high-efficiency and high-power-density scaling up scheme for energy harvesting based on nanofluidic reverse electrodialysis

    International Nuclear Information System (INIS)

    Ouyang Wei; Wang Wei; Zhang Haixia; Wu Wengang; Li Zhihong

    2013-01-01

    The great advances in nanotechnology call for advances in miniaturized power sources for micro/nano-scale systems. Nanofluidic channels have received great attention as promising high-power-density substitutes for ion exchange membranes for use in energy harvesting from ambient ionic concentration gradient, namely reverse electrodialysis. This paper proposes the nanofluidic crystal (NFC), of packed nanoparticles in micro-meter-sized confined space, as a facile, high-efficiency and high-power-density scaling-up scheme for energy harvesting by nanofluidic reverse electrodialysis (NRED). Obtained from the self-assembly of nanoparticles in a micropore, the NFC forms an ion-selective network with enormous nanochannels due to electrical double-layer overlap in the nanoparticle interstices. As a proof-of-concept demonstration, a maximum efficiency of 42.3 ± 1.84%, a maximum power density of 2.82 ± 0.22 W m −2 , and a maximum output power of 1.17 ± 0.09 nW/unit (nearly three orders of magnitude of amplification compared to other NREDs) were achieved in our prototype cell, which was prepared within 30 min. The current NFC-based prototype cell can be parallelized and cascaded to achieve the desired output power and open circuit voltage. This NFC-based scaling-up scheme for energy harvesting based on NRED is promising for the building of self-powered micro/nano-scale systems. (paper)

  9. Oscillatory vapour shielding of liquid metal walls in nuclear fusion devices

    NARCIS (Netherlands)

    van Eden, G.G.; Kvon, V.; Van De Sanden, M.C.M.; Morgan, T.W.

    2017-01-01

    Providing an efficacious plasma facing surface between the extreme plasma heat exhaust and the structural materials of nuclear fusion devices is a major challenge on the road to electricity production by fusion power plants. The performance of solid plasma facing surfaces may become critically

  10. Diagnostics Development towards Steady State Operation in Fusion Devices

    Energy Technology Data Exchange (ETDEWEB)

    Burhenn, R.; Baldzuhn, J.; Dreier, H.; Endler, M.; Hartfuss, H.J.; Hildebrandt, D.; Hirsch, M.; Koenig, R.; Kornejev, P.; Krychowiak, M.; Laqua, H.P.; Laux, M.; Oosterbeek, J.W.; Pasch, E.; Schneider, W.; Thomsen, H.; Weller, A.; Werner, A.; Wolf, R.; Zhang, D. [Max-Planck-Institute fuer Plasmaphysik, EURATOM Association, D-17491 Greifswald (Germany); Biel, W. [Institut fuer Energieforschung - Plasmaphysik, Forschungszentrum Juelich GmbH EURATOM Association, Trilateral Euregio Cluster, D-52425 Juelich (Germany)

    2011-07-01

    The stellarator Wendelstein 7-X (W7-X) is being presently under construction and is already equipped with superconducting coil systems and principally is capable of quasi-continuous operation. However, W7-X is faced with new enhanced technical requirements which have to be met by plasma facing components as well as the diagnostic systems in general. Depending on the available heating power, the continuous heat flux to plasma facing components during long pulse operation might lead to unacceptable local thermal overload and necessitates sufficient but often complicate active cooling precautions. Fusion devices with electron cyclotron frequency heating (ECRH) are concerned with significant stray radiation, depending on the chosen heating scheme and the plasma parameters. The required shielding is often not compatible with optimal UHV-consistent design and high intensity throughput. For machine safety, diagnostics are required which are able to identify enhanced plasma wall interaction on a fast time scale in order to prevent damage in time. For W7-X, video camera systems covering most of the inner wall, fast IR-camera systems with coating-resistant pinhole-optics for the observation of the divertor surface temperature and spectrometers with large spectral survey covering relevant spectral lines of all intrinsic impurities with sufficient spectral resolution and sensitivity are necessary. In combination with energy integrating but spatially resolving diagnostics like bolometers and soft-X cameras slow impurity accumulation phenomena on a time scale much larger than flat-top times typically achieved in short-pulse operation can be identified and a radiative plasma collapse possibly be avoided by counteractive measures. Longer port dimensions due to thermal insulation of the cryogenic coil system and high density operation with strong density gradients necessitate the choice of shorter wavelengths for interferometer laser beams. This complicates the avoidance of fringe

  11. Contributions to the 7th International Conference on plasma surface interactions in controlled fusion devices

    International Nuclear Information System (INIS)

    1986-01-01

    The report contains three papers presented in the 7th International Conference on plasma surface interactions in controlled fusion devices held in Princeton (USA) 5-9 May 1986, all referred to the FT Tokamak

  12. Design, construction, and characterization of high-performance membrane fusion devices with target-selectivity.

    Science.gov (United States)

    Kashiwada, Ayumi; Yamane, Iori; Tsuboi, Mana; Ando, Shun; Matsuda, Kiyomi

    2012-01-31

    Membrane fusion proteins such as the hemagglutinin glycoprotein have target recognition and fusion accelerative domains, where some synergistically working elements are essential for target-selective and highly effective native membrane fusion systems. In this work, novel membrane fusion devices bearing such domains were designed and constructed. We selected a phenylboronic acid derivative as a recognition domain for a sugar-like target and a transmembrane-peptide (Leu-Ala sequence) domain interacting with the target membrane, forming a stable hydrophobic α-helix and accelerating the fusion process. Artificial membrane fusion behavior between the synthetic devices in which pilot and target liposomes were incorporated was characterized by lipid-mixing and inner-leaflet lipid-mixing assays. Consequently, the devices bearing both the recognition and transmembrane domains brought about a remarkable increase in the initial rate for the membrane fusion compared with the devices containing the recognition domain alone. In addition, a weakly acidic pH-responsive device was also constructed by replacing three Leu residues in the transmembrane-peptide domain by Glu residues. The presence of Glu residues made the acidic pH-dependent hydrophobic α-helix formation possible as expected. The target-selective liposome-liposome fusion was accelerated in a weakly acidic pH range when the Glu-substituted device was incorporated in pilot liposomes. The use of this pH-responsive device seems to be a potential strategy for novel applications in a liposome-based delivery system. © 2011 American Chemical Society

  13. Integrative Multi-Spectral Sensor Device for Far-Infrared and Visible Light Fusion

    Science.gov (United States)

    Qiao, Tiezhu; Chen, Lulu; Pang, Yusong; Yan, Gaowei

    2018-06-01

    Infrared and visible light image fusion technology is a hot spot in the research of multi-sensor fusion technology in recent years. Existing infrared and visible light fusion technologies need to register before fusion because of using two cameras. However, the application effect of the registration technology has yet to be improved. Hence, a novel integrative multi-spectral sensor device is proposed for infrared and visible light fusion, and by using the beam splitter prism, the coaxial light incident from the same lens is projected to the infrared charge coupled device (CCD) and visible light CCD, respectively. In this paper, the imaging mechanism of the proposed sensor device is studied with the process of the signals acquisition and fusion. The simulation experiment, which involves the entire process of the optic system, signal acquisition, and signal fusion, is constructed based on imaging effect model. Additionally, the quality evaluation index is adopted to analyze the simulation result. The experimental results demonstrate that the proposed sensor device is effective and feasible.

  14. Tungsten as First Wall Material in Fusion Devices

    International Nuclear Information System (INIS)

    Kaufmann, M.

    2006-01-01

    In the PLT tokamak with a tungsten limiter strong cooling of the central plasma was observed. Since then mostly graphite has been used as limiter or target plate material. Only a few tokamaks (limiter: FTU, TEXTOR; divertor: Alcator C-Mod, ASDEX Upgrade) gained experience with high-Z-materials. With the observed strong co- deposition of tritium together with carbon in JET and as a result of design studies of fusion reactors, it became clear that in the long run tungsten is the favourite for the first-wall material. Tungsten as a plasma facing material requires intensive research in all areas, i.e. in plasma physics, plasma wall-interaction and material development. Tungsten as an impurity in the confined plasma reveals considerable differences to carbon. Strong radiation at high temperatures, in connection with mostly a pronounced inward drift forms a particular challenge. Turbulent transport plays a beneficial role in this regard. The inward drift is an additional problem in the pedestal region of H-mode plasmas in ITER-like configurations. The erosion by low energy hydrogen atoms is in contrast to carbon small. However, erosion by fast particles from heating measures and impurity ions, accelerated in the sheath potential, play an important role in the case of tungsten. Radiation by carbon in the plasma boundary reduces the load to the target plates. Neon or Argon as substitutes will increase the erosion of tungsten. So far experiments have demonstrated that in most scenarios the tungsten content in the central plasma can be kept sufficiently small. The material development is directed to the specific needs of existing or future devices. In ASDEX Upgrade, which will soon be a divertor experiment with a complete tungsten first-wall, graphite tiles are coated with tungsten layers. In ITER, the solid tungsten armour of the target plates has to be castellated because of its difference in thermal expansion compared to the cooling structure. In a reactor the technical

  15. Flexible rechargeable Ni//Zn battery based on self-supported NiCo2O4 nanosheets with high power density and good cycling stability

    Directory of Open Access Journals (Sweden)

    Haozhe Zhang

    2018-01-01

    Full Text Available The overall electrochemical performances of Ni–Zn batteries are still far from satisfactory, specifically for rate performance and cycling stability Herein, we demonstrated a high-performance flexible Ni//Zn battery with outstanding durability and high power density based on self-supported NiCo2O4 nanosheets as cathode and Zn nanosheets as anode. This Ni//Zn battery is able to deliver a remarkable capacity of 183.1 mAh g−1 and a good cycling performance (82.7% capacity retention after 3500 cycles. More importantly, this battery achieves an admirable power density of 49.0 kW kg−1 and energy density of 303.8 Wh kg−1, substantially higher than most recently reported batteries. With such excellent electrochemical performance, this battery will have great potential as an ultrafast power source in practical application.

  16. Plasma facing materials and components for future fusion devices - development, characterization and performance under fusion specific loading conditions

    Energy Technology Data Exchange (ETDEWEB)

    Linke, J. [Forschungszentrum Juelich (Germany). Inst. fuer Plasmaphysik

    2006-04-15

    The plasma exposed components in existing and future fusion devices are strongly affected by the plasma material interaction processes. These mechanisms have a strong influence on the plasma performance; in addition they have major impact on the lifetime of the plasma facing armour and the joining interface between the plasma facing material (PFM) and the heat sink. Besides physical and chemical sputtering processes, high heat quasi-stationary fluxes during normal and intense thermal transients are of serious concern for the engineers who develop reliable wall components. In addition, the material and component degradation due to intense fluxes of energetic neutrons is another critical issue in D-T-burning fusion devices which requires extensive RandD. This paper presents an overview on the materials development and joining, the testing of PFMs and components, and the analysis of the neutron irradiation induced degradation.

  17. Plasma facing materials and components for future fusion devices - development, characterization and performance under fusion specific loading conditions

    International Nuclear Information System (INIS)

    Linke, J.

    2006-01-01

    The plasma exposed components in existing and future fusion devices are strongly affected by the plasma material interaction processes. These mechanisms have a strong influence on the plasma performance; in addition they have major impact on the lifetime of the plasma facing armour and the joining interface between the plasma facing material (PFM) and the heat sink. Besides physical and chemical sputtering processes, high heat quasi-stationary fluxes during normal and intense thermal transients are of serious concern for the engineers who develop reliable wall components. In addition, the material and component degradation due to intense fluxes of energetic neutrons is another critical issue in D-T-burning fusion devices which requires extensive RandD. This paper presents an overview on the materials development and joining, the testing of PFMs and components, and the analysis of the neutron irradiation induced degradation

  18. Safety considerations in the design of the Fusion Engineering Device

    International Nuclear Information System (INIS)

    Barrett, R.J.

    1983-01-01

    The US Department of Energy (DOE) regulations and guidelines for radiation protection have been reviewed and are being applied to the device design. Direct radiation protection is provided by the device shield and the reactor building walls. Radiation from the activated device components and the tritium fuel is to be controlled with shielding, contamination control, and ventilation. The potential release of tritium from the plant has influenced the selection of reactor building and plant designs and specifications. The safety of the plant workers is affected primarily by the radiation from the activated device components and from plasma chamber debris

  19. AxiaLIF system: minimally invasive device for presacral lumbar interbody spinal fusion

    Directory of Open Access Journals (Sweden)

    Rapp SM

    2011-08-01

    Full Text Available Steven M Rapp1, Larry E Miller2,3, Jon E Block31Michigan Spine Institute, Waterford, MI, USA; 2Miller Scientific Consulting Inc, Biltmore Lake, NC, USA; 3Jon E. Block, Ph.D., Inc., San Francisco, CA, USAAbstract: Lumbar fusion is commonly performed to alleviate chronic low back and leg pain secondary to disc degeneration, spondylolisthesis with or without concomitant lumbar spinal stenosis, or chronic lumbar instability. However, the risk of iatrogenic injury during traditional anterior, posterior, and transforaminal open fusion surgery is significant. The axial lumbar interbody fusion (AxiaLIF system is a minimally invasive fusion device that accesses the lumbar (L4–S1 intervertebral disc spaces via a reproducible presacral approach that avoids critical neurovascular and musculoligamentous structures. Since the AxiaLIF system received marketing clearance from the US Food and Drug Administration in 2004, clinical studies of this device have reported high fusion rates without implant subsidence, significant improvements in pain and function, and low complication rates. This paper describes the design and approach of this lumbar fusion system, details the indications for use, and summarizes the clinical experience with the AxiaLIF system to date.Keywords: AxiaLIF, fusion, lumbar, minimally invasive, presacral

  20. Experimental laser fusion devices and related vacuum problems

    International Nuclear Information System (INIS)

    O'Neal, W.C.; Campbell, D.E.; Glaros, S.S.; Hurley, C.A.; Kobierecki, M.W.; McFann, C.B. Jr.; Monjes, J.A.; Patton, H.G.; Rienecker, F. Jr.

    1977-01-01

    Laser fusion experiments require hard vacuum in the laser-beam spatial filters, target chambers and for target diagnostics instruments. Laser focusing lenses and windows, and target alignment windows must hold vacuum without optical distortion, and must be protected from target debris. The vacuum must be sufficient to prevent residual gas breakdown in focused laser light, avoid arcing at high voltage terminals, minimize contamination and melting of cryogenic targets, and prevent adsorption of the target's microfusion radiation before it reaches the diagnostics instruments

  1. Applicability of the PHITS code to a tokamak fusion device

    International Nuclear Information System (INIS)

    Sukegawa, Atsuhiko; Okuno, Koichi; Kawasaki, Hiromitsu

    2011-01-01

    The three-dimensional Monte-Carlo code PHITS (particle and Heavy Ion Transport code System) has been developed to perform the radiation transport analysis, design of the radiation shields and neutronics calculations for tokamak-type D-D fusion reactors. A subroutine was included in PHITS to represent the toroidal neutron source of 2.45 MeV neutrons from the D-D reaction. Here, an example of preliminary tests using PHITS is given. (author)

  2. Spatial heterogeneity of tungsten transmutation in a fusion device

    Science.gov (United States)

    Gilbert, M. R.; Sublet, J.-Ch.; Dudarev, S. L.

    2017-04-01

    Accurately quantifying the transmutation rate of tungsten (W) under neutron irradiation is a necessary requirement in the assessment of its performance as an armour material in a fusion power plant. The usual approach of calculating average responses, assuming large, homogenised material volumes, is insufficient to capture the full complexity of the transmutation picture in the context of a realistic fusion power plant design, particularly for rhenium (Re) production from W. Combined neutron transport and inventory simulations for representative spatially heterogeneous high-resolution models of a fusion power plant show that the production rate of Re is strongly influenced by the surrounding local spatial environment. Localised variation in neutron moderation (slowing down) due to structural steel and coolant, particularly water, can dramatically increase Re production because of the huge cross sections of giant resolved resonances in the neutron-capture reaction of 186W at low neutron energies. Calculations using cross section data corrected for temperature (Doppler) effects suggest that temperature may have a relatively lesser influence on transmutation rates.

  3. Utilization of a Network of Small Magnetic Confinement Fusion Devices for Mainstream Fusion Research. Report of a Coordinated Research Project 2011–2016

    International Nuclear Information System (INIS)

    2016-12-01

    The IAEA actively promotes the development of controlled fusion as a source of energy. Through its coordinated research activities, the IAEA helps Member States to exchange and establish scientific and technical knowledge required for the design, construction and operation of a fusion reactor. Due to their compactness, flexibility and low operation costs, small fusion devices are a great resource for supporting and accelerating the development of mainstream fusion research on large fusion devices such as the International Thermonuclear Experimental Reactor. They play an important role in investigating the physics of controlled fusion, developing innovative technologies and diagnostics, testing new materials, training highly qualified personnel for larger fusion facilities, and supporting educational programmes for young scientists. This publication reports on the research work accomplished within the framework of the Coordinated Research Project (CRP) on Utilization of the Network of Small Magnetic Confinement Fusion Devices for Mainstream Fusion Research, organized and conducted by the IAEA in 2011–2016. The CRP has contributed to the coordination of a network of research institutions, thereby enhancing international collaboration through scientific visits, joint experiments and the exchange of information and equipment. A total of 16 institutions and 14 devices from 13 Member States participated in this CRP (Belgium, Bulgaria, Canada, China, Costa Rica, the Czech Republic, the Islamic Republic of Iran, Kazakhstan, Pakistan, Portugal, the Russian Federation, Ukraine and the United Kingdom).

  4. Core-shell N-doped active carbon fiber@graphene composites for aqueous symmetric supercapacitors with high-energy and high-power density

    Science.gov (United States)

    Xie, Qinxing; Bao, Rongrong; Xie, Chao; Zheng, Anran; Wu, Shihua; Zhang, Yufeng; Zhang, Renwei; Zhao, Peng

    2016-06-01

    Graphene wrapped nitrogen-doped active carbon fibers (ACF@GR) of a core-shell structure were successfully prepared by a simple dip-coating method using natural silk as template. Compared to pure silk active carbon, the as-prepared ACF@GR composites exhibit high specific surface area in a range of 1628-2035 m2 g-1, as well as superior energy storage capability, an extremely high single-electrode capacitance of 552.8 F g-1 was achieved at a current density of 0.1 A g-1 in 6 M KOH aqueous electrolyte. The assembled aqueous symmetric supercapacitors are capable of deliver both high energy density and high power density, for instance, 17.1 Wh kg-1 at a power density of 50.0 W kg-1, and 12.2 Wh kg-1 at 4.7 kW kg-1 with a retention rate of 71.3% for ACF@GR1-based supercapacitor.

  5. High power density cell using nanostructured Sr-doped SmCoO3 and Sm-doped CeO2 composite powder synthesized by spray pyrolysis

    Science.gov (United States)

    Shimada, Hiroyuki; Yamaguchi, Toshiaki; Suzuki, Toshio; Sumi, Hirofumi; Hamamoto, Koichi; Fujishiro, Yoshinobu

    2016-01-01

    High power density solid oxide electrochemical cells were developed using nanostructure-controlled composite powder consisting of Sr-doped SmCoO3 (SSC) and Sm-doped CeO2 (SDC) for electrode material. The SSC-SDC nano-composite powder, which was synthesized by spray pyrolysis, had a narrow particle size distribution (D10, D50, and D90 of 0.59, 0.71, and 0.94 μm, respectively), and individual particles were spherical, composing of nano-size SSC and SDC fragments (approximately 10-15 nm). The application of the powder to a cathode for an anode-supported solid oxide fuel cell (SOFC) realized extremely fine cathode microstructure and excellent cell performance. The anode-supported SOFC with the SSC-SDC cathode achieved maximum power density of 3.65, 2.44, 1.43, and 0.76 W cm-2 at 800, 750, 700, and 650 °C, respectively, using humidified H2 as fuel and air as oxidant. This result could be explained by the extended electrochemically active region in the cathode induced by controlling the structure of the starting powder at the nano-order level.

  6. Ion surface collisions on surfaces relevant for fusion devices

    International Nuclear Information System (INIS)

    Rasul, B.; Endstrasser, N.; Zappa, F.; Grill, V.; Scheier, P.; Mark, T.

    2006-01-01

    Full text: One of the great challenges of fusion research is the compatibility of reactor grade plasmas with plasma facing materials coating the inner walls of a fusion reactor. The question of which surface coating should be used is of particular interest for the design of ITER. The impact of energetic plasma particles leads to sputtering of wall material into the plasma. A possible solution for the coating of plasma facing walls would be the use of special carbon surfaces. Investigations of these various surfaces have been started at BESTOF ion-surface collision apparatus. Experiment beam of singly charged molecular ions of hydrocarbon molecules, i.e. C 2 H + 4 , is generated in a Nier-type electron impact ionization source at an electron energy of about 70 eV. In the first double focusing mass spectrometer the ions are mass and energy analyzed and afterwards refocused onto a surface. The secondary reaction products are monitored using a Time Of Flight mass spectrometer. The secondary ion mass spectra are recorded as a function of the collision energy for different projectile ions and different surfaces. A comparison of these spectra show for example distinct changes in the survival probability of the same projectile ion C 2 H + 4 for different surfaces. (author)

  7. First fusion neutrons from a thermonuclear weapon device

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    An account of the first observation of thermonuclear neutrons from a hydrogen weapon, the George shot, is presented. A personal narrative by the researchers J. Allred and L. Rosen includes such topics as the formation of the experimental team, description of the experimental technique, testing the experimental apparatus, testing the effects of a blast, a description of the test area, and the observation of neutrons from fusion. Excerpts are presented from several chapters of the Scientific Director's report on the atomic weapons tests of 1951. Also included is a brief description of the basic design of the hydrogen bomb, a recounting of subsequent developments, and short scientific biographies of the researchers. 21 figures, 2 tables

  8. IAEA technical committee meeting on research using small fusion devices (abstracts)

    International Nuclear Information System (INIS)

    1999-12-01

    The thirteenth IAEA technical committee meeting on research using small fusion devices are held in Chengdu, P. R. China on 18-20 Oct. , 1999. 41 articles are received and the content includes toroidal systems, helical systems, plasma focus, diagnostic systems, theory and modeling, improving confinement, numerical simulation, innovative concepts and others

  9. Non-superconducting magnet structures for near-term, large fusion experimental devices

    International Nuclear Information System (INIS)

    File, J.; Knutson, D.S.; Marino, R.E.; Rappe, G.H.

    1980-10-01

    This paper describes the magnet and structural design in the following American tokamak devices: the Princeton Large Torus (PLT), the Princeton Divertor Experiment (PDX), and the Tokamak Fusion Test Reactor (TFTR). The Joint European Torus (JET), also presented herein, has a magnet structure evolved from several European programs and, like TFTR, represents state of the art magnet and structure design

  10. Coils in a fusion device and its fabrication method

    International Nuclear Information System (INIS)

    Maeda, Hideto; Moritani, Einoshin.

    1975-01-01

    Object: To provide a coil for nuclear fusion equipment, which coil has superior rigidity and strength and is separable into two sections and used for removing impurity ions from high temperature plasma such as heavy hydrogen and tritium. Structure: The coil according to the invention is manufactured by (1) a step of insulating horseshoe-shaped conductors one from another and bundling them into coil halves. (2) a step of assembling a flange on a coil case accommodating each coil half and hermetically welding a lid to each end of the coil half, (3) a step of evacuating the interior of each coil case, (4) a step of pouring a thermosetting resin into each evacuated coil case and hardening the resin, (5) a step of connecting the two coil halves with their ends not covered with resin held in abutting relation to each other, (6) a step of coupling coil case joint pieces to the joined portions and covering the joint pieces with a seal box and hermetically welding the box to the joint pieces, and (7) a step of pouring a thermosetting resin into each evacuated joint portion and hardening the resin. (Kamimura, M.)

  11. A review of fusion device fuel cleanup systems

    International Nuclear Information System (INIS)

    Dombra, A.H.; Carney, M.

    1985-01-01

    Design options for a small fusion fuel purification system are assessed by comparing six conceptual system designs based on one of the following: a Zr/Al getter pump for in vacuo applications, a cryogenic molecular sieve adsorber at 77K, a palladium-alloy membrane diffuser, a U-bed reactor at 1170K, a two-compartment cryogenic freezer at 27-50K and 50-300K, a U-bed and non-regenerative Zr/Al gas purifier. The latter system introduces a new concept of fuel purification based on well-established techniques: recovery of purified D 2 -DT-T 2 from a helium carrier gas with the U-bed, followed by the removal of impurities from the carrier gas with the non-regenerative Zr/Al gas purifier. The main advantages of this system are simplicity, safety and relatively small quantity of tritiated waste produced by the process. The tritium in the waste is immobilized as a stable tritide of Zr/Al

  12. Preparation of Ni-Fe bimetallic porous anode support for solid oxide fuel cells using LaGaO{sub 3} based electrolyte film with high power density

    Energy Technology Data Exchange (ETDEWEB)

    Ju, Young-Wan; Ida, Shintaro; Ishihara, Tatsumi [Department of Applied Chemistry, Faculty of Engineering, Kyushu University, Motooka 744, Nishi-Ku, Fukuoka 819-0395 (Japan); Eto, Hiroyuki [Mitsubishi Materials Corporation, Central Research Institute, 1002-14 Mukohyama, Naka-Shi, Ibaraki 311-0102 (Japan); Inagaki, Toru [The Kansai Electric Power Co., Inc., 11-20 Nakoji 3-Chome, Amagasaki, Hyogo 661-0974 (Japan)

    2010-10-01

    Optimization of sintering temperature for NiO-Fe{sub 2}O{sub 3} composite oxide substrate was studied in order to obtain a dense substrate with smooth surface. By in situ reduction, the substrate was changed to a porous Ni-Fe alloy metal. The volumetric shrinkage and porosity of the substrate were also studied systematically with the Ni-Fe substrate reduced at different temperatures. A Sr and Mg-doped LaGaO{sub 3} (LSGM) thin film was prepared on dense substrate by the pulsed laser deposition (PLD) method. The LSGM film with stoichiometric composition was successfully prepared under optimal deposition parameters and a target composition. Sm{sub 0.5}Sr{sub 0.5}CoO{sub 3} (SSC55) cathode was prepared by the slurry coating method on the deposited film. Prepared SOFC single cell shows high power density and the maximum power density (MPD) achieved was 1.79, 0.82 and 0.29 W cm{sup -2} at 973, 873 and 773 K, respectively. After thermal cycle from 973 to 298 K, the cell shows almost theoretical open circuit potential (1.1 V) and the power density of 1.62 W cm{sup -2}, which is almost the same as that at first cycles. Therefore, the Ni-Fe porous metal support made by the selective reduction is highly promising as a metal anode substrate for SOFC using LaGaO{sub 3} thin film. (author)

  13. Metal/graphite-composite materials for fusion device

    International Nuclear Information System (INIS)

    Kneringer, G.; Kny, E.; Fischer, W.; Reheis, N.; Staffler, R.; Samm, U.; Winter, J.

    1995-01-01

    The utilization of graphite as a structural material depends to an important extent on the availability of a joining technique suitable for the production of reliable large scale metal/graphite-composites. This study has been conducted to evaluate vacuum brazes and procedures for graphite and metals which can be used in fusion applications up to about 1500 degree C. The braze materials included: AgCuTi, CuTi, NiTi, Ti, ZrTi, Zr. Brazing temperatures ranged from 850 degree C to 1900 degree C. The influence of graphite quality on wettability and pore-penetration of the braze has been investigated. Screening tests of metal/graphite-assemblies with joint areas exceeding some square-centimeters have shown that they can only successfully be produced when graphite is brazed to a metal, such as tungsten or molybdenum with a coefficient of thermal expansion closely matching that of graphite. Therefore all experimental work on evaluation of joints has been concentrated on molybdenum/graphite brazings. The tensile strength of molybdenum/graphite-composites compares favorably with the tensile strength of bulk graphite from room temperature close to the melting temperature of the braze. In electron beam testing the threshold damage line for molybdenum/graphite-composites has been evaluated. Results show that even composites with the low melting AgCuTi-braze are expected to withstand 10 MW/m 2 power density for at least 10 3 cycles. Limiter testing in TEXTOR shows that molybdenum/graphite-segments with 3 mm graphite brazed on molybdenum-substrate withstand severe repeated TEXTOR plasma discharge conditions without serious damage. Results prove that actively cooled components on the basis of a molybdenum/graphite-composite can sustain a higher heat flux than bulk graphite alone. (author)

  14. Fusion energy in an inertial electrostatic confinement device using a magnetically shielded grid

    Energy Technology Data Exchange (ETDEWEB)

    Hedditch, John, E-mail: john.hedditch@sydney.edu.au; Bowden-Reid, Richard, E-mail: rbow3948@physics.usyd.edu.au; Khachan, Joe, E-mail: joe.khachan@sydney.edu.au [School of Physics, The University of Sydney, Sydney, New South Whales 2006 (Australia)

    2015-10-15

    Theory for a gridded inertial electrostatic confinement (IEC) fusion system is presented, which shows a net energy gain is possible if the grid is magnetically shielded from ion impact. A simplified grid geometry is studied, consisting of two negatively biased coaxial current-carrying rings, oriented such that their opposing magnetic fields produce a spindle cusp. Our analysis indicates that better than break-even performance is possible even in a deuterium-deuterium system at bench-top scales. The proposed device has the unusual property that it can avoid both the cusp losses of traditional magnetic fusion systems and the grid losses of traditional IEC configurations.

  15. Development of electrical insulation and conduction coating for fusion experimental devices

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, M. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan); Tsujimura, S. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan); Toyoda, M. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan); Inoue, M. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan); Abe, T. [Japan Atomic Energy Research Inst., Naka (Japan); Murakami, Y. [Japan Atomic Energy Research Inst., Naka (Japan)

    1995-12-31

    Development of electrical insulation and conduction methods that can be applied for large components have been investigated for future large fusion experimental devices. A thermal spraying method is employed to coat the insulation or conduction materials on the structural components. Al{sub 2}O{sub 3} has been selected as an insulation material, while Cr{sub 3}C{sub 2}-NiCr and WC-NiCr have been chosen as conduction materials. These materials were coated on stainless steel base plates to examine the basic characteristics of the coated layers, such as their adhesive strength to the base plate and electrical resistance. It was found that they have sufficient electrical insulation and conduction properties, respectively. In addition, the sliding tests of the coated layers showed sufficient frictional properties. The applicability of the spraying method was examined on a 100mm x 1000mm surface and found to be applicable for large surfaces in fusion experimental devices. (orig.).

  16. Development of electrical insulation and conduction coating for fusion experimental devices

    International Nuclear Information System (INIS)

    Onozuka, M.; Tsujimura, S.; Toyoda, M.; Inoue, M.; Abe, T.; Murakami, Y.

    1995-01-01

    Development of electrical insulation and conduction methods that can be applied for large components have been investigated for future large fusion experimental devices. A thermal spraying method is employed to coat the insulation or conduction materials on the structural components. Al 2 O 3 has been selected as an insulation material, while Cr 3 C 2 -NiCr and WC-NiCr have been chosen as conduction materials. These materials were coated on stainless steel base plates to examine the basic characteristics of the coated layers, such as their adhesive strength to the base plate and electrical resistance. It was found that they have sufficient electrical insulation and conduction properties, respectively. In addition, the sliding tests of the coated layers showed sufficient frictional properties. The applicability of the spraying method was examined on a 100mm x 1000mm surface and found to be applicable for large surfaces in fusion experimental devices. (orig.)

  17. Fusion

    CERN Document Server

    Mahaffey, James A

    2012-01-01

    As energy problems of the world grow, work toward fusion power continues at a greater pace than ever before. The topic of fusion is one that is often met with the most recognition and interest in the nuclear power arena. Written in clear and jargon-free prose, Fusion explores the big bang of creation to the blackout death of worn-out stars. A brief history of fusion research, beginning with the first tentative theories in the early 20th century, is also discussed, as well as the race for fusion power. This brand-new, full-color resource examines the various programs currently being funded or p

  18. THIDA: code system for calculation of the exposure dose rate around a fusion device

    International Nuclear Information System (INIS)

    Iida, Hiromasa; Igarashi, Masahito.

    1978-12-01

    A code system THIDA has been developed for calculation of the exposure dose rates around a fusion device. It consists of the following: one- and two-dimensional discrete ordinate transport codes; induced activity calculation code; activation chain, activation cross section, radionuclide gamma-ray energy/intensity and gamma-ray group constant files; and gamma ray flux to exposure dose rate conversion coefficients. (author)

  19. Plasma sprayed TiC coatings for first wall protection in fusion devices

    International Nuclear Information System (INIS)

    Groot, P.; Laan, J.G. van der; Laas, L.; Mack, M.; Dvorak, M.

    1989-01-01

    For protection of plasma facing components in nuclear fusion devices thick titanium carbide coatings are being developed. Coatings have been produced by plasma spraying at atmospheric pressure (APS) and low pressure (LPPS) and analyzed with respect to microstructure and chemical composition. Thermo-mechanical evaluation has been performed by applying short pulse laser heat flux tests. The influence of coating thickness and porosity on the resistance to spalling by thermal shocks appears to be more important than aspects of chemical composition. (author)

  20. Irradiation devices for fusion reactor materials results obtained from irradiated lithium aluminate at the OSIRIS reactor

    International Nuclear Information System (INIS)

    Lefevre, F.; Thevenot, G.; Rasneur, B.; Botter, F.

    1986-06-01

    Studies about controlled fusion reactor of the Tokamak type require the examination of the radiation effects on the behaviour of various potential materials. Thus, in the first part of this paper, are presented the devices adapted to these materials studies and used in the OSIRIS reactor. In a second part, is described an experiment of irradiation ceramics used as candidates for breeding material and are given the first results

  1. Sausage instability of Z-discharged plasma channel in LIB-fusion device

    International Nuclear Information System (INIS)

    Murakami, H.; Kawata, S.; Niu, K.

    1982-07-01

    Current-carring plasma channels have been proposed for transporting intense ion beams from diodes to a target in a LIB-fusion device. In this paper, the growth rate of the most dangerous surface mode, that is, axisymmetric sausage instability is examined for the plasma channel. The growth rate is shown to be smaller than that of the plasma channel with no fluid motion in a sharp boundary. It is concluded that the stable plasma channel can be formed. (author)

  2. Characterization of the Plasma Edge for Technique of Atomic Helium Beam in the CIEMAT Fusion Device

    International Nuclear Information System (INIS)

    Hidalgo, A.

    2003-01-01

    In this report, the measurement of Electron Temperature and Density in the Boundary Plasma of TJ-II with a Supersonic Helium Beam Diagnostic and work devoted to the upgrading of this technique are described. Also, simulations of Laser Induced Fluorescence (LIF) studies of level populations of electronically excited He atoms are shown. This last technique is now being installed in the CIEMAT fusion device. (Author )

  3. The measurement of potential distribution of plasma in MM-4 fusion device

    International Nuclear Information System (INIS)

    Tian Zhongyu; Ming Linzhou; Feng Xiaozhen; Feng Chuntang; Yi Youjun; Wang Jihai; Liu Yihua

    1988-11-01

    Some experimental results of the potential distribution in MM-4 fusion device are presented by measuring the floating potential of probe. The results showed that the distribution of axial potential is asymmetrical, but the radial potential is symmetrical. There are double ion potential wells in the plasma. The depth of the deepest potential well become deeper is the strength of the magnetic field and injection current are increasing. The location of the deepest well is moved towards the device center along with the increasing of injection energy. This is different from others results. The mechanism of causing this distribution in also discussed

  4. Properties of plasma sheath with ion temperature in magnetic fusion devices

    International Nuclear Information System (INIS)

    Liu Jinyuan; Wang Feng; Sun Jizhong

    2011-01-01

    The plasma sheath properties in a strong magnetic field are investigated in this work using a steady state two-fluid model. The motion of ions is affected heavily by the strong magnetic field in fusion devices; meanwhile, the effect of ion temperature cannot be neglected for the plasma in such devices. A criterion for the plasma sheath in a strong magnetic field, which differs from the well-known Bohm criterion for low temperature plasma sheath, is established theoretically with a fluid model. The fluid model is then solved numerically to obtain detailed sheath information under different ion temperatures, plasma densities, and magnetic field strengths.

  5. DD fusion neutron production at UW-Madison using IEC devices

    Science.gov (United States)

    Fancher, Aaron; Michalak, Matt; Kulcinski, Gerald; Santarius, John; Bonomo, Richard

    2017-10-01

    An inertial electrostatic confinement (IEC) device using spherical, gridded electrodes at high voltage accelerates deuterium ions, allowing for neutrons to be produced within the device from DD fusion reactions. The effects of the device cathode voltage (30-170 kV), current (30-100 mA), and pressure (0.15-1.25 mTorr) on the neutron production rate have been measured. New high voltage capabilities have resulted in the achievement of a steady state neutron production rate of 3.3x108 n/s at 175 kV, 100 mA, and 1.0 mTorr of deuterium. Applications of IEC devices include the production of DD neutrons to detect chemical explosives and special nuclear materials using active interrogation methods. Research supported by US Dept. of Homeland Security Grant 2015-DN-077-AR1095 and the Grainger Foundation.

  6. Characterization of size, composition and origins of dust in fusion devices. Summary report of the 1. research coordination meeting

    International Nuclear Information System (INIS)

    Clark, R.E.H.

    2009-03-01

    Nine experts on dust formation and their physical and behavioural characteristics attended the first Research Coordination Meeting (RCM) on Characterization of Size, Composition and Origins of Dust in Fusion Devices held at IAEA Headquarters on 10-12 December 2008. Participants summarized recent relevant developments related to dust in fusion devices. The specific objectives of the CRP and a detailed work plan were formulated. Discussions, conclusions and recommendations of the RCM are briefly described in this report. (author)

  7. The Roles and Developments needed for Diagnostics in the ITER Fusion Device

    Energy Technology Data Exchange (ETDEWEB)

    Walsh, Michael [ITER Organization, Route de Vinon-sur-Verdon - CS 90046, 13067 St Paul-lez-Durance Cedex (France)

    2015-07-01

    Harnessing the power from Fusion on earth is an important and challenging task. Excellent work has been carried out in this area over the years with several demonstrations of the ability to produce power. Now, a new large device is being constructed in the south of France. This is called ITER. ITER is a large-scale scientific experiment that aims to demonstrate a possibility to produce commercial energy from fusion. This project is now well underway with the many teams working on the construction and completing various aspects of the design. This device will carry up to 15 MA of plasma current and produce about 500 MW of power, 400 MW approximately in high energy neutrons. The typical temperatures of the electrons inside this device are in the region of a few hundred million Kelvin. It is maintained using a magnetic field. This device is pushing several boundaries from those currently existing. As a result of this, several technologies need to be developed or extended. This is especially true for the systems or diagnostics that measure the performance and provide the control signals for this device. A diagnostic set will be installed on the ITER machine to provide the measurements necessary to control, evaluate and optimize plasma performance in ITER and to further the understanding of plasma physics. These include amongst others, measurements of the plasma shape, temperature, density, impurity concentration, and particle and energy confinement times. The system will comprise about 45 individual measuring systems drawn from the full range of modern plasma diagnostic techniques, including magnetics, lasers, X-rays, neutron cameras, impurity monitors, particle spectrometers, radiation bolometers, pressure and gas analysis, and optical fibres. These devices will have to be made to work in the new and challenging environment inside the vacuum vessel. These systems will have to cope with a range of phenomena that extend the current knowledge in the Fusion field. One

  8. Studies on advanced superconductors for fusion device. Pt. 1. Present status of Nb3Sn conductors

    International Nuclear Information System (INIS)

    Tachikawa, Kyoji; Yamamoto, Junya

    1996-03-01

    Nb 3 Sn conductors have been developed with great expectation as an advanced high-field superconductor to be used in fusion devices of next generation. Furthermore, Nb 3 Sn conductors are being developed for NMR magnet and superconducting generator as well as for cryogen-free superconducting magnet. A variety of fabrication procedures, such as bronze process, internal tin process and Nb tube method, have been developed based on the diffusion reaction. Recently, Nb 3 Sn conductors with ultra-thin filaments have been fabricated for AC use. Both high-field and AC performances of Nb 3 Sn conductors have been significantly improved by alloying addition. The Ti-doped Nb 3 Sn conductor has generated 21.5T at 1.8K operation. This report summarizes manufacturing procedures, superconducting performances and applications of Nb 3 Sn conductors fabricated through different processes in different countries. More detailed subjects included in this report are high-field properties, AC properties, conductors for fusion with large current capacities, stress-strain effect and irradiation effect as well as standardization of critical current measurement method regarding to Nb 3 Sn conductors. Comprehensive grasp on the present status of Nb 3 Sn conductors provided by this report will act as a useful data base for the future planning of fusion devices. (author). 172 refs

  9. Role of inert gases in first wall phenomena in fusion devices

    International Nuclear Information System (INIS)

    Das, S.K.

    1979-01-01

    The first wall surfaces of fusion devices will be exposed to bombardment by inert gaseous projectiles such as helium. The flux, energy and angular distribution of the helium radiation will depend not only on the type of device but also on its design parameters. For near term tokamak devices, the first wall surface phenomena caused by helium bombardment that appear to be quite important are physical sputtering and radiation blistering. Examples of these processes for a number of first wall candidate materials are discussed. While the physical sputtering phenomen is well understood, the mechanism of blister formation is still not fully understood. The various models proposed for radiation blistering of metal during helium bombardment is critically reviewed in the light of most recent experimental results

  10. Summary of the 16th IAEA Technical Meeting on 'Research using Small Fusion Devices'

    International Nuclear Information System (INIS)

    Gribkov, V.; Oost, G. van; Malaquias, A.; Herrera, J.

    2006-01-01

    Common research topics that are being studied in small, medium and large devices such as H-mode like or improved confinement, turbulence and transport are reported. These included modelling and diagnostic developments for edge and core, to characterize plasma density, temperature, electric potential, plasma flows, turbulence scale, etc. Innovative diagnostic methods were designed and implemented which could be used to develop experiments in small devices (in some cases not possible in large devices due to higher power deposition) to allow a better understanding of plasma edge and core properties. Reports are given addressing research in linear devices that can be used to study particular plasma physics topics relevant for other magnetic confinement devices such as the radial transport and the modelling of self-organized plasma jets involved in spheromak-like plasma formation. Some aspects of the work presented are of interest to the astrophysics community since they are believed to shed light on the basis of the physics of stellar jets. On the dense magnetized plasmas (DMP) topic, the present status of research, operation of new devices, plasma dynamics modelling and diagnostic developments is reported. The main devices presented belong to the class of Z-pinches, mostly plasma foci, and several papers were presented under this topic. The physics of DMP is important both for the main-stream fusion investigations as well as for providing the basis for elaboration of new concepts. New high-current technology introduced in the DMP devices design and construction make these devices nowadays more reliably fitted to various applications and give the possibility to widen the energy range used by them in both directions-to the multi-MJ level facilities and down to miniature plasma focus devices with energy of just a few J. (conference report)

  11. TASKA-M - a low cost, near term tandem mirror device for fusion technology testing

    International Nuclear Information System (INIS)

    Badger, B.; Corradini, M.L.; El-Guebaly, L.; Emmert, G.A.; Kulcinski, G.L.; Larsen, E.M.; Maynard, C.W.; Perkins, L.J.; Peterson, R.R.; Plute, K.E.; Santarius, J.F.; Sawan, M.E.; Scharer, J.E.; Sviatoslavsky, I.N.; Sze, D.K.; Vogelsang, W.F.; Wittenberg, L.J.; Leppelmeier, G.W.; Grover, J.M.; Opperman, E.K.; Vogel, M.A.; Borie, E.; Taczanowski, S.; Arendt, F.; Dittrich, H.G.; Fett, T.; Haferkamp, B.; Heinz, W.; Hoelzchen, E.; Kleefeldt, K.; Klingelhoefer, R.; Komarek, P.; Kuntze, M.; Leiste, H.G.; Link, W.; Malang, S.; Manes, B.M.; Maurer, W.; Michael, I.; Mueller, R.A.; Neffe, G.; Schramm, K.; Suppan, A.; Weinberg, D.

    1984-04-01

    TASKA-M (Modifizierte Tandem Spiegelmaschine Karlsruhe) is a study of a dedicated fusion technology device based on the mirror principle, in continuation of the 1981/82 TASKA study. The main objective is to minimize cost while retaining key requirements of neutron flux and fluence for blanket and material development and for component testing in a nuclear environment. Direct costs are reduced to about 400 M$ by dropping reactor-relevant aspects not essential to technology testing: No thermal barrier and electrostatic plugging of the plasma; fusion power of 7 MW at an injected power of 44 MW; tritium supply from external sources. All technologies for operating the machine are expected to be available by 1990; the plasma physics relies on microstabilization in a sloshing ion population. (orig.) [de

  12. Multilayer mirror based monitors for impurity controls in large fusion reactor type devices

    International Nuclear Information System (INIS)

    Regan, S.P.; May, M.J.; Soukhanovskii, V.; Finkenthal, M.; Moos, H.W.

    1995-01-01

    Multilayer Mirror (MLM) based monitors are compact, high throughput diagnostics capable of extracting XUV emissions (the wavelength range including the soft-x-ray and the extreme ultraviolet, 10 angstrom to 304 angstrom) of impurities from the harsh environment of large fusion reactor type devices. For several years the Plasma Spectroscopy Group at Johns Hopkins University has investigated the application of MLM based XUV spectroscopic diagnostics for magnetically confined fusion plasmas. MLM based monitors have been constructed for and extensively used on DIII-D, Alcator C-mod, TEXT, Phaedrus-T, and CDX-U tokamaks to study the impurity behavior of elements ranging from He to Mo. On ITER MLM based devices would be used to monitor the spectral line emissions from Li I-like to F I-like charge states of Fe, Cr, and Ni, as well as extractors for the bands of emissions from high Z elements such as Mo or W for impurity controls of the fusion plasma. In addition to monitoring the impurity emissions from the main plasma, MLM based devices can also be adapted for radiation measurements of low Z elements in the divertor. The concepts and designs of these MLM based monitors for impurity controls in ITER will be presented. The results of neutron irradiation experiments of the MLMs performed in the Los Alamos Spallation Radiation Effects Facility (LASREF) at the Los Alamos National Laboratory will also be discussed. These preliminary neutron exposure studies show that the dispersive and reflective qualities of the MLMs were not affected in a significant manner

  13. Development and Testing of Atomic Beam-Based Plasma Edge Diagnostics in the CIEMAT Fusion Devices

    International Nuclear Information System (INIS)

    Tafalla, D.; Tabares, F.L.; Ortiz, P.; Herrero, V.J.; Tanarro, I.

    1998-01-01

    In this report the development of plasma edge diagnostic based on atomic beam techniques fir their application in the CIEMAT fusion devices is described. The characterisation of the beams in laboratory experiments at the CSIC, together with first results in the Torsatron TJ-II are reported. Two types of beam diagnostics have been developed: a thermal (effusive) Li and a supersonic, pulsed He beams. This work has been carried out in collaboration between the institutions mentioned above under partial financial support by EURATOM. (Author) 17 refs

  14. Energy system for the generation of divertor magnetic fields in the PDX fusion research device

    International Nuclear Information System (INIS)

    Turitzin, N.M.

    1975-01-01

    One of the major problems encountered in the development of Tokamak type fusion reactors is the presence of impurities in the plasma. The PDX device is designed to study the operation of poloidal magnetic field divertors and consequent magnetic limiters for controlling and reducing the amount of impurities. A system of coils placed at specific locations produces a required field configuration for the poloidal divertor. This paper describes the system of energy supplies required and the interrelations of field coil currents during plasma current initiation, growth and steady state

  15. Fusion Engineering Device. Volume VI. Complementary development plan for engineering development

    International Nuclear Information System (INIS)

    1981-10-01

    The basic approach followed in this volume is to define key technical issues for several fusion reactor technologies and to device program strategies to resolve each of these issues. Particular attention has been paid to elucidating the role of FED vis-a-vis complementary (non-FED) facilities in this process. The remainder of this chapter consists of summaries of the major conclusions of the technology plans in each of the areas studied, i.e., plasma heating, magnetics, nuclear, and systems considerations

  16. IAEA technical meeting on nuclear data library for advanced systems - Fusion devices

    International Nuclear Information System (INIS)

    Forrest, R.; Mengoni, A.

    2008-04-01

    A Technical Meeting on 'Nuclear Data Library for Advanced Systems - Fusion Devices' was held at the IAEA Headquarters in Vienna from 31 October to 2 November 2007. The main objective of the initiative has been to define a proposal and detailed plan of activities for a Co-ordinated Research Project on this subject. Details of the discussions which took place at the meeting, including a review of the current activities in the field, a list of recommendations and a proposed timeline schedule for the CRP are summarized in this report. (author)

  17. Performance of large-scale helium refrigerators subjected to pulsed heat load from fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Dutta, R.; Ghosh, P.; Chowdhury, K. [Cryogenic Engineering Centre, Indian Institute of Technology, Kharagpur (India)

    2012-07-01

    The immediate effect of pulsed heat load from fusion devices in helium refrigerators is wide variation in mass flow rate of low pressure stream returning to the cold-box. In this paper, a four expander based modified Claude cycle has been analyzed in quasi steady and dynamic simulations using Aspen HYSYS to identify critical equipment that may be affected due to such flow rate fluctuations at the return stream and their transient performance. Additional constraints on process parameters over steady state design have been identified. Suitable techniques for mitigation of fluctuation of return stream have also been explored. (author)

  18. Performance of large-scale helium refrigerators subjected to pulsed heat load from fusion devices

    International Nuclear Information System (INIS)

    Dutta, R.; Ghosh, P.; Chowdhury, K.

    2012-01-01

    The immediate effect of pulsed heat load from fusion devices in helium refrigerators is wide variation in mass flow rate of low pressure stream returning to the cold-box. In this paper, a four expander based modified Claude cycle has been analyzed in quasi steady and dynamic simulations using Aspen HYSYS to identify critical equipment that may be affected due to such flow rate fluctuations at the return stream and their transient performance. Additional constraints on process parameters over steady state design have been identified. Suitable techniques for mitigation of fluctuation of return stream have also been explored. (author)

  19. Energy system for the generation of divertor magnetic fields in the PDX fusion research device

    International Nuclear Information System (INIS)

    Turitzin, N.M.

    1976-05-01

    One of the major problems encountered in the development of Tokamak type fusion reactors is the presence of impurities in the plasma. The PDX device is designed to study the operation of poloidal magnetic field divertors and consequent magnetic limiters for controlling and reducing the amount of impurities. A system of coils placed at specific locations produces a required field configuration for the poloidal divertor. This paper describes the system of energy supplies required and the interrelations of field coil currents during plasma current initiation, growth and steady state

  20. 7. IAEA Technical Meeting on Steady State Operation of Magnetic Fusion Devices - Booklet of abstracts

    International Nuclear Information System (INIS)

    2015-01-01

    This meeting has provided an appropriate forum to discuss current issues covering a wide range of technical topics related to the steady state operation issues and also to encourage forecast of the ITER performances. The technical meeting includes invited and contributed papers. The topics that have been dealt with are: 1) Superconducting devices (ITER, KSTAR, Tore-Supra, HT-7U, EAST, LHD, Wendelstein-7-X,...); 2) Long-pulse operation and advanced tokamak physics; 3) steady state fusion technologies; 4) Long pulse heating and current drive; 5) Particle control and power exhaust, and 6) ITER-related research and development issues. This document gathers the abstracts

  1. On fractal properties of equipotentials over a real rough surface faced to plasma in fusion devices

    International Nuclear Information System (INIS)

    Budaev, V.P.; Yakovlev, M.

    2008-01-01

    We consider a sheath region bounded by a corrugated surface of material conductor and a flat boundary held to a constant voltage bias. The real profile of the film deposited from plasma on a limiter in a fusion device was used in numerical solving of the Poisson's equation to find a profile of electrostatic potential. The rough surface influences the equipotential lines over the surface. We characterized a shape of equipotential lines by a fractal dimension. The long-range correlation in the potential field is imposed by the non-trivial fractal structure of the surface. Dust particles bounced in such irregular potential field can accelerate due to the Fermi acceleration. (author)

  2. Device for supporting a toroidal coil in a toroidal type nuclear fusion device

    International Nuclear Information System (INIS)

    Kitazawa, Hakaru; Sato, Hiroshi.

    1975-01-01

    Object: To easily manufacture a center block having a strength sufficient to withstand an electromagnetic force exerted on the center of toroidal of a toroidal coil and to increase its reliability. Structure: In a device for supporting toroidal coils wherein the electromagnetic force exerted on the center of toroidal of a plurality of toroidal coils arranged in toroidal fashion, the contact surface between the toroidal coil and the center block is arranged parallel to the center axis of toroidal so as to receive the electromagnetic force exerted on the center of toroidal of the toroidal coil as the component of force in a radial direction. (Taniai, N.)

  3. Effects of Lumbar Fusion Surgery with ISOBAR Devices Versus Posterior Lumbar Interbody Fusion Surgery on Pain and Disability in Patients with Lumbar Degenerative Diseases: A Meta-Analysis.

    Science.gov (United States)

    Su, Shu-Fen; Wu, Meng-Shan; Yeh, Wen-Ting; Liao, Ying-Chin

    2018-06-01

    Purpose/Aim: Lumbar degenerative diseases (LDDs) cause pain and disability and are treated with lumbar fusion surgery. The aim of this study was to evaluate the efficacy of lumbar fusion surgery with ISOBAR devices versus posterior lumbar interbody fusion (PLIF) surgery for alleviating LDD-associated pain and disability. We performed a literature review and meta-analysis conducted in accordance with Cochrane methodology. The analysis included Group Reading Assessment and Diagnostic Evaluation assessments, Jadad Quality Score evaluations, and Risk of Bias in Non-randomized Studies of Interventions assessments. We searched PubMed, MEDLINE, the Cumulative Index to Nursing and Allied Health Literature, the Cochrane Library, ProQuest, the Airiti Library, and the China Academic Journals Full-text Database for relevant randomized controlled trials and cohort studies published in English or Chinese between 1997 and 2017. Outcome measures of interest included general pain, lower back pain, and disability. Of the 18 studies that met the inclusion criteria, 16 examined general pain (802 patients), 5 examined lower back pain (274 patients), and 15 examined disability (734 patients). General pain, lower back pain, and disability scores were significantly lower after lumbar fusion surgery with ISOBAR devices compared to presurgery. Moreover, lumbar fusion surgery with ISOBAR devices was more effective than PLIF for decreasing postoperative disability, although it did not provide any benefit in terms of general pain or lower back pain. Lumbar fusion surgery with ISOBAR devices alleviates general pain, lower back pain, and disability in LDD patients and is superior to PLIF for reducing postoperative disability. Given possible publication bias, we recommend further large-scale studies.

  4. Oscillatory vapour shielding of liquid metal walls in nuclear fusion devices.

    Science.gov (United States)

    van Eden, G G; Kvon, V; van de Sanden, M C M; Morgan, T W

    2017-08-04

    Providing an efficacious plasma facing surface between the extreme plasma heat exhaust and the structural materials of nuclear fusion devices is a major challenge on the road to electricity production by fusion power plants. The performance of solid plasma facing surfaces may become critically reduced over time due to progressing damage accumulation. Liquid metals, however, are now gaining interest in solving the challenge of extreme heat flux hitting the reactor walls. A key advantage of liquid metals is the use of vapour shielding to reduce the plasma exhaust. Here we demonstrate that this phenomenon is oscillatory by nature. The dynamics of a Sn vapour cloud are investigated by exposing liquid Sn targets to H and He plasmas at heat fluxes greater than 5 MW m -2 . The observations indicate the presence of a dynamic equilibrium between the plasma and liquid target ruled by recombinatory processes in the plasma, leading to an approximately stable surface temperature.Vapour shielding is one of the interesting mechanisms for reducing the heat load to plasma facing components in fusion reactors. Here the authors report on the observation of a dynamic equilibrium between the plasma and the divertor liquid Sn surface leading to an overall stable surface temperature.

  5. Probabilistic Multi-Sensor Fusion Based Indoor Positioning System on a Mobile Device

    Directory of Open Access Journals (Sweden)

    Xiang He

    2015-12-01

    Full Text Available Nowadays, smart mobile devices include more and more sensors on board, such as motion sensors (accelerometer, gyroscope, magnetometer, wireless signal strength indicators (WiFi, Bluetooth, Zigbee, and visual sensors (LiDAR, camera. People have developed various indoor positioning techniques based on these sensors. In this paper, the probabilistic fusion of multiple sensors is investigated in a hidden Markov model (HMM framework for mobile-device user-positioning. We propose a graph structure to store the model constructed by multiple sensors during the offline training phase, and a multimodal particle filter to seamlessly fuse the information during the online tracking phase. Based on our algorithm, we develop an indoor positioning system on the iOS platform. The experiments carried out in a typical indoor environment have shown promising results for our proposed algorithm and system design.

  6. High-Z plasma facing components in fusion devices: boundary conditions and operational experiences

    Science.gov (United States)

    Neu, R.

    2006-04-01

    In present day fusion devices optimization of the performance and experimental freedom motivates the use of low-Z plasma facing materials (PFMs). However, in a future fusion reactor, for economic reasons, a sufficient lifetime of the first wall components is essential. Additionally, tritium retention has to be small to meet safety requirements. Tungsten appears to be the most realistic material choice for reactor plasma facing components (PFCs) because it exhibits the lowest erosion. But besides this there are a lot of criteria which have to be fulfilled simultaneously in a reactor. Results from present day devices and from laboratory experiments confirm the advantages of high-Z PFMs but also point to operational restrictions, when using them as PFCs. These are associated with the central impurity concentration, which is determined by the sputtering yield, the penetration of the impurities and their transport within the confined plasma. The restrictions could exclude successful operation of a reactor, but concomitantly there exist remedies to ameliorate their impact. Obviously some price has to be paid in terms of reduced performance but lacking of materials or concepts which could substitute high-Z PFCs, emphasis has to be put on the development and optimization of reactor-relevant scenarios which incorporate the experiences and measures.

  7. High-Z plasma facing components in fusion devices: boundary conditions and operational experiences

    International Nuclear Information System (INIS)

    Neu, R.

    2006-01-01

    In present day fusion devices optimization of the performance and experimental freedom motivates the use of low-Z plasma facing materials (PFMs). However, in a future fusion reactor, for economic reasons, a sufficient lifetime of the first wall components is essential. Additionally, tritium retention has to be small to meet safety requirements. Tungsten appears to be the most realistic material choice for reactor plasma facing components (PFCs) because it exhibits the lowest erosion. But besides this there are a lot of criteria which have to be fulfilled simultaneously in a reactor. Results from present day devices and from laboratory experiments confirm the advantages of high-Z PFMs but also point to operational restrictions, when using them as PFCs. These are associated with the central impurity concentration, which is determined by the sputtering yield, the penetration of the impurities and their transport within the confined plasma. The restrictions could exclude successful operation of a reactor, but concomitantly there exist remedies to ameliorate their impact. Obviously some price has to be paid in terms of reduced performance but lacking of materials or concepts which could substitute high-Z PFCs, emphasis has to be put on the development and optimization of reactor-relevant scenarios which incorporate the experiences and measures

  8. A study of hydrogen isotopes fuel control by wall effect in magnetic fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Motevalli, S.M., E-mail: motavali@umz.ac.ir; Safari, M.

    2016-11-15

    Highlights: • A particle balance model for the main plasma and wall inventory in magnetic fusion device has been represented. • The dependence of incident particles energy on the wall has been considered in 10–300 eV for the sputtering yield and recycling coefficient. • The effect of fueling methods on plasma density behavior has been studied. - Abstract: Determination of plasma density behavior in magnetic confinement system needs to study the plasma materials interaction in the facing components such as first wall, limiter and divertor. Recycling of hydrogen isotope is an effective parameter in plasma density rate and plasma fueling. Recycling coefficient over the long pulse operation, gets to the unity, so it has a significant effect on steady state in magnetic fusion devices. Typically, sputtered carbon atoms from the plasma facing components form hydrocarbons and they redeposit on the wall. In this case little rate of hydrogen loss occurs. In present work a zero dimensional particle equilibrium model has been represented to determine particles density rate in main plasma and wall inventory under recycling effect and codeposition of hydrogen in case of continues and discontinues fueling methods and effective parameters on the main plasma decay has been studied.

  9. Low-Z coating as a first wall of nuclear fusion devices

    International Nuclear Information System (INIS)

    Shikama, Tatsuo; Okada, Masatoshi

    1984-01-01

    The tokamak nuclear fusion devices of the largest scale in the world, TFTR in USA and JET in Europe, started the operation from the end of 1982 to 1983. Also in Japan, the tokamak JT-60 is scheduled to begin the operation in 1985. One of the technological obstacles is the problem of first walls facing directly to plasma and subjected to high particle loading and thermal loading. Moreover, first walls achieve the active role of controlling impurities in plasma and recycling hydrogen isotopes. It is impossible to find a single material which satisfies all these requirements. The compounding of materials can create a material having new function, but also has the meaning of expanding the range of material selection. One of the material compounding methods is surface coating. In this paper, as the materials for first walls, the characteristics of low Z materials are discussed from the design examples of actual takamak nuclear fusion devices. The outline of first walls is explained. High priority is given to the impurity control in plasma, and in view of plasma energy emissivity and the rate of self sputtering, low Z material coating seems to be the solution. The merits and the problems of such low Z material coating are discussed. (Kako, I.)

  10. Modelling of surface evolution of rough surface on divertor target in fusion devices

    International Nuclear Information System (INIS)

    Dai, Shuyu; Liu, Shengguang; Sun, Jizhong; Kirschner, A.; Kawamura, G.; Tskhakaya, D.; Ding, Rui; Luo, Guangnan; Wang, Dezhen

    2015-01-01

    Highlights: • We study the surface evolution of rough surface on divertor target in fusion devices. • The effects of gyration motion and E × B drift affect 3D angular distribution. • A larger magnetic field angle leads to a reduced net eroded areal density. • The rough surface evolution affects the physical sputtering yield. - Abstract: The 3D Monte-Carlo code SURO has been used to study the surface evolution of rough surface on the divertor target in fusion devices. The edge plasma at divertor region is modelled by the SDPIC code and used as input data for SURO. Coupled with SDPIC, SURO can perform more sophisticated simulations to calculate the local angle and surface evolution of rough surface. The simulation results show that the incident direction of magnetic field, gyration and E × B force has a significant impact on 3D angular distribution of background plasma and accordingly on the erosion of rough surface. The net eroded areal density of rough surface is studied by varying the magnetic field angle with surface normal. The evolution of the microscopic morphology of rough surface can lead to a significant change in the physical sputtering yield

  11. EMP Fusion

    OpenAIRE

    KUNTAY, Isık

    2010-01-01

    This paper introduces a novel fusion scheme, called EMP Fusion, which has the promise of achieving breakeven and realizing commercial fusion power. The method is based on harnessing the power of an electromagnetic pulse generated by the now well-developed flux compression technology. The electromagnetic pulse acts as a means of both heating up the plasma and confining the plasma, eliminating intermediate steps. The EMP Fusion device is simpler compared to other fusion devices and this reduces...

  12. Impaction durability of porous polyether-ether-ketone (PEEK) and titanium-coated PEEK interbody fusion devices.

    Science.gov (United States)

    Torstrick, F Brennan; Klosterhoff, Brett S; Westerlund, L Erik; Foley, Kevin T; Gochuico, Joanna; Lee, Christopher S D; Gall, Ken; Safranski, David L

    2018-05-01

    Various surface modifications, often incorporating roughened or porous surfaces, have recently been introduced to enhance osseointegration of interbody fusion devices. However, these topographical features can be vulnerable to damage during clinical impaction. Despite the potential negative impact of surface damage on clinical outcomes, current testing standards do not replicate clinically relevant impaction loading conditions. The purpose of this study was to compare the impaction durability of conventional smooth polyether-ether-ketone (PEEK) cervical interbody fusion devices with two surface-modified PEEK devices that feature either a porous structure or plasma-sprayed titanium coating. A recently developed biomechanical test method was adapted to simulate clinically relevant impaction loading conditions during cervical interbody fusion procedures. Three cervical interbody fusion devices were used in this study: smooth PEEK, plasma-sprayed titanium-coated PEEK, and porous PEEK (n=6). Following Kienle et al., devices were impacted between two polyurethane blocks mimicking vertebral bodies under a constant 200 N preload. The posterior tip of the device was placed at the entrance between the polyurethane blocks, and a guided 1-lb weight was impacted upon the anterior face with a maximum speed of 2.6 m/s to represent the strike force of a surgical mallet. Impacts were repeated until the device was fully impacted. Porous PEEK durability was assessed using micro-computed tomography (µCT) pre- and postimpaction. Titanium-coating coverage pre- and postimpaction was assessed using scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy. Changes to the surface roughness of smooth and titanium-coated devices were also evaluated. Porous PEEK and smooth PEEK devices showed minimal macroscopic signs of surface damage, whereas the titanium-coated devices exhibited substantial visible coating loss. Quantification of the porous PEEK deformation

  13. Motion-sensor fusion-based gesture recognition and its VLSI architecture design for mobile devices

    Science.gov (United States)

    Zhu, Wenping; Liu, Leibo; Yin, Shouyi; Hu, Siqi; Tang, Eugene Y.; Wei, Shaojun

    2014-05-01

    With the rapid proliferation of smartphones and tablets, various embedded sensors are incorporated into these platforms to enable multimodal human-computer interfaces. Gesture recognition, as an intuitive interaction approach, has been extensively explored in the mobile computing community. However, most gesture recognition implementations by now are all user-dependent and only rely on accelerometer. In order to achieve competitive accuracy, users are required to hold the devices in predefined manner during the operation. In this paper, a high-accuracy human gesture recognition system is proposed based on multiple motion sensor fusion. Furthermore, to reduce the energy overhead resulted from frequent sensor sampling and data processing, a high energy-efficient VLSI architecture implemented on a Xilinx Virtex-5 FPGA board is also proposed. Compared with the pure software implementation, approximately 45 times speed-up is achieved while operating at 20 MHz. The experiments show that the average accuracy for 10 gestures achieves 93.98% for user-independent case and 96.14% for user-dependent case when subjects hold the device randomly during completing the specified gestures. Although a few percent lower than the conventional best result, it still provides competitive accuracy acceptable for practical usage. Most importantly, the proposed system allows users to hold the device randomly during operating the predefined gestures, which substantially enhances the user experience.

  14. Improved zero dimensional model of a reversed field pinch fusion device

    International Nuclear Information System (INIS)

    Haynes, K.E.

    1987-01-01

    A zero-dimensional model has been developed which accurately predicts conditions observed during several runs of the ZT-40M reversed field pinch fusion device at Los Alamos National Laboratory. The model is based on a physical model developed by E.H. Klevans at Penn State University. Improvements made to this model included the use of coronal non-equilibrium equations for predicting impurity effects, the inclusion of an exponentially decaying ion heating term, and the relaxation of the assumption that ion and electron densities are equal in the device. The model has been used to simulate ZT-40M in both flat-top and slowly ramped current modes. Using experimentally measured density and current evolutions, the model accurately predicts observed tau/sub E/, β/sub Θ/, T/sub e/, T/sub i/, Z/sub eff/, and radiated power. The continuing goal of this work is to predict conditions in the ZT-H device, which is under construction. 28 refs., 18 figs

  15. Implications of fusion results for a reactor: a proposed next step device-JIT

    International Nuclear Information System (INIS)

    Rebut, P.H.

    1989-01-01

    Simulations with a critical-temperature model have been made of proposed future devices (NET, ITER, JIT, etc.). These show that only machines with a current capability of ∼ 30MA have a sufficient ignition domain to cope with more realistic operating conditions (i.e. taking into account sawteeth effects, impurity dilution and semi-continuous operation). The importance of dilution and Bremsstrahlung radiation are clearly demonstrated; a mean temperature > 7keV is required for ignition. This prevents higher field, lower current devices from reaching ignition. Transient operations with monster sawteeth or H-mode allow such devices (>30MA) to reach ignition at lower density without additional heating. To investigate the problems of a controlled burning plasma for days in semi-continuous operation, the plasma of the next-step tokamak should be similar in size and performance to an energy producing reactor. The scientific and technical aims of such a machine should be to study burning plasma, test wall technology, provide a test-bed for breeding blankets and most importantly to demonstrate the potential and viability of fusion as an energy source. The main design characteristics of a Thermonuclear Furnace-JIT-dedicated to these objectives are presented. Watercooled copper magnets are used to benefit from proven technology. A single-null divertor configuration ensures helium exhaust and possibly benefits from an H-mode to reach the ignition domain. The X-point position relative to the dump plates would be swept to limit wall loading

  16. Computerized cost estimation spreadsheet and cost data base for fusion devices

    International Nuclear Information System (INIS)

    Hamilton, W.R.; Rothe, K.E.

    1985-01-01

    An automated approach to performing and cataloging cost estimates has been developed at the Fusion Engineering Design Center (FEDC), wherein the cost estimate record is stored in the LOTUS 1-2-3 spreadsheet on an IBM personal computer. The cost estimation spreadsheet is based on the cost coefficient/cost algorithm approach and incorporates a detailed generic code of cost accounts for both tokamak and tandem mirror devices. Component design parameters (weight, surface area, etc.) and cost factors are input, and direct and indirect costs are calculated. The cost data base file derived from actual cost experience within the fusion community and refined to be compatible with the spreadsheet costing approach is a catalog of cost coefficients, algorithms, and component costs arranged into data modules corresponding to specific components and/or subsystems. Each data module contains engineering, equipment, and installation labor cost data for different configurations and types of the specific component or subsystem. This paper describes the assumptions, definitions, methodology, and architecture incorporated in the development of the cost estimation spreadsheet and cost data base, along with the type of input required and the output format

  17. The first operation of the superconducting optimized stellarator fusion device Wendelstein 7-X

    Energy Technology Data Exchange (ETDEWEB)

    Klinger, Thomas [Max-Planck-Institut fuer Plasmaphysik, Greifswald (Germany); Ernst-Moritz-Arndt Universitaet, Greifswald (Germany)

    2016-07-01

    The confinement of a high-temperature plasma by a suitable magnetic field is the most promising path to master nuclear fusion of Deuterium and Tritium on the scale of a reasonable power station. The two leading confinement concepts are the tokamak and the stellarator. Different from a tokamak, the stellarator does not require a strong current in the plasma but generates the magnetic field by external coils only. This has significant advantages, e.g. better stability properties and inherent steady-state capability. But stellarators need optimization, since ad hoc chosen magnetic field geometries lead to insufficient confinement properties, unfavourable plasma equilibria, and loss of fast particles. Wendelstein 7-X is a large (plasma volume 30 m{sup 3}) stellarator device with shaped superconducting coils that were determined via pure physics optimization criteria. After 19 years of construction, Wendelstein 7-X has now started operation. This talk introduces into the stellarator concept as a candidate for a future fusion power plant, summarizes the optimization principles, and presents the first experimental results with Helium and Hydrogen high temperature plasmas. An outlook on the physics program and the main goals of the project is given, too.

  18. Proceedings of the Japan-U.S. workshop P-118 on vacuum technologies for fusion devices

    International Nuclear Information System (INIS)

    Miyahara, A.

    1989-01-01

    Fusion community does not appreciate vacuum technologies to the same extent as accelerator community does. This is because, in the case of accelerators, in particular storage ring systems, the requirement of attaining ultrahigh vacuum in order to avoid collisional loss is well defined, on the other hand, it is not possible to define the requirement so precisely in the case of fusion devices. One of the reasons is that core plasma interacts with vessel wall so strongly and unpredictably that it becomes difficult to identify the role played by individual components. However, in the next step and the next generation machines like CIT, LHS, ITER, FER and NET, vacuum technologies would play more significant roles, because the CIT will introduce tritium in a vacuum vessel, and the aim of the ITER project is to demonstrate particle balance, namely, to achieve steady state operation with D-T fuel. The Japan-U.S. workshop P-118 was held at the Institute of Plasma Physics, Nagoya University, from August 1 to 5, 1988. 33 participants including 4 from the U.S. took part in the workshop. In the plenary session, 12 lectures were given, and also the topics-oriented session on pumping, gauging, remote maintenance, first wall, pump limiter, divertor and others was held. (K.I.)

  19. Developing Boundary/PMI Solutions for Next-Step Fusion Devices

    Science.gov (United States)

    Guo, H. Y.; Leonard, A. W.; Thomas, D. M.; Allen, S. L.; Hill, D. N.; Unterberg, Z.

    2014-10-01

    The path towards next-step fusion development requires increased emphasis on the boundary/plasma-material interface. The new DIII-D Boundary/Plasma-Material Interactions (PMI) Center has been established to address these critical issues on a timescale relevant to the design of FNSF, adopting the following transformational approaches: (1) Develop and test advanced divertor configurations on DIII-D compatible with core plasma high performance operational scenarios in FNSF; (2) Validate candidate reactor PFC materials at reactor-relevant temperatures in DIII-D high-performance plasmas, in collaboration with the broad material research/development community; (3) Integrate validated boundary-materials interface with high performance plasmas to provide viable boundary/PMI solutions for next-step fusion devices. This program leverages unique DIII-D capabilities, promotes synergistic programs within the broad PMI community, including linear material research facilities. It will also enable us to build a compelling bridge for the US research on long-pulse facilities. Work supported by the US DOE under DE-FC02-04ER54698 and DE-AC52-07NA27344, DE-AC05-00OR2725.

  20. Calculational models for the treatment of pulsed/intermittent activation within fusion energy devices

    International Nuclear Information System (INIS)

    Spangler, S.E.; Sisolak, J.E.; Henderson, D.L.

    1993-01-01

    Two calculationally efficient methods have been developed to compute the induced radioactivity due to pulsed/intermittent irradiation histories as encountered in both magnetic and inertial fusion energy devices. The numerical algorithms are based on the linear chain method (Bateman Equations) and employ series reduction and matrix algebra. The first method models the case in which the irradiated materials are present throughout a series of irradiation pulses. The second method treats the case where a fixed amount of radioactive and transmuted material is created during each pulse. Analytical solutions are given for each method for a three nuclide linear chain. Numerical results and comparisons are presented for a select number of linear chains. (orig.)

  1. Characterization of Size, Composition and Origins of Dust in Fusion Devices. Summary Report of the Second Research Coordination Meeting

    International Nuclear Information System (INIS)

    Braams, B.J.; Skinner, C.H.

    2010-11-01

    Eleven experts on processes of dust in fusion experiments met for the 2nd Research Coordination Meeting (RCM) of the Coordinated Research Project (CRP) on 'Characterization of size, composition and origins of dust in fusion devices' held at IAEA Headquarters 21-23 June 2010. Participants summarized their studies on dust in fusion experiments and reviewed progress made since the first RCM. Gaps in knowledge were identified and a plan of work for the remainder of the CRP was developed. Presentations, discussions and recommendations of the RCM are summarized in this report. Eleven experts on processes of dust in fusion experiments met for the 2nd Research Coordination Meeting (RCM) of the Coordinated Research Project (CRP) on 'Characterization of size, composition and origins of dust in fusion devices' held at IAEA Headquarters 21-23 June 2010. Participants summarized their studies on dust in fusion experiments and reviewed progress made since the first RCM. Gaps in knowledge were identified and a plan of work for the remainder of the CRP was developed. Presentations, discussions and recommendations of the RCM are summarized in this report. (author)

  2. Dust in fusion devices-a multi-faceted problem connecting high- and low-temperature plasma physics

    International Nuclear Information System (INIS)

    Winter, J

    2004-01-01

    Small particles with sizes between a few nanometers and a few 10 μm (dust) are formed in fusion devices by plasma-surface interaction processes. Though it is not a major problem today, dust is considered a problem that could arise in future long pulse fusion devices. This is primarily due to its radioactivity and due to its very high chemical reactivity. Dust formation is particularly pronounced when carbonaceous wall materials are used. Dust particles can be transported in the tokamak over significant distances. Radioactivity leads to electrical charging of dust and to its interaction with plasmas and electric fields. This may cause interference with the discharge but may also result in options for particle removal. This paper discusses some of the multi-faceted problems using information both from fusion research and from low-temperature dusty plasma work

  3. Suppression of hydrogenated carbon film deposition by scavenger techniques and their application to the tritium inventory control of fusion devices

    International Nuclear Information System (INIS)

    Tabares, F.L.; Tafalla, D.; Tanarro, I.; Herrero, V.J.; Islyaikin, A.; Maffiotte, C.

    2002-01-01

    The well-known radical and ion scavenger techniques of application in amorphous hydrogenated carbon film deposition studies are investigated in relation to the mechanism of tritium and deuterium co-deposition in carbon-dominated fusion devices. A particularly successful scheme results from the injection of nitrogen into methane/hydrogen plasmas for conditions close to those prevailing in the divertor region of present fusion devices. A complete suppression of the a-C : H film deposition has been achieved for N 2 /CH 4 ratios close to one in methane (5%)/hydrogen DC plasma. The implications of these findings in the tritium retention control in future fusion reactors are addressed. (author). Letter-to-the-editor

  4. Development of laser-based techniques for in situ characterization of the first wall in ITER and future fusion devices

    NARCIS (Netherlands)

    Philipps, V.; Malaquias, A.; Hakola, A.; Karhunen, J.; Maddaluno, G.; Almaviva, S.; Caneve, L.; Colao, F.; Fortuna, E.; Gasior, P.; Kubkowska, M.; Czarnecka, A.; Laan, M.; Lissovski, A.; Paris, P.; van der Meiden, H. J.; Petersson, P.; Rubel, M.; Huber, A.; Zlobinski, M.; Schweer, B.; Gierse, N.; Xiao, Q.; Sergienko, G.

    2013-01-01

    Analysis and understanding of wall erosion, material transport and fuel retention are among the most important tasks for ITER and future devices, since these questions determine largely the lifetime and availability of the fusion reactor. These data are also of extreme value to improve the

  5. High-throughput deterministic single-cell encapsulation and droplet pairing, fusion, and shrinkage in a single microfluidic device

    NARCIS (Netherlands)

    Schoeman, R.M.; Kemna, Evelien; Wolbers, F.; van den Berg, Albert

    In this article, we present a microfluidic device capable of successive high-yield single-cell encapsulation in droplets, with additional droplet pairing, fusion, and shrinkage. Deterministic single-cell encapsulation is realized using Dean-coupled inertial ordering of cells in a Yin-Yang-shaped

  6. Development of MW gyrotrons for fusion devices by University of Tsukuba

    International Nuclear Information System (INIS)

    Minami, R.; Kariya, T.; Imai, T.; Numakura, T.; Endo, Y.; Nakabayashi, H.; Eguchi, T.; Shimozuma, T.; Kubo, S.; Yoshimura, Y.; Igami, H.; Takahashi, H.; Mutoh, T.; Ito, S.; Idei, H.; Zushi, H.; Yamaguchi, Y.; Sakamoto, Keishi; Mitsunaka, Y.

    2012-11-01

    Over-1 MW power gyrotrons for electron cyclotron heating (ECH) have been developed in the joint program of NIFS and University of Tsukuba. The obtained maximum outputs are 1.9 MW for 0.1 s on the 77 GHz Large Helical Device (LHD) tube and 1.0 MW for 1 ms on the 28 GHz GAMMA 10 one, which are new records in these frequency ranges. In long pulse operation, 300 kW for 40 min at 77 GHz and 540 kW for 2 s at 28 GHz were achieved. A new program of 154 GHz 1 MW development has started for high density plasma heating in LHD and the first tube has been fabricated. These lower frequency tubes like 77 GHz or 28 GHz one are also important for advanced magnetic fusion devices, which use Electron Bernstein Wave (EBW) heating / current drive. As a next activity of 28 GHz gyrotron, we have already started the development of over-1.5 MW gyrotron and a new design study of 28 GHz / 35 GHz dual frequency gyrotron, which indicates the practicability of the multi-purpose gyrotron. (author)

  7. A comparison of hydrogen vs. helium glow discharge effects on fusion device first-wall conditioning

    International Nuclear Information System (INIS)

    Dylla, H.F.

    1989-09-01

    Hydrogen- and deuterium-fueled glow discharges are used for the initial conditioning of magnetic fusion device vacuum vessels following evacuation from atmospheric pressure. Hydrogenic glow discharge conditioning (GDC) significantly reduces the near-surface concentration of simple adsorbates, such as H 2 O, CO, and CH 4 , and lowers ion-induced desorption coefficients by typically three orders of magnitude. The time evolution of the residual gas production observed during hydrogen-glow discharge conditioning of the carbon first-wall structure of the TFTR device is similar to the time evolution observed during hydrogen GDC of the initial first-wall configuration in TFTR, which was primarily stainless steel. Recently, helium GDC has been investigated for several wall-conditioning tasks on a number of tokamaks including TFTR. Helium GDC shows negligible impurity removal with stainless steel walls. For impurity conditioning with carbon walls, helium GDC shows significant desorption of H 2 O, CO, and CO 2 ; however, the total desorption yield is limited to the monolayer range. In addition, helium GDC can be used to displace hydrogen isotopes from the near-surface region of carbon first-walls in order to lower hydrogenic retention and recycling. 38 refs., 6 figs

  8. Design study of an indirect cooling superconducting magnet for a fusion device

    International Nuclear Information System (INIS)

    Mito, Toshiyuki; Hemmi, Tsutomu

    2009-01-01

    The design study of superconducting magnets adapting a new coil winding scheme of an indirect cooling method is reported. The superconducting magnet system for the spherical tokamak (ST), which is proposed to study the steady state plasma experiment with Q - equiv-1, requires high performances with a high current density compared to the ordinal magnet design because of its tight spatial restriction. The superconducting magnet system for the fusion device has been used in the condition of high magnetic field, high electromagnetic force, and high heat load. The pool boiling liquid helium cooling outside of the conductor or the forced flow of supercritical helium cooling inside of the conductor, such as cable-in-conduit conductors, were used so far for the cooling method of the superconducting magnet for a fusion application. The pool cooling magnet has the disadvantages of low mechanical rigidities and low withstand voltages of the coil windings. The forced flow cooling magnet with cable-in-conduit conductors has the disadvantages of the restriction of the coil design because of the path of the electric current must be the same as that of the cooling channel for refrigerant. The path of the electric current and that of the cooling channel for refrigerant can be independently designed by adopting the indirect cooling method that inserts the independent cooling panel in the coil windings and cools the conductor from the outside. Therefore the optimization of the coil windings structure can be attempted. It was shown that the superconducting magnet design of the high current density became possible by the indirect cooling method compared with those of the conventional cooling scheme. (author)

  9. Self-shielding characteristics of aqueous self-cooled blankets for next generation fusion devices

    International Nuclear Information System (INIS)

    Pelloni, S.; Cheng, E.T.; Embrechts, M.J.

    1987-11-01

    The present study examines self-shielding characteristics for two aqueous self-cooled tritium producing driver blankets for next generation fusion devices. The aqueous Self-Cooled Blanket concept (ASCB) is a very simple blanket concept that relies on just structural material and coolant. Lithium compounds are dissolved in water to provide for tritium production. An ASCB driver blanket would provide a low technology and low temperature environment for blanket test modules in a next generation fusion reactor. The primary functions of such a blanket would be shielding, energy removal and tritium production. One driver blanket considered in this study concept relates to the one proposed for the Next European Torus (NET), while the second concept is indicative for the inboard shield design for the Engineering Test Reactor proposed by the USA (TIBER II/ETR). The driver blanket for NET is based on stainless steel for the structural material and aqueous solution, while the inboard shielding blanket for TIBER II/ETR is based on a tungsten/aqueous solution combination. The purpose of this study is to investigate self-shielding and heterogeneity effects in aqueous self-cooled blankets. It is found that no significant gains in tritium breeding can be achieved in the stainless steel blanket if spatial and energy self-shielding effects are considered, and the heterogeneity effects are also insignificant. The tungsten blanket shows a 5 percent increase in tritium production in the shielding blanket when energy and spatial self-shielding effects are accounted for. However, the tungsten blanket shows a drastic increase in the tritium breeding ratio due to heterogeneity effects. (author) 17 refs., 9 figs., 9 tabs

  10. Characterization of Size, Composition and Origins of Dust in Fusion Devices. Summary Report of the Third Research Coordination Meeting

    International Nuclear Information System (INIS)

    Braams, B.J.

    2013-02-01

    Twelve experts on processes of dust in fusion experiments met at IAEA Headquarters 30 November - 02 December 2011 for the 3rd Research Coordination Meeting (RCM) of the Coordinated Research Project (CRP) on ''Characterization of size, composition and origins of dust in fusion devices.'' Participants reviewed their work done in the course of the CRP and the current state of knowledge, and they made plans for a dust database and a final CRP report. Presentations, discussions and recommendations of the RCM are summarized here. (author)

  11. A study to compare the efficacy of polyether ether ketone rod device with titanium devices in posterior spinal fusion in a canine model.

    Science.gov (United States)

    Wang, Nanxiang; Xie, Huanxin; Xi, Chunyang; Zhang, Han; Yan, Jinglong

    2017-03-09

    The benefits of posterior lumbar fusion surgery with orthotopic paraspinal muscle-pediculated bone flaps are well established. However, the problem of non-union due to mechanical support is not completely resolved. The aim of the study was to compare the efficacy of polyether ether ketone (PEEK) rod device with conventional titanium devices in the posterior lumbar fusion surgery with orthotopic paraspinal muscle-pediculated bone flaps. This was a randomized controlled study with an experimental animal model. Thirty-two mongrel dogs were randomly divided into two groups-control group (n = 16), which received the titanium device and the treatment group (n = 16), which received PEEK rods. The animals were sacrificed 8 or 16 weeks after surgery. Lumbar spines of dogs in both groups were removed, harvested, and assessed for radiographic, biomechanical, and histological changes. Results in the current study indicated that there was no significant difference in the lumbar spine of the control and treatment groups in terms of radiographic, manual palpation, and gross examination. However, certain parameters of biomechanical testing showed significant differences (p < 0.05) in stiffness and displacement, revealing a better fusion (treatment group showed decreased stiffness with decreased displacement) of the bone graft. Similarly, the histological analysis also revealed a significant fusion mass in both treatment and control groups (p < 0.05). These findings revealed that fixation using PEEK connecting rod could improve the union of the bone graft in the posterior lumbar spine fusion surgery compared with that of the titanium rod fixation.

  12. High power density carbonate fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Yuh, C.; Johnsen, R.; Doyon, J.; Allen, J. [Energy Research Corp., Danbury, CT (United States)

    1996-12-31

    Carbonate fuel cell is a highly efficient and environmentally clean source of power generation. Many organizations worldwide are actively pursuing the development of the technology. Field demonstration of multi-MW size power plant has been initiated in 1996, a step toward commercialization before the turn of the century, Energy Research Corporation (ERC) is planning to introduce a 2.85MW commercial fuel cell power plant with an efficiency of 58%, which is quite attractive for distributed power generation. However, to further expand competitive edge over alternative systems and to achieve wider market penetration, ERC is exploring advanced carbonate fuel cells having significantly higher power densities. A more compact power plant would also stimulate interest in new markets such as ships and submarines where space limitations exist. The activities focused on reducing cell polarization and internal resistance as well as on advanced thin cell components.

  13. A carbon-metal brazing for divertor plates in fusion devices

    International Nuclear Information System (INIS)

    Matsuda, T.; Matsumoto, T.; Miki, S.; Sogabe, T.; Okada, M.; Kubota, Y.; Sagara, A.; Noda, N.; Motojima, O.; Hino, T.; Yamashina, T.

    1993-01-01

    A divertor unit, which consists of carbon armors brazed to a copper cooling channel, is under development for fusion devices. Isotropic graphite (IG-430U) and CFC (CX-2002U) are used for the armor, and a copper for the cooling tube. A technique named as dissolution and deposit of base metal was employed for brazing. The reliability of the brazed components was evaluated both by 4-point bending test and thermal shock test. According to the results of a 4-point bending test under the temperature ranged from RT to 800 C in a vacuum, it was found that the strength of the brazed surface at RT was maintained up to the higher temperature, 600 C. High heat load test has been also performed on the brazed sample in order to find whether the samples meet the requirement of the divertor plates of LHD (Large Helical Device). Active Cooling Teststand (ACT:NIFS) with electron beam power of 100kW was used. In LHD, it is presumed that the maximum heat flux is 10MW/m 2 . In addition, the surface temperature of divertor has to be kept below 1,200 C to avoid RES, by active cooling. The heat load test showed that the brazing components of CX-2002U (flat plate type CFC-Cu brazed) was stable at 1,300 C under a heat flux of 10MW/m 2 , when the flow velocity of cooling water was 6m/s. No damage nor deterioration was found at the brazed zone after the heat load test

  14. Radio-frequency-assisted current startup in the Fusion Engineering Device

    International Nuclear Information System (INIS)

    Borowski, S.K.; Kammash, T.; Martin Peng, Y.K.

    1984-01-01

    Auxiliary radio-frequency (RF) heating of electrons before and during the current rise phase of a large tokamak, such as the Fusion Engineering Device (FED) (R 0 = 4.8 m, a = 1.3 m, sigma = 1.6, B(R 0 ) = 3.62 T), is examined as a means of reducing both the initiation loop voltage and resistive flux expenditure during startup. Prior to current initiation, 1 to 2 MW of electron cyclotron resonance heating power at about90 GHz is used to create a small volume of high conductivity plasma (T /sub e/ approx. = 100 eV, n /sub e/ approx. = 10 19 m -3 ) near the upper hybrid resonance (UHR) region. This plasma conditioning, referred to as preheating, permits a small radius (a 0 approx. = 0.2 to 0.4 m) current channel to be established with a relatively low initial loop voltage (less than or equal to 25 V as opposed to about 100 V without rf assist). During the subsequent plasma expansion and current rise phase, a combination of rf heating (up to 5 MW) and linear current ramping leads to a substantial savings in voltseconds by (a) minimizing the resistive flux consumption and (b) producing broad current density profiles. (With such broad profiles, the internal flux requirements are maintained at or near the flat profile limit.)

  15. Experimental evaluation of torsional fatigue strength of welded bellows and application to design of fusion device

    International Nuclear Information System (INIS)

    Takatsu, Hideyuki; Yamamoto, Masahiro; Shimizu, Masatsugu; Suzuki, Kazuo; Sonobe, Tadashi; Hayashi, Yuzo; Mizuno, Gen-ichiro.

    1984-01-01

    Torsional fatigue strength of the welded bellows was evaluated experimentally, aiming the application to a port of a fusion device. The welded bellows revealed elastic torsional buckling and spiral distorsion even under a small angle of torsion. Twisting load never leads the welded bellows to fracture easily so far as the angle of torsion is not excessively large, and the welded bellows has the torsional fatigue strength much larger than that expected so far. Two formulae were proposed to evaluate the stress of the welded bellows under the forced angle of torsion; shearing stress evaluation formula in the case that torsional buckling does not occur and the axial bending stress evaluation formula in the case that torsional buckling occurs. And the results of the torsional fatigue experiments showed that the former is reasonably conservative and simulates the actual behavior of the welded bellows better than the latter in the high cycle fatigue region and vice versa in the low cycle fatigue region from the viewpoint of the mechanical design. The present evaluation method of the torsional fatigue strength was applied to the welded bellows for the port of the JT-60 vacuum vessel and its structural integrity was confirmed under the design load condition. (author)

  16. Encoding technique for high data compaction in data bases of fusion devices

    International Nuclear Information System (INIS)

    Vega, J.; Cremy, C.; Sanchez, E.; Portas, A.; Dormido, S.

    1996-01-01

    At present, data requirements of hundreds of Mbytes/discharge are typical in devices such as JET, TFTR, DIII-D, etc., and these requirements continue to increase. With these rates, the amount of storage required to maintain discharge information is enormous. Compaction techniques are now essential to reduce storage. However, general compression techniques may distort signals, but this is undesirable for fusion diagnostics. We have developed a general technique for data compression which is described here. The technique, which is based on delta compression, does not require an examination of the data as in delayed methods. Delta values are compacted according to general encoding forms which satisfy a prefix code property and which are defined prior to data capture. Several prefix codes, which are bit oriented and which have variable code lengths, have been developed. These encoding methods are independent of the signal analog characteristics and enable one to store undistorted signals. The technique has been applied to databases of the TJ-I tokamak and the TJ-IU torsatron. Compaction rates of over 80% with negligible computational effort were achieved. Computer programs were written in ANSI C, thus ensuring portability and easy maintenance. We also present an interpretation, based on information theory, of the high compression rates achieved without signal distortion. copyright 1996 American Institute of Physics

  17. Benefits and drawbacks of low magnetic shears on the confinement in magnetic fusion toroidal devices

    Science.gov (United States)

    Firpo, Marie-Christine; Constantinescu, Dana

    2012-10-01

    The issue of confinement in magnetic fusion devices is addressed within a purely magnetic approach. As it is well known, the magnetic field being divergence-free, the equations of its field lines can be cast in Hamiltonian form. Using then some Hamiltonian models for the magnetic field lines, the dual impact of low magnetic shear is demonstrated. Away from resonances, it induces a drastic enhancement of magnetic confinement that favors robust internal transport barriers (ITBs) and turbulence reduction. However, when low-shear occurs for values of the winding of the magnetic field lines close to low-order rationals, the amplitude thresholds of the resonant modes that break internal transport barriers by allowing a radial stochastic transport of the magnetic field lines may be much lower than the ones obtained for strong shear profiles. The approach can be applied to assess the robustness versus magnetic perturbations of general almost-integrable magnetic steady states, including non-axisymmetric ones such as the important single helicity steady states. This analysis puts a constraint on the tolerable mode amplitudes compatible with ITBs and may be proposed as a possible explanation of diverse experimental and numerical signatures of their collapses.

  18. Ohmic heating coil power supply using thyristor circuit breaker in a thermonuclear fusion device

    International Nuclear Information System (INIS)

    Tani, Keiji; Shimada, Ryuichi; Tamura, Sanae; Yabuno, Kohei; Koseki, Shoichiro.

    1982-01-01

    In a large scale Tokamak thermonuclear fusion device such as the critical plasma testing facility (JT60) presently under construction, mechanical breakers such as vacuum and air breakers are mostly used for interrupting DC heavy current which is supplied to the ohmic heating coils of inductive energy accumulation method. The practical use of the DC breakers employing thyristors has just been started because the history of thyristor development is short and thristors are still expensive, in spite of the advantages. In this paper, the circuit is investigated in which the excellent high speed controllability of thyristors is fully utilized, while the economy is taken into accout, and the experiment carried out with a unit model is described. It was found that a thyristor switch, which was constructed by connecting the high speed thyristors of peak off-state voltage rating 2,000 V and mean current rating 500 A in direct parallel, was able to interrupt 12.7 kA current in the power supply circuit of ohmic heating coils developed this time. In addition, the switch configuration was able to be greatly simplified. When the multistage raising of plasma current is required, the raise can be performed with a single thyristor breaker because it can make high speed control. Therefore, the capacity of the breaker can be doubly and drastically reduced. Also, if current unbalance might occur between thyristor switch units, it gives no problem since the time of reverse voltage after current interruption dispersed smaller as current increased. (Wakatsuki, Y.)

  19. Surface temperature measurements by means of pulsed photothermal effects in fusion devices

    International Nuclear Information System (INIS)

    Loarer, Th.; Brygo, F.; Gauthier, E.; Grisolia, C.; Le Guern, F.; Moreau, F.; Murari, A.; Roche, H.; Semerok, A.

    2007-01-01

    In fusion devices, the surface temperature of plasma facing components is measured using infrared cameras. This method requires a knowledge of the emissivity of the material, the reflected and parasitic fluxes (Bremsstrahlung). For carbon, the emissivity is known and constant over the detection wavelength (∼3-5 μm). For beryllium and tungsten, the reflected flux could contribute significantly to the collected flux. The pulsed photothermal method described in this paper allows temperature measurements independently of both reflected and parasitic fluxes. A local increase of the surface temperature (ΔT ∼ 10-15 K) introduced by a laser pulse (few ns) results in an additional component of the photon flux collected by the detector. Few μs after the pulse, a filtering of the signal allows to extract a temporal flux proportional only to the variation of the emitted flux, the emissivity and ΔT. The ratio of simultaneous measurements at two wavelengths leads to the elimination of ΔT and emissivity. The range of application increases for measurements at short wavelengths (1-1.7 μm) with no limitation due to the Bremsstrahlung emission

  20. Engineering challenges encountered in the design of the ELMO BUMPY TORUS proof-of-principle fusion device

    International Nuclear Information System (INIS)

    Dillow, C.F.; Imster, H.F.

    1982-01-01

    This paper first provides a summary of the history and current status of the Elmo Bumpy Torus (EBT) fusion concept. A brief description of the EBT-P is then provided in which the many unique features of this fusion device are highlighted. This description will provide the technical background for the following discussions of some of the more challenging mechanical engineering problems encountered to date in the evolution of the EBT-P design. The problems discussed are: optimization of the device primary structure design, optimization of the superconducting magnet x-ray shield design, design of the liquid helium supply and distribution system, and selection of high vacuum seals and pumps and their protection from the high power microwave environment. The common challenge in each of these design issues was to assure adequate performance at minimum cost

  1. Nuclear Data Libraries for Advanced Systems - Fusion Devices (FENDL 3.0). Summary report of the Third Research Coordination Meeting

    International Nuclear Information System (INIS)

    Sawan, Mohamed E.

    2012-03-01

    The third Research Co-ordination Meeting of the Nuclear Data Libraries for Advanced Systems - Fusion Devices (FENDL-3) was held at IAEA Headquarters in Vienna from 6 to 9 December 2011. A summary of the presentations given during meeting is given in this report along with the discussions that took place. A list of actions necessary to complete the library production, processing and testing are given. Details of the documents arising from the CRP were agreed. (author)

  2. Interacting with mobile devices by fusion eye and hand gestures recognition systems based on decision tree approach

    Science.gov (United States)

    Elleuch, Hanene; Wali, Ali; Samet, Anis; Alimi, Adel M.

    2017-03-01

    Two systems of eyes and hand gestures recognition are used to control mobile devices. Based on a real-time video streaming captured from the device's camera, the first system recognizes the motion of user's eyes and the second one detects the static hand gestures. To avoid any confusion between natural and intentional movements we developed a system to fuse the decision coming from eyes and hands gesture recognition systems. The phase of fusion was based on decision tree approach. We conducted a study on 5 volunteers and the results that our system is robust and competitive.

  3. Properties of an interspinous fixation device (ISD) in lumbar fusion constructs: a biomechanical study.

    Science.gov (United States)

    Techy, Fernando; Mageswaran, Prasath; Colbrunn, Robb W; Bonner, Tara F; McLain, Robert F

    2013-05-01

    Segmental fixation improves fusion rates and promotes patient mobility by controlling instability after lumbar surgery. Efforts to obtain stability using less invasive techniques have lead to the advent of new implants and constructs. A new interspinous fixation device (ISD) has been introduced as a minimally invasive method of stabilizing two adjacent interspinous processes by augmenting an interbody cage in transforaminal interbody fusion. The ISD is intended to replace the standard pedicle screw instrumentation used for posterior fixation. The purpose of this study is to compare the rigidity of these implant systems when supplementing an interbody cage as used in transforaminal lumbar interbody fusion. An in vitro human cadaveric biomechanical study. Seven human cadaver spines (T12 to the sacrum) were mounted in a custom-designed testing apparatus, for biomechanical testing using a multiaxial robotic system. A comparison of segmental stiffness was carried out among five conditions: intact spine control; interbody spacer (IBS), alone; interbody cage with ISD; IBS, ISD, and unilateral pedicle screws (unilat); and IBS, with bilateral pedicle screws (bilat). An industrial robot (KUKA, GmbH, Augsburg, Germany) applied a pure moment (±5 Nm) in flexion-extension (FE), lateral bending (LB), and axial rotation (AR) through an anchor to the T12 vertebral body. The relative vertebral motion was captured using an optoelectronic camera system (Optotrak; Northern Digital, Inc., Waterloo, Ontario, Canada). The load sensor and the camera were synchronized. Maximum rotation was measured at each level and compared with the intact control. Implant constructs were compared with the control and with each other. A statistical analysis was performed using analysis of variance. A comparison between the intact spine and the IBS group showed no significant difference in the range of motion (ROM) in FE, LB, or AR for the operated level, L3-L4. After implantation of the ISD to augment

  4. Study of lower hybrid current drive system in tokamak fusion devices

    International Nuclear Information System (INIS)

    Maebara, Sunao

    2001-01-01

    This report describes R and D of a high-power klystron, RF vacuum window, low-outgassing antenna and a front module for a plasma-facing antenna aiming the 5 GHz Lower Hybrid Current Drive (LHCD) system for the next Tokamak Fusion Device. 5 GHz klystron with a low-perveances of 0.7 μP is designed for a high-power and a high-efficiency, the output-power of 715 kW and the efficiency of 63%, which are beyond the conventional design scaling of 450 kW-45%, are performed using the prototype klystron which operates at the pulse duration of 15 μsec. A new pillbox window, which has an oversized length in both the axial and the radial direction, are designed to reduce the RF power density and the electric field strength at the ceramics. It is evaluated that the power capability by cooling edge of ceramics is 1 MW with continuous-wave operation. The antenna module using Dispersion Strengthened Copper which combines high mechanical property up to 500degC with high thermal conductivity, are developed for a low-outgassing antenna in a steady state operation. It is found that the outgassing rate is in the lower range of 4x10 -6 Pam 3 /sm 2 at the module temperature of 300degC, which requires no active vacuum pumping of the LHCD antenna. A front module using Carbon Fiber Composite (CFC) are fabricated and tested for a plasma facing antenna which has a high heat-resistive. Stationary operation of the CFC module with water cooling is performed at the RF power of 46 MWm -2 (about 2 times higher than the design value) during 1000 sec, it is found that the outgassing rate is less than 10 -5 Pam 3 /sm 2 which is low enough for an antenna material. (author)

  5. On the role of turbulence on momentum redistribution in fusion devices

    International Nuclear Information System (INIS)

    Hidalgo, C.

    2005-01-01

    flows. These findings provide the first experimental evidence of the important role of parallel turbulence forces on edge momentum dynamic in fusion devices. (author)

  6. From Data Acquisition to Data Fusion: A Comprehensive Review and a Roadmap for the Identification of Activities of Daily Living Using Mobile Devices

    Science.gov (United States)

    Pires, Ivan Miguel; Garcia, Nuno M.; Pombo, Nuno; Flórez-Revuelta, Francisco

    2016-01-01

    This paper focuses on the research on the state of the art for sensor fusion techniques, applied to the sensors embedded in mobile devices, as a means to help identify the mobile device user’s daily activities. Sensor data fusion techniques are used to consolidate the data collected from several sensors, increasing the reliability of the algorithms for the identification of the different activities. However, mobile devices have several constraints, e.g., low memory, low battery life and low processing power, and some data fusion techniques are not suited to this scenario. The main purpose of this paper is to present an overview of the state of the art to identify examples of sensor data fusion techniques that can be applied to the sensors available in mobile devices aiming to identify activities of daily living (ADLs). PMID:26848664

  7. From Data Acquisition to Data Fusion: A Comprehensive Review and a Roadmap for the Identification of Activities of Daily Living Using Mobile Devices

    Directory of Open Access Journals (Sweden)

    Ivan Miguel Pires

    2016-02-01

    Full Text Available This paper focuses on the research on the state of the art for sensor fusion techniques, applied to the sensors embedded in mobile devices, as a means to help identify the mobile device user’s daily activities. Sensor data fusion techniques are used to consolidate the data collected from several sensors, increasing the reliability of the algorithms for the identification of the different activities. However, mobile devices have several constraints, e.g., low memory, low battery life and low processing power, and some data fusion techniques are not suited to this scenario. The main purpose of this paper is to present an overview of the state of the art to identify examples of sensor data fusion techniques that can be applied to the sensors available in mobile devices aiming to identify activities of daily living (ADLs.

  8. Review of compact, alternate concepts for magnetic confinement fusion

    International Nuclear Information System (INIS)

    Nickerson, S.B.; Shmayda, W.T.; Dinner, P.J.; Gierszewski, P.

    1984-06-01

    This report documents a study of compact alternate magnetic confinement fusion experiments and conceptual reactor designs. The purpose of this study is to identify those devices with a potential to burn tritium in the near future. The bulk of the report is made up of a review of the following compact alternates: compact toroids, high power density tokamaks, linear magnetic systems, compact mirrors, reversed field pinches and some miscellaneous concepts. Bumpy toruses and stellarators were initially reviewed but were not pursued since no compact variations were found. Several of the concepts show promise of either burning tritium or evolving into tritium burning devices by the early 1990's: RIGGATRON, Ignitor, OHTE, Frascati Tokamak upgrade, several driven (low or negative net power) mirror experiments and several Reversed Field Pinch experiments that may begin operation around 1990. Of the above only the Frascati Tokamak Upgrade has had funds allocated. Also identified in this report are groups who may have tritium burning experiments in the mid to late 1990's. There is a discussion of the differences between the reviewed devices and the mainline tokamak experiments. This discussion forms the basis of recommendations for R and D aimed at the compact alternates and the applicability of the present CFFTP program to the needs of the compact alternates. These recommendations will be presented in a subsequent report

  9. Critical plasma-wall interaction issues for plasma-facing materials and components in near-term fusion devices

    International Nuclear Information System (INIS)

    Federici, G.; Coad, J.P.; Haasz, A.A.; Janeschitz, G.; Noda, N.; Philipps, V.; Roth, J.; Skinner, C.H.; Tivey, R.; Wu, C.H.

    2000-01-01

    The increase in pulse duration and cumulative run-time, together with the increase of the plasma energy content, will represent the largest changes in operation conditions in future fusion devices such as the International Thermonuclear Experimental Reactor (ITER) compared to today's experimental facilities. These will give rise to important plasma-physics effects and plasma-material interactions (PMIs) which are only partially observed and accessible in present-day experiments and will open new design, operation and safety issues. For the first time in fusion research, erosion and its consequences over many pulses (e.g., co-deposition and dust) may determine the operational schedule of a fusion device. This paper identifies the most critical issues arising from PMIs which represent key elements in the selection of materials, the design, and the optimisation of plasma-facing components (PFCs) for the first-wall and divertor. Significant advances in the knowledge base have been made recently, as part of the R and D supporting the engineering design activities (EDA) of ITER, and some of the most relevant data are reviewed here together with areas where further R and D work is urgently needed

  10. Exploitation of a Breakthrough in Magnetic Confinement Fusion to Improve Transuranic Incineration

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, Erich [Nuclear and Radiation Engineering Program, The University of Texas at Austin, Austin, TX 78712 (United States); Kotschenreuther, Mike; Mahajan, Swadesh; Valanju, Prashant [Institute for Fusion Studies, The University of Texas at Austin, Austin, TX 78712 (United States)

    2009-06-15

    A fusion-assisted transmutation system for the destruction of transuranic nuclear waste is developed by combining a subcritical fusion-fission hybrid assembly uniquely equipped to burn the worst thermal non-fissile transuranic isotopes with a new fuel cycle that uses cheaper light water reactors for most of the transmutation. The centerpiece of this fuel cycle, the high power density compact fusion neutron source (CFNS, 100 MW, outer radius <3 m), is made possible by a new divertor with a heat-handling capacity five times that of the standard alternative. The number of hybrids needed to destroy a given amount of waste is about an order of magnitude below the corresponding number of critical fast spectrum reactors (FR) as the latter cannot fully exploit the new fuel cycle. Also, the time needed for 99% transuranic waste destruction reduces from centuries (with FR) to decades. The generic Hybrid, combining neutron-rich fusion with energy-rich fission, was first conceptualized several decades ago. However, it is only now that accumulated advances in fusion science and technology allow designing a neutron source like CFNS that is simultaneously compact and high power density, offering a neutron source an order of magnitude stronger than that obtained from accelerator driven systems. The former is essential for efficient coupling to the fission blanket, and the latter is key to efficient neutron production necessary to yield high neutron fluxes needed for effective transmutation. The recent invention of the SuperX-Divertor (SXD)1, a new magnetic configuration that allows the system to safely exhaust large heat and particle fluxes peculiar to CFNS-like devices, is a crucial addition to the underlying knowledge base. The subcritical FFTS acquires a definite advantage over the critical FR approach because of its ability to support an innovative fuel cycle that makes the cheaper LWR do the bulk (75%) of the transuranic transmutation via deep burn in an inert matrix fuel

  11. Magnetic Fusion Energy Plasma Interactive and High Heat Flux Components: Volume 5, Technical assessment of critical issues in the steady state operation of fusion confinement devices

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-01

    Critical issues for the steady state operation of plasma confinement devices exist in both the physics and technology fields of fusion research. Due to the wide range and number of these issues, this technical assessment has focused on the crucial issues associated with the plasma physics and the plasma interactive components. The document provides information on the problem areas that affect the design and operation of a steady state ETR or ITER type confinement device. It discusses both tokamaks and alternative concepts, and provides a survey of existing and planned confinement machines and laboratory facilities that can address the identified issues. A universal definition of steady state operation is difficult to obtain. From a physics point of view, steady state is generally achieved when the time derivatives approach zero and the operation time greatly exceeds the characteristic time constants of the device. Steady state operation for materials depends on whether thermal stress, creep, fatigue, radiation damage, or power removal are being discussed. For erosion issues, the fluence and availability of the machine for continuous operation are important, assuming that transient events such as disruptions do not limit the component lifetimes. The panel suggests, in general terms, that steady state requires plasma operation from 100 to 1000 seconds and an availability of more than a few percent, which is similar to the expectations for an ETR type device. The assessment of critical issues for steady state operation is divided into four sections: physics issues; technology issues; issues in alternative concepts; and devices and laboratory facilities that can address these problems.

  12. Magnetic Fusion Energy Plasma Interactive and High Heat Flux Components: Volume 5, Technical assessment of critical issues in the steady state operation of fusion confinement devices

    International Nuclear Information System (INIS)

    1988-01-01

    Critical issues for the steady state operation of plasma confinement devices exist in both the physics and technology fields of fusion research. Due to the wide range and number of these issues, this technical assessment has focused on the crucial issues associated with the plasma physics and the plasma interactive components. The document provides information on the problem areas that affect the design and operation of a steady state ETR or ITER type confinement device. It discusses both tokamaks and alternative concepts, and provides a survey of existing and planned confinement machines and laboratory facilities that can address the identified issues. A universal definition of steady state operation is difficult to obtain. From a physics point of view, steady state is generally achieved when the time derivatives approach zero and the operation time greatly exceeds the characteristic time constants of the device. Steady state operation for materials depends on whether thermal stress, creep, fatigue, radiation damage, or power removal are being discussed. For erosion issues, the fluence and availability of the machine for continuous operation are important, assuming that transient events such as disruptions do not limit the component lifetimes. The panel suggests, in general terms, that steady state requires plasma operation from 100 to 1000 seconds and an availability of more than a few percent, which is similar to the expectations for an ETR type device. The assessment of critical issues for steady state operation is divided into four sections: physics issues; technology issues; issues in alternative concepts; and devices and laboratory facilities that can address these problems

  13. High-throughput deterministic single-cell encapsulation and droplet pairing, fusion, and shrinkage in a single microfluidic device.

    Science.gov (United States)

    Schoeman, Rogier M; Kemna, Evelien W M; Wolbers, Floor; van den Berg, Albert

    2014-02-01

    In this article, we present a microfluidic device capable of successive high-yield single-cell encapsulation in droplets, with additional droplet pairing, fusion, and shrinkage. Deterministic single-cell encapsulation is realized using Dean-coupled inertial ordering of cells in a Yin-Yang-shaped curved microchannel using a double T-junction, with a frequency over 2000 Hz, followed by controlled droplet pairing with a 100% success rate. Subsequently, droplet fusion is realized using electrical actuation resulting in electro-coalescence of two droplets, each containing a single HL60 cell, with 95% efficiency. Finally, volume reduction of the fused droplet up to 75% is achieved by a triple pitchfork structure. This droplet volume reduction is necessary to obtain close cell-cell membrane contact necessary for final cell electrofusion, leading to hybridoma formation, which is the ultimate aim of this research. © 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  14. A problem to be solved for tungsten diagnostics through EUV spectroscopy in fusion devices

    International Nuclear Information System (INIS)

    Morita, S.; Murakami, I.; Sakaue, H.A.; Dong, C.F.; Goto, M.; Kato, D.; Oishi, T.; Huang, X.L.; Wang, E.H.

    2013-01-01

    Tungsten spectra have been observed from Large Helical Device (LHD) in extreme ultraviolet (EUV) wavelength ranges of 10-650Å. When the electron temperature is less than 2keV, the EUV spectra from plasma core are dominated by unresolved transition array (UTA) composing of a lot of spectral lines, e.g., 6g-4f, 5g-4f, 5f-4d and 5p-4d transitions for W"+"2"4"-"+"3"3 in 15-35Å. In order to understand the UTA spectrum, the EUV spectra measured from LHD plasmas are compared to those measured from Compact electron Beam Ion Trap (CoBIT), in which the electron beam is operated with monoenergetic energy of E_e ≤ 2keV. The tungsten spectra from LHD are well analyzed based on the knowledge from CoBIT tungsten spectra. The collisional-radiative (C-R) model has been developed to explain the UTA spectra from LHD in details. Radial profiles of EUV spectra from highly ionized tungsten ions have been measured and analyzed by impurity transport simulation code with ADPAK atomic database to examine the ionization balance determined by ionization and recombination rate coefficients. If the electron temperature is higher than 2keV, Zn-like WXLV (W"4"4"+) and Cu-like WXLVI (W"4"5"+) spectra can be observed in LHD. Such ions of W"4"4"+ and W"4"5"+ can exhibit much simpler atomic configuration compared to other ionization stages of tungsten. Quantitative analysis of the tungsten density is attempted for the first time on the radial profile of Zn-like WXLV (W"4"4"+) 4p-4s transition measured at 60.9Å, based on the emission rate coefficient calculated with HULLAC code. As a result, a total tungsten ion density of 3.5x10"1"0 cm"-"3 at the plasma center of LHD is reasonably obtained. Finally, the present problem for tungsten diagnostics in fusion plasmas is summarized. (author)

  15. Study of lower hybrid current drive system in tokamak fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Maebara, Sunao [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-01-01

    This report describes R and D of a high-power klystron, RF vacuum window, low-outgassing antenna and a front module for a plasma-facing antenna aiming the 5 GHz Lower Hybrid Current Drive (LHCD) system for the next Tokamak Fusion Device. 5 GHz klystron with a low-perveances of 0.7 {mu}P is designed for a high-power and a high-efficiency, the output-power of 715 kW and the efficiency of 63%, which are beyond the conventional design scaling of 450 kW-45%, are performed using the prototype klystron which operates at the pulse duration of 15 {mu}sec. A new pillbox window, which has an oversized length in both the axial and the radial direction, are designed to reduce the RF power density and the electric field strength at the ceramics. It is evaluated that the power capability by cooling edge of ceramics is 1 MW with continuous-wave operation. The antenna module using Dispersion Strengthened Copper which combines high mechanical property up to 500degC with high thermal conductivity, are developed for a low-outgassing antenna in a steady state operation. It is found that the outgassing rate is in the lower range of 4x10{sup -6} Pam{sup 3}/sm{sup 2} at the module temperature of 300degC, which requires no active vacuum pumping of the LHCD antenna. A front module using Carbon Fiber Composite (CFC) are fabricated and tested for a plasma facing antenna which has a high heat-resistive. Stationary operation of the CFC module with water cooling is performed at the RF power of 46 MWm{sup -2} (about 2 times higher than the design value) during 1000 sec, it is found that the outgassing rate is less than 10{sup -5} Pam{sup 3}/sm{sup 2} which is low enough for an antenna material. (author)

  16. Summary report from 1. research coordination meeting on nuclear data libraries for advance systems - fusion devices (FENDL - 3)

    International Nuclear Information System (INIS)

    Trkov, A.; Forrest, R.; Mengoni, A.

    2009-03-01

    The first Research Co-ordination Meeting of the Nuclear Data Libraries for Advance Systems - Fusion Devices (FENDL - 3) was held at the IAEA Headquarters in Vienna from 2 to 5 December 2008. A summary of the meeting is given in this report along with discussions which took place. An important outcome of the meeting was the agreement to create a new FENDL-3.0 Starter Library. Finally, a list of task assignments was prepared together with the plan for future CRP activities. (author)

  17. Nuclear Data Libraries for Advanced Systems - Fusion Devices (FENDL-3). Summary report from the Second Research Coordination Meeting

    International Nuclear Information System (INIS)

    Sawan, Mohamed E.

    2010-06-01

    The second Research Co-ordination Meeting of the Nuclear Data Libraries for Advanced Systems - Fusion Devices (FENDL - 3) was held at the IAEA Headquarters in Vienna from 23 to 26 March 2010. A summary of the meeting is given in this report along with the discussions which took place. An important outcome of the meeting was the decision to provide ENDF data libraries (FENDL-3/T) by April 2011. Finally, a list of task assignments was prepared together with the plan for future CRP activities. (author)

  18. Automated pose estimation of objects using multiple ID devices for handling and maintenance task in nuclear fusion reactor

    International Nuclear Information System (INIS)

    Umetani, Tomohiro; Morioka, Jun-ichi; Tamura, Yuichi; Inoue, Kenji; Arai, Tatsuo; Mae, Yasusi

    2011-01-01

    This paper describes a method for the automated estimation of three-dimensional pose (position and orientation) of objects by autonomous robots, using multiple identification (ID) devices. Our goal is to estimate the object pose for assembly or maintenance tasks in a real nuclear fusion reactor system, with autonomous robots cooperating in a virtual assembly system. The method estimates the three-dimensional pose for autonomous robots. This paper discusses a method of motion generation for ID acquisition using the sensory data acquired by the measurement system attached to the robots and from the environment. Experimental results show the feasibility of the proposed method. (author)

  19. Proceedings of US/Japan workshop, Q219 on high heat flux components and plasma surface interactions for next fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Ulrickson, M.A.; Stevens, P.L.; Hino, T.; Hirohata, Y. [eds.

    1996-12-01

    This report contains the viewgraphs from the proceedings of US/Japan Workshop on High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices. Some of the general topics covered by this report are: PFC/PSI in tokamak and helical devices; development of high heat flux components; PSIS and plasma facing materials;tritium; and material damage.

  20. Proceedings of US/Japan workshop, Q219 on high heat flux components and plasma surface interactions for next fusion devices

    International Nuclear Information System (INIS)

    Ulrickson, M.A.; Stevens, P.L.; Hino, T.; Hirohata, Y.

    1996-12-01

    This report contains the viewgraphs from the proceedings of US/Japan Workshop on High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices. Some of the general topics covered by this report are: PFC/PSI in tokamak and helical devices; development of high heat flux components; PSIS and plasma facing materials;tritium; and material damage

  1. High-speed repetitive pellet injector prototype for magnetic confinement fusion devices

    International Nuclear Information System (INIS)

    Frattolillo, A.; Gasparotto, M.; Migliori, S.; Angelone, G.; Baldarelli, M.; Scaramuzzi, F.; Ronci, G.; Reggiori, A.; Riva, G.; Carlevaro, R.; Daminelli, G.B.

    1992-01-01

    The design of a test facility aimed at demonstrating the feasibility of high-speed repetitive acceleration of solid D 2 pellets for fusion applications, developed in a collaboration between Oak Ridge National Laboratory and ENEA Frascati, is presented. The results of tests performed at the CNPM/CNR on the piston wear in a repetitively operating two-stage gun are also reported

  2. Performance of Hall sensor-based devices for magnetic field diagnosis at fusion reactors

    Czech Academy of Sciences Publication Activity Database

    Bolshakova, I.; Ďuran, Ivan; Holyaka, R.; Hristoforou, E.; Marusenkov, A.

    2007-01-01

    Roč. 5, č. 1 (2007), s. 283-288 ISSN 1546-198X R&D Projects: GA AV ČR KJB100430504 Institutional research plan: CEZ:AV0Z20430508 Keywords : Galvanomagnetic * Sensor * Fusion Reactor * Magnetic Diagnostics * Radiation Hardness Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.587, year: 2007

  3. Studies on advanced superconductors for fusion device. Pt. 2. Metallic superconductors other than Nb{sub 3}Sn

    Energy Technology Data Exchange (ETDEWEB)

    Tachikawa, K.; Yamamoto, J.; Mito, T. [eds.

    1997-03-01

    A comprehensive report on the present status of the development of Nb{sub 3}Sn superconductors was published as the NIFS-MEMO-20 in March, 1996 (Part 1 of this report series). The second report of this study covers various progress so far achieved in the research and development on advanced metallic superconductors other than Nb{sub 3}Sn. Among different A15 crystal-type compounds, Nb{sub 3}Al has been fabricated into cables with large current-carrying capacity for fusion device referring its smaller sensitivity to mechanical strain than Nb{sub 3}Sn. Other high-field A15 superconductors, e.g. V{sub 3}Ga, Nb{sub 3}Ge and Nb{sub 3}(Al,Ge), have been also fabricated through different novel processes as promising alternatives to Nb{sub 3}Sn conductors. Meanwhile, B1 crystal-type NbN and C15 crystal-type V{sub 2}(Hf,Zr) high-field superconductors are characterized by their excellent tolerance to mechanical strain and neutron irradiation. Chevrel-type PbMo{sub 6}S{sub 8} compound has gained much interests due to its extremely high upper critical field. In addition, this report includes the recent progress in ultra-fine filamentary NbTi wires for AC use, and that in NbTi/Cu magnetic shields necessary in the application of high magnetic field. The data on the decay of radioactivity in a variety of metals relating to fusion superconducting magnet are also attached as appendices. We hope that this report might contribute substantially as a useful reference for the planning of fusion apparatus of next generation as well as that of other future superconducting devices. (author)

  4. An in situ accelerator-based diagnostic for plasma-material interactions science on magnetic fusion devices.

    Science.gov (United States)

    Hartwig, Zachary S; Barnard, Harold S; Lanza, Richard C; Sorbom, Brandon N; Stahle, Peter W; Whyte, Dennis G

    2013-12-01

    This paper presents a novel particle accelerator-based diagnostic that nondestructively measures the evolution of material surface compositions inside magnetic fusion devices. The diagnostic's purpose is to contribute to an integrated understanding of plasma-material interactions in magnetic fusion, which is severely hindered by a dearth of in situ material surface diagnosis. The diagnostic aims to remotely generate isotopic concentration maps on a plasma shot-to-shot timescale that cover a large fraction of the plasma-facing surface inside of a magnetic fusion device without the need for vacuum breaks or physical access to the material surfaces. Our instrument uses a compact (~1 m), high-current (~1 milliamp) radio-frequency quadrupole accelerator to inject 0.9 MeV deuterons into the Alcator C-Mod tokamak at MIT. We control the tokamak magnetic fields--in between plasma shots--to steer the deuterons to material surfaces where the deuterons cause high-Q nuclear reactions with low-Z isotopes ~5 μm into the material. The induced neutrons and gamma rays are measured with scintillation detectors; energy spectra analysis provides quantitative reconstruction of surface compositions. An overview of the diagnostic technique, known as accelerator-based in situ materials surveillance (AIMS), and the first AIMS diagnostic on the Alcator C-Mod tokamak is given. Experimental validation is shown to demonstrate that an optimized deuteron beam is injected into the tokamak, that low-Z isotopes such as deuterium and boron can be quantified on the material surfaces, and that magnetic steering provides access to different measurement locations. The first AIMS analysis, which measures the relative change in deuterium at a single surface location at the end of the Alcator C-Mod FY2012 plasma campaign, is also presented.

  5. Evaluation of efficacy of a new hybrid fusion device: a randomized, two-centre controlled trial.

    Science.gov (United States)

    Siewe, Jan; Bredow, Jan; Oppermann, Johannes; Koy, Timmo; Delank, Stefan; Knoell, Peter; Eysel, Peer; Sobottke, Rolf; Zarghooni, Kourosh; Röllinghoff, Marc

    2014-09-05

    The 360° fusion of lumbar segments is a common and well-researched therapy to treat various diseases of the spine. But it changes the biomechanics of the spine and may cause adjacent segment disease (ASD). Among the many techniques developed to avoid this complication, one appears promising. It combines a rigid fusion with a flexible pedicle screw system (hybrid instrumentation, "topping off"). However, its clinical significance is still uncertain due to the lack of conclusive data. The study is a randomized, therapy-controlled, two-centre trial conducted in a clinical setting at two university hospitals. If they meet the criteria, outpatients presenting with degenerative disc disease, facet joint arthrosis or spondylolisthesis will be included in the study and randomized into two groups: a control group undergoing conventional fusion surgery (PLIF - posterior lumbar intervertebral fusion), and an intervention group undergoing fusion surgery using a new flexible pedicle screw system (PLIF + "topping off"), which was brought on the market in 2013. Follow-up examination will take place immediately after surgery, after 6 weeks and after 6, 12, 24 and 36 months. An ongoing assessment will be performed every year.Outcome measurements will include quality of life and pain assessments using validated questionnaires (ODI - Ostwestry Disability Index, SF-36™ - Short Form Health Survey 36, COMI - Core Outcome Measure Index). In addition, clinical and radiologic ASD, sagittal balance parameters and duration of work disability will be assessed. Inpatient and 6-month mortality, surgery-related data (e.g., intraoperative complications, blood loss, length of incision, surgical duration), postoperative complications (e.g. implant failure), adverse events, and serious adverse events will be monitored and documented throughout the study. New hybrid "topping off" systems might improve the outcome of lumbar spine fusion. But to date, there is a serious lack of and a great need

  6. AxiaLIF system: minimally invasive device for presacral lumbar interbody spinal fusion

    OpenAIRE

    Block, Jon; Rapp,; Miller,Larry E.

    2011-01-01

    Steven M Rapp1, Larry E Miller2,3, Jon E Block31Michigan Spine Institute, Waterford, MI, USA; 2Miller Scientific Consulting Inc, Biltmore Lake, NC, USA; 3Jon E. Block, Ph.D., Inc., San Francisco, CA, USAAbstract: Lumbar fusion is commonly performed to alleviate chronic low back and leg pain secondary to disc degeneration, spondylolisthesis with or without concomitant lumbar spinal stenosis, or chronic lumbar instability. However, the risk of iatrogenic injury during traditional anterior, post...

  7. Development of a swelling equation for 20%-CW 316 in a fusion device

    International Nuclear Information System (INIS)

    1980-01-01

    The difficulties involved in the development of swelling correlations for AISI 316 in fusion environments are discussed. A set of void and bubble-swelling correlations has been developed which incorporates the limited available data from EBR-II and HFIR irradiations. It appears that at high fluences helium may play a minor role in the determination of total swelling over a considerable temperature range

  8. Assessment of the slowly-imploding liner (LINUS) fusion reactor concept

    International Nuclear Information System (INIS)

    Miller, R.L.; Krakowski, R.A.

    1980-01-01

    Prospects for the slowly-imploding liner (LINUS) fusion reactor concept are reviewed. The concept envisages the nondestructive, repetitive and reversible implosion of a liquid-metal cylindrical annulus (liner) onto field-reversed DT plasmoids. Adiabatic heating of the plasmoid to ignition at ultra-high magnetic fields results in a compact, high power density fusion reactor with unique solutions to several technological problems and potentially favorable economics

  9. Bipolar sealer device reduces blood loss and transfusion requirements in posterior spinal fusion for adolescent idiopathic scoliosis.

    Science.gov (United States)

    Gordon, Zachary L; Son-Hing, Jochen P; Poe-Kochert, Connie; Thompson, George H

    2013-01-01

    Reducing perioperative blood loss and transfusion requirements is important in the operative treatment of idiopathic scoliosis. This can be achieved with special frames, cell saver systems, pharmacologic aspects, and other techniques. Recently there has been interest in bipolar sealer devices as an adjunct to traditional monopolar electrocautery. However, there is limited information on this device in pediatric spinal deformity surgery. We reviewed our experience with this device in a setting of a standard institutional operative carepath. Perioperative blood loss and transfusion requirements of 50 consecutive patients with adolescent idiopathic scoliosis undergoing a posterior spinal fusion and segmental spinal instrumentation and who had a bipolar sealer device used during their surgery was compared with a control group of the 50 preceding consecutive patients who did not. Anesthesia, surgical technique, use of intraoperative epsilon aminocaproic acid (Amicar), postoperative protocol, and indications for transfusions (hemoglobin≤7.0 g/dL) were identical in both groups. The preoperative demographics for the patients in both groups were statistically the same. The bipolar sealer group demonstrated a significant reduction in intraoperative estimated blood loss, total perioperative blood loss, volume of blood products transfused, and overall transfusion rate when compared with the control group. When subgroups consisting of only hybrid or all-pedicle screw constructs were considered individually, these findings remained consistent. There were no complications associated with the use of this device. Using the bipolar sealer device is a significant adjunct in decreasing perioperative blood loss and transfusion requirements in patients undergoing surgery for adolescent idiopathic scoliosis. Level III-retrospective comparative study.

  10. Fusion neutronics

    CERN Document Server

    Wu, Yican

    2017-01-01

    This book provides a systematic and comprehensive introduction to fusion neutronics, covering all key topics from the fundamental theories and methodologies, as well as a wide range of fusion system designs and experiments. It is the first-ever book focusing on the subject of fusion neutronics research. Compared with other nuclear devices such as fission reactors and accelerators, fusion systems are normally characterized by their complex geometry and nuclear physics, which entail new challenges for neutronics such as complicated modeling, deep penetration, low simulation efficiency, multi-physics coupling, etc. The book focuses on the neutronics characteristics of fusion systems and introduces a series of theories and methodologies that were developed to address the challenges of fusion neutronics, and which have since been widely applied all over the world. Further, it introduces readers to neutronics design’s unique principles and procedures, experimental methodologies and technologies for fusion systems...

  11. Plasma Equilibrium Control in Nuclear Fusion Devices 2. Plasma Control in Magnetic Confinement Devices 2.1 Plasma Control in Tokamaks

    Science.gov (United States)

    Fukuda, Takeshi

    The plasma control technique for use in large tokamak devices has made great developmental strides in the last decade, concomitantly with progress in the understanding of tokamak physics and in part facilitated by the substantial advancement in the computing environment. Equilibrium control procedures have thereby been established, and it has been pervasively recognized in recent years that the real-time feedback control of physical quantities is indispensable for the improvement and sustainment of plasma performance in a quasi-steady-state. Further development is presently undertaken to realize the “advanced plasma control” concept, where integrated fusion performance is achieved by the simultaneous feedback control of multiple physical quantities, combined with equilibrium control.

  12. Experimental study of potential structure in a spherical IEC fusion device

    International Nuclear Information System (INIS)

    Gu, Y.; Miley, G.H.

    2000-01-01

    The spherical inertial-electrostatic confinement (SIEC) concept is designed to focus and accelerate ions and electrons radially inward towards the center of a negatively biased, highly transparent spherical grid. The converging ions create a high-density plasma core where a high fusion rate occurs. In addition, under proper conditions, the ion and electron flows create a space-charge induced double potential well (a negative potential well nested inside a positive potential well). This structure traps high-energy ions within the virtual anode created by the double potential, providing a high fusion density in the trap volume. The present experiment was designed to verify double potential well formation and trapping by a measurement of the radial birth profile of energetic (3-MeV) protons produced by D-D fusion reactions in a deuterium discharge. This experiment was designed to operate at high perveance (0.4 to 1.4 mA/kV 3/2 ), where formation of a double well is predicted theoretically. Additional steps to aid well formation included: use of the unique Star mode of operation to obtain ion beam focusing down to approximately 1.6 H the ballistic limit and the incorporation of a second electrically floating grid (in addition to the focusing/accelerating cathode grid) to reduce the ion radial energy spread to 0.34 mA/kV 3/2 . As the perveance increased, the depth of the double well also increased. At the maximum perveance studied, 1.38 mA/kV 3/2 (corresponding to 80 mA and 15 kV), the negative potential well depth, corresponding to the measured proton-rate density, was estimated to be 22%--27% of the applied cathode voltage. This represents the first conclusive demonstration of double well formation in an SIEC, since prior measurements by other researchers typically yielded marginal or negative results

  13. Development of functional ceramics for nuclear fusion devices and their property measurements in radiation environment

    International Nuclear Information System (INIS)

    Ohno, Hideo; Kondo, Tatsuo

    1989-01-01

    The research and development of high performance ceramics related to nuclear energy increase their importance. Especially innovation and application of ceramics are needed in fusion reactors. Necessity of the selection of composite elements for low activation ceramics and transmutation effects with high energy neutron are summarized in general requirements. The development of new materials such as Si 3 N 4 with good dielectric properties and the application of zirconia for high temperature electrolysis of tritiated water in tritium recycling system are summarized as topical issues. (author)

  14. Apparatus and method for removing particle species from fusion-plasma-confinement devices

    Science.gov (United States)

    Hamilton, G.W.

    1981-10-26

    In a mirror fusion plasma confinement apparatus, method and apparatus are provided for selectively removing (pumping) trapped low energy (thermal) particle species from the end cell region, without removing the still useful high energy particle species, and without requiring large power input to accomplish the pumping. Perturbation magnets are placed in the thermal barrier region of the end cell region at the turning point characteristic of trapped thermal particles, thus deflecting the thermal particles from their closed trajectory, causing them to drift sufficiently to exit the thermal barrier.

  15. 8th international workshop on plasma edge theory in fusion devices. Abstracts of invited and contributed papers

    International Nuclear Information System (INIS)

    Sipilae, S.K.; Heikkinen, J.A.

    2001-01-01

    The 8th International Workshop on Plasma Edge Theory in Fusion Devices, held at Dipoli Congress Centre, Espoo, Finland, is organised on behalf of the International Scientific Committee by Helsinki University of Technology and VTT (Technical Research Centre of Finland). Similar to the seven preceding Workshops, it addresses the theory for the boundary layer of magnetically confined fusion plasmas. It reflects the present status of the theory for the edge region of fusion plasmas. Emphasis is placed on the development of theory and of appropriate numerical methods as well as on self-consistent modelling of experimental data (including also empirical elements). The following topics are covered: basic edge plasma theory, models of special phenomena and edge control, and integrated edge plasma modelling. The International Scientific Committee has selected the papers and compiled the scientific programme. All other arrangements have been made by the Local Organising Committee. The Workshop is supported by the European Commission, High-Level Scientific Conferences. This Book of Abstracts contains the scientific programme and the abstracts of the invited and contributed papers. The Workshop has seven invited lectures of 60 minutes duration (including 10 minutes for discussion). In addition, 10 contributed papers were selected for oral presentation of 30 minutes duration (including five minutes for discussion). All oral presentations are given in plenary sessions. The remaining 34 contributed papers are presented as posters in three sessions. The invited lectures and contributed oral papers are presented also as posters. All invited and contributed papers will be refereed and published also as a regular issue of the journal Contributions to Plasma Physics. (orig.)

  16. Development of a diagnostic technique based on Cherenkov effect for measurements of fast electrons in fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Plyusnin, V. V.; Duarte, P.; Fernandes, H.; Silva, C. [Association Euratom/IST, Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, Av. Rovisco Pais, 1049-001 Lisboa (Portugal); Jakubowski, L.; Zebrowski, J.; Malinowski, K.; Rabinski, M.; Sadowski, M. J. [National Centre for Nuclear Research (NCBJ), 7 Andrzeja Soltana Str., 05-400 Otwock (Poland)

    2012-08-15

    A diagnostic technique based on the Cherenkov effect is proposed for detection and characterization of fast (super-thermal and runaway) electrons in fusion devices. The detectors of Cherenkov radiation have been specially designed for measurements in the ISTTOK tokamak. Properties of several materials have been studied to determine the most appropriate one to be used as a radiator of Cherenkov emission in the detector. This technique has enabled the detection of energetic electrons (70 keV and higher) and the determination of their spatial and temporal variations in the ISTTOK discharges. Measurement of hard x-ray emission has also been carried out in experiments for validation of the measuring capabilities of the Cherenkov-type detector and a high correlation was found between the data of both diagnostics. A reasonable agreement was found between experimental data and the results of numerical modeling of the runaway electron generation in ISTTOK.

  17. Integrated Prediction and Mitigation Methods of Materials Damage and Lifetime Assessment during Plasma Operation and Various Instabilities in Fusion Devices

    Energy Technology Data Exchange (ETDEWEB)

    Hassanein, Ahmed [Purdue Univ., West Lafayette, IN (United States)

    2015-03-31

    This report describes implementation of comprehensive and integrated models to evaluate plasma material interactions during normal and abnormal plasma operations. The models in full3D simulations represent state-of-the art worldwide development with numerous benchmarking of various tokamak devices and plasma simulators. In addition, significant number of experimental work has been performed in our center for materials under extreme environment (CMUXE) at Purdue to benchmark the effect of intense particle and heat fluxes on plasma-facing components. This represents one-year worth of work and resulted in more than 23 Journal Publications and numerous conferences presentations. The funding has helped several students to obtain their M.Sc. and Ph.D. degrees and many of them are now faculty members in US and around the world teaching and conducting fusion research. Our work has also been recognized through many awards.

  18. Integrated Prediction and Mitigation Methods of Materials Damage and Lifetime Assessment during Plasma Operation and Various Instabilities in Fusion Devices

    International Nuclear Information System (INIS)

    Hassanein, Ahmed

    2015-01-01

    This report describes implementation of comprehensive and integrated models to evaluate plasma material interactions during normal and abnormal plasma operations. The models in full3D simulations represent state-of-the art worldwide development with numerous benchmarking of various tokamak devices and plasma simulators. In addition, significant number of experimental work has been performed in our center for materials under extreme environment (CMUXE) at Purdue to benchmark the effect of intense particle and heat fluxes on plasma-facing components. This represents one-year worth of work and resulted in more than 23 Journal Publications and numerous conferences presentations. The funding has helped several students to obtain their M.Sc. and Ph.D. degrees and many of them are now faculty members in US and around the world teaching and conducting fusion research. Our work has also been recognized through many awards.

  19. Mechanical performance of cervical intervertebral body fusion devices: A systematic analysis of data submitted to the Food and Drug Administration.

    Science.gov (United States)

    Peck, Jonathan H; Sing, David C; Nagaraja, Srinidhi; Peck, Deepa G; Lotz, Jeffrey C; Dmitriev, Anton E

    2017-03-21

    Cervical intervertebral body fusion devices (IBFDs) are utilized to provide stability while fusion occurs in patients with cervical pathology. For a manufacturer to market a new cervical IBFD in the United States, substantial equivalence to a cervical IBFD previously cleared by FDA must be established through the 510(k) regulatory pathway. Mechanical performance data are typically provided as part of the 510(k) process for IBFDs. We reviewed all Traditional 510(k) submissions for cervical IBFDs deemed substantially equivalent and cleared for marketing from 2007 through 2014. To reduce sources of variability in test methods and results, analysis was restricted to cervical IBFD designs without integrated fixation, coatings, or expandable features. Mechanical testing reports were analyzed and results were aggregated for seven commonly performed tests (static and dynamic axial compression, compression-shear, and torsion testing per ASTM F2077, and subsidence testing per ASTM F2267), and percentile distributions of performance measurements were calculated. Eighty-three (83) submissions met the criteria for inclusion in this analysis. The median device yield strength was 10,117N for static axial compression, 3680N for static compression-shear, and 8.6Nm for static torsion. Median runout load was 2600N for dynamic axial compression, 1400N for dynamic compression-shear, and ±1.5Nm for dynamic torsion. In subsidence testing, median block stiffness (Kp) was 424N/mm. The mechanical performance data presented here will aid in the development of future cervical IBFDs by providing a means for comparison for design verification purposes. Published by Elsevier Ltd.

  20. The Catchment Feature Model: A Device for Multimodal Fusion and a Bridge between Signal and Sense

    Science.gov (United States)

    Quek, Francis

    2004-12-01

    The catchment feature model addresses two questions in the field of multimodal interaction: how we bridge video and audio processing with the realities of human multimodal communication, and how information from the different modes may be fused. We argue from a detailed literature review that gestural research has clustered around manipulative and semaphoric use of the hands, motivate the catchment feature model psycholinguistic research, and present the model. In contrast to "whole gesture" recognition, the catchment feature model applies a feature decomposition approach that facilitates cross-modal fusion at the level of discourse planning and conceptualization. We present our experimental framework for catchment feature-based research, cite three concrete examples of catchment features, and propose new directions of multimodal research based on the model.

  1. The Catchment Feature Model: A Device for Multimodal Fusion and a Bridge between Signal and Sense

    Directory of Open Access Journals (Sweden)

    Francis Quek

    2004-09-01

    Full Text Available The catchment feature model addresses two questions in the field of multimodal interaction: how we bridge video and audio processing with the realities of human multimodal communication, and how information from the different modes may be fused. We argue from a detailed literature review that gestural research has clustered around manipulative and semaphoric use of the hands, motivate the catchment feature model psycholinguistic research, and present the model. In contrast to “whole gesture” recognition, the catchment feature model applies a feature decomposition approach that facilitates cross-modal fusion at the level of discourse planning and conceptualization. We present our experimental framework for catchment feature-based research, cite three concrete examples of catchment features, and propose new directions of multimodal research based on the model.

  2. Evaluation of an Electrostatic Dust Removal System with Potential Application in Next-Step Fusion Devices

    International Nuclear Information System (INIS)

    Friesen, F.Q.L.; John, B.; Skinner, C.H.; Roquemore, A.L.; Calle, C.I.

    2011-01-01

    The ability to manage inventories of carbon, tritium, and high-Z elements in fusion plasmas depends on means for effective dust removal. A dust conveyor, based on a moving electrostatic potential well, was tested with particles of tungsten, carbon, glass and sand. A digital microscope imaged a representative portion of the conveyor, and dust particle size and volume distributions were derived before and after operation. About 10 mm3 volume of carbon and tungsten particles were moved in under 5 seconds. The highest driving amplitude tested of 3 kV was the most effective. The optimal driving frequency was 210 Hz (maximum tested) for tungsten particles, decreasing to below 60 Hz for the larger sand particles. Measurements of particle size and volume distributions after 10 and 100 cycles show the breaking apart of agglomerated carbon, and the change in particle distribution over short timescales (<1 s).

  3. Liquid metals as alternative solution for the power exhaust of future fusion devices: status and perspective

    International Nuclear Information System (INIS)

    Coenen, J W; Philipps, V; Sergienko, G; Terra, A; Unterberg, B; Wegener, T; De Temmerman, G; Van den Bekerom, D C M; Federici, G; Strohmayer, G

    2014-01-01

    Applying liquid metals as plasma facing components for fusion power-exhaust can potentially ameliorate lifetime issues as well as limitations to the maximum allowed surface heat loads by allowing for a more direct contact with the coolant. The material choice has so far been focused on lithium (Li), as it showed beneficial impact on plasma operation. Here materials such as tin (Sn), gallium (Ga) and aluminum (Al) are discussed as alternatives potentially allowing higher operating temperatures without strong evaporation. Power loads of up to 25 MW m −2 for a Sn/W component can be envisioned based on calculations and modeling. Reaching a higher operating temperature due to material re-deposition will be discussed. Liquids typically face stability issues due to j × B forces, potential pressure and magnetohydrodynamic driven instabilities. The capillary porous system is used for stabilization by a mesh (W and Mo) substrate and replenishment by means of capillary action. (paper)

  4. Computerized cost estimation spreadsheet and cost data base for fusion devices

    International Nuclear Information System (INIS)

    Hamilton, W.R.; Rothe, K.E.

    1985-01-01

    Component design parameters (weight, surface area, etc.) and cost factors are input and direct and indirect costs are calculated. The cost data base file derived from actual cost experience within the fusion community and refined to be compatible with the spreadsheet costing approach is a catalog of cost coefficients, algorithms, and component costs arranged into data modules corresponding to specific components and/or subsystems. Each data module contains engineering, equipment, and installation labor cost data for different configurations and types of the specific component or subsystem. This paper describes the assumptions, definitions, methodology, and architecture incorporated in the development of the cost estimation spreadsheet and cost data base, along with the type of input required and the output format

  5. Assessment of martensitic steels as structural materials in magnetic fusion devices

    International Nuclear Information System (INIS)

    Rawls, J.M.; Chen, W.Y.K.; Cheng, E.T.; Dalessandro, J.A.; Miller, P.H.; Rosenwasser, S.N.; Thompson, L.D.

    1980-01-01

    This manuscript documents the results of preliminary experiments and analyses to assess the feasibility of incorporating ferromagnetic martensitic steels in fusion reactor designs and to evaluate the possible advantages of this class of material with respect to first wall/blanket lifetime. The general class of alloys under consideration are ferritic steels containing from about 9 to 13 percent Cr with some small additions of various strengthening elements such as Mo. These steels are conventionally used in the normalized and tempered condition for high temperature applications and can compete favorably with austenitic alloys up to about 600 0 C. Although the heat treatment can result in either a tempered martensite or bainite structure, depending on the alloy and thermal treatment parameters, this general class of materials will be referred to as martensitic stainless steels for simplicity

  6. The manufacture of carbon armoured plasma-facing components for fusion devices

    International Nuclear Information System (INIS)

    Schedler, B.; Huber, T.; Zabernig, A.; Rainer, F.; Scheiber, K.H.; Schedle, D.

    2001-01-01

    Within the last decade Plansee has been active in the development and manufacture of different plasma-facing-components for nuclear fusion experiments consisting in a tungsten or CFC-armor joined onto metallic substrates like TZM, stainless steel or copper-alloys. The manufacture of these components requires unique joining technologies in order to obtain reliable thermo mechanical stable joints able to withstand highest heat fluxes without any deterioration of the joint. In an overview the different techniques will be presented by some examples of components already manufactured and successfully tested under high heat flux conditions. Furthermore an overview will be given on the manufacture of different high heat flux components for TORE SUPRA, Wendelstein 7-X and ITER. (author)

  7. Fusion Physics

    Energy Technology Data Exchange (ETDEWEB)

    Kikuchi, Mitsuru; Lackner, Karl; Tran, Minh Quang [eds.

    2012-09-15

    Recreating the energy production process of the Sun - nuclear fusion - on Earth in a controlled fashion is one of the greatest challenges of this century. If achieved at affordable costs, energy supply security would be greatly enhanced and environmental degradation from fossil fuels greatly diminished. Fusion Physics describes the last fifty years or so of physics and research in innovative technologies to achieve controlled thermonuclear fusion for energy production. The International Atomic Energy Agency (IAEA) has been involved since its establishment in 1957 in fusion research. It has been the driving force behind the biennial conferences on Plasma Physics and Controlled Thermonuclear Fusion, today known as the Fusion Energy Conference. Hosted by several Member States, this biennial conference provides a global forum for exchange of the latest achievements in fusion research against the backdrop of the requirements for a net energy producing fusion device and, eventually, a fusion power plant. The scientific and technological knowledge compiled during this series of conferences, as well as by the IAEA Nuclear Fusion journal, is immense and will surely continue to grow in the future. It has led to the establishment of the International Thermonuclear Experimental Reactor (ITER), which represents the biggest experiment in energy production ever envisaged by humankind.

  8. Neutronic design of pulse operation simulating device for in-pile functional test of fusion blanket by MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Nagao, Yoshiharu; Nakamichi, Masaru; Kawamura, Hiroshi [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan)

    2000-03-01

    The pulse operation of a fusion reactor can be simulated in a fission reactor by controlling the neutron flux entering a test section by using a rotating 'hollow cylinder with window' made of hafnium. The rotating cylinder is installed between the test section and the fixed outer neutron absorber cylinder and is also made of hafnium with an opening in the direction to the core center. For gathering engineering data for the tritium breeding blanket such as characteristics of temperature change, tritium release and recovery, etc., it is desirable that the ratio of minimum to maximum thermal neutron fluxes is greater than 1:10. Design calculations were performed for the test assembly which considered local neutronic effects and the mechanical constraints of the device. From the results of these calculations, the ratio of minimum to maximum thermal neutron flux under irradiation would be about 1:10 using a pulse operation simulating device which has a thickness of 6.5 mm and a 150deg window angle for the rotating hollow cylinder and 5.0 mm in thickness of fixed neutron absorber. (author)

  9. Multi-Sensor Fusion for Enhanced Contextual Awareness of Everyday Activities with Ubiquitous Devices

    Directory of Open Access Journals (Sweden)

    John J. Guiry

    2014-03-01

    Full Text Available In this paper, the authors investigate the role that smart devices, including smartphones and smartwatches, can play in identifying activities of daily living. A feasibility study involving N = 10 participants was carried out to evaluate the devices’ ability to differentiate between nine everyday activities. The activities examined include walking, running, cycling, standing, sitting, elevator ascents, elevator descents, stair ascents and stair descents. The authors also evaluated the ability of these devices to differentiate indoors from outdoors, with the aim of enhancing contextual awareness. Data from this study was used to train and test five well known machine learning algorithms: C4.5, CART, Naïve Bayes, Multi-Layer Perceptrons and finally Support Vector Machines. Both single and multi-sensor approaches were examined to better understand the role each sensor in the device can play in unobtrusive activity recognition. The authors found overall results to be promising, with some models correctly classifying up to 100% of all instances.

  10. Effects of DD and DT neutron irradiation on some Si devices for fusion diagnostics

    International Nuclear Information System (INIS)

    Tanimura, Y.; Iida, T.

    1998-01-01

    In order to examine the difference in the irradiation effects on Si devices between DT and DD neutrons, CCD image sensors, memory ICs and a Si detector were irradiated with neutrons from a deuteron accelerator. The transient effects (i.e. neutron-induced background noises) and permanent effects (i.e. neutron damage) on them were in situ measured during irradiation. Regarding the transient effects, brightening spot noises, soft-error upsets and induced-charge noises were measured for the CCDs, memory ICs and Si detector, respectively. As for the permanent effect, the number of damaged cells of the CCDs and the leakage current of the Si detector increased with neutron fluence. Also we developed a Monte-Carlo code with the TRIM code to evaluate the correlation of DT and DD neutron effects on Si devices. The calculated correlation factor of DT and DD neutron damage for Si devices agreed approximately with the correlation factor obtained from the irradiation experiments on the CCDs and Si detector. (orig.)

  11. Effects of DD and DT neutron irradiation on some Si devices for fusion diagnostics

    Science.gov (United States)

    Tanimura, Yoshihiko; Iida, Toshiyuki

    1998-10-01

    In order to examine the difference in the irradiation effects on Si devices between DT and DD neutrons, CCD image sensors, memory ICs and a Si detector were irradiated with neutrons from a deuteron accelerator. The transient effects (i.e. neutron-induced background noises) and permanent effects (i.e. neutron damage) on them were in situ measured during irradiation. Regarding the transient effects, brightening spot noises, soft-error upsets and induced-charge noises were measured for the CCDs, memory ICs and Si detector, respectively. As for the permanent effect, the number of damaged cells of the CCDs and the leakage current of the Si detector increased with neutron fluence. Also we developed a Monte-Carlo code with the TRIM code to evaluate the correlation of DT and DD neutron effects on Si devices. The calculated correlation factor of DT and DD neutron damage for Si devices agreed approximately with the correlation factor obtained from the irradiation experiments on the CCDs and Si detector.

  12. Conceptual design of a generic pulse schedule and event handling editor for improved fusion device operation

    International Nuclear Information System (INIS)

    Barana, Oliviero; Nouailletas, Rémy; Brémond, Sylvain; Moreau, Philippe; Allegretti, Ludovic; Balme, Stéphane; Ravenel, Nathalie; Mannori, Simone; Guillerminet, Bernard; Leroux, Fabrice; Douai, David; Nardon, Eric; Hertout, Patrick; Saint-Laurent, François

    2013-01-01

    Highlights: ► Real-time event handling requires extended functionalities of pulse schedule editors and plasma control systems ► A new pulse schedule editor, conceived for parameterization of systematic off-normal event handling, is described ► A global, generic approach on off-normal event handling is highlighted ► The functional architecture of an off-normal event handling oriented plasma control system is discussed ► The main objects of the pulse schedule editor are the segment-descriptor object and the scenario-descriptor object. -- Abstract: Coping with unexpected events is an important issue of nuclear fusion experiments. The future machines, characterized by very long plasma discharges and actively cooled metallic plasma-facing components, will require a systematic intervention in real time, in order to maximize the performance and protect the investment. The real-time management of events will require extending the functionalities of the current pulse schedule editors with the possibility of using reference waveforms provided with acceptability margins and setting up advanced mitigation strategies and event countermeasures. With this purpose, a new pulse schedule editor, based on a time-segment approach for the preparation of experimental scenarios, is being conceived on Tore Supra, together with a new plasma control system. This paper will report on their conceptual design and give account of the preliminary results of a feasibility study currently under way in order to prepare a possible implementation on Tore Supra

  13. Conceptual design of a generic pulse schedule and event handling editor for improved fusion device operation

    Energy Technology Data Exchange (ETDEWEB)

    Barana, Oliviero, E-mail: oliviero.barana@cea.fr [CEA, IRFM, F-13108 Saint-Paul-Lez Durance (France); Nouailletas, Rémy; Brémond, Sylvain; Moreau, Philippe; Allegretti, Ludovic; Balme, Stéphane; Ravenel, Nathalie [CEA, IRFM, F-13108 Saint-Paul-Lez Durance (France); Mannori, Simone [ENEA C.R. Brasimone, 40032 Camugnano (Italy); Guillerminet, Bernard; Leroux, Fabrice; Douai, David; Nardon, Eric; Hertout, Patrick; Saint-Laurent, François [CEA, IRFM, F-13108 Saint-Paul-Lez Durance (France)

    2013-10-15

    Highlights: ► Real-time event handling requires extended functionalities of pulse schedule editors and plasma control systems ► A new pulse schedule editor, conceived for parameterization of systematic off-normal event handling, is described ► A global, generic approach on off-normal event handling is highlighted ► The functional architecture of an off-normal event handling oriented plasma control system is discussed ► The main objects of the pulse schedule editor are the segment-descriptor object and the scenario-descriptor object. -- Abstract: Coping with unexpected events is an important issue of nuclear fusion experiments. The future machines, characterized by very long plasma discharges and actively cooled metallic plasma-facing components, will require a systematic intervention in real time, in order to maximize the performance and protect the investment. The real-time management of events will require extending the functionalities of the current pulse schedule editors with the possibility of using reference waveforms provided with acceptability margins and setting up advanced mitigation strategies and event countermeasures. With this purpose, a new pulse schedule editor, based on a time-segment approach for the preparation of experimental scenarios, is being conceived on Tore Supra, together with a new plasma control system. This paper will report on their conceptual design and give account of the preliminary results of a feasibility study currently under way in order to prepare a possible implementation on Tore Supra.

  14. Computation of stationary 3D halo currents in fusion devices with accuracy control

    Science.gov (United States)

    Bettini, Paolo; Specogna, Ruben

    2014-09-01

    This paper addresses the calculation of the resistive distribution of halo currents in three-dimensional structures of large magnetic confinement fusion machines. A Neumann electrokinetic problem is solved on a geometry so complicated that complementarity is used to monitor the discretization error. An irrotational electric field is obtained by a geometric formulation based on the electric scalar potential, whereas three geometric formulations are compared to obtain a solenoidal current density: a formulation based on the electric vector potential and two geometric formulations inspired from mixed and mixed-hybrid Finite Elements. The electric vector potential formulation is usually considered impractical since an enormous computing power is wasted by the topological pre-processing it requires. To solve this challenging problem, we present novel algorithms based on lazy cohomology generators that enable to save orders of magnitude computational time with respect to all other state-of-the-art solutions proposed in literature. Believing that our results are useful in other fields of scientific computing, the proposed algorithm is presented as a detailed pseudocode in such a way that it can be easily implemented.

  15. Two-dimensional cross-section and SED uncertainty analysis for the Fusion Engineering Device (FED)

    International Nuclear Information System (INIS)

    Embrechts, M.J.; Urban, W.T.; Dudziak, D.J.

    1982-01-01

    The theory of two-dimensional cross-section and secondary-energy-distribution (SED) sensitivity was implemented by developing a two-dimensional sensitivity and uncertainty analysis code, SENSIT-2D. Analyses of the Fusion Engineering Design (FED) conceptual inboard shield indicate that, although the calculated uncertainties in the 2-D model are of the same order of magnitude as those resulting from the 1-D model, there might be severe differences. The more complex the geometry, the more compulsory a 2-D analysis becomes. Specific results show that the uncertainty for the integral heating of the toroidal field (TF) coil for the FED is 114.6%. The main contributors to the cross-section uncertainty are chromium and iron. Contributions to the total uncertainty were smaller for nickel, copper, hydrogen and carbon. All analyses were performed with the Los Alamos 42-group cross-section library generated from ENDF/B-V data, and the COVFILS covariance matrix library. The large uncertainties due to chromium result mainly from large convariances for the chromium total and elastic scattering cross sections

  16. Particle and energy transport studies on TFTR and implications for helium ash in future fusion devices

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Efthimion, P.C.; Rewoldt, G.; Stratton, B.C.; Tang, W.M.; Bell, R.E.; Grek, B.; Hulse, R.A.; Johnson, D.W.; Hill, K.W.; Mansfield, D.K.; McCune, D.; Mikkelsen, D.R.; Park, H.K.; Ramsey, A.T.; Scott, S.D.; Taylor, G.; Timberlake, J.; Zarnstorff, M.C.

    1992-01-01

    Particle and energy transport in tokamak plasmas have long been subjects of vigorous investigation. Present-day measurement techniques permit radially resolved studies of the transport of electron perturbations, low- and high-Z impurities, and energy. In addition, developments in transport theory provide tools that can be brought to bear on transport issues. Here, we examine local particle transport measurements of electrons, fully-stripped thermal helium, and helium-like iron in balanced-injection L-mode and enhanced confinement deuterium plasmas on TFTR of the same plasma current, toroidal field, and auxiliary heating power. He 2+ and Fe 24+ transport has been studied with charge exchange recombination spectroscopy, while electron transport has been studied by analyzing the perturbed electron flux following the same helium puff used for the He 2+ studies. By examining the electron and He 2+ responses following the same gas puff in the same plasmas, an unambiguous comparison of the transport of the two species has been made. The local energy transport has been examined with power balance analysis, allowing for comparisons to the local thermal fluxes. Some particle and energy transport results from the Supershot have been compared to a transport model based on a quasilinear picture of electrostatic toroidal drift-type microinstabilities. Finally, implications for future fusion reactors of the observed correlation between thermal transport and helium particle transport is discussed

  17. The long way to steady state fusion plasmas - the superconducting stellarator device Wendelstein 7-X

    CERN Multimedia

    CERN. Geneva

    2016-01-01

    The stable generation of high temperature Hydrogen plasmas (ion and electron temperature in the range 10-20 keV) is the basis for the use of nuclear fusion to generate heat and thereby electric power. The most promising path is to use strong, toroidal, twisted magnetic fields to confine the electrically charged plasma particles in order to avoid heat losses to the cold, solid wall elements. Two magnetic confinement concepts have been proven to be most suitable: (a) the tokamak and (b) the stellarator. The stellarator creates the magnetic field by external coils only, the tokamak by combining the externally created field with the magnetic field generated by a strong current in the plasma. “Wendelstein 7-X” is the name of a large superconducting stellarator that went successfully into operation after 15 years of construction. With 30 m3 plasma volume, 3 T magnetic field on axis, and 10 MW micro wave heating power, Hydrogen plasmas are generated that allow one to establish a scientific basis for the extrapol...

  18. Multi parametric sensitivity study applied to temperature measurement of metallic plasma facing components in fusion devices

    International Nuclear Information System (INIS)

    Aumeunier, M-H.; Corre, Y.; Firdaouss, M.; Gauthier, E.; Loarer, T.; Travere, J-M.; Gardarein, J-L.; EFDA JET Contributor

    2013-06-01

    In nuclear fusion experiments, the protection system of the Plasma Facing Components (PFCs) is commonly ensured by infrared (IR) thermography. Nevertheless, the surface monitoring of new metallic plasma facing component, as in JET and ITER is being challenging. Indeed, the analysis of infrared signals is made more complicated in such a metallic environment since the signals will be perturbed by the reflected photons coming from high temperature regions. To address and anticipate this new measurement environment, predictive photonic models, based on Monte-Carlo ray tracing (SPEOS R CAA V5 Based), have been performed to assess the contribution of the reflective part in the total flux collected by the camera and the resulting temperature error. This paper deals with the effects of metals features, as the emissivity and reflectivity models, on the accuracy of the surface temperature estimation. The reliability of the features models is discussed by comparing the simulation with experimental data obtained with the wide angle IR thermography system of JET ITER like wall. The impact of the temperature distribution is studied by considering two different typical plasma scenarios, in limiter (ITER start-up scenario) and in X-point configurations (standard divertor scenario). The achievable measurement performances of IR system and risks analysis on its functionalities are discussed. (authors)

  19. Detective studies of soft X-ray tomography on controlled thermonuclear fusion device

    International Nuclear Information System (INIS)

    Li Linzhong; Su Fei

    2004-01-01

    In is necessary to design tomographic detective system with very high accuracy and high quality. It is such a detective system that its five resolutions are all very high quality. The five resolutions are: the radial resolution, the angular resolution, the spatial resolution of detector, the resolution of detector array, and the time resolution. The radial resolution is decided by the number of detectors in detector array. The angular resolutions depend on the number of detector arrays. According to the concrete condition of controlled device, through making special rectangular detector the optimum spatial resolution of detector and the optimum spatial resolution of detector array can be obtained. The high time resolution can be got by making wide-band ampli-filter circuit system. The tomographic system with high quality can use the multi-angle multi-array mode or perfect single array mode. The soft X-ray tomographic system with high sensitivity can measure the stable signal and perform the tomography under the conditions of Te ∼150 eV, ne ∼1013 cm-3 on the small Tokamak devices. (authors)

  20. Material options for a commercial fusion reactor first wall

    International Nuclear Information System (INIS)

    Dabiri, A.E.

    1986-05-01

    A study has been conducted to evaluate the potential of various materials for use as first walls in high-power-density commercial fusion reactors. Operating limits for each material were obtained based on a number of criteria, including maximum allowable structural temperatures, critical heat flux, ultimate tensile strength, and design-allowable stress. The results with water as a coolant indicate that a modified alloy similar to HT-9 may be a suitable candidate for low- and medium-power-density reactor first walls with neutron loads of up to 6 MW/m 2 . A vanadium or copper alloy must be used for high-power-density reactors. The neutron wall load limit for vanadium alloys is about 14 MW 2 , provided a suitable coating material is chosen. The extremely limited data base for radiation effects hinders any quantitative assessment of the limits for copper alloys

  1. Core configuration of a gas-cooled reactor as a tritium production device for fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakaya, H., E-mail: nakaya@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Matsuura, H.; Nakao, Y. [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Shimakawa, S.; Goto, M.; Nakagawa, S. [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki (Japan); Nishikawa, M. [Malaysia-Japan International Institute of Technology, UTM, Kuala Lumpur 54100 (Malaysia)

    2014-05-01

    The performance of a high-temperature gas-cooled reactor as a tritium production device is examined, assuming the compound LiAlO{sub 2} as the tritium-producing material. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations are carried out. To load sufficient Li into the core, LiAlO{sub 2} is loaded into the removable reflectors that surround the ring-shaped fuel blocks in addition to the burnable poison insertion holes. It is shown that module high-temperature gas-cooled reactors with a total thermal output power of 3 GW can produce almost 8 kg of tritium in a year.

  2. Practical sublimation source for large-scale chromium gettering in fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Simpkins, J E; Gabbard, W A; Emerson, L C; Mioduszewski, P K [Oak Ridge National Lab., TN (USA)

    1984-05-01

    This paper describe the fabrication and testing of a large-scale chromium sublimation source that resembles the VARIAN Ti-ballsup(TM) in its design. The device consists of a hollow chromium sphere with a diameter of approximately 3 cm and an incandescent filament for radiation heating from inside the ball. We also discuss the gettering technique utilizing this source. The experimental arrangement consists of an ultrahigh vacuum (UHV) system instrumented for total and partial pressure measurements, a film thickness monitor, thermocouples, an optical pyrometer, and appropriate instrumentation to measure the heating power. The results show the temperature and corresponding sublimation rate of the Cr-ball as functions of input power. In addition, an example of the total pumping speed of a gettered surface is shown.

  3. A practical sublimation source for large-scale chromium gettering in fusion devices

    International Nuclear Information System (INIS)

    Simpkins, J.E.; Gabbard, W.A.; Emerson, L.C.; Mioduszewski, P.K.

    1984-01-01

    This paper describe the fabrication and testing of a large-scale chromium sublimation source that resembles the VARIAN Ti-ballsup(TM) in its design. The device consists of a hollow chromium sphere with a diameter of approximately 3 cm and an incandescent filament for radiation heating from inside the ball. We also discuss the gettering technique utilizing this source. The experimental arrangement consists of an ultrahigh vacuum (UHV) system instrumented for total and partial pressure measurements, a film thickness monitor, thermocouples, an optical pyrometer, and appropriate instrumentation to measure the heating power. The results show the temperature and corresponding sublimation rate of the Cr-ball as functions of input power. In addition, an example of the total pumping speed of a gettered surface is shown. (orig.)

  4. Characterization of the Plasma Edge for Technique of Atomic Helium Beam in the CIEMAT Fusion Device; Caracterizacion del Borde del Plasma del Dispositivo de Fusion TJ-II del CIEMAT mediante el Diagnostico del Haz Supersonico de Helio

    Energy Technology Data Exchange (ETDEWEB)

    Hidalgo, A.

    2003-07-01

    In this report, the measurement of Electron Temperature and Density in the Boundary Plasma of TJ-II with a Supersonic Helium Beam Diagnostic and work devoted to the upgrading of this technique are described. Also, simulations of Laser Induced Fluorescence (LIF) studies of level populations of electronically excited He atoms are shown. This last technique is now being installed in the CIEMAT fusion device. (Author ) 36 refs.

  5. Research on nuclear fusion reactor - Development of mm-wve (Electron cyclotron) heating device

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sae Young; Myung, Jung Su; Lee, Keun Ho; Lee, Myung Jae; Kim, Hyung Suk; Hur, Jin Woo; Song, Ho Young [Institute for Advanced Engineering, Seoul (Korea, Republic of)

    1996-08-01

    To establish cooperating system with foreign relevant research institutes, consultation has been given to IAE by Dr. T. V. George regarding ECRH and gyrotron development plan. Discussions with Prof. Temkin and Dr. Kreisher at MIT, who are working for ITER gyrotron development, were made and those helped IAE to collect necessary information for fundamental parameters of ECCD. Also, Prof. Vic Granatstein, U. of Maryland, and Dr. Baruch Levush, NRL, were consulted for computer codes of the gyrotron R and D. It will also be prepared for cooperation in ECCD and mm-wave heating with device research teams of General Atomics and Russia. By visiting various University labs and research institutes and investigating the up-to-date research results, the basic operating parameters of gyrotron for KSTAR project has been determined. By cooperation with MIT, a conceptual design has been made for the KSTAR gyrotron that should generate 1 MW and 110 GHz CW waves. The simulation result of EGUN using self-consistent theory shows that 1.2 MW power with the efficiency of 42.8% can be obtained for TE22,6,1 mode where the average ohmic loss is 0.54 kW/cm{sup 2} assuming 77 kV cathode voltage, 34 A beam current, velocity ratio of 1.62 and perpendicular velocity spread of 6.5%. 9 refs., 5 figs., 3 tabs. (author)

  6. Radio frequency siliconization: An approach to the coating for the future large superconducting fusion devices

    International Nuclear Information System (INIS)

    Li, J.; Zhao, Y.P.; Wan, B.N.; Gong, X.Z.; Zhen, M.; Gu, X.M.; Zhang, X.D.; Luo, J.R.; Wan, Y.X.; Xie, J.K.; Li, C.F.; Chen, J.L.; Toi, K.; Noda, N.; Watari, T.

    2001-01-01

    Radio frequency (rf) siliconization has been carried out on the HT-7 superconducting tokamak in the presence of a high magnetic field, which is a try on superconducting tokamaks. Three different procedures of rf siliconization have been tested and a very promising method to produce high quality silicon films was found after comparing the film properties and plasma performance produced by these three different procedures. The Si/C films are amorphous, semitransparent, and homogeneous throughout the layer and adhere firmly to all the substrates. The advantages of silicon atoms as a powerful radiator and a good oxygen getter have been proved. An outstanding merit of rf siliconization to superconducting devices is its fast recovery after a serious degradation of the condition due to the leakage of air to good wall conditions. A wider stable operation region has been obtained and plasma performance is improved immediately after each siliconization due to significant reduction of impurities. Energy confinement time increases more than 50% and particle confinement time increases by a factor of 2. The lifetime of the silicon film is more than 400 standard ohmic heated plasma discharges. Simulation shows that the confinement improvement is due to the reduction of the electron thermal diffusivity in the outer region of the plasma

  7. An electrically conducting first wall for the fusion engineering device-A (FED-A) tokamak

    International Nuclear Information System (INIS)

    Cramer, B.A.; Fuller, G.M.

    1983-01-01

    The first wall of the tokamak FED-A device was designed to satisfy two conflicting requirements. They are a low electrical resistance to give a long eddy-current decay time and a high neutron transparency to give a favorable tritium breeding ratio. The tradeoff between these conflicting requirements resulted in a copper alloy first wall that satisfied the specific goals for FED-A, i.e., a minimum eddy-current decay time of 0.5 sec and a tritium breeding ratio of at least 1.2. Aluminum alloys come close to meeting the requirements and would also probably work. Stainless steel will not work in this application because shells thin enough to satisfy temperature and stress limits are not thick enough to give a long eddy-current decay time and to avoid disruption induced melting. The baseline first wall design is a rib-stiffened, double-wall construction. The total wall thickness is 1.5 cm, including a water coolant thickness of 0.5 cm. The first wall is divided into twelve 30-degree sectors. Flange rings at the ends of each sector are bolted together to form the torus. Structural support is provided at the top center of each sector

  8. Tungsten: An option for divertor and main chamber plasma facing components in future fusion devices

    International Nuclear Information System (INIS)

    Neu, R.; Dux, R.; Kallenbach, A.; Maggi, C.F.; Puetterich, T.; Balden, M.; Eich, T.; Fuchs, J.C.; Gruber, O.; Herrmann, A.; Maier, H.; Mueller, H.W.; Pugno, R.; Radivojevic, I.; Rohde, V.; Sips, A.C.C.; Suttrop, W.; Ye, M.Y.; O'Mullane, M.; Whiteford, A.

    2005-01-01

    The tungsten programme in ASDEX Upgrade is pursued towards a full high-Z device. The spectroscopic diagnostic and the cooling factor of W have been extended and refined. The W-coated surfaces represent now a fraction of 65% (24.8 m2). The only two major components which are not yet coated are the strikepoint region of the lower divertor as well as the limiters at the low field side. While extending the W surfaces, the W concentration and the discharge behaviour have changed gradually pointing to critical issues when operating with a W wall: anomalous transport in the plasma centre should not be too low, otherwise neoclassical accumulation can occur. A very successful remedy is the addition of central RF heating at the 20-30% level. Regimes with low ELM activity show increased impurity concentration over the whole plasma radius. These discharges can be cured by increasing the ELM frequency through pellet ELM pacemaking or by higher heating power. Moderate gas puffing also mitigates the impurity influx and penetration, however at the expense of lower confinement. The erosion yield at the low field side guard limiter can be as high as 10 -3 and fast particle losses from NBI were identified to contribute a significant part to the W sputtering. Discharges run in the upper, W coated divertor do not show higher W concentrations than comparable discharges in the lower C-based divertor. (author)

  9. Rotor-dynamic design aspects for a variable frequency drive based high speed cryogenic centrifugal pump in fusion devices

    International Nuclear Information System (INIS)

    Das, Jotirmoy; Vaghela, Hitensinh; Bhattacharya, Ritendra; Patel, Pratik; Shukla, Vinit; Shah, Nitin; Sarkar, Biswanath

    2015-01-01

    Superconducting magnets of large size are inevitable for fusion devices due to high magnetic field requirements. Forced flow cooling of the superconducting magnets with high mass flowrate of the order ∼3 kg/s is required to keep superconducting magnets within its safe operational boundaries during various plasma scenarios. This important requirement can be efficiently fulfilled by employing high capacity and high efficiency cryogenic centrifugal pumps. The efficiency > 70% will ensure overall lower heat load to the cryoplant. Thermo-hydraulic design of cryogenic centrifugal pump revealed that to achieve the operational regime with high efficiency, the speed should be ∼ 10,000 revolutions per minute. In this regard, the rotor-dynamic design aspect is quite critical from the operational stability point of view. The rotor shaft design of the cryogenic pump is primarily an outcome of optimization between thermal heat-in leak at cryogenic temperature level from ambient, cryogenic fluid impedance and designed rotation speed of the impeller wheel. The paper describes the basic design related to critical speed of the rotor shaft, rotor whirl and system instability prediction to explore the ideal operational range of the pump from the system stability point of view. In the rotor-dynamic analysis, the paper also describes the Campbell plots to ensure that the pump is not disturbed by any of the critical speeds, especially while operating near the nominal and enhanced operating modes. (author)

  10. The role and application of ion beam analysis for studies of plasma-facing components in controlled fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Rubel, Marek, E-mail: Marek.Rubel@ee.kth.se [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Fusion Plasma Physics, Royal Institute of Technology (KTH), 100 44 Stockholm (Sweden); Petersson, Per [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Fusion Plasma Physics, Royal Institute of Technology (KTH), 100 44 Stockholm (Sweden); Alves, Eduardo [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Instituto Superior Técnico, Universidade de Lisboa, 1049-001 Lisbon (Portugal); Brezinsek, Sebastijan [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Institut für Klima- und Energieforschung, Forschungszentrum Jülich, D-52425 Jülich (Germany); Coad, Joseph Paul [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Heinola, Kalle [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); University of Helsinki, 00014 Helsinki (Finland); Mayer, Matej [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Max-Planck-Institut für Plasmaphysik, 85478 Garching (Germany); Widdowson, Anna [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)

    2016-03-15

    First wall materials in controlled fusion devices undergo serious modification by several physical and chemical processes arising from plasma–wall interactions. Detailed information is required for the assessment of material lifetime and accumulation of hydrogen isotopes in wall materials. The intention of this work is to give a concise overview of key issues in the characterization of plasma-facing materials and components in tokamaks, especially in JET with an ITER-Like Wall. IBA techniques play a particularly prominent role here because of their isotope selectivity in the low-Z range (1–10), high sensitivity and combination of several methods in a single run. The role of {sup 3}He-based NRA, RBS (standard and micro-size beam) and HIERDA in fuel retention and material migration studies is presented. The use of tracer techniques with rare isotopes (e.g. {sup 15}N) or marker layers on wall diagnostic components is described. Special instrumentation, development of equipment to enhance research capabilities and issues in handling of contaminated materials are addressed.

  11. The role and application of ion beam analysis for studies of plasma-facing components in controlled fusion devices

    Science.gov (United States)

    Rubel, Marek; Petersson, Per; Alves, Eduardo; Brezinsek, Sebastijan; Coad, Joseph Paul; Heinola, Kalle; Mayer, Matej; Widdowson, Anna

    2016-03-01

    First wall materials in controlled fusion devices undergo serious modification by several physical and chemical processes arising from plasma-wall interactions. Detailed information is required for the assessment of material lifetime and accumulation of hydrogen isotopes in wall materials. The intention of this work is to give a concise overview of key issues in the characterization of plasma-facing materials and components in tokamaks, especially in JET with an ITER-Like Wall. IBA techniques play a particularly prominent role here because of their isotope selectivity in the low-Z range (1-10), high sensitivity and combination of several methods in a single run. The role of 3He-based NRA, RBS (standard and micro-size beam) and HIERDA in fuel retention and material migration studies is presented. The use of tracer techniques with rare isotopes (e.g. 15N) or marker layers on wall diagnostic components is described. Special instrumentation, development of equipment to enhance research capabilities and issues in handling of contaminated materials are addressed.

  12. Irradiation creep at temperatures of 400 degrees C and below for application to near-term fusion devices

    International Nuclear Information System (INIS)

    Grossbeck, M.L.; Gibson, L.T.; Mansur, L.K.

    1996-01-01

    To study irradiation creep at 400 degrees C and below, a series of six austenitic stainless steels and two ferritic alloys was irradiated sequentially in two research reactors where the neutron spectrum was tailored to produce a He production rate typical of a fusion device. Irradiation began in the Oak Ridge Research Reactor; and, after an atomic displacement level of 7.4 dpa, the specimens were moved to the High Flux Isotope Reactor for the remainder of the 19 dpa accumulated. Irradiation temperatures of 60, 200, 330, and 400 degrees C were studied with internally pressurized tubes of type 316 stainless steel, PCA, HT 9, and a series of four laboratory heats of: Fe-13.5Cr-15Ni, Fe-13.5Cr-35Ni, Fe-1 3.5Cr-1 W-0.18Ti, and Fe-16Cr. At 330 degrees C, irradiation creep was shown to be linear in fluence and stress. There was little or no effect of cold-work on creep under these conditions at all temperatures investigated. The HT9 demonstrated a large deviation from linearity at high stress levels, and a minimum in irradiation creep with increasing stress was observed in the Fe-Cr-Ni ternary alloys

  13. The life test of a DC circuit breaker of tokamak device JT-60 for a nuclear fusion research

    International Nuclear Information System (INIS)

    Shimada, Ryuichi; Tani, Keiji; Kishimoto, Hiroshi; Tamura, Sanae; Yanabu, Satoru.

    1979-01-01

    In the Tokamak devices for nuclear fusion research, the construction of the current transformer circuits having plasma as the secondary circuit and the change of the primary circuit current are necessary for generating current in the plasma. This is considered to be fairly difficult in practice if conventional methods using capacitor discharge and iron core coils are employed. Considering such circumstances, it was decided for JT-60 to use an air-core current transformer coil and to employ the method of storing energy in the form of current in the coil inductance instead of a capacitor. For this reason, a DC circuit breaker is required to interrupt coil current. The authors improved an AV vacuum breaker, which had been developed as the vacuum breaker of longitudinal magnetic field type applying a magnetic field in parallel with an arc, to get the one for DC circuit for the purpose of applying it to JT-60. In this paper, the operational characteristic of the DC breaker is described, the construction and function of the life test circuit is explained, and the test results are reported. Finally, interruptions of 10,000 times at 20 kA were carried out. It is successful that the restrike of arc occurring during tens of milli-seconds after interruptions was improved to 0.05% or less for 10,000 times operations. Further, it was found that the generation of arc restrike can be reduced practically to zero with two breakers in series. (Wakatsuki, Y.)

  14. High-speed repetitive pellet injector for plasma fueling of magnetic confinement fusion devices

    International Nuclear Information System (INIS)

    Combs, S.K.; Baylor, L.R.; Foust, C.R.

    1993-01-01

    The projected fueling requirements of future magnetic confinement devices for controlled thermonuclear research [e.g., the International Thermonuclear Experimental Reactor (ITER)] indicate that a flexible plasma fueling capability is required. This includes a mix of traditional gas puffing and low- and high-velocity deuterium-tritium pellets. Conventional pellet injectors (based on light gas guns or centrifugal accelerators) can reliably provide frozen hydrogen pellets (1- to 6-mm-diam sizes tested) up to ∼1.3-km/s velocity at the appropriate pellet fueling rates (1 to 10 Hz or greater). For long-pulse operation in a higher velocity regime (>2 km/s), an experiment in collaboration between Oak Ridge National Laboratory (ORNL) and ENEA Frascati is under way. This activity will be carried out in the framework of a collaborative agreement between the US Department of Energy and European Atomic Energy Community -- ENEA Association. In this experiment, an existing ORNL hydrogen extruder (equipped with a pellet chambering mechanism/gun barrel assembly) and a Frascati two-stage light gas gun driver have been combined on a test facility at ORNL. Initial testing has been carried out with single deuterium pellets accelerated up to 2.05 km/s with the two-stage driver; in addition, some preliminary repetitive testing (to commission the diagnostics) was performed at reduced speeds, including sequences at 0.5 to 1 Hz and 10 to 30 pellets. The primary objective of this study is to demonstrate repetitive operation (up to ∼1 Hz) with speeds in the 2- to 3-km/s range. In addition, the strength of extruded hydrogen ice as opposed to that produced in situ by direct condensation in pipe guns can be investigated. The equipment and initial experimental results are described

  15. Collection of Summaries of reports on result of research at basic experiment device for nuclear fusion reactor blanket design, 1994

    International Nuclear Information System (INIS)

    1995-07-01

    The development of nuclear fusion reactors reached such stage that the generation of fusion power output comparable with the input power into core plasma is possible. At present, the engineering design of the international thermonuclear fusion experimental reactor, ITER, is advanced by the cooperation of Japan, USA, Europe and Russia, aiming at the start of operation at the beginning of 21st century. This meeting for reporting the results has been held every year, and this time, it was held on May 19, 1995 at University of Tokyo with the theme ''The interface properties of fusion reactor materials and the control of particle transport''. About 50 participants from academic, governmental and industrial circles discussed actively on the theme. Three lectures on the topics of fusion reactor engineering and materials and seven lectures on the basic experiment of fusion reactor blanket design related to the next period project were given at the meeting. (K.I.)

  16. Posterior lumbar interbody fusion using nonresorbable poly-ether-ether-ketone versus resorbable poly-L-lactide-co-D,L-lactide fusion devices: a prospective, randomized study to assess fusion and clinical outcome

    NARCIS (Netherlands)

    Jiya, T.U.; Smit, T.H.; Deddens, J.; Mullender, M.G.

    2009-01-01

    STUDY DESIGN: A prospective randomized clinical study. OBJECTIVE.: To assess fusion, clinical outcome, and complications. SUMMARY OF BACKGROUND DATA: Resorbable poly-L- lactide-co-D,L-lactide (PLDLLA) cages intended to aid spinal interbody fusion have been introduced into clinical practice within

  17. Inertial confinement fusion with direct electric generation by magnetic flux comparession

    International Nuclear Information System (INIS)

    Lasche, G.P.

    1983-01-01

    A high-power-density laser-fusion-reactor concept in investigated in which directed kinetic enery imparted to a large mass of liquid lithium--in which the fusion target is centrally located--is maximized. In turn, this kinetic energy is converted directly to electricity with, potentially, very high efficiency by work done against a pulsed magnetic field applied exterior to the lithium. Because the concept maximizes the blanket thickness per unit volume of lithium, neutron-induced radioactivities in the reaction chamber wall can be many orders of magnitude less than is typical of D-T fusion reactor concepts

  18. First wall and blanket stresses induced by cyclic fusion core operations

    International Nuclear Information System (INIS)

    Bohachevsky, I.O.; Kostoff, R.N.

    1981-01-01

    An analysis is made of cyclic thermal loads and stresses for the complete range of operating conditions. Two critical components were examined; the solid wall adjacent to the fusion plasma (first wall) and the fuel elements in the high power density region of the blanket. Simple closed form expressions were derived for temperature increases and thermal stresses that may be evaluated conveniently and rapidly and the values compared for different systems

  19. Chamber and Wall Response to Target Implosion in Inertial and Z-Pinch Fusion and Lithography Devices

    International Nuclear Information System (INIS)

    Hassanein, A.; Konkashbaev, I.; Morozov, V.; Sizyuk, V.

    2006-01-01

    The chamber walls, both solid and liquid, in inertial fusion energy (IFE) and Z-pinch reactors and Lithography devices are exposed to harsh conditions following each target implosion or pinching of plasma. Key issues of the cyclic IFE operation include intense photon and ion deposition, wall thermal and hydrodynamic evolution, wall erosion and fatigue lifetime, and chamber clearing and evacuation to ensure desirable conditions prior to target implosion. Detailed models have been developed for reflected laser light, emitted photons, neutrons, and target debris deposition and interaction with chamber components and have been implemented in the comprehensive HEIGHTS software package. The hydrodynamic response of chamber walls in bare or in gas-filled cavities and the photon transport of the deposited energy has been calculated by means of new and advanced numerical techniques for accurate shock treatment and propagation. These models include detail media hydrodynamics, non-LTE multi-group for both continuum and line radiation transport, and dynamics of eroded debris resulting from the intense energy deposition. The focus of this study is to critically assess the reliability and the dynamic response of chamber walls in various proposed protection methods for IFE systems. Key requirements are that: (i) the chamber wall accommodates the cyclic energy deposition while providing the required lifetime due to various erosion mechanisms, such as vaporization, chemical and physical sputtering, melt/liquid splashing and explosive erosion, and fragmentation of liquid walls, and (ii) after each shot the chamber is cleared and returned to a quiescent state in preparation for the target injection and the firing of the driver for the subsequent shot. This paper investigates in details these two important issues and found that the required operating frequency of the IFE reactors for power production may be severely limited due to these two requirements. (author)

  20. Performance test of dual modulator polarimeters in two different configurations for magneto-optic measurement of fusion devices

    International Nuclear Information System (INIS)

    Kenji Higuchi; Tsuyoshi Akiyama; Yoshifumi Azuma; Shunji Tsuji-Iio; Hiroaki Tsutsui; Ryuichi Shimada

    2006-01-01

    Accurate measurement of the magnetic field around plasma is indispensable for real-time control and data analysis on magnetic fusion devices such as tokamaks. Instead of commonly used pick-up loops, which have the problems of zero-point drifts, we proposed and tested a magneto-optic polarimeter based on the polarization modulation method using two photoelastic modulators (PEMs). Polarization detection using a pair of PEMs has been applied to the motional Stark effect (MSE) measurements in some tokamaks. The CO 2 laser polarimeter for electron density measurement on JT-60U adopted this method and demonstrated long time stability for several hours. However, this method requires the same number of pairs of PEMs, which are delicate and expensive, as that of channels so that this method is not easy to apply to multi-point measurements of magnetic fields around tokamaks. To cope with this problem, the two PEMs, which are conventionally placed behind each magnetic sensor, are used to modulate the incident beam before split for each magneto-optic sensor. This configuration can reduce the number of PEMs drastically and the optical system becomes simple. In this new optical configuration, the polarization angle resolution comparable to the conventional optical configuration of 0.002 o with response time of 10 ms was achieved at an incident polarization angle of about 0 o while that at 21 o was 0.07 o . The resolution of 0.07 o corresponds to 7 gauss when a 40-mm-long ZnSe sensing rod is used. Performance test between the two optical configurations were also made on the long-time stability and the accuracy with increasing numbers of beam splitters and/or mirrors for multi-point measurements. (author)

  1. Image fusion analysis of 99mTc-HYNIC-Tyr3-octreotide SPECT and diagnostic CT using an immobilisation device with external markers in patients with endocrine tumours

    International Nuclear Information System (INIS)

    Gabriel, Michael; Hausler, Florian; Moncayo, Roy; Decristoforo, Clemens; Virgolini, Irene; Bale, Reto; Kovacs, Peter

    2005-01-01

    The aim of this study was to assess the value of multimodality imaging using a novel repositioning device with external markers for fusion of single-photon emission computed tomography (SPECT) and computed tomography (CT) images. The additional benefit derived from this methodological approach was analysed in comparison with SPECT and diagnostic CT alone in terms of detection rate, reliability and anatomical assignment of abnormal findings with SPECT. Fifty-three patients (30 males, 23 females) with known or suspected endocrine tumours were studied. Clinical indications for somatostatin receptor (SSTR) scintigraphy (SPECT/CT image fusion) included staging of newly diagnosed tumours (n=14) and detection of unknown primary tumour in the presence of clinical and/or biochemical suspicion of neuroendocrine malignancy (n=20). Follow-up studies after therapy were performed in 19 patients. A mean activity of 400 MBq of 99m Tc-EDDA/HYNIC-Tyr 3 -octreotide was given intravenously. SPECT using a dual-detector scintillation camera and diagnostic multi-detector CT were sequentially performed. To ensure reproducible positioning, patients were fixed in an individualised vacuum mattress with modality-specific external markers for co-registration. SPECT and CT data were initially interpreted separately and the fused images were interpreted jointly in consensus by nuclear medicine and diagnostic radiology physicians. SPECT was true-positive (TP) in 18 patients, true-negative (TN) in 16, false-negative (FN) in ten and false-positive (FP) in nine; CT was TP in 18 patients, TN in 21, FP in ten and FN in four. With image fusion (SPECT and CT), the scan result was TP in 27 patients (50.9%), TN in 25 patients (47.2%) and FN in one patient, this FN result being caused by multiple small liver metastases; sensitivity was 95% and specificity, 100%. The difference between SPECT and SPECT/CT was statistically as significant as the difference between CT and SPECT/CT image fusion (P<0

  2. Fusion reactors - types - problems

    International Nuclear Information System (INIS)

    Schmitter, K.H.

    1979-07-01

    A short account is given of the principles of fusion reactions and of the expected advantages of fusion reactors. Descriptions are presented of various Tokamak experimental devices being developed in a number of countries and of some mirror machines. The technical obstacles to be overcome before a fusion reactor could be self-supporting are discussed. (U.K.)

  3. Surface modification study of zirconium on exposure to fusion grade plasma in an 11.5 kJ plasma focus device

    International Nuclear Information System (INIS)

    Srivastava, Rohit; Niranjan, Ram; Rout, R.K.; Kaushik, T.C.; Chakravarthy, Y.; Mishra, P.

    2017-01-01

    In continuation of our investigation on effect of fusion grade plasma produced in an existing MEPF-12 (11.5 kJ, 40 μF, 24 kV) plasma focus (PF) facility on different materials, likely to be used in future fusion reactors, we have reported here the study on Zirconium (Zr) metal. In the present work, the Zr sample in disc (2 mm thick, 10 mm diameter) form was exposed to twenty shots of plasma focus operated at 4 mbar deuterium gas filling pressure and 11.5 kJ bank energy. The samples were placed at a distance of 6 cm from the tip of the anode in the MEPF-12 PF device. The emissions from the device comprise of deuterium ions in wide energy range (a few keV to several hundreds of keV), high temperature plasma (in general a few keV) and neutrons of 2.45 MeV energy produced due to D(D, 3 He)n fusion reactions

  4. Collection of summaries of reports on result of research at basic experiment device for nuclear fusion reactor blanket design, 1995

    International Nuclear Information System (INIS)

    1996-07-01

    This report meeting was held on May 22, 1995 at University of Tokyo by about 40 participants. As the topics on the fusion reactor engineering research in Japan, lectures were given on the present state and future of nuclear fusion networks and on the strong magnetic field tokamak using electromagnetic force-balanced coils being planned. Thereafter, the reports of the results of the researches which were carried out by using this experimental facility were made, centering around the subject related to the future conception 'The interface properties of fusion reactor materials and particle transport control'. The publication was made on the future conception of the basic experiment setup for fusion reactor blanket design, the application of high temperature superconductors to the advancement of nuclear fusion reactors, the modeling of the dynamic irradiation behavior of fusion reactor materials, the interface particle behavior in plasma-wall interaction, the behavior of tritium on the surface of breeding materials, and breeding materials and the behavior of tritium in plasma-wall interaction. (K.I.)

  5. Rational decision making in a wide scenario of different minimally invasive lumbar interbody fusion approaches and devices.

    Science.gov (United States)

    Pimenta, Luiz; Tohmeh, Antoine; Jones, David; Amaral, Rodrigo; Marchi, Luis; Oliveira, Leonardo; Pittman, Bruce C; Bae, Hyun

    2018-03-01

    With the proliferation of a variety of modern MIS spine surgery procedures, it is mandatory that the surgeon dominate all aspects involved in surgical indication. The information related to the decision making in patient selection for specific procedures is mandatory for surgical success. The objective of this study is to present decision-making criteria in minimally invasive surgery (MIS) selection for a variety of patients and pathologies. In this article, practicing surgeons who specialize in various MIS approaches for spinal fusion were engaged to provide expert opinion and literature review on decision making criteria for several MIS procedures. Pros, cons, relative limitations, and case examples are provided for patient selection in treatment with MIS posterolateral fusion (MIS-PLF), mini anterior lumbar interbody fusion (mini-ALIF), lateral interbody fusion (LLIF), MIS posterior lumbar interbody fusion (MIS-PLIF) and MIS transforaminal lumbar interbody fusion (MIS-TLIF). There is a variety of aspects to consider when deciding which modern MIS surgical approach is most appropriate to use based on patient and pathologic characteristics. The surgeon must adapt them to the characteristic of each type of patients, helping them to choose the most effective and efficient therapeutic option for each case.

  6. Proceedings of US/Japan Workshop (97FT5-06) on High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices

    International Nuclear Information System (INIS)

    Nygren, Richard; Kureczko, Diana

    1998-10-01

    The 1997 US-Japan Workshop on High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices was held at the Warwick Regis Hotel in San Francisco, California, on December 8-11, 1997. There were 53 presentations as well as discussions on technical issues and on planning for future collaborations, and 35 researchers from japan and the US participated in the workshop. Over the last few years, with the strong emphasis in the US on technology for ITER, there has been less work done in the US fusion program on basic plasma materials interaction and this change in emphasis workshops. The program this year emphasized activities that were not carried out under the ITER program and a new element this year in the US program was planning and some analysis on liquid surface concepts for advanced plasma facing components. The program included a ceremony to honor Professor Yamashina, who was retiring this year and a special presentation on his career

  7. Effects of fusion relevant transient energetic radiation, plasma and thermal load on PLANSEE double forged tungsten samples in a low-energy plasma focus device

    Science.gov (United States)

    Javadi, S.; Ouyang, B.; Zhang, Z.; Ghoranneviss, M.; Salar Elahi, A.; Rawat, R. S.

    2018-06-01

    Tungsten is the leading candidate for plasma facing component (PFC) material for thermonuclear fusion reactors and various efforts are ongoing to evaluate its performance or response to intense fusion relevant radiation, plasma and thermal loads. This paper investigates the effects of hot dense decaying pinch plasma, highly energetic deuterium ions and fusion neutrons generated in a low-energy (3.0 kJ) plasma focus device on the structure, morphology and hardness of the PLANSEE double forged tungsten (W) samples surfaces. The tungsten samples were provided by Forschungszentrum Juelich (FZJ), Germany via International Atomic Energy Agency, Vienna, Austria. Tungsten samples were irradiated using different number of plasma focus (PF) shots (1, 5 and 10) at a fixed axial distance of 5 cm from the anode top and also at various distances from the top of the anode (5, 7, 9 and 11 cm) using fixed number (5) of plasma focus shots. The virgin tungsten sample had bcc structure (α-W phase). After PF irradiation, the XRD analysis showed (i) the presence of low intensity new diffraction peak corresponding to β-W phase at (211) crystalline plane indicating the partial structural phase transition in some of the samples, (ii) partial amorphization, and (iii) vacancy defects formation and compressive stress in irradiated tungsten samples. Field emission scanning electron microscopy showed the distinctive changes to non-uniform surface with nanometer sized particles and particle agglomerates along with large surface cracks at higher number of irradiation shots. X-ray photoelectron spectroscopy analysis demonstrated the reduction in relative tungsten oxide content and the increase in metallic tungsten after irradiation. Hardness of irradiated samples initially increased for one shot exposure due to reduction in tungsten oxide phase, but then decreased with increasing number of shots due to increasing concentration of defects. It is demonstrated that the plasma focus device provides

  8. Fusion facility siting considerations

    International Nuclear Information System (INIS)

    Bussell, G.T.

    1985-01-01

    Inherent in the fusion program's transition from hydrogen devices to commercial power machines is a general increase in the size and scope of succeeding projects. This growth will lead to increased emphasis on safety, environmental impact, and the external effects of fusion in general, and of each new device in particular. A critically important consideration in this regard is site selection. The purpose of this paper is to examine major siting issues that may affect the economics, safety, and environmental impact of fusion

  9. Graphite for fusion energy applications

    International Nuclear Information System (INIS)

    Eatherly, W.P.; Clausing, R.E.; Strehlow, R.A.; Kennedy, C.R.; Mioduszewski, P.K.

    1987-03-01

    Graphite is in widespread and beneficial use in present fusion energy devices. This report reflects the view of graphite materials scientists on using graphite in fusion devices. Graphite properties are discussed with emphasis on application to fusion reactors. This report is intended to be introductory and descriptive and is not intended to serve as a definitive information source

  10. Fusion technology 1992

    International Nuclear Information System (INIS)

    Ferro, C.; Gasparatto, M.; Knoepfel, H.

    1993-01-01

    The aim of the biennial series of symposia on the title subject, organized by the European Fusion Laboratories, is the exchange of information on the design, construction and operation of fusion experiments and on the technology being developed for the next step devices and fusion reactors. The coverage of the volume includes the technological aspects of fusion reactors in relation to new developments, this forming a guideline for the definition of future work. These proceedings comprise three volumes and contain both the invited lectures and contributed papers presented at the symposium which was attended by 569 participants from around the globe. The 343 papers, including 12 invited papers, characterize the increasing interest of industry in the fusion programme, giving a broad and current overview on the progress and trends fusion technology is experiencing now, as well as indicating the future for fusion devices

  11. Engineering design of a toroidal divertor for the EBT-S fusion device. Final report, Phase II. EBT-S divertor project

    International Nuclear Information System (INIS)

    Mai, L.P.; Malick, F.S.

    1981-01-01

    The mechanical, structural, thermal, electrical, and vacuum design of a magnetic toroidal divertor system for the Elmo Bumpy Torus (EBT-S) is presented. The EBT-S is a toroidal magnetic fusion device located at the ORNL that operates under steady state conditions. The engineering of the divertor was performed during the second of three phases of a program aimed at the selection, design, fabrication, and installation of a magnetic divertor for EBT-S. The magnetic analysis of the toroidal divertor was performed during Phase I of the program and has been reported in a separate document. In addition to the details of the divertor design, the modest modifications that are required to the EBT-S device and facility to accommodate the divertor system are presented

  12. Canada's Fusion Program

    International Nuclear Information System (INIS)

    Jackson, D. P.

    1990-01-01

    Canada's fusion strategy is based on developing specialized technologies in well-defined areas and supplying these technologies to international fusion projects. Two areas are specially emphasized in Canada: engineered fusion system technologies, and specific magnetic confinement and materials studies. The Canadian Fusion Fuels Technology Project focuses on the first of these areas. It tritium and fusion reactor fuel systems, remote maintenance and related safety studies. In the second area, the Centre Canadian de fusion magnetique operates the Tokamak de Varennes, the main magnetic fusion device in Canada. Both projects are partnerships linking the Government of Canada, represented by Atomic Energy of Canada Limited, and provincial governments, electrical utilities, universities and industry. Canada's program has extensive international links, through which it collaborates with the major world fusion programs, including participation in the International Thermonuclear Experimental Reactor project

  13. Low prepulse, high power density water dielectric switching

    International Nuclear Information System (INIS)

    Johnson, D.L.; VanDevender, J.P.; Martin, T.H.

    1979-01-01

    Prepulse voltage suppression has proven difficult in high power, high voltage accelerators employing self-breakdown water dielectric switches. A novel and cost effective water switch has been developed at Sandia Laboratories which reduces prepulse voltage by reducing the capacity across the switch. This prepulse suppression switch causes energy formerly stored in the switch capacity and dissipated in the arc to be useful output energy. The switching technique also allows the pulse forming lines to be stacked in parallel and electrically isolated from the load after the line has been discharged. The switch consists of a ground plane, with several holes, inserted between the switch electrodes. The output line switch electrodes extend through the holes and face electrodes on the pulse forming line (PFL). The capacity between the PFL and the output transmission line is reduced by about 80%. The gap spacing between the output line electrode and the hole in the ground plane is adjusted so that breakdown occurs after the main pulse and provides a crow bar between the load and the source. Performance data from the Proto II, Mite and Ripple test facilities are presented

  14. High power density capacitor and method of fabrication

    Science.gov (United States)

    Tuncer, Enis

    2012-11-20

    A ductile preform for making a drawn capacitor includes a plurality of electrically insulating, ductile insulator plates and a plurality of electrically conductive, ductile capacitor plates. Each insulator plate is stacked vertically on a respective capacitor plate and each capacitor plate is stacked on a corresponding insulator plate in alignment with only one edge so that other edges are not in alignment and so that each insulator plate extends beyond the other edges. One or more electrically insulating, ductile spacers are disposed in horizontal alignment with each capacitor plate along the other edges and the pattern is repeated so that alternating capacitor plates are stacked on alternating opposite edges of the insulator plates. A final insulator plate is positioned at an extremity of the preform. The preform may then be drawn to fuse the components and decrease the dimensions of the preform that are perpendicular to the direction of the draw.

  15. High power density superconducting motor for control applications

    International Nuclear Information System (INIS)

    Lopez, J; Granados, X; Lloberas, J; Torres, R; Grau, J; Maynou, R; Bosch, R

    2008-01-01

    A high dynamics superconducting low power motor for control applications has been considered for design. The rotor is cylindrical with machined bulks that generate the field by trapping flux in a four poles configuration. The toothless iron armature is wound by copper, acting iron only as magnetic screen. Details of the magnetic assembling, cryogenics and electrical supply conditioning will be reported. Improvements due to the use of a superconducting set are compared with performances of equivalent conventional motors

  16. High Power Density, Lightweight Thermoelectric Metamaterials for Energy Harvesting

    Data.gov (United States)

    National Aeronautics and Space Administration — Thermoelectric energy harvesting utilizes materials that generate an electrical current when subjected to a temperature gradient, or simply, a hot and cold source of...

  17. ICRF array module development and optimization for high power density

    International Nuclear Information System (INIS)

    Ryan, P.M.; Swain, D.W.

    1997-02-01

    This report describes the analysis and optimization of the proposed International Thermonuclear Experimental Reactor (ITER) Antenna Array for the ion cyclotron range of frequencies (ICRF). The objectives of this effort were to: (1) minimize the applied radiofrequency rf voltages occurring in vacuum by proper layout and shape of components, limit the component's surface/volumes where the rf voltage is high; (2) study the effects of magnetic insulation, as applied to the current design; (3) provide electrical characteristics of the antenna for the development and analysis of tuning, arc detection/suppression, and systems for discriminating between arcs and edge-localized modes (ELMs); (4) maintain close interface with mechanical design

  18. High power density reactors based on direct cooled particle beds

    Science.gov (United States)

    Powell, J. R.; Horn, F. L.

    Reactors based on direct cooled High Temperature Gas Cooled Reactor (HTGR) type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out along the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBRs) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed.

  19. High power density reactors based on direct cooled particle beds

    International Nuclear Information System (INIS)

    Powell, J.R.; Horn, F.L.

    1985-01-01

    Reactors based on direct cooled HTGR type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out long the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBR's) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed. 12 figs

  20. Proceedings of 1999 U.S./Japan Workshop (99FT-05) On High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices

    Energy Technology Data Exchange (ETDEWEB)

    NYGREN,RICHARD E.; STAVROS,DIANA T.

    2000-06-01

    The 1999 US-Japan Workshop on High Heat Flux Components and Plasma Surface Interactions in Next Step Fusion Devices was held at the St. Francis Hotel in Santa Fe, New Mexico, on November 1-4, 1999. There were 42 presentations as well as discussion on technical issues and planning for future collaborations. The participants included 22 researchers from Japan and the United States as well as seven researchers from Europe and Russia. There have been important changes in the programs in both the US and Japan in the areas of plasma surface interactions and plasma facing components. The US has moved away from a strong focus on the ITER Project and has introduced new programs on use of liquid surfaces for plasma facing components, and operation of NSTX has begun. In Japan, the Large Helical Device began operation. This is the first large world-class confinement device operating in a magnetic configuration different than a tokamak. In selecting the presentations for this workshop, the organizers sought a balance between research in laboratory facilities or confinement devices related to plasma surface interactions and experimental research in the development of plasma facing components. In discussions about the workshop itself, the participants affirmed their preference for a setting where ''work-in-progress'' could be informally presented and discussed.

  1. Proceedings of 1999 U.S./Japan Workshop (99FT-05) On High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices

    International Nuclear Information System (INIS)

    NYGREN, RICHARD E.; STAVROS, DIANA T.

    2000-01-01

    The 1999 US-Japan Workshop on High Heat Flux Components and Plasma Surface Interactions in Next Step Fusion Devices was held at the St. Francis Hotel in Santa Fe, New Mexico, on November 1-4, 1999. There were 42 presentations as well as discussion on technical issues and planning for future collaborations. The participants included 22 researchers from Japan and the United States as well as seven researchers from Europe and Russia. There have been important changes in the programs in both the US and Japan in the areas of plasma surface interactions and plasma facing components. The US has moved away from a strong focus on the ITER Project and has introduced new programs on use of liquid surfaces for plasma facing components, and operation of NSTX has begun. In Japan, the Large Helical Device began operation. This is the first large world-class confinement device operating in a magnetic configuration different than a tokamak. In selecting the presentations for this workshop, the organizers sought a balance between research in laboratory facilities or confinement devices related to plasma surface interactions and experimental research in the development of plasma facing components. In discussions about the workshop itself, the participants affirmed their preference for a setting where ''work-in-progress'' could be informally presented and discussed

  2. The TITAN Reversed-Field Pinch fusion reactor study

    International Nuclear Information System (INIS)

    1988-03-01

    The TITAN Reversed-Field Pinch (RFP) fusion reactor study is a multi-institutional research effort to determine the technical feasibility and key developmental issues of an RFP fusion reactor, especially at high power density, and to determine the potential economics, operations, safety, and environmental features of high-mass-power-density fusion systems. The TITAN conceptual designs are DT burning, 1000 MWe power reactors based on the RFP confinement concept. The designs are compact, have a high neutron wall loading of 18 MW/m 2 and a mass power density of 700 kWe/tonne. The inherent characteristics of the RFP confinement concept make fusion reactors with such a high mass power density possible. Two different detailed designs have emerged: the TITAN-I lithium-vanadium design, incorporating the integrated-blanket-coil concept; and the TITAN-II aqueous loop-in-pool design with ferritic steel structure. This report contains a collection of 16 papers on the results of the TITAN study which were presented at the International Symposium on Fusion Nuclear Technology. This collection describes the TITAN research effort, and specifically the TITAN-I and TITAN-II designs, summarizing the major results, the key technical issues, and the central conclusions and recommendations. Overall, the basic conclusions are that high-mass power-density fusion reactors appear to be technically feasible even with neutron wall loadings up to 20 MW/m 2 ; that single-piece maintenance of the FPC is possible and advantageous; that the economics of the reactor is enhanced by its compactness; and the safety and environmental features need not to be sacrificed in high-power-density designs. The fact that two design approaches have emerged, and others may also be possible, in some sense indicates the robustness of the general findings

  3. Characteristics of a high-power RF source of negative hydrogen ions for neutral beam injection into controlled fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Abdrashitov, G. F.; Belchenko, Yu. I.; Gusev, I. A.; Ivanov, A. A.; Kondakov, A. A.; Sanin, A. L.; Sotnikov, O. Z., E-mail: O.Z.Sotnikov@inp.nsk.su; Shikhovtsev, I. V. [Russian Academy of Sciences, Budker Institute of Nuclear Physics, Siberian Branch (Russian Federation)

    2017-01-15

    An injector of hydrogen atoms with an energy of 0.5–1 MeV and equivalent current of up to 1.5 A for purposes of controlled fusion research is currently under design at the Budker Institute of Nuclear Physics, Siberian Branch, Russian Academy of Sciences. Within this project, a multiple-aperture RF surface-plasma source of negative hydrogen ions is designed. The source design and results of experiments on the generation of a negative ion beam with a current of >1 A in the long-pulse mode are presented.

  4. Application of structural mechanics methods to the design of large tandem mirror fusion devices (MFTF-B)

    International Nuclear Information System (INIS)

    Karpenko, V.N.; Ng, D.S.

    1985-01-01

    The Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory requires state-of-the-art structural-mechanics methods to deal with access constraints for plasma heating and diagnostics, alignment requirements, and load complexity and variety. Large interactive structures required an integrated analytical approach to achieve a resonable level of overall system optimization. The Tandem Magnet Generator (TMG) creates a magnet configuration for the EFFI calculation of electromagnetic-field forces that, coupled with other loads, form the input loading to magnetic and vessel finite-element models. The anlytical results provide the data base for detailed design of magnet, vessel, foundation, and interaction effects. (orig.)

  5. Selected topics on surface effects in fusion devices: neutral-beam injectors and beam-direct converters

    International Nuclear Information System (INIS)

    Kaminsky, M.

    1978-01-01

    Neutral-beam injectors are being used for the heating and fueling of plasmas in existing devices such as PLT (Princeton), ISX (Oak Ridge) and 2XIIB (Lawrence Livermore Laboratory) and will be used in devices such as TFTR (Princeton), MX (Livermore) and Doublet III (Gulf Atomic). For example, TFTR has been designed to receive a total of 20 MW of 120-keV deuterium atoms in pulses of 0.5-sec duration from 12 neutral beam injectors; for the MX experiment it is planned to inject a total of 750A (equivalent) of deuterium atoms with a mean energy of 56 keV in 0.5-sec pulses. The interaction of energetic deuterium atoms with exposed surfaces of device components such as beam dumps, beam-direct-convertors collectors, beam calorimeters, and armor plates, cause a variety of surface effects which affect deleteriously the operation of such devices. Some of the major effects will be discussed

  6. A Novel Remote Rehabilitation System with the Fusion of Noninvasive Wearable Device and Motion Sensing for Pulmonary Patients.

    Science.gov (United States)

    Tey, Chuang-Kit; An, Jinyoung; Chung, Wan-Young

    2017-01-01

    Chronic obstructive pulmonary disease is a type of lung disease caused by chronically poor airflow that makes breathing difficult. As a chronic illness, it typically worsens over time. Therefore, pulmonary rehabilitation exercises and patient management for extensive periods of time are required. This paper presents a remote rehabilitation system for a multimodal sensors-based application for patients who have chronic breathing difficulties. The process involves the fusion of sensory data-captured motion data by stereo-camera and photoplethysmogram signal by a wearable PPG sensor-that are the input variables of a detection and evaluation framework. In addition, we incorporated a set of rehabilitation exercises specific for pulmonary patients into the system by fusing sensory data. Simultaneously, the system also features medical functions that accommodate the needs of medical professionals and those which ease the use of the application for patients, including exercises for tracking progress, patient performance, exercise assignments, and exercise guidance. Finally, the results indicate the accurate determination of pulmonary exercises from the fusion of sensory data. This remote rehabilitation system provides a comfortable and cost-effective option in the healthcare rehabilitation system.

  7. Ultrafine tungsten as a plasma-facing component in fusion devices: effect of high flux, high fluence low energy helium irradiation

    International Nuclear Information System (INIS)

    El-Atwani, O.; Gonderman, Sean; Allain, J.P.; Efe, Mert; Klenosky, Daniel; Qiu, Tian; De Temmerman, Gregory; Morgan, Thomas; Bystrov, Kirill

    2014-01-01

    This work discusses the response of ultrafine-grained tungsten materials to high-flux, high-fluence, low energy pure He irradiation. Ultrafine-grained tungsten samples were exposed in the Pilot-PSI (Westerhout et al 2007 Phys. Scr. T128 18) linear plasma device at the Dutch Institute for Fundamental Energy Research (DIFFER) in Nieuwegein, the Netherlands. The He flux on the tungsten samples ranged from 1.0 × 10 23 –2.0 × 10 24  ions m −2  s −1 , the sample bias ranged from a negative (20–65) V, and the sample temperatures ranged from 600–1500 °C. SEM analysis of the exposed samples clearly shows that ultrafine-grained tungsten materials have a greater fluence threshold to the formation of fuzz by an order or magnitude or more, supporting the conjecture that grain boundaries play a major role in the mechanisms of radiation damage. Pre-fuzz damage analysis is addressed, as in the role of grain orientation on structure formation. Grains of (1 1 0) and (1 1 1) orientation showed only pore formation, while (0 0 1) oriented grains showed ripples (higher structures) decorated with pores. Blistering at the grain boundaries is also observed in this case. In situ TEM analysis during irradiation revealed facetted bubble formation at the grain boundaries likely responsible for blistering at this location. The results could have significant implications for future plasma-burning fusion devices given the He-induced damage could lead to macroscopic dust emission into the fusion plasma. (paper)

  8. Finite element and node point generation computer programs used for the design of toroidal field coils in tokamak fusion devices

    International Nuclear Information System (INIS)

    Smith, R.A.

    1975-06-01

    The structural analysis of toroidal field coils in Tokamak fusion machines can be performed with the finite element method. This technique has been employed for design evaluations of toroidal field coils on the Princeton Large Torus (PLT), the Poloidal Diverter Experiment (PDX), and the Tokamak Fusion Test Reactor (TFTR). The application of the finite element method can be simplified with computer programs that are used to generate the input data for the finite element code. There are three areas of data input where significant automation can be provided by supplementary computer codes. These concern the definition of geometry by a node point mesh, the definition of the finite elements from the geometric node points, and the definition of the node point force/displacement boundary conditions. The node point forces in a model of a toroidal field coil are computed from the vector cross product of the coil current and the magnetic field. The computer programs named PDXNODE and ELEMENT are described. The program PDXNODE generates the geometric node points of a finite element model for a toroidal field coil. The program ELEMENT defines the finite elements of the model from the node points and from material property considerations. The program descriptions include input requirements, the output, the program logic, the methods of generating complex geometries with multiple runs, computational time and computer compatibility. The output format of PDXNODE and ELEMENT make them compatible with PDXFORC and two general purpose finite element computer codes: (ANSYS) the Engineering Analysis System written by the Swanson Analysis Systems, Inc., and (WECAN) the Westinghouse Electric Computer Analysis general purpose finite element program. The Fortran listings of PDXNODE and ELEMENT are provided

  9. The role of bone SPECT/CT in the evaluation of lumbar spinal fusion with metallic fixation devices

    DEFF Research Database (Denmark)

    Damgaard, Morten; Nimb, Lars; Madsen, Jan L

    2010-01-01

    PURPOSE: It is difficult to evaluate the stability of the lumbar spondylodesis with metallic fixation devices by conventional imaging methods such as radiography or magnetic resonance imaging. It is unknown whether single photon emission computed tomography/computed tomography (SPECT/CT) may be u...

  10. Flexible GaN for High Performance, Strainable Radio Frequency Devices (Postprint)

    Science.gov (United States)

    2017-11-02

    wireless systems where consumers will benefit significantly from the high power densities achievable in GaN devices.[8] Further complicating the...future strainable and conformal devices for transmission of radio-frequency (RF) signals over large distances for more efficient wireless communication... power density of traditional RF amplifier materials at different frequencies and wireless generation bands, as well as an image of the flexible GaN

  11. Fundamental processes of plasma and reactive gas surface treatment for the recovery of hydrogen isotopes from carbon co-deposits in fusion devices

    International Nuclear Information System (INIS)

    Moeller, Soeren

    2014-01-01

    The use of carbon-based plasma-facing wall components offers many advantages for plasma operation in magnetic confinement nuclear fusion devices. However, through reactions with the hydrogen based fusion plasma, carbon forms amorphous hydrogenated carbon co-deposits (a-C:H) in the vacuum vessels. If tritium is used to fuel the reactor, this co-deposition can quickly lead to an inacceptable high tritium inventory. Through co-deposition with carbon about 10% of the tritium injected into the reactor can be trapped. Even with other wall materials co-deposition can be significant. A method to recover the hydrogen isotopes from the co-deposits is necessary. The method has to be compatible with the requirements of the devices and nuclear fusion plasma operation. In this work thermo-chemical removal by neutral gases (TCR) and removal by plasmas is investigated. Models are developed to describe the involved processes of both removal methods. TCR is described using a reaction-diffusion model. Within this model the reactive gas diffuses into the co-deposits and subsequently reacts in a thermally activated process. The co-deposits are pyrolysed, forming volatile gases, e.g. CO 2 and H 2 O. These gases are pumped from the vacuum vessel and recycled. Applying the model to literature observations enables to connect data on exposure temperature, pressure, time and co-deposit properties. Two limits of TCR (reaction- or diffusion-limited) are identified. Plasma removal sputters co-deposits by their chemical and physical interaction with the impinging ions. The description uses a 0D plasma model from the literature which derives plasma parameters from the balance of input power to plasma power losses. The model is extended with descriptions of the plasma sheath and ion-surface interactions to derive the co-deposit removal rates. Plasma removal can be limited by this ion induced surface release rate or the rate of pumping of the released species. To test the models dedicated

  12. PREFACE: 15th Latin American Workshop on Plasma Physics (LAWPP 2014) and 21st IAEA TM on Research Using Small Fusion Devices (RUSFD)

    Science.gov (United States)

    Iván Vargas-Blanco, V.; Herrera-Velázquez, J. Julio E.

    2015-03-01

    Written contributions from participants of the Joint 15th Latin American Workshop on Plasma Physics (LAWPP 2014) - 21st IAEA Technical Meeting on Research Using Small Fusion Devices (21st IAEA TM RUSFD). The International Advisory Committees of the 15th Latin American Workshop on Plasma Physics (LAWPP 2014) and the 21st IAEA TM on Research Using Small Fusion Devices (RUSFD), agreed to carry out together this Joint LAWPP 2014 - 21st RUSFD in San José, Costa Rica, on 27-31 January 2014. The Joint LAWPP 2014 - 21st RUSFD meeting, organized by the Instituto Tecnológico de Costa Rica, Universidad Nacional de Costa Rica, and Ad Astra Rocket Company in collaboration with the International Atomic Energy Agency (IAEA). The Latin American Workshop on Plasma Physics (LAWPP) is a series of events which has been held periodically since 1982, with the purpose of providing a forum in which the research of the Latin American plasma physics community can be displayed, as well as fostering collaborations among plasma scientists within the region and with researchers from the rest of the world. Recognized plasma scientists from developed countries are specially invited to the meeting to present the state of the art on several "hot" topics related to plasma physics. It is an open meeting, with an International Advisory Committee, in which the working language is English. It was firstly held in 1982 in Cambuquira, Brazil, followed by workshops in Medellín, Colombia (1985), Santiago de Chile, Chile (1988), Buenos Aires, Argentina (1990), Mexico City, Mexico (1992), Foz do Iguaçu, Brazil (1994, combined with the International Congress on Plasma Physics (ICPP)), Caracas, Venezuela (1997), Tandil, Argentina (1998), La Serena, Chile (2000), Sao Pedro, Brazil (2003), Mexico City, Mexico (2005), Caracas, Venezuela (2007), Santiago de Chile, Chile (2010, combined with the ICPP) and Mar de Plata, Argentina (2011). The 21st IAEA TM on Research Using Small Fusion Devices is an ideal forum for

  13. Fundamental processes of plasma and reactive gas surface treatment for the recovery of hydrogen isotopes from carbon co-deposits in fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Moeller, Soeren

    2014-11-01

    The use of carbon-based plasma-facing wall components offers many advantages for plasma operation in magnetic confinement nuclear fusion devices. However, through reactions with the hydrogen based fusion plasma, carbon forms amorphous hydrogenated carbon co-deposits (a-C:H) in the vacuum vessels. If tritium is used to fuel the reactor, this co-deposition can quickly lead to an inacceptable high tritium inventory. Through co-deposition with carbon about 10% of the tritium injected into the reactor can be trapped. Even with other wall materials co-deposition can be significant. A method to recover the hydrogen isotopes from the co-deposits is necessary. The method has to be compatible with the requirements of the devices and nuclear fusion plasma operation. In this work thermo-chemical removal by neutral gases (TCR) and removal by plasmas is investigated. Models are developed to describe the involved processes of both removal methods. TCR is described using a reaction-diffusion model. Within this model the reactive gas diffuses into the co-deposits and subsequently reacts in a thermally activated process. The co-deposits are pyrolysed, forming volatile gases, e.g. CO{sub 2} and H{sub 2}O. These gases are pumped from the vacuum vessel and recycled. Applying the model to literature observations enables to connect data on exposure temperature, pressure, time and co-deposit properties. Two limits of TCR (reaction- or diffusion-limited) are identified. Plasma removal sputters co-deposits by their chemical and physical interaction with the impinging ions. The description uses a 0D plasma model from the literature which derives plasma parameters from the balance of input power to plasma power losses. The model is extended with descriptions of the plasma sheath and ion-surface interactions to derive the co-deposit removal rates. Plasma removal can be limited by this ion induced surface release rate or the rate of pumping of the released species. To test the models dedicated

  14. Steady-state operation of magnetic fusion devices: Plasma control and plasma facing components. Report on the IAEA technical committee meeting held at Fukuoka, 25-29 October 1999

    International Nuclear Information System (INIS)

    Engelmann, F.

    2000-01-01

    An IAEA Technical Committee Meeting on Steady-State Operation of Magnetic Fusion Devices - Plasma Control and Plasma Facing Components was held at Fukuoka, Japan, from 25 to 29 October 1999. The meeting was the second IAEA Techical Committee Meeting on the subject, following the one held at Hefei, China, a year earlier. The meeting was attended by over 150 researchers from 10 countries

  15. Fusion energy division computer systems network

    International Nuclear Information System (INIS)

    Hammons, C.E.

    1980-12-01

    The Fusion Energy Division of the Oak Ridge National Laboratory (ORNL) operated by Union Carbide Corporation Nuclear Division (UCC-ND) is primarily involved in the investigation of problems related to the use of controlled thermonuclear fusion as an energy source. The Fusion Energy Division supports investigations of experimental fusion devices and related fusion theory. This memo provides a brief overview of the computing environment in the Fusion Energy Division and the computing support provided to the experimental effort and theory research

  16. Demonstration tests of tritium removal device under the conditions of nuclear fusion reactor. Cooperation test between Japan and USA

    International Nuclear Information System (INIS)

    Hayashi, Takumi; Kobayashi, Kazuhiro; Nishi, Masataka

    2001-01-01

    Performance of oxidation catalysis in emergency tritium removal device was tested in Los Alamos National Laboratory by cooperation between Japan and USA on November 8, 2000. To reduce the effects of tritium on the environment, a plan of the closed space for trapping tritium was made. A tritium removal device using oxidation catalysis and water vapor adsorption removes the tritium in the closed space. The treatment flow rate of the device is about 2,500 m 3 /h, the same as ITER(3,000 to 4,500 m 3 /h). Catalysis is Pt/ alumina. The closed space is 3,000m 2 . The initial concentration of tritium was about 7 Bq/cm 2 , ten times as large as the concentration limit in atmosphere. The concentration of tritium in the test laboratory decreased linearly with time and attained to the limit value after about 200 min. Residue of tritium on the wall had been removed and the significant quantity was not detected after three days. The results proved to satisfy safety of ITER. (S.Y.)

  17. Busbar arcs at large fusion magnets: Conductor to feeder tube arcing model experiments with the LONGARC device

    Energy Technology Data Exchange (ETDEWEB)

    Klimenko, Dmitry, E-mail: dmitry.klimenko@kit.edu; Pasler, Volker

    2014-10-15

    Highlights: •The LONGARC device was successfully implemented for busbar to feeder tubes arcing model experiments. •Arcing at an ITER busbar inside its feeder tube was simulated in scaled model experiments. •The narrower half tubes imply a slight increase of the arc propagation speed in compare to full tube experiments. •All simulated half tubes experiments show severe damage indicating that the ITER inner feeder tube will not withstand a busbar arc. -- Abstract: Electric arcs moving along the power cables (the so-called busbars) of the toroidal field (TF) coils of ITER may reach and penetrate the cryostat wall. Model experiments with the new LONGARC device continue the VACARC (VACuum ARC) experiments that were initiated to investigate the propagation and destruction mechanisms of busbar arcs in small scale [1]. The experiments are intended to support the development and validation of a numerical model. LONGARC overcomes the space limitations inside VACARC and allows also for advanced 1:3 (vs. ITER full scale) model setups. The LONGARC device and first results are presented below.

  18. Compact fusion reactors

    CERN Multimedia

    CERN. Geneva

    2015-01-01

    Fusion research is currently to a large extent focused on tokamak (ITER) and inertial confinement (NIF) research. In addition to these large international or national efforts there are private companies performing fusion research using much smaller devices than ITER or NIF. The attempt to achieve fusion energy production through relatively small and compact devices compared to tokamaks decreases the costs and building time of the reactors and this has allowed some private companies to enter the field, like EMC2, General Fusion, Helion Energy, Lawrenceville Plasma Physics and Lockheed Martin. Some of these companies are trying to demonstrate net energy production within the next few years. If they are successful their next step is to attempt to commercialize their technology. In this presentation an overview of compact fusion reactor concepts is given.

  19. The role of bone SPECT/CT in the evaluation of lumbar spinal fusion with metallic fixation devices

    DEFF Research Database (Denmark)

    Damgaard, Morten; Nimb, Lars; Madsen, Jan L

    2010-01-01

    PURPOSE: It is difficult to evaluate the stability of the lumbar spondylodesis with metallic fixation devices by conventional imaging methods such as radiography or magnetic resonance imaging. It is unknown whether single photon emission computed tomography/computed tomography (SPECT/CT) may......, whereas in 1 case loose pedicle screws were detected at a wrong vertebral level. CONCLUSION: SPECT/CT may be useful to detect a lack of fixation of the metallic implants, and hence instability of the spondylodesis by evaluating the focal bone mineralization activity in relation to the pedicle screws....

  20. Conceptual design studies of special-purpose equipment for Fusion Engineering Device torus-sector remote maintenance

    International Nuclear Information System (INIS)

    Masson, L.S.; Watts, K.D.; Aldrich, W.E.; McPherson, R.S.

    1982-01-01

    One of the major maintenance operations anticipated for fusion reactors of the Tokamak configuration is remote removal and replacement of torus sectors. This operation will be difficult due to the massive nature of the sector (375 tonnes), and also due to the precision with which it must be positioned within the fixed structure. The same problem, only to a lesser degree, applies to sub-components of the sector such as the limiter blades, shielding, test assemblies, etc. General and specific design requirements have been generated and trade studies conducted on reactor interfacing details as well as handling machine concepts. On the basis of the design requirements and trade studies, a perferred concept for the sector handling machine was developed. In addition, a similar machine was developed for handling the intermediate sized sector sub-components. While most operations will be performed by special purpose machines such as described above, there is a need for a versatile, relatively high capacity mobile system. A concept suitable for this mobile application was also developed as part of these studies. The general conclusion, to the extent these studies have been completed, was that special single-purpose machines will be required to perform the operations requiring high load capacity and handling precision. The machine concepts developed were felt to be within the state-of-the-art, and will make extensive use of commercially available components. The most serious problem was felt to be development of simple methods to obtain the required precision in positioning massive objects such as the torus sector

  1. GEM gas detectors for soft X-ray imaging in fusion devices with neutron–gamma background

    Energy Technology Data Exchange (ETDEWEB)

    Pacella, Danilo, E-mail: danilo.pacella@enea.it [Associazione EURATOM-ENEA, C.R. Frascati, Via E. Fermi 45, 00044 Frascati, Roma (Italy); Romano, Afra; Gabellieri, Lori [Associazione EURATOM-ENEA, C.R. Frascati, Via E. Fermi 45, 00044 Frascati, Roma (Italy); Murtas, Fabrizio [Istituto Nazionale di Fisica Nucleare, Via E. Fermi 45, 00044 Frascati, Roma (Italy); Mazon, Didier [Association EURATOM-CEA, CEA Cadarache, DSM/IRFM, 13108 St. Paul Lez Durance Cedex (France)

    2013-08-21

    A triple gas electron multiplier (GEM) detector has been built and characterized in a collaboration between ENEA, INFN and CEA to develop a soft X-ray imaging diagnostic for magnetic fusion plasmas. It has an active area of 5×5 cm{sup 2}, 128 pixels and electronics in counting mode. Since burning plasma experiments will have a very large background of radiation, this prototype has been tested with contemporary X-ray, neutron and gamma irradiation, to study the detection efficiencies, and the discrimination capabilities. The detector has been preliminarily characterized under DD neutron irradiation (2.45 MeV) up to 2.2×10{sup 6} n/s on the detector active area, showing a detection efficiency of about 10{sup −4}, while the detection efficiency of X-rays is more than three orders of magnitude higher. The detector has been also tested under DT neutron flux (14 MeV) up to 2.8×10{sup 8} n/s on the whole detector, with a detection efficiency of about 10{sup −5}. The calibration of the γ-rays detection has been done by means of a source of {sup 60}Co (gamma rays of energy 1.17 MeV and 1.33 MeV) and the detection efficiency was found of the order of 10{sup −4}. Thanks to the adjustable gain of the detector and the discrimination threshold of the electronics, it is possible to minimize the sensitivity to neutrons and gamma, and discriminate the X-ray signals even with very high radiative background.

  2. Fusion energy

    International Nuclear Information System (INIS)

    Gross, R.A.

    1984-01-01

    This textbook covers the physics and technology upon which future fusion power reactors will be based. It reviews the history of fusion, reaction physics, plasma physics, heating, and confinement. Descriptions of commercial plants and design concepts are included. Topics covered include: fusion reactions and fuel resources; reaction rates; ignition, and confinement; basic plasma directory; Tokamak confinement physics; fusion technology; STARFIRE: A commercial Tokamak fusion power plant. MARS: A tandem-mirror fusion power plant; and other fusion reactor concepts

  3. Health physics around a controlled fusion research device: the Tokamak at Fontenay-aux-Roses (T.F.R.)

    International Nuclear Information System (INIS)

    1977-01-01

    The X and neutron dosimetry measurement near the magnetic confinement device for hot plasma, called T.F.R. (Tokamak, Fontenay-aux-Roses) are presented. The biological shielding consists of an ordinary concrete wall 30 cm thick; the dose rate is thus limited at 10 -1 mrem per discharge (corresponding to 10 mrem per day) in the whole area frequented by people during T.F.R. operation. A numerical calculation, taking into account the true geometry and X ray reflexion by the walls and roof, and normalized to the measurements, gives some indications on the electron beam which produces X rays. The photoneutron source (up to 10 10 neutrons per dischage) and the activation of the vacuum vessel result from high energy electrons (>= 10 MeV) supporting a 10 to 1,000 A current [fr

  4. 21 CFR 886.1880 - Fusion and stereoscopic target.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Fusion and stereoscopic target. 886.1880 Section... (CONTINUED) MEDICAL DEVICES OPHTHALMIC DEVICES Diagnostic Devices § 886.1880 Fusion and stereoscopic target. (a) Identification. A fusion and stereoscopic target is a device intended for use as a viewing object...

  5. Sensor Fusion of a Mobile Device to Control and Acquire Videos or Images of Coffee Branches and for Georeferencing Trees

    Directory of Open Access Journals (Sweden)

    Paula Jimena Ramos Giraldo

    2017-04-01

    Full Text Available Smartphones show potential for controlling and monitoring variables in agriculture. Their processing capacity, instrumentation, connectivity, low cost, and accessibility allow farmers (among other users in rural areas to operate them easily with applications adjusted to their specific needs. In this investigation, the integration of inertial sensors, a GPS, and a camera are presented for the monitoring of a coffee crop. An Android-based application was developed with two operating modes: (i Navigation: for georeferencing trees, which can be as close as 0.5 m from each other; and (ii Acquisition: control of video acquisition, based on the movement of the mobile device over a branch, and measurement of image quality, using clarity indexes to select the most appropriate frames for application in future processes. The integration of inertial sensors in navigation mode, shows a mean relative error of ±0.15 m, and total error ±5.15 m. In acquisition mode, the system correctly identifies the beginning and end of mobile phone movement in 99% of cases, and image quality is determined by means of a sharpness factor which measures blurriness. With the developed system, it will be possible to obtain georeferenced information about coffee trees, such as their production, nutritional state, and presence of plagues or diseases.

  6. Sensor Fusion of a Mobile Device to Control and Acquire Videos or Images of Coffee Branches and for Georeferencing Trees.

    Science.gov (United States)

    Giraldo, Paula Jimena Ramos; Aguirre, Álvaro Guerrero; Muñoz, Carlos Mario; Prieto, Flavio Augusto; Oliveros, Carlos Eugenio

    2017-04-06

    Smartphones show potential for controlling and monitoring variables in agriculture. Their processing capacity, instrumentation, connectivity, low cost, and accessibility allow farmers (among other users in rural areas) to operate them easily with applications adjusted to their specific needs. In this investigation, the integration of inertial sensors, a GPS, and a camera are presented for the monitoring of a coffee crop. An Android-based application was developed with two operating modes: ( i ) Navigation: for georeferencing trees, which can be as close as 0.5 m from each other; and ( ii ) Acquisition: control of video acquisition, based on the movement of the mobile device over a branch, and measurement of image quality, using clarity indexes to select the most appropriate frames for application in future processes. The integration of inertial sensors in navigation mode, shows a mean relative error of ±0.15 m, and total error ±5.15 m. In acquisition mode, the system correctly identifies the beginning and end of mobile phone movement in 99% of cases, and image quality is determined by means of a sharpness factor which measures blurriness. With the developed system, it will be possible to obtain georeferenced information about coffee trees, such as their production, nutritional state, and presence of plagues or diseases.

  7. Contribution of the different erosion processes to material release from the vessel walls of fusion devices during plasma operation

    International Nuclear Information System (INIS)

    Behrisch, R.

    2002-01-01

    In high temperature plasma experiments several processes contribute to erosion and loss of material from the vessel walls. This material may enter the plasma edge and the central plasma where it acts as impurities. It will finally be re-deposited at other wall areas. These erosion processes are: evaporation due to heating of wall areas. At very high power deposition evaporation may become very large, which has been named ''blooming''. Large evaporation and melting at some areas of the vessel wall surface may occur during heat pulses, as observed in plasma devices during plasma disruptions. At tips on the vessel walls and/or hot spots on the plasma exposed solid surfaces electrical arcs between the plasma and the vessel wall may ignite. They cause the release of ions, atoms and small metal droplets, or of carbon dust particles. Finally, atoms from the vessel walls are removed by physical and chemical sputtering caused by the bombardment of the vessel walls with ions as well as energetic neutral hydrogen atoms from the boundary plasma. All these processes have been, and are, observed in today's plasma experiments. Evaporation can in principle be controlled by very effective cooling of the wall tiles, arcing is reduced by very stable plasma operation, and sputtering by ions can be reduced by operating with a cold plasma in front of the vessel walls. However, sputtering by energetic neutrals, which impinge on all areas of the vessel walls, is likely to be the most critical process because ions lost from the plasma recycle as neutrals or have to be refuelled by neutrals leading to the charge exchange processes in the plasma. In order to quantify the wall erosion, ''materials factors'' (MF) have been introduced in the following for the different erosion processes. (orig.)

  8. Accelerators for heavy ion fusion

    International Nuclear Information System (INIS)

    Bangerter, R.O.

    1985-10-01

    Large fusion devices will almost certainly produce net energy. However, a successful commercial fusion energy system must also satisfy important engineering and economic constraints. Inertial confinement fusion power plants driven by multi-stage, heavy-ion accelerators appear capable of meeting these constraints. The reasons behind this promising outlook for heavy-ion fusion are given in this report. This report is based on the transcript of a talk presented at the Symposium on Lasers and Particle Beams for Fusion and Strategic Defense at the University of Rochester on April 17-19, 1985

  9. Fusion reactor materials

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    The following topics are briefly discussed: (1) surface blistering studies on fusion reactor materials, (2) TFTR design support activities, (3) analysis of samples bombarded in-situ in PLT, (4) chemical sputtering effects, (5) modeling of surface behavior, (6) ion migration in glow discharge tube cathodes, (7) alloy development for irradiation performance, (8) dosimetry and damage analysis, and (9) development of tritium migration in fusion devices and reactors

  10. Fusion-Fission Transmutation Scheme-Efficient destruction of nuclear waste

    International Nuclear Information System (INIS)

    Kotschenreuther, M.; Valanju, P.M.; Mahajan, S.M.; Schneider, E.A.

    2009-01-01

    A fusion-assisted transmutation system for the destruction of transuranic nuclear waste is developed by combining a subcritical fusion-fission hybrid assembly uniquely equipped to burn the worst thermal nonfissile transuranic isotopes with a new fuel cycle that uses cheaper light water reactors for most of the transmutation. The center piece of this fuel cycle, the high power density compact fusion neutron source (100 MW, outer radius <3 m), is made possible by a new divertor with a heat-handling capacity five times that of the standard alternative. The number of hybrids needed to destroy a given amount of waste is an order of magnitude below the corresponding number of critical fast-spectrum reactors (FRs) as the latter cannot fully exploit the new fuel cycle. Also, the time needed for 99% transuranic waste destruction reduces from centuries (with FR) to decades

  11. A Review on the Potential Use of Austenitic Stainless Steels in Nuclear Fusion Reactors

    Science.gov (United States)

    Şahin, Sümer; Übeyli, Mustafa

    2008-12-01

    Various engineering materials; austenitic stainless steels, ferritic/martensitic steels, vanadium alloys, refractory metals and composites have been suggested as candidate structural materials for nuclear fusion reactors. Among these structural materials, austenitic steels have an advantage of extensive technological database and lower cost compared to other non-ferrous candidates. Furthermore, they have also advantages of very good mechanical properties and fission operation experience. Moreover, modified austenitic stainless (Ni and Mo free) have relatively low residual radioactivity. Nevertheless, they can't withstand high neutron wall load which is required to get high power density in fusion reactors. On the other hand, a protective flowing liquid wall between plasma and solid first wall in these reactors can eliminate this restriction. This study presents an overview of austenitic stainless steels considered to be used in fusion reactors.

  12. Overview of the TITAN-II reversed-field pinch aqueous fusion power core design

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.C.; Creedon, R.L.; Grotz, S.; Cheng, E.T.; Sharafat, S.; Cooke, P.I.H.

    1988-03-01

    TITAN-II is a compact, high power density Reversed-Field Pinch fusion power reactor design based on the aqueous lithium solution fusion power core concept. The selected breeding and structural materials are LiNO/sub 3/ and 9-C low activation ferritic steel, respectively. TITAN-II is a viable alternative to the TITAN-I lithium self-cooled design for the Reversed-Field Pinch reactor to operate at a neutron wall loading of 18 MWm/sup 2/. Submerging the complete fusion power core and the primary loop in a large pool of cool water will minimize the probability of radioactivity release. Since the protection of the large pool integrity is the only requirement for the protection of the public, TITAN-II is a passive safety assurance design. 13 refs., 3 figs., 1 tab.

  13. Overview of the TITAN-II reversed-field pinch aqueous fusion power core design

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.C.; Creedon, R.L.; Cheng, E.T. (General Atomic Co., San Diego, CA (USA)); Grotz, S.P.; Sharafat, S.; Cooke, P.I.H. (California Univ., Los Angeles (USA). Dept. of Mechanical, Aerospace and Nuclear Engineering; California Univ., Los Angeles, CA (USA). Inst. for Plasma and Fusion Research); TITAN Research Group

    1989-04-01

    TITAN-II is a compact, high-power-density Reversed-Field Pinch fusion power reactor design based on the aqueous lithium solution fusion power core concept. The selected breeding and structural materials are LiNO/sub 3/ and 9-C low activation ferritic steel, respectively. TITAN-II is a viable alternative to the TITAN-I lithium self-cooled design for the Reversed-Field Pinch reactor to operate at a neutron wall loading of 18 MW/m/sup 2/. Submerging the complete fusion power core and the primary loop in a large pool of cool water will minimize the probability of radioactivity release. Since the protection of the large pool integrity is the only requirement for the protection of the public, TITAN-II is a level 2 of passive safety assurance design. (orig.).

  14. Fusion power and its prospects

    International Nuclear Information System (INIS)

    Kammash, T.

    1981-01-01

    Recent progress in research towards the development of fusion power is reviewed. In the magnetic approach, the impressive advances made in Tokamak research in the past few years have bolstered the confidence that experimental Tokamak devices currently under construction will demonstrate the break-even condition or scientific feasibility of fusion power. Exciting and innovative ideas in mirror magnetic confinement are expected to culminate in high-Q devices which will make open-ended confinement a serious contender for fusion reactors. In the inertial confinement approach, conflicting pellet temperature requirements have placed severe constraints on useful laser intensities and wavelengths for laser-driven fusion. Relativistic electron beam fusion must solve critical focusing and pellet coupling problems, and the newly proposed heavy ion beam fusion, though feasible and attractive in principle, requires very high energy particles for which the accelerator technology may not be available for some time to come

  15. Vacuum fusion of uranium

    International Nuclear Information System (INIS)

    Stohr, J.A.

    1957-01-01

    After having outlined that vacuum fusion and moulding of uranium and of its alloys have some technical and economic benefits (vacuum operations avoid uranium oxidation and result in some purification; precision moulding avoids machining, chip production and chemical reprocessing of these chips; direct production of the desired shape is possible by precision moulding), this report presents the uranium fusion unit (its low pressure enclosure and pumping device, the crucible-mould assembly, and the MF supply device). The author describes the different steps of cast production, and briefly comments the obtained results

  16. Revitalizing Fusion via Fission Fusion

    Science.gov (United States)

    Manheimer, Wallace

    2001-10-01

    Existing tokamaks could generate significant nuclear fuel. TFTR, operating steady state with DT might generate enough fuel for a 300 MW nuclear reactor. The immediate goals of the magnetic fusion program would necessarily shift from a study of advanced plasma regimes in larger sized devices, to mostly known plasmas regimes, but at steady state or high duty cycle operation in DT plasmas. The science and engineering of breeding blankets would be equally important. Follow on projects could possibly produce nuclear fuel in large quantity at low price. Although today there is strong opposition to nuclear power in the United States, in a 21st century world of 10 billion people, all of whom will demand a middle class life style, nuclear energy will be important. Concern over greenhouse gases will also drive the world toward nuclear power. There are studies indicating that the world will need 10 TW of carbon free energy by 2050. It is difficult to see how this can be achieved without the breeding of nuclear fuel. By using the thorium cycle, proliferation risks are minimized. [1], [2]. 1 W. Manheimer, Fusion Technology, 36, 1, 1999, 2.W. Manheimer, Physics and Society, v 29, #3, p5, July, 2000

  17. The fusion-fission hybrid

    International Nuclear Information System (INIS)

    Teller, E.

    1985-01-01

    As the history of the development of fusion energy shows, a sustained controlled fusion reaction is much more difficult to produce than rapid uncontrolled release of fusion energy. Currently, the ''magnetic bottle'' technique shows sufficient progress that it might applied for the commercial fuel production of /sup 233/U, suitable for use in fission reactors, by developing a fusion-fission hybrid. Such a device would consist of a fusion chamber core surrounded by a region containing cladded uranium pellets cooled by helium, with lithium salts also present to produce tritium to refuel the fusion process. Successful development of this hybrid might be possible within 10 y, and would provide both experience and funds for further development of controlled fusion energy

  18. Osteoclast Fusion

    DEFF Research Database (Denmark)

    Marie Julie Møller, Anaïs; Delaissé, Jean-Marie; Søe, Kent

    2017-01-01

    on the nuclearity of fusion partners. While CD47 promotes cell fusions involving mono-nucleated pre-osteoclasts, syncytin-1 promotes fusion of two multi-nucleated osteoclasts, but also reduces the number of fusions between mono-nucleated pre-osteoclasts. Furthermore, CD47 seems to mediate fusion mostly through...... individual fusion events using time-lapse and antagonists of CD47 and syncytin-1. All time-lapse recordings have been studied by two independent observers. A total of 1808 fusion events were analyzed. The present study shows that CD47 and syncytin-1 have different roles in osteoclast fusion depending...... broad contact surfaces between the partners' cell membrane while syncytin-1 mediate fusion through phagocytic-cup like structure. J. Cell. Physiol. 9999: 1-8, 2016. © 2016 Wiley Periodicals, Inc....

  19. Posterior lumbar interbody fusion using non resorbable poly-ether-ether-ketone versus resorbable poly-L-lactide-co-D,L-lactide fusion devices. Clinical outcome at a minimum of 2-year follow-up

    NARCIS (Netherlands)

    Jiya, T.U.; Smit, T.H.; van Royen, B.J.; Mullender, M.G.

    2011-01-01

    Previous papers on resorbable poly-L-lactideco-D,L-lactide (PLDLLA) cages in spinal fusion have failed to report adequately on patient-centred clinical outcome measures. Also comparison of PLDLLA cage with a traditionally applicable counterpart has not been previously reported. This is the first

  20. Fusion-power demonstration

    International Nuclear Information System (INIS)

    Henning, C.D.; Logan, B.G.; Carlson, G.A.; Neef, W.S.; Moir, R.W.; Campbell, R.B.; Botwin, R.; Clarkson, I.R.; Carpenter, T.J.

    1983-01-01

    As a satellite to the MARS (Mirror Advanced Reactor Study) a smaller, near-term device has been scoped, called the FPD (Fusion Power Demonstration). Envisioned as the next logical step toward a power reactor, it would advance the mirror fusion program beyond MFTF-B and provide an intermediate step toward commercial fusion power. Breakeven net electric power capability would be the goal such that no net utility power would be required to sustain the operation. A phased implementation is envisioned, with a deuterium checkout first to verify the plasma systems before significant neutron activation has occurred. Major tritium-related facilities would be installed with the second phase to produce sufficient fusion power to supply the recirculating power to maintain the neutral beams, ECRH, magnets and other auxiliary equipment

  1. Fusion power demonstration

    International Nuclear Information System (INIS)

    Henning, C.D.; Logan, B.G.

    1983-01-01

    As a satellite to the MARS (Mirror Advanced Reactor Study) a smaller, near-term device has been scoped, called the FPD (Fusion Power Demonstration). Envisioned as the next logical step toward a power reactor, it would advance the mirror fusion program beyond MFTF-B and provide an intermediate step toward commercial fusion power. Breakeven net electric power capability would be the goal such that no net utility power would be required to sustain the operation. A phased implementation is envisioned, with a deuterium checkout first to verify the plasma systems before significant neutron activation has occurred. Major tritium-related facilities would be installed with the second phase to produce sufficient fusion power to supply the recirculating power to maintain the neutral beams, ECRH, magnets and other auxiliary equipment

  2. Conceptual design of the SlimCS fusion DEMO reactor

    International Nuclear Information System (INIS)

    Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Nakamura, Hirofumi; Hayashi, Takumi; Asakura, Nobuyuki; Utoh, Hiroyasu; Tanigawa, Hiroyasu; Nishitani, Takeo; Isono, Takaaki; Sakurai, Shinji; Kurita, Genichi; Hayashi, Takao; Oyama, Naoyuki; Liu Changle; Hamamatsu, Kiyotaka; Inoue, Takashi; Ozeki, Takahisa; Sato, Masayasu; Suzuki, Satoshi; Kawashima, Hisato; Ezato, Koichiro; Tsuru, Daigo; Koizumi, Norikiyo; Sakamoto, Keiji; Ando, Masami; Sakamoto, Yoshiteru; Shibama, Yusuke; Suzuki, Takahiro; Takechi, Manabu; Takahashi, Koji; Hirose, Takanori; Sato, Satoru; Nozawa, Takashi; Tanigawa, Hisashi; Kakudate, Satoshi; Kawamura, Yoshinori; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Ochiai, Kentaro; Ide, Shunsuke; Aiba, Nobuyuki; Shimizu, Katsuhiro; Honda, Mitsuru; Nakamichi, Masaru; Nishi, Hiroshi; Seki, Yoji; Nakamura, Yukiharu; Tsuchiya, Kunihiko; Yoshida, Tohru; Song Yuntao

    2010-08-01

    This report describes the results of the conceptual design study of the SlimCS fusion DEMO reactor aiming at demonstrating fusion power production in a plant scale and allowing to assess the economic prospects of a fusion power plant. The design study has focused on a compact and low aspect ratio tokamak reactor concept with a reduced-sized central solenoid, which is novel compared with previous tokamak reactor concept such as SSTR (Steady State Tokamak Reactor). Owing to low aspect ratio, the reactor will be capable of having comparatively high beta limit and high elongation (which can elevate the Greenwald density limit), having potential for high power density. The reactor has the main parameters of a major radius of 5.5 m, aspect ratio of 2.6, elongation of 2.0, normalized beta of 4.3, fusion out put of 2.95 GW and average neutron wall load of 3 MW/m 2 . This report covers various aspects of design study including systematic design, physics design, torus configuration, blanket, superconducting magnet, maintenance and building, which were carried out increase the engineering feasibility of the concept. (author)

  3. Fusion power

    International Nuclear Information System (INIS)

    Hancox, R.

    1981-01-01

    The principles of fusion power, and its advantages and disadvantages, are outlined. Present research programmes and future plans directed towards the development of a fusion power reactor, are summarized. (U.K.)

  4. Fusion rings and fusion ideals

    DEFF Research Database (Denmark)

    Andersen, Troels Bak

    by the so-called fusion ideals. The fusion rings of Wess-Zumino-Witten models have been widely studied and are well understood in terms of precise combinatorial descriptions and explicit generating sets of the fusion ideals. They also appear in another, more general, setting via tilting modules for quantum......This dissertation investigates fusion rings, which are Grothendieck groups of rigid, monoidal, semisimple, abelian categories. Special interest is in rational fusion rings, i.e., fusion rings which admit a finite basis, for as commutative rings they may be presented as quotients of polynomial rings...

  5. Fusion Simulation Program

    International Nuclear Information System (INIS)

    Greenwald, Martin

    2011-01-01

    Many others in the fusion energy and advanced scientific computing communities participated in the development of this plan. The core planning team is grateful for their important contributions. This summary is meant as a quick overview the Fusion Simulation Program's (FSP's) purpose and intentions. There are several additional documents referenced within this one and all are supplemental or flow down from this Program Plan. The overall science goal of the DOE Office of Fusion Energy Sciences (FES) Fusion Simulation Program (FSP) is to develop predictive simulation capability for magnetically confined fusion plasmas at an unprecedented level of integration and fidelity. This will directly support and enable effective U.S. participation in International Thermonuclear Experimental Reactor (ITER) research and the overall mission of delivering practical fusion energy. The FSP will address a rich set of scientific issues together with experimental programs, producing validated integrated physics results. This is very well aligned with the mission of the ITER Organization to coordinate with its members the integrated modeling and control of fusion plasmas, including benchmarking and validation activities. (1). Initial FSP research will focus on two critical Integrated Science Application (ISA) areas: ISA1, the plasma edge; and ISA2, whole device modeling (WDM) including disruption avoidance. The first of these problems involves the narrow plasma boundary layer and its complex interactions with the plasma core and the surrounding material wall. The second requires development of a computationally tractable, but comprehensive model that describes all equilibrium and dynamic processes at a sufficient level of detail to provide useful prediction of the temporal evolution of fusion plasma experiments. The initial driver for the whole device model will be prediction and avoidance of discharge-terminating disruptions, especially at high performance, which are a critical

  6. Fusion: introduction

    International Nuclear Information System (INIS)

    Decreton, M.

    2006-01-01

    The article gives an overview and introduction to the activities of SCK-CEN's research programme on fusion. The decision to construct the ITER international nuclear fusion experiment in Cadarache is highlighted. A summary of the Belgian contributions to fusion research is given with particular emphasis on studies of radiation effects on diagnostics systems, radiation effects on remote handling sensing systems, fusion waste management and socio-economic studies

  7. Rencontre on fusion technology

    International Nuclear Information System (INIS)

    Read, S.F.J.

    1979-02-01

    This report of a rencontre held to consider the technology of magnetic confinement fusion devices gives the agenda for the meeting and lists those topics which were identified as areas of research. These topics included materials, tritium, structures and heat transfer, neutronics and nuclear data, and corrosion problems. (UK)

  8. Membrane fusion

    DEFF Research Database (Denmark)

    Bendix, Pól Martin

    2015-01-01

    At Stanford University, Boxer lab, I worked on membrane fusion of small unilamellar lipid vesicles to flat membranes tethered to glass surfaces. This geometry closely resembles biological systems in which liposomes fuse to plasma membranes. The fusion mechanism was studied using DNA zippering...... between complementary strands linked to the two apposing membranes closely mimicking the zippering mechanism of SNARE fusion complexes....

  9. Fusion Canada

    International Nuclear Information System (INIS)

    1987-07-01

    This first issue of a quarterly newsletter announces the startup of the Tokamak de Varennes, describes Canada's national fusion program, and outlines the Canadian Fusion Fuels Technology Program. A map gives the location of the eleven principal fusion centres in Canada. (L.L.)

  10. Controlled fusion and plasma physics

    International Nuclear Information System (INIS)

    1995-01-01

    This document presents the several speeches that took place during the 22nd European Physical Society conference on Controlled Fusion and Plasma Physics in Bournemouth, UK, between the 2nd and 7th July 1995. The talks deal with new experiments carried out on several tokamaks, particularly Tore Supra, concerning plasma confinement and fusion. Some information on specific fusion devices or tokamak devices is provided, as well as results of experiments concerning plasma instability. Separate abstracts were prepared for all the 31 papers in this volume. (TEC)

  11. Controlled fusion and plasma physics

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-31

    This document presents the several speeches that took place during the 22nd European Physical Society conference on Controlled Fusion and Plasma Physics in Bournemouth, UK, between the 2nd and 7th July 1995. The talks deal with new experiments carried out on several tokamaks, particularly Tore Supra, concerning plasma confinement and fusion. Some information on specific fusion devices or tokamak devices is provided, as well as results of experiments concerning plasma instability. Separate abstracts were prepared for all the 31 papers in this volume. (TEC).

  12. Controlled fusion and plasma physics

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-31

    This document presents the several speeches that took place during the 22nd European Physical Society conference on Controlled Fusion and Plasma Physics in Bournemouth, UK, between the 2nd and 7th July 1995. The talks deal with new experiments carried out on several tokamaks, particularly Tore Supra, concerning plasma confinement and fusion. Some information on specific fusion devices or tokamak devices is provided, as well as results of experiments concerning plasma instability. Separate abstracts were prepared for all the 31 papers in this volume. (TEC).

  13. Device description and general physics requirements

    International Nuclear Information System (INIS)

    Neilson, G.H. Jr.

    1992-01-01

    To accomplish the dual goals set forth in the mission statement - determination of burning plasma physics and demonstration of fusion power production - requires a tokamak with special characteristics. A conceptual design for such a facility has been developed by the CIT/BPX Project team over a period of about 5 years. The process has drawn extensively upon the world tokamak physics data base as well as engineering experience gained on actual machines like Tokamak Fusion Test Reactor (TFTR) and the Alcators and on design studies for machines like the Long-Pulse Ignited Test experiment (LITE) and the Tokamak Fusion Core Experiment (TFCX). The Joint European Torus (JET) and Doublet III-D (DIII-D) experiments in particular have had a significant influence on the physics base. The resulting design incorporates features that have proven successful in tokamak experiments, combined with new features that are needed in the regime of high-power-density deuterium-tritium (D-T) fusion plasmas

  14. Posterior lumbar interbody fusion using non resorbable poly-ether-ether-ketone versus resorbable poly-L-lactide-co-D,L-lactide fusion devices. Clinical outcome at a minimum of 2-year follow-up.

    Science.gov (United States)

    Jiya, Timothy U; Smit, T; van Royen, B J; Mullender, M

    2011-04-01

    Previous papers on resorbable poly-L-lactide-co-D,L-lactide (PLDLLA) cages in spinal fusion have failed to report adequately on patient-centred clinical outcome measures. Also comparison of PLDLLA cage with a traditionally applicable counterpart has not been previously reported. This is the first randomized prospective study that assesses clinical outcome of PLDLLA cage compared with a poly-ether-ether-ketone (PEEK) implant. Twenty-six patients were randomly assigned to undergo instrumented posterior lumbar interbody fusion (PLIF) whereby either a PEEK cage or a PLDLLA cage was implanted. Clinical outcome based on visual analogue scale scores for leg pain and back pain, as well as Oswestry Disability Index (ODI) and SF-36 questionnaires were documented and analysed. When compared with preoperative values, all clinical parameters have significantly improved in the PEEK group at 2 years after surgery with the exception of SF-36 general health, SF-36 mental health and SF-36 role emotional scores. No clinical parameter showed significant improvement at 2 years after surgery compared with preoperative values in the PLDLLA patient group. Only six patients (50%) in the PLDLLA group showed improvement in the VAS scores for leg and back pain as well as the ODI, as opposed to 10 patients (71%) in the PEEK group. One-third of the patients in the PLDLLA group actually reported worsening of their pain scores and ODI. Three cases of mild to moderate osteolysis were seen in the PLDLLA group. Following up on our preliminary report, these 2-year results confirm the superiority of the PEEK implant to the resorbable PLDLLA implant in aiding spinal fusion and alleviating symptoms following PLIF in patients with degenerative spondylolisthesis associated with either canal stenosis or foramen stenosis or both and emanating from a single lumbar segment.

  15. Controlled thermonuclear fusion reactors

    International Nuclear Information System (INIS)

    Walstrom, P.L.

    1976-01-01

    Controlled production of energy by fusion of light nuclei has been the goal of a large portion of the physics community since the 1950's. In order for a fusion reaction to take place, the fuel must be heated to a temperature of 100 million degrees Celsius. At this temperature, matter can exist only in the form of an almost fully ionized plasma. In order for the reaction to produce net power, the product of the density and energy confinement time must exceed a minimum value of 10 20 sec m -3 , the so-called Lawson criterion. Basically, two approaches are being taken to meet this criterion: inertial confinement and magnetic confinement. Inertial confinement is the basis of the laser fusion approach; a fuel pellet is imploded by intense laser beams from all sides and ignites. Magnetic confinement devices, which exist in a variety of geometries, rely upon electromagnetic forces on the charged particles of the plasma to keep the hot plasma from expanding. Of these devices, the most encouraging results have been achieved with a class of devices known as tokamaks. Recent successes with these devices have given plasma physicists confidence that scientific feasibility will be demonstrated in the next generation of tokamaks; however, an even larger effort will be required to make fusion power commercially feasible. As a result, emphasis in the controlled thermonuclear research program is beginning to shift from plasma physics to a new branch of nuclear engineering which can be called fusion engineering, in which instrumentation and control engineers will play a major role. Among the new problem areas they will deal with are plasma diagnostics and superconducting coil instrumentation

  16. Fusion breeder

    International Nuclear Information System (INIS)

    Moir, R.W.

    1982-01-01

    The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the US fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the US fusion program and the US nuclear energy program. The purpose of this paper is to suggest this policy change be made and tell why it should be made, and to outline specific research and development goals so that the fusion breeder will be developed in time to meet fissile fuel needs

  17. Fusion Implementation

    International Nuclear Information System (INIS)

    Schmidt, J.A.

    2002-01-01

    If a fusion DEMO reactor can be brought into operation during the first half of this century, fusion power production can have a significant impact on carbon dioxide production during the latter half of the century. An assessment of fusion implementation scenarios shows that the resource demands and waste production associated with these scenarios are manageable factors. If fusion is implemented during the latter half of this century it will be one element of a portfolio of (hopefully) carbon dioxide limiting sources of electrical power. It is time to assess the regional implications of fusion power implementation. An important attribute of fusion power is the wide range of possible regions of the country, or countries in the world, where power plants can be located. Unlike most renewable energy options, fusion energy will function within a local distribution system and not require costly, and difficult, long distance transmission systems. For example, the East Coast of the United States is a prime candidate for fusion power deployment by virtue of its distance from renewable energy sources. As fossil fuels become less and less available as an energy option, the transmission of energy across bodies of water will become very expensive. On a global scale, fusion power will be particularly attractive for regions separated from sources of renewable energy by oceans

  18. Fusion blanket design and optimization techniques

    International Nuclear Information System (INIS)

    Gohar, Y.

    2005-01-01

    In fusion reactors, the blanket design and its characteristics have a major impact on the reactor performance, size, and economics. The selection and arrangement of the blanket materials, dimensions of the different blanket zones, and different requirements of the selected materials for a satisfactory performance are the main parameters, which define the blanket performance. These parameters translate to a large number of variables and design constraints, which need to be simultaneously considered in the blanket design process. This represents a major design challenge because of the lack of a comprehensive design tool capable of considering all these variables to define the optimum blanket design and satisfying all the design constraints for the adopted figure of merit and the blanket design criteria. The blanket design techniques of the First Wall/Blanket/Shield Design and Optimization System (BSDOS) have been developed to overcome this difficulty and to provide the state-of-the-art techniques and tools for performing blanket design and analysis. This report describes some of the BSDOS techniques and demonstrates its use. In addition, the use of the optimization technique of the BSDOS can result in a significant blanket performance enhancement and cost saving for the reactor design under consideration. In this report, examples are presented, which utilize an earlier version of the ITER solid breeder blanket design and a high power density self-cooled lithium blanket design for demonstrating some of the BSDOS blanket design techniques

  19. Recent fusion research in the National Institute for Fusion Science

    International Nuclear Information System (INIS)

    Komori, Akio; Sakakibara, Satoru; Sagara, Akio; Horiuchi, Ritoku; Yamada, Hiroshi; Takeiri, Yasuhiko

    2011-01-01

    The National Institute for Fusion Science (NIFS), which was established in 1989, promotes academic approaches toward the exploration of fusion science for steady-state helical reactor and realizes the establishment of a comprehensive understanding of toroidal plasmas as an inter-university research organization and a key center of worldwide fusion research. The Large Helical Device (LHD) Project, the Numerical Simulation Science Project, and the Fusion Engineering Project are organized for early realization of net current free fusion reactor, and their recent activities are described in this paper. The LHD has been producing high-performance plasmas comparable to those of large tokamaks, and several new findings with regard to plasma physics have been obtained. The numerical simulation science project contributes understanding and systemization of the physical mechanisms of plasma confinement in fusion plasmas and explores complexity science of a plasma for realization of the numerical test reactor. In the fusion engineering project, the design of the helical fusion reactor has progressed based on the development of superconducting coils, the blanket, fusion materials and tritium handling. (author)

  20. Controlled thermonuclear fusion

    International Nuclear Information System (INIS)

    Trocheris, M.

    1975-01-01

    An outline is given of the present position of research into controlled fusion. After a brief reminder of the nuclear reactions of fusion and the principle of their use as a source of energy, the results obtained by the method of magnetic confinement are summarized. Among the many solutions that have been imagined and tried out to achieve a magnetic containing vessel capable of holding the thermonuclear plasma, the devices of the Tokamak type have a good lead and that is why they are described in greater detail. An idea is then given of the problems that arise when one intends conceiving the thermonuclear reactor based on the principle of the Tokamaks. The last section deals with fusion by lasers which is a new and most attractive alternative, at least from the viewpoint of basis physics. The report concludes with an indication of the stages to be passed through to reach production of energy on an industrial scale [fr

  1. Peaceful Uses of Fusion

    Science.gov (United States)

    Teller, E.

    1958-07-03

    Applications of thermonuclear energy for peaceful and constructive purposes are surveyed. Developments and problems in the release and control of fusion energy are reviewed. It is pointed out that the future of thermonuclear power reactors will depend upon the construction of a machine that produces more electric energy than it consumes. The fuel for thermonuclear reactors is cheap and practically inexhaustible. Thermonuclear reactors produce less dangerous radioactive materials than fission reactors and, when once brought under control, are not as likely to be subject to dangerous excursions. The interaction of the hot plasma with magnetic fields opens the way for the direct production of electricity. It is possible that explosive fusion energy released underground may be harnessed for the production of electricity before the same feat is accomplished in controlled fusion processes. Applications of underground detonations of fission devices in mining and for the enhancement of oil flow in large low-specific-yield formations are also suggested.

  2. Ceramics for fusion applications

    International Nuclear Information System (INIS)

    Clinard, F.W. Jr.

    1987-01-01

    Ceramics are required for a variety of uses in both near-term fusion devices and in commercial powerplants. These materials must retain adequate structural and electrical properties under conditions of neutron, particle and ionizing irradiation; thermal and applied stresses; and physical and chemical sputtering. Ceramics such as Al 2 O 3 , MgAl 2 O 4 , BeO, Si 3 N 4 and SiC are currently under study for fusion applications, and results to date show widely-varying responses to the fusion environment. Materials can be identified today that will meet initial operating requirements, but improvements in physical properties are needed to achieve satisfactory lifetimes for critical applications. (author)

  3. Ceramics for fusion applications

    International Nuclear Information System (INIS)

    Clinard, F.W. Jr.

    1986-01-01

    Ceramics are required for a variety of uses in both near-term fusion devices and in commercial powerplants. These materials must retain adequate structural and electrical properties under conditions of neutron, particle, and ionizing irradiation; thermal and applied stresses; and physical and chemical sputtering. Ceramics such as Al 2 O 3 , MgAl 2 O 4 , BeO, Si 3 N 4 and SiC are currently under study for fusion applications, and results to date show widely-varying response to the fusion environment. Materials can be identified today which will meet initial operating requirements, but improvements in physical properties are needed to achieve satisfactory lifetimes for critical applications

  4. Fusion research at ORNL

    International Nuclear Information System (INIS)

    1982-03-01

    The ORNL Fusion Program includes the experimental and theoretical study of two different classes of magnetic confinement schemes - systems with helical magnetic fields, such as the tokamak and stellarator, and the ELMO Bumpy Torus (EBT) class of toroidally linked mirror systems; the development of technologies, including superconducting magnets, neutral atomic beam and radio frequency (rf) heating systems, fueling systems, materials, and diagnostics; the development of databases for atomic physics and radiation effects; the assessment of the environmental impact of magnetic fusion; and the design of advanced demonstration fusion devices. The program involves wide collaboration, both within ORNL and with other institutions. The elements of this program are shown. This document illustrates the program's scope; and aims by reviewing recent progress

  5. Neutrons and fusion

    International Nuclear Information System (INIS)

    Maynard, C.W.

    1976-01-01

    The production of energy from fusion reactions does not require neutrons in the fundamental sense that they are required in a fission reactor. Nevertheless, the dominant fusion reaction, that between deuterium and tritium, yields a 14 MeV neutron. To contrast a fusion reactor based on this reaction with the fission case, 3 x 10 20 such neutrons produced per gigawatt of power. This is four times as many neutrons as in an equivalent fission reactor and they carry seven times the energy of the fission neutrons. Thus, they dominate the energy recovery problem and create technological problems comparable to the original plasma confinement problem as far as a practical power producing device is concerned. Further contrasts of the fusion and fission cases are presented to establish the general role of neutrons in fusion devices. Details of the energy deposition processes are discussed and those reactions necessary for producing additional tritium are outlined. The relatively high energy flux with its large intensity will activate almost any materials of which the reactor may be composed. This activation is examined from the point of view of decay heat, radiological safety, and long-term storage. In addition, a discussion of the deleterious effects of neutron interactions on materials is given in some detail; this includes the helium and hydrogen producing reactions and displacement rate of the lattice atoms. The various materials that have been proposed for structural purposes, for breeding, reflecting, and moderating neutrons, and for radiation shielding are reviewed from the nuclear standpoint. The specific reactions of interest are taken up for various materials and finally a report is given on the status and prospects of data for fusion studies

  6. Insulators for fusion applications

    International Nuclear Information System (INIS)

    1987-04-01

    Design studies for fusion devices and reactors have become more detailed in recent years and with this has come a better understanding of requirements and operating conditions for insulators in these machines. Ceramic and organic insulators are widely used for many components of fusion devices and reactors namely: radio frequency (RF) energy injection systems (BeO, Al 2 O 3 , Mg Al 2 O 4 , Si 3 N 4 ); electrical insulation for the torus structure (SiC, Al 2 O 3 , MgO, Mg Al 2 O 4 , Si 4 Al 2 O 2 N 6 , Si 3 N 4 , Y 2 O 3 ); lightly-shielded magnetic coils (MgO, MgAl 2 O 4 ); the toroidal field coil (epoxies, polyimides), neutron shield (B 4 C, TiH 2 ); high efficiency electrical generation; as well as the generation of very high temperatures for high efficiency hydrogen production processes (ZrO 2 and Al 2 O 3 - mat, graphite and carbon - felt). Timely development of insulators for fusion applications is clearly necessary. Those materials to be used in fusion machines should show high resistance to radiation damage and maintain their structural integrity. Now the need is urgent for a variety of radiation resistant materials, but much effort in these areas is required for insulators to be considered seriously by the design community. This document contains 14 papers from an IAEA meeting. It was the objective of this meeting to identify existing problems in analysing various situations of applications and requirements of electrical insulators and ceramics in fusion and to recommend strategies and different stages of implementation. This meeting was endorsed by the International Fusion Research Council

  7. Thermonuclear fusion

    International Nuclear Information System (INIS)

    Weisse, J.

    2000-01-01

    This document takes stock of the two ways of thermonuclear fusion research explored today: magnetic confinement fusion and inertial confinement fusion. The basic physical principles are recalled first: fundamental nuclear reactions, high temperatures, elementary properties of plasmas, ignition criterion, magnetic confinement (charged particle in a uniform magnetic field, confinement and Tokamak principle, heating of magnetized plasmas (ohmic, neutral particles, high frequency waves, other heating means), results obtained so far (scale laws and extrapolation of performances, tritium experiments, ITER project), inertial fusion (hot spot ignition, instabilities, results (Centurion-Halite program, laser experiments). The second part presents the fusion reactor and its associated technologies: principle (tritium production, heat source, neutron protection, tritium generation, materials), magnetic fusion (superconducting magnets, divertor (role, principle, realization), inertial fusion (energy vector, laser adaptation, particle beams, reaction chamber, stresses, chamber concepts (dry and wet walls, liquid walls), targets (fabrication, injection and pointing)). The third chapter concerns the socio-economic aspects of thermonuclear fusion: safety (normal operation and accidents, wastes), costs (costs structure and elementary comparison, ecological impact and external costs). (J.S.)

  8. Fusion technology 1998

    International Nuclear Information System (INIS)

    Beaumont, B.; Libeyre, P.; Gentile, B. de; Tonon, G.

    1998-01-01

    The Symposium On Fusion Technology (SOFT) is held every two years with the objective to set the stage for the exchange of information on the design, construction and operation of fusion experiments and on the technology which is being developed for the next step devices and fusion reactors. By decision of the International Organizing Committee, the 20. SOFT includes invited talks, and oral and poster contributions in the following topics: plasma facing components, plasma heating and current drive, plasma engineering and control, experimental systems and diagnostics, magnets and power supplies, fuel technologies, remote operation, blanket and shield technologies, safety and environment, and system engineering and future devices. This symposium differs from the previous ones of this series by the way the present proceedings are produced. In order to have the written material available to the participants and the community at the nearest to the conference event, the papers have been collected 2 months in advance and printed in the present books. The goal was to deliver them to each participant upon arrival to the conference centre. These books contain all the papers corresponding to poster presentation, and the abstracts of the oral contributions and invited papers. The papers corresponding to these presentations, both oral and invited, will be published in 1999, after a standard review process, in a supplement of Fusion Engineering and Design. (author)

  9. Controlled nuclear fusion apparatus

    International Nuclear Information System (INIS)

    Bussard, R.W.; Coppi, B.

    1982-01-01

    A fusion power generating device is disclosed having a relatively small and inexpensive core region which may be contained within an energy absorbing blanket region. The fusion power core region contains apparatus of the toroidal type for confining a high density plasma. The fusion power core is removable from the blanket region and may be disposed and/or recycled for subsequent use within the same blanket region. Thermonuclear ignition of the plasma is obtained by feeding neutral fusible gas into the plasma in a controlled manner such that charged particle heating produced by the fusion reaction is utilized to bootstrap the device to a region of high temperatures and high densities wherein charged particle heating is sufficient to overcome radiation and thermal conductivity losses. The high density plasma produces a large radiation and particle flux on the first wall of the plasma core region thereby necessitating replacement of the core from the blanket region from time to time. A series of potentially disposable and replaceable central core regions are disclosed for a large-scale economical electrical power generating plant

  10. The scientific status of fusion

    International Nuclear Information System (INIS)

    Crandall, D.H.

    1989-01-01

    The development of fusion energy has been a large-scale scientific undertaking of broad interest. The magnetic plasma containment in tokamaks and the laser-drive ignition of microfusion capsules appear to be scientifically feasible sources of energy. These concepts are bounded by questions of required intensity in magnetid field and plasma currents or in drive energy and, for both concepts, by issues of plasma stability and energy transport. The basic concept and the current scientific issues are described for magnetic fusion and for the interesting, but likely infeasible, muon-catalyzed fusion concept. Inertial fusion is mentioned, qualitatively, to complete the context. For magnetic fusion, the required net energy production within the plasma may be accomplished soon, but the more useful goal of self-sustained plasma ignition requires a new device of somewhat uncertain (factor of 2) cost and size. (orig.)

  11. Application of structural-mechanics methods to the design of large tandem-mirror fusion devices (MFTF-B). Revision 1

    International Nuclear Information System (INIS)

    Karpenko, V.N.; Ng, D.S.

    1985-01-01

    The Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory requires state-of-the-art structural-mechanics methods to deal with access constraints for plasma heating and diagnostics, alignment requirements, and load complexity and variety. Large interactive structures required an integrated analytical approach to achieve a reasonable level of overall system optimization. The Tandem Magnet Generator (TMG) creates a magnet configuration for the EFFI calculation of electromagnetic-field forces that, coupled with other loads, form the input loading to magnet and vessel finite-element models. The analytical results provide the data base for detailed design of magnet, vessel, foundation, and interaction effects. 13 refs

  12. On fusion driven systems (FDS) for transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Aagren, O (Uppsala Univ., Aangstroem laboratory, div. of electricity, Uppsala (Sweden)); Moiseenko, V.E. (Inst. of Plasma Physics, National Science Center, Kharkov Inst. of Physics and Technology, Kharkov (Ukraine)); Noack, K. (Forschungszentrum Dresden-Rossendorf (Germany))

    2008-10-15

    This report gives a brief description of ongoing activities on fusion driven systems (FDS) for transmutation of the long-lived radioactive isotopes in the spent nuclear waste from fission reactors. Driven subcritical systems appears to be the only option for efficient minor actinide burning. Driven systems offer a possibility to increase reactor safety margins. A comparatively simple fusion device could be sufficient for a fusion-fission machine, and transmutation may become the first industrial application of fusion. Some alternative schemes to create strong fusion neutron fluxes are presented

  13. On fusion driven systems (FDS) for transmutation

    International Nuclear Information System (INIS)

    Aagren, O; Moiseenko, V.E.; Noack, K.

    2008-10-01

    This report gives a brief description of ongoing activities on fusion driven systems (FDS) for transmutation of the long-lived radioactive isotopes in the spent nuclear waste from fission reactors. Driven subcritical systems appears to be the only option for efficient minor actinide burning. Driven systems offer a possibility to increase reactor safety margins. A comparatively simple fusion device could be sufficient for a fusion-fission machine, and transmutation may become the first industrial application of fusion. Some alternative schemes to create strong fusion neutron fluxes are presented

  14. Alternative fusion concepts

    International Nuclear Information System (INIS)

    Rostagni, G.

    1981-01-01

    The paper reports the discussions and statements made by the participants on the actual state and future of five different approaches on the fusion concept; they are the following: bumpy torus, reversed-field pinch, open-ended configurations, compact toroids and stellarators. Tables show for each concept parameters that represent the achieved results; data expected for future devices and extrapolations on reactor requirements are included

  15. Fusion development and technology

    International Nuclear Information System (INIS)

    Montgomery, D.B.

    1992-01-01

    This report discusses the following: superconducting magnet technology; high field superconductors; advanced magnetic system and divertor development; poloidal field coils; gyrotron development; commercial reactor studies--aries; ITER physics: alpha physics and alcator R ampersand D for ITER; lower hybrid current drive and heating in the ITER device; ITER superconducting PF scenario and magnet analysis; ITER systems studies; and safety, environmental and economic factors in fusion development

  16. (Fusion energy research)

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, C.A. (ed.)

    1988-01-01

    This report discusses the following topics: principal parameters achieved in experimental devices (FY88); tokamak fusion test reactor; Princeton beta Experiment-Modification; S-1 Spheromak; current drive experiment; x-ray laser studies; spacecraft glow experiment; plasma deposition and etching of thin films; theoretical plasma; tokamak modeling; compact ignition tokamak; international thermonuclear experimental reactor; Engineering Department; Project Planning and Safety Office; quality assurance and reliability; and technology transfer.

  17. [Fusion energy research

    International Nuclear Information System (INIS)

    Phillips, C.A.

    1988-01-01

    This report discusses the following topics: principal parameters achieved in experimental devices (FY88); tokamak fusion test reactor; Princeton beta Experiment-Modification; S-1 Spheromak; current drive experiment; x-ray laser studies; spacecraft glow experiment; plasma deposition and etching of thin films; theoretical plasma; tokamak modeling; compact ignition tokamak; international thermonuclear experimental reactor; Engineering Department; Project Planning and Safety Office; quality assurance and reliability; and technology transfer

  18. Atomic fusion, Gerrard atomic fusion

    International Nuclear Information System (INIS)

    Gerrard, T.H.

    1980-01-01

    In the approach to atomic fusion described here the heat produced in a fusion reaction, which is induced in a chamber by the interaction of laser beams and U.H.F. electromagnetic beams with atom streams, is transferred to a heat exchanger for electricity generation by a coolant flowing through a jacket surrounding the chamber. (U.K.)

  19. Canadian contributions to the safety and environmental aspects of fusion

    International Nuclear Information System (INIS)

    Stasko, R.; Wong, K.

    1987-05-01

    Since next-step fusion devices will be fuelled with mixtures of tritium and deuterium, the knowledge base and tritium handling experience associated with the operation of CANDU reactors is viewed as relevant to the development of safe fusion technology. Fusion safety issues will be compared with fission safety experience, after which specific Canadian activities in support of fusion safety will be overviewed. In addition, recommendations for appropriate fusion safety criteria will be summarized. 18 refs

  20. Device configuration-management system

    International Nuclear Information System (INIS)

    Nowell, D.M.

    1981-01-01

    The Fusion Chamber System, a major component of the Magnetic Fusion Test Facility, contains several hundred devices which report status to the Supervisory Control and Diagnostic System for control and monitoring purposes. To manage the large number of diversity of devices represented, a device configuration management system was required and developed. Key components of this software tool include the MFTF Data Base; a configuration editor; and a tree structure defining the relationships between the subsystem devices. This paper will describe how the configuration system easily accomodates recognizing new devices, restructuring existing devices, and modifying device profile information

  1. Peaceful fusion

    Energy Technology Data Exchange (ETDEWEB)

    Englert, Matthias [IANUS, TU Darmstadt (Germany)

    2014-07-01

    Like other intense neutron sources fusion reactors have in principle a potential to be used for military purposes. Although the use of fissile material is usually not considered when thinking of fusion reactors (except in fusion-fission hybrid concepts) quantitative estimates about the possible production potential of future commercial fusion reactor concepts show that significant amounts of weapon grade fissile materials could be produced even with very limited amounts of source materials. In this talk detailed burnup calculations with VESTA and MCMATH using an MCNP model of the PPCS-A will be presented. We compare different irradiation positions and the isotopic vectors of the plutonium bred in different blankets of the reactor wall with the liquid lead-lithium alloy replaced by uranium. The technical, regulatory and policy challenges to manage the proliferation risks of fusion power will be addressed as well. Some of these challenges would benefit if addressed at an early stage of the research and development process. Hence, research on fusion reactor safeguards should start as early as possible and accompany the current research on experimental fusion reactors.

  2. Lasers and particle beam for fusion and strategic defense

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    This special issue of the Journal of Fusion Energy consists of the edited transscripts of a symposium on the applications of laser and particle beams to fusion and strategic defense. Its eleven papers discuss these topics: the Strategic Defense Initiative; accelerators for heavy ion fusion; rf accelerators for fusion and strategic defense; Pulsed power, ICF, and the Strategic Defense Initiative; chemical lasers; the feasibility of KrF lasers for fusion; the damage resistance of coated optic; liquid crystal devices for laser systems; fusion neutral-particle beam research and its contribution to the Star Wars program; and induction linacs and free electron laser amplifiers for ICF devices and directed-energy weapons

  3. Economic analysis of fusion breeders

    International Nuclear Information System (INIS)

    Delene, J.G.

    1985-01-01

    This paper presents a study of the economic performance of Fission/Fusion Hybrid devices. This work takes fusion breeder cost estimates and applies methodology and cost factors used in the fission reactor programs to compare fusion breeders with Liquid Metal Fast Breeder Reactors (LMFBR). The results of the analysis indicate that the Hybrid will be in the same competitive range as proposed LMFBRs and have the potential to provide economically competitive power in a future of rising uranium prices. The sensitivity of the results to variations in key parameters is included

  4. The management of fusion waste

    International Nuclear Information System (INIS)

    Hancox, R.; Butterworth, G.J.

    1990-01-01

    Fusion reactors based on the deuterium-tritium fuel cycle will generate radioactive waste as a result of neutron irradiation of the structural materials and absorption of the tritium fuel. An important issue is whether the volume of this waste and the risks associated with it can be reduced to a sufficiently low level that the environmental advantage of fusion can be maintained without incurring unacceptable additional costs. Information is presented on the radioactive waste expected from the decommissioning of three generations of fusion devices - the JET experiment, NET, and power reactors. The characteristics and probable volumes of this waste are considered, together with the risks associated with its disposal. (author)

  5. The management of fusion waste

    International Nuclear Information System (INIS)

    Hancox, R.; Butterworth, G.J.

    1991-01-01

    Fusion reactors based on the deuterium-tritium fuel cycle will generate radioactive waste as a result of neutron irradiation of the structural materials and absorption of the tritium fuel. An important issue is whether the volume of this waste and the risks associated with it can be reduced to a sufficiently low level that the environmental advantage of fusion can be maintained without incurring unacceptable additional costs. Information is presented on the radioactive waste expected from the decommissioning of three generations of fusion devices - the JET experiment, NET, and power reactors. The characteristics and probable volumes of this waste are considered, together with the risks associated with its disposal. (orig.)

  6. A new deflection technique applied to an existing scheme of electrostatic accelerator for high energy neutral beam injection in fusion reactor devices

    Science.gov (United States)

    Pilan, N.; Antoni, V.; De Lorenzi, A.; Chitarin, G.; Veltri, P.; Sartori, E.

    2016-02-01

    A scheme of a neutral beam injector (NBI), based on electrostatic acceleration and magneto-static deflection of negative ions, is proposed and analyzed in terms of feasibility and performance. The scheme is based on the deflection of a high energy (2 MeV) and high current (some tens of amperes) negative ion beam by a large magnetic deflector placed between the Beam Source (BS) and the neutralizer. This scheme has the potential of solving two key issues, which at present limit the applicability of a NBI to a fusion reactor: the maximum achievable acceleration voltage and the direct exposure of the BS to the flux of neutrons and radiation coming from the fusion reactor. In order to solve these two issues, a magnetic deflector is proposed to screen the BS from direct exposure to radiation and neutrons so that the voltage insulation between the electrostatic accelerator and the grounded vessel can be enhanced by using compressed SF6 instead of vacuum so that the negative ions can be accelerated at energies higher than 1 MeV. By solving the beam transport with different magnetic deflector properties, an optimum scheme has been found which is shown to be effective to guarantee both the steering effect and the beam aiming.

  7. A new deflection technique applied to an existing scheme of electrostatic accelerator for high energy neutral beam injection in fusion reactor devices

    Energy Technology Data Exchange (ETDEWEB)

    Pilan, N., E-mail: nicola.pilan@igi.cnr.it; Antoni, V.; De Lorenzi, A.; Chitarin, G.; Veltri, P.; Sartori, E. [Consorzio RFX—Associazione EURATOM-ENEA per la Fusione, Corso Stati Uniti 4, 35127 Padova (Italy)

    2016-02-15

    A scheme of a neutral beam injector (NBI), based on electrostatic acceleration and magneto-static deflection of negative ions, is proposed and analyzed in terms of feasibility and performance. The scheme is based on the deflection of a high energy (2 MeV) and high current (some tens of amperes) negative ion beam by a large magnetic deflector placed between the Beam Source (BS) and the neutralizer. This scheme has the potential of solving two key issues, which at present limit the applicability of a NBI to a fusion reactor: the maximum achievable acceleration voltage and the direct exposure of the BS to the flux of neutrons and radiation coming from the fusion reactor. In order to solve these two issues, a magnetic deflector is proposed to screen the BS from direct exposure to radiation and neutrons so that the voltage insulation between the electrostatic accelerator and the grounded vessel can be enhanced by using compressed SF{sub 6} instead of vacuum so that the negative ions can be accelerated at energies higher than 1 MeV. By solving the beam transport with different magnetic deflector properties, an optimum scheme has been found which is shown to be effective to guarantee both the steering effect and the beam aiming.

  8. Fusion Power measurement at ITER

    Energy Technology Data Exchange (ETDEWEB)

    Bertalot, L.; Barnsley, R.; Krasilnikov, V.; Stott, P.; Suarez, A.; Vayakis, G.; Walsh, M. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

    2015-07-01

    Nuclear fusion research aims to provide energy for the future in a sustainable way and the ITER project scope is to demonstrate the feasibility of nuclear fusion energy. ITER is a nuclear experimental reactor based on a large scale fusion plasma (tokamak type) device generating Deuterium - Tritium (DT) fusion reactions with emission of 14 MeV neutrons producing up to 700 MW fusion power. The measurement of fusion power, i.e. total neutron emissivity, will play an important role for achieving ITER goals, in particular the fusion gain factor Q related to the reactor performance. Particular attention is given also to the development of the neutron calibration strategy whose main scope is to achieve the required accuracy of 10% for the measurement of fusion power. Neutron Flux Monitors located in diagnostic ports and inside the vacuum vessel will measure ITER total neutron emissivity, expected to range from 1014 n/s in Deuterium - Deuterium (DD) plasmas up to almost 10{sup 21} n/s in DT plasmas. The neutron detection systems as well all other ITER diagnostics have to withstand high nuclear radiation and electromagnetic fields as well ultrahigh vacuum and thermal loads. (authors)

  9. Cold fusion

    International Nuclear Information System (INIS)

    Suh, Suk Yong; Sung, Ki Woong; Kang, Joo Sang; Lee, Jong Jik

    1995-02-01

    So called 'cold fusion phenomena' are not confirmed yet. Excess heat generation is very delicate one. Neutron generation is most reliable results, however, the records are erratic and the same results could not be repeated. So there is no reason to exclude the malfunction of testing instruments. The same arguments arise in recording 4 He, 3 He, 3 H, which are not rich in quantity basically. An experiment where plenty of 4 He were recorded is attached in appendix. The problem is that we are trying to search cold fusion which is permitted by nature or not. The famous tunneling effect in quantum mechanics will answer it, however, the most fusion rate is known to be negligible. The focus of this project is on the theme that how to increase that negligible fusion rate. 6 figs, 4 tabs, 1512 refs. (Author)

  10. Cold fusion

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Suk Yong; Sung, Ki Woong; Kang, Joo Sang; Lee, Jong Jik [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-02-01

    So called `cold fusion phenomena` are not confirmed yet. Excess heat generation is very delicate one. Neutron generation is most reliable results, however, the records are erratic and the same results could not be repeated. So there is no reason to exclude the malfunction of testing instruments. The same arguments arise in recording {sup 4}He, {sup 3}He, {sup 3}H, which are not rich in quantity basically. An experiment where plenty of {sup 4}He were recorded is attached in appendix. The problem is that we are trying to search cold fusion which is permitted by nature or not. The famous tunneling effect in quantum mechanics will answer it, however, the most fusion rate is known to be negligible. The focus of this project is on the theme that how to increase that negligible fusion rate. 6 figs, 4 tabs, 1512 refs. (Author).

  11. Laser fusion

    International Nuclear Information System (INIS)

    Ashby, D.E.T.F.

    1976-01-01

    A short survey is given on laser fusion its basic concepts and problems and the present theoretical and experimental methods. The future research program of the USA in this field is outlined. (WBU) [de

  12. Fusion energy

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    The efforts of the Chemical Technology Division in fusion energy include the areas of fuel handling, processing, and containment. Current studies are concerned largely with the development of vacuum pumps for fusion reactors and experiments and with development and evaluation of techniques for recovering tritium from solid or liquid breeding blankets. In addition, a small effort is devoted to support of the ORNL design of a major Tokamak experiment, The Next Step (TNS)

  13. Laser fusion

    International Nuclear Information System (INIS)

    Key, M.H.; Oxford Univ.

    1990-04-01

    The use of lasers to drive implosions for the purpose of inertially confined fusion is an area of intense activity where progress compares favourably with that made in magnetic fusion and there are significant prospects for future development. In this brief review the basic concept is summarised and the current status is outlined both in the area of laser technology and in the most recent results from implosion experiments. Prospects for the future are also considered. (author)

  14. Devices for launching 0.1-g projectiles to 150 km/s or more to initiate fusion. Part 2. Railgun accelerators

    International Nuclear Information System (INIS)

    Hawke, R.S.

    1979-01-01

    The possibility of using a railgun accelerator to launch 0.1-g projectiles to hypervelocities (150 km/s or more) to initiate thermonuclear fusion is studied. The analysis revealed that a railgun with a plasma-arc armature is a viable approach to the goal. When calculating the railgun's probable performance, it was discovered that this launch system might possibly be designed to avoid adverse effects from boundary layer drag. An appendix provided by A.C. Buckingham summarizes his calculations that predict the amount of erosive drag between projectile and rail. Finally, it was found that certain properties of railgun and projectile materials can impose operational limits. Using these limits, single- and multistage accelerators were designed. Within such limits, a railgun could accelerate a 0.1-g projectile to hypervelocities

  15. Devices for launching 0.1-g projectiles to 150 km/s or more to initiate fusion. Part 1. Magnetic-gradient and electrostatic accelerators

    International Nuclear Information System (INIS)

    Brittingham, J.N.

    1979-01-01

    The feasibility of using magnetic-gradient and electrostatic accelerators to launch a 0.1-g projectile to hypervelocities (150 km/s or more) is studied. Such hypervelocity projectiles could be used to ignite deuterium-tritium fuel pellets in a fusion reactor. For the magnetic-gradient accelerator, several types of projectile were studied: shielded and unshielded copper, ferromagnetic, and superconducting. The calculations revealed the superconducting projectile to be the best of those materials. It would require a 3.2-km-long magnetic-gradient accelerator and achieve a 92% efficiency. This accelerator-projectile combination would be the one most likely to launch a 0.1-g projectile to 150 km/s or more. Its components would cost $58.9 million. The electrostatic accelerator was found to be impractical because of its excessive length of 23 km

  16. Nuclear fusion

    International Nuclear Information System (INIS)

    Al-zaelic, M.M.

    2013-01-01

    Nuclear fusion can be relied on to solve the global energy crisis if the process of limiting the heat produced by the fusion reaction (Plasma) is successful. Currently scientists are progressively working on this aspect whereas there are two methods to limit the heat produced by fusion reaction, the two methods are auto-restriction using laser beam and magnetic restriction through the use of magnetic fields and research is carried out to improve these two methods. It is expected that at the end of this century the nuclear fusion energy will play a vital role in overcoming the global energy crisis and for these reasons, acquiring energy through the use of nuclear fusion reactors is one of the most urge nt demands of all mankind at this time. The conclusion given is that the source of fuel for energy production is readily available and inexpensive ( hydrogen atoms) and whole process is free of risks and hazards, especially to general health and the environment . Nuclear fusion importance lies in the fact that energy produced by the process is estimated to be about four to five times the energy produced by nuclear fission. (author)

  17. A hybrid electrochemical device based on a synergetic inner combination of Li ion battery and Li ion capacitor for energy storage.

    Science.gov (United States)

    Zheng, Jun-Sheng; Zhang, Lei; Shellikeri, Annadanesh; Cao, Wanjun; Wu, Qiang; Zheng, Jim P

    2017-02-07

    Li ion battery (LIB) and electrochemical capacitor (EC) are considered as the most widely used energy storage systems (ESSs) because they can produce a high energy density or a high power density, but it is a huge challenge to achieve both the demands of a high energy density as well as a high power density on their own. A new hybrid Li ion capacitor (HyLIC), which combines the advantages of LIB and Li ion capacitor (LIC), is proposed. This device can successfully realize a potential match between LIB and LIC and can avoid the excessive depletion of electrolyte during the charge process. The galvanostatic charge-discharge cycling tests reveal that at low current, the HyLIC exhibits a high energy density, while at high current, it demonstrates a high power density. Ragone plot confirms that this device can make a synergetic balance between energy and power and achieve a highest energy density in the power density range of 80 to 300 W kg -1 . The cycle life test proves that HyLIC exhibits a good cycle life and an excellent coulombic efficiency. The present study shows that HyLIC, which is capable of achieving a high energy density, a long cycle life and an excellent power density, has the potential to achieve the winning combination of a high energy and power density.

  18. Bridge between fusion plasma and plasma processing

    International Nuclear Information System (INIS)

    Ohno, Noriyasu; Takamura, Shuichi

    2008-01-01

    In the present review, relationship between fusion plasma and processing plasma is discussed. From boundary-plasma studies in fusion devices new applications such as high-density plasma sources, erosion of graphite in a hydrogen plasma, formation of helium bubbles in high-melting-point metals and the use of toroidal plasmas for plasma processing are emerging. The authors would like to discuss a possibility of knowledge transfer from fusion plasmas to processing plasmas. (T. Ikehata)

  19. EURATOM strategy towards fusion energy

    International Nuclear Information System (INIS)

    Varandas, C.

    2007-01-01

    Research and development (Research and Development) activities in controlled thermonuclear fusion have been carried out since the 60's of the last century aiming at providing a new clean, powerful, practically inexhaustive, safe, environmentally friend and economically attractive energy source for the sustainable development of our society.The EURATOM Fusion Programme (EFP) has the leadership of the magnetic confinement Research and Development activities due to the excellent results obtained on JET and other specialized devices, such as ASDEX-Upgrade, TORE SUPRA, FTU, TCV, TEXTOR, CASTOR, ISTTOK, MAST, TJ-II, W7-X, RFX and EXTRAP. JET is the largest tokamak in operation and the single device that can use deuterium and tritium mixes. It has produced 16 MW of fusion power, during 3 seconds, with an energy amplification of 0.6. The next steps of the EFP strategy towards fusion energy are ITER complemented by a vigorous Accompanying Programme, DEMO and a prototype of a fusion power plant. ITER, the first experimental fusion reactor, is a large-scale project (35-year duration, 10000 MEuros budget), developed in the frame of a very broad international collaboration, involving EURATOM, Japan, Russia Federation, United States of America, Korea, China and India. ITER has two main objectives: (i) to prove the scientific and technical viability of fusion energy by producing 500 MW, during 300 seconds and a energy amplification between 10 and 20; and (ii) to test the simultaneous and integrated operation of the technologies needed for a fusion reactor. The Accompanying Programme aims to prepare the ITER scientific exploitation and the DEMO design, including the development of the International Fusion Materials Irradiation Facility (IFMIF). A substantial part of this programme will be carried out in the frame of the Broader Approach, an agreement signed by EURATOM and Japan. The main goal of DEMO is to produce electricity, during a long time, from nuclear fusion reactions. The

  20. Data security on the national fusion grid

    International Nuclear Information System (INIS)

    Burruss, Justine R.; Fredian, Tom W.; Thompson, Mary R.

    2005-01-01

    The National Fusion Collaboratory project is developing and deploying new distributed computing and remote collaboration technologies with the goal of advancing magnetic fusion energy research. This work has led to the development of the US Fusion Grid (FusionGrid), a computational grid composed of collaborative, compute, and data resources from the three large US fusion research facilities and with users both in the US and in Europe. Critical to the development of FusionGrid was the creation and deployment of technologies to ensure security in a heterogeneous environment. These solutions to the problems of authentication, authorization, data transfer, and secure data storage, as well as the lessons learned during the development of these solutions, may be applied outside of FusionGrid and scale to future computing infrastructures such as those for next-generation devices like ITER

  1. Security on the US Fusion Grid

    International Nuclear Information System (INIS)

    Burruss, Justin R.; Fredian, Tom W.; Thompson, Mary R.

    2005-01-01

    The National Fusion Collaboratory project is developing and deploying new distributed computing and remote collaboration technologies with the goal of advancing magnetic fusion energy research. This work has led to the development of the US Fusion Grid (FusionGrid), a computational grid composed of collaborative, compute, and data resources from the three large US fusion research facilities and with users both in the US and in Europe. Critical to the development of FusionGrid was the creation and deployment of technologies to ensure security in a heterogeneous environment. These solutions to the problems of authentication, authorization, data transfer, and secure data storage, as well as the lessons learned during the development of these solutions, may be applied outside of FusionGrid and scale to future computing infrastructures such as those for next-generation devices like ITER

  2. Security on the US Fusion Grid

    Energy Technology Data Exchange (ETDEWEB)

    Burruss, Justin R.; Fredian, Tom W.; Thompson, Mary R.

    2005-06-01

    The National Fusion Collaboratory project is developing and deploying new distributed computing and remote collaboration technologies with the goal of advancing magnetic fusion energy research. This work has led to the development of the US Fusion Grid (FusionGrid), a computational grid composed of collaborative, compute, and data resources from the three large US fusion research facilities and with users both in the US and in Europe. Critical to the development of FusionGrid was the creation and deployment of technologies to ensure security in a heterogeneous environment. These solutions to the problems of authentication, authorization, data transfer, and secure data storage, as well as the lessons learned during the development of these solutions, may be applied outside of FusionGrid and scale to future computing infrastructures such as those for next-generation devices like ITER.

  3. Data security on the national fusion grid

    Energy Technology Data Exchange (ETDEWEB)

    Burruss, Justine R.; Fredian, Tom W.; Thompson, Mary R.

    2005-06-01

    The National Fusion Collaboratory project is developing and deploying new distributed computing and remote collaboration technologies with the goal of advancing magnetic fusion energy research. This work has led to the development of the US Fusion Grid (FusionGrid), a computational grid composed of collaborative, compute, and data resources from the three large US fusion research facilities and with users both in the US and in Europe. Critical to the development of FusionGrid was the creation and deployment of technologies to ensure security in a heterogeneous environment. These solutions to the problems of authentication, authorization, data transfer, and secure data storage, as well as the lessons learned during the development of these solutions, may be applied outside of FusionGrid and scale to future computing infrastructures such as those for next-generation devices like ITER.

  4. Security on the US fusion grid

    International Nuclear Information System (INIS)

    Burruss, J.R.; Fredian, T.W.; Thompson, M.R.

    2006-01-01

    The National Fusion Collaboratory project is developing and deploying new distributed computing and remote collaboration technologies with the goal of advancing magnetic fusion energy research. This has led to the development of the U.S. fusion grid (FusionGrid), a computational grid composed of collaborative, compute, and data resources from the three large U.S. fusion research facilities and with users both in the U.S. and in Europe. Critical to the development of FusionGrid was the creation and deployment of technologies to ensure security in a heterogeneous environment. These solutions to the problems of authentication, authorization, data transfer, and secure data storage, as well as the lessons learned during the development of these solutions, may be applied outside of FusionGrid and scale to future computing infrastructures such as those for next-generation devices like ITER

  5. Advanced lasers for fusion

    International Nuclear Information System (INIS)

    Krupke, W.F.; George, E.V.; Haas, R.A.

    1979-01-01

    Laser drive systems' performance requirements for fusion reactors are developed following a review of the principles of inertial confinement fusion and of the technical status of fusion research lasers (Nd:glass; CO 2 , iodine). These requirements are analyzed in the context of energy-storing laser media with respect to laser systems design issues: optical damage and breakdown, medium excitation, parasitics and superfluorescence depumping, energy extraction physics, medium optical quality, and gas flow. Three types of energy-storing laser media of potential utility are identified and singled out for detailed review: (1) Group VI atomic lasers, (2) rare earth solid state hybrid lasers, and (3) rare earth molecular vapor lasers. The use of highly-radiative laser media, particularly the rare-gas monohalide excimers, are discussed in the context of short pulse fusion applications. The concept of backward wave Raman pulse compression is considered as an attractive technique for this purpose. The basic physics and device parameters of these four laser systems are reviewed and conceptual designs for high energy laser systems are presented. Preliminary estimates for systems efficiencies are given. (Auth.)

  6. Identification of future engineering-development needs of alternative concepts for magnetic-fusion energy

    International Nuclear Information System (INIS)

    Krakowski, R.A.

    1982-01-01

    A qualitative identification of future engineering needs of alternative fusion concepts (AFCs) is presented. These needs are assessed relative to the similar needs of the tokamak in order to emphasize differences in required technology with respect to the well documented mainline approach. Although nearly thirty AFCs can be identified as being associated with some level of reactor projection, redirection, refocusing, and general similarities can be used to generate a reduced AFC list that includes only the bumpy tori, stellarators, reversed-field pinches, and compact toroids. Furthermore, each AFC has the potential of operating as a conventional (low power density, superconducting magnets) or a compact, high-power-density (HPD) system. Hence, in order to make tractable an otherwise difficult task, the future engineering needs for the AFCs are addressed here for conventional versus compact approaches, with the latter being treated as a generic class and the former being composed of bumpy tori, stellarators, reversed-field pinches, and compact toroids

  7. General characteristics and assessment of the scientific/technical feasibility of the next major device in the tokamak fusion program. Summary of the US contributions to the INTOR workshop

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Gilleland, J.R.; Kulcinski, G.L.; Rutherford, P.H.

    1979-09-01

    A substantial physics and technology data base for INTOR exists today, and this data base will be expanded over the next few years by currently planned programs. However, certain crucial information will not be developed by currently planned programs. Much of this missing information could be developed on the INTOR time scale by the expansion and/or acceleration of existing R and D programs and by the establishment of new R and D programs. On this basis, it is concluded that it is scientifically and technologically feasible to undertake the construction of an INTOR-like device to operate in the early 1990s, provided that the supporting R and D effort is expanded immediately to provide an adequate data base within the next few years in a few critical areas. Furthermore, it is concluded that the construction of an INTOR-like device to operate in the early 1990s is the appropriate next major step in the development of fusion power

  8. Characterization of a segmented plasma torch assisted High Heat Flux (HHF) system for performance evaluation of plasma facing components in fusion devices

    International Nuclear Information System (INIS)

    Ngangom, Aomoa; Sarmah, Trinayan; Sah, Puspa; Kakati, Mayur; Ghosh, Joydeep

    2015-01-01

    A wide variety of high heat and particle flux test facilities are being used by the fusion community to evaluate the thermal performance of plasma facing materials/components, which includes electron beam, ion beam, neutral beam and thermal plasma assisted sources. In addition to simulate heat loads, plasma sources have the additional advantage of reproducing exact fusion plasma like conditions, in terms of plasma density, temperature and particle flux. At CPP-IPR, Assam, we have developed a high heat and particle flux facility using a DC, non-transferred, segmented thermal plasma torch system, which can produce a constricted, stabilized plasma jet with high ion density. In this system, the plasma torch exhausts into a low pressure chamber containing the materials to be irradiated, which produces an expanded plasma jet with more uniform profiles, compared to plasma torches operated at atmospheric pressure. The heat flux of the plasma beam was studied by using circular calorimeters of different diameters (2 and 3 cm) for different input power (5-55 kW). The effect of the change in gas (argon) flow rate and mixing of gases (argon + hydrogen) was also studied. The heat profile of the plasma beam was also studied by using a pipe calorimeter. From this, the radial heat flux was calculated by using Abel inversion. It is seen that the required heat flux of 10 MW/m 2 is achievable in our system for pure argon plasma as well as for plasma with gas mixtures. The plasma parameters like the temperature, density and the beam velocity were studied by using optical emission spectroscopy. For this, a McPherson made 1.33 meter focal length spectrometer; model number 209, was used. A plane grating with 1800 g/mm was used which gave a spectral resolution of 0.007 nm. A detailed characterization with respect to these plasma parameters for different gas (argon) flow rate and mixing of gases (argon+hydrogen) for different input power will be presented in this paper. The plasma

  9. Tritium accountancy in fusion systems

    Energy Technology Data Exchange (ETDEWEB)

    Klein, J.E.; Clark, E.A.; Harvel, C.D.; Farmer, D.A.; Tovo, L.L.; Poore, A.S. [Savannah River National Laboratory, Aiken, SC (United States); Moore, M.L. [Savannah River Nuclear Solutions, Aiken, SC (United States)

    2015-03-15

    The US Department of Energy (DOE) has clearly defined requirements for nuclear material control and accountability (MCA) of tritium whereas the International Atomic Energy Agency (IAEA) does not since tritium is not a fissile material. MCA requirements are expected for tritium fusion machines and will be dictated by the host country or regulatory body where the machine is operated. Material Balance Areas (MBA) are defined to aid in the tracking and reporting of nuclear material movements and inventories. Material sub-accounts (MSA) are established along with key measurement points (KMP) to further subdivide a MBA to localize and minimize uncertainties in the inventory difference (ID) calculations for tritium accountancy. Fusion systems try to minimize tritium inventory which may require continuous movement of material through the MSA. The ability of making meaningful measurements of these material transfers is described in terms of establishing the MSA structure to perform and reconcile ID calculations. For fusion machines, changes to the traditional ID equation will be discussed which includes breeding, burn-up, and retention of tritium in the fusion device. The concept of 'net' tritium quantities consumed or lost in fusion devices is described in terms of inventory taking strategies and how it is used to track the accumulation of tritium in components or fusion machines. (authors)

  10. Survey of fusion reactor technology

    International Nuclear Information System (INIS)

    Chung, M.K.; Kang, H.D.; Oh, Y.K.; Lee, K.W.; In, S.Y.; Kim, Y.C.

    1983-01-01

    The present object of the fusion research is to accomplish the scientific break even by the year of 1986. In view of current progress in the field of Fusion reactor development, we decided to carry out the conceptual design of Tokamak-type fusion reactor during the year of 82-86 in order to acquire the principles of the fusion devices, find the engineering problems and establish the basic capabilities to develop the key techniques with originality. In this year the methods for calculating the locations of the poloidal coils and distribution of the magnetic field, which is one of the most essential and complicated task in the fusion reactor design works, were established. Study on the optimization of the design method of toroidal field coil was also done. Through this work, we established the logic for the design of the toroidal field coil in tokamak and utilize this technique to the design of small compact tokamak. Apart from the development work as to the design technology of tokamak, accelerating column and high voltage power supply (200 KVDC, 100 mA) for intense D-T neutron generator were constructed and now beam transport systems are under construction. This device will be used to develop the materials and the components for the tokamak fusion reactor. (Author)

  11. Cold fusion

    International Nuclear Information System (INIS)

    Koster, J.

    1989-01-01

    In this contribution the author the phenomenom of so-called cold fusion, inspired by the memorable lecture of Moshe Gai on his own search for this effect. Thus much of what follows was presented by Dr. Gai; the rest is from independent reading. What is referred to as cold fusion is of course the observation of possible products of deuteron-deuteron (d-d) fusion within deuterium-loaded (dentended) electrodes. The debate over the two vanguard cold fusion experiments has raged under far more public attention than usually accorded new scientific phenomena. The clamor commenced with the press conference of M. Fleishmann and S. Pons on March 23, 1989 and the nearly simultaneous wide circulation of a preprint of S. Jones and collaborators. The majority of work attempting to confirm these observations has at the time of this writing yet to appear in published form, but contributions to conferences and electronic mail over computer networks were certainly filled with preliminary results. To keep what follows to a reasonable length the author limit this discussion to the searches for neutron (suggested by ref. 2) or for excessive heat production (suggested by ref. 1), following a synopsis of the hypotheses of cold fusion

  12. Controlled thermonuclear fusion. Present state and prospective

    International Nuclear Information System (INIS)

    Consoli, T.

    1976-01-01

    The interest of thermonuclear fusion for energy production is underlined. The present state of the research in this field is presented, emphasis being given to Tokamak configurations. The problems concerning confinement and additional heating in these devices are presented [fr

  13. Asymmetric Supercapacitor Electrodes and Devices.

    Science.gov (United States)

    Choudhary, Nitin; Li, Chao; Moore, Julian; Nagaiah, Narasimha; Zhai, Lei; Jung, Yeonwoong; Thomas, Jayan

    2017-06-01

    The world is recently witnessing an explosive development of novel electronic and optoelectronic devices that demand more-reliable power sources that combine higher energy density and longer-term durability. Supercapacitors have become one of the most promising energy-storage systems, as they present multifold advantages of high power density, fast charging-discharging, and long cyclic stability. However, the intrinsically low energy density inherent to traditional supercapacitors severely limits their widespread applications, triggering researchers to explore new types of supercapacitors with improved performance. Asymmetric supercapacitors (ASCs) assembled using two dissimilar electrode materials offer a distinct advantage of wide operational voltage window, and thereby significantly enhance the energy density. Recent progress made in the field of ASCs is critically reviewed, with the main focus on an extensive survey of the materials developed for ASC electrodes, as well as covering the progress made in the fabrication of ASC devices over the last few decades. Current challenges and a future outlook of the field of ASCs are also discussed. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  14. New trends in fusion research

    CERN Multimedia

    CERN. Geneva

    2004-01-01

    The efforts of the international fusion community aim at demonstrating the scientific feasibility of thermonuclear fusion energy power plants. Understanding the behavior of burning plasmas, i.e. plasmas with strong self-heating, represents a primary scientific challenge for fusion research and a new science frontier. Although integrated studies will only be possible, in new, dedicated experimental facilities, such as the International Tokamak Experimental Reactor (ITER), present devices can address specific issues in regimes relevant to burning plasmas. Among these are an improvement of plasma performance via a reduction of the energy and particle transport, an optimization of the path to ignition or to sustained burn using additional heating and a control of plasma-wall interaction and energy and particle exhaust. These lectures address recent advances in plasma science and technology that are relevant to the development of fusion energy. Mention will be made of the inertial confinement line of research, but...

  15. Overview of fusion reactor safety

    International Nuclear Information System (INIS)

    Cohen, S.; Crocker, J.G.

    1981-01-01

    Use of deuterium-tritium burning fusion reactors requires examination of several major safety and environmental issues: (1) tritium inventory control, (2) neutron activation of structural materials, fluid streams and reactor hall environment, (3) release of radioactivity from energy sources including lithium spill reactions, superconducting magnet stored energy release, and plasma disruptions, (4) high magnetic and electromagnetic fields associated with fusion reactor superconducting magnets and radio frequency heating devices, and (5) handling and disposal of radioactive waste. Early recognition of potential safety problems with fusion reactors provides the opportunity for improvement in design and materials to eliminate or greatly reduce these problems. With an early start in this endeavor, fusion should be among the lower risk technologies for generation of commercial electrical power

  16. Fusion events

    International Nuclear Information System (INIS)

    Aboufirassi, M; Angelique, J.C.; Bizard, G.; Bougault, R.; Brou, R.; Buta, A.; Colin, J.; Cussol, D.; Durand, D.; Genoux-Lubain, A.; Horn, D.; Kerambrun, A.; Laville, J.L.; Le Brun, C.; Lecolley, J.F.; Lefebvres, F.; Lopez, O.; Louvel, M.; Meslin, C.; Metivier, V.; Nakagawa, T.; Peter, J.; Popescu, R.; Regimbart, R.; Steckmeyer, J.C.; Tamain, B.; Vient, E.; Wieloch, A.; Yuasa-Nakagawa, K.

    1998-01-01

    The fusion reactions between low energy heavy ions have a very high cross section. First measurements at energies around 30-40 MeV/nucleon indicated no residue of either complete or incomplete fusion, thus demonstrating the disappearance of this process. This is explained as being due to the high amount o energies transferred to the nucleus, what leads to its total dislocation in light fragments and particles. Exclusive analyses have permitted to mark clearly the presence of fusion processes in heavy systems at energies above 30-40 MeV/nucleon. Among the complete events of the Kr + Au reaction at 60 MeV/nucleon the majority correspond to binary collisions. Nevertheless, for the most considerable energy losses, a class of events do occur for which the detected fragments appears to be emitted from a unique source. These events correspond to an incomplete projectile-target fusion followed by a multifragmentation. Such events were singled out also in the reaction Xe + Sn at 50 MeV/nucleon. For the events in which the energy dissipation was maximal it was possible to isolate an isotropic group of events showing all the characteristics of fusion nuclei. The fusion is said to be incomplete as pre-equilibrium Z = 1 and Z = 2 particles are emitted. The cross section is of the order of 25 mb. Similar conclusions were drown for the systems 36 Ar + 27 Al and 64 Zn + nat Ti. A cross section value of ∼ 20 mb was determined at 55 MeV/nucleon in the first case, while the measurement of evaporation light residues in the last system gave an upper limit of 20-30 mb for the cross section at 50 MeV/nucleon

  17. Fusion through the NET

    International Nuclear Information System (INIS)

    Spears, B.

    1987-01-01

    The paper concerns the next generation of fusion machines which are intended to demonstrate the technical viability of fusion. In Europe, the device that will follow on from JET is known as NET - the Next European Torus. If the design programme for NET proceeds, Europe could start to build the machine in 1994. The present JET programme hopes to achieve breakeven in the early 1990's. NET hopes to reach ignition in the next century, and so lay the foundation for a demonstration reactor. A description is given of the technical specifications of the components of NET, including: the first wall, the divertors to protect the wall, the array of magnets that provide the fields containing the plasma, the superconducting magnets, and the shield of the machine. NET's research programme is briefly outlined, including the testing programme to optimise conditions in the machine to achieve ignition, and its safety work. (U.K.)

  18. Complications of Lumbar Artificial Disc Replacement Compared to Fusion: Results From the Prospective, Randomized, Multicenter US Food and Drug Administration Investigational Device Exemption Study of the Charité Artificial Disc

    Science.gov (United States)

    Majd, Mohammed E.; Isaza, Jorge E.; Blumenthal, Scott L.; McAfee, Paul C.; Guyer, Richard D.; Hochschuler, Stephen H.; Geisler, Fred H.; Garcia, Rolando; Regan, John J.

    2007-01-01

    Background Previous reports of lumbar total disc replacement (TDR) have described significant complications. The US Food and Drug Administration (FDA) investigational device exemption (IDE) study of the Charité artificial disc represents the first level I data comparison of TDR to fusion. Methods In the prospective, randomized, multicenter IDE study, patients were randomized in a 2:1 ratio, with 205 patients in the Charité group and 99 patients in the control group (anterior lumbar interbody fusion [ALIF] with BAK cages). Inclusion criteria included confirmed single-level degenerative disc disease at L4-5 or L5-S1 and failure of nonoperative treatment for at least 6 months. Complications were reported throughout the study. Results The rate of approach-related complications was 9.8% in the investigational group and 10.1% in the control group. The rate of major neurological complications was similar between the 2 groups (investigational = 4.4%, control = 4.0%). There was a higher rate of superficial wound infection in the investigational group but no deep wound infections in either group. Pseudarthrosis occurred in 9.1% of control group patients. The rate of subsidence in the investigational group was 3.4%. The reoperation rate was 5.4% in the investigational group and 9.1% in the control group. Conclusions The incidence of perioperative and postoperative complications for lumbar TDR was similar to that of ALIF. Vigilance is necessary with respect to patient indications, training, and correct surgical technique to maintain TDR complications at the levels experienced in the IDE study. PMID:25802575

  19. A conceptual design study of a reversed field pinch fusion reactor

    International Nuclear Information System (INIS)

    Kondo, S.; Tanaka, S.; Terai, T.; Hashizume, H.

    1989-01-01

    A conceptual design of a Reversed-Field Pinch (RFP) fusion reactor with a solid breeder blanket REPUTER-1 has been studied through parametric system studies and detailed design and analysis in order to clarify the technical feasibility of a compact fusion reactor. F-θ pumping is used for driving the plasma current necessary for steady state operation. A maintenance policy of replacing a whole fusion power core including TF coils is proposed to cope with the requirements of high wall loading and high mass power density. For the same reason a normal conductor is selected for most of the coils. The first wall is structurally independent of the blanket. The blanket module is composed of SiC reinforced blocks which form a stable arch so as to keep the stresses in SiC basically compressive. The coolant for the first wall and the limiter is pressurized water, while the coolant for the blanket is helium gas. A number of thin Li 2 O and thick beryllium tiles are packed into the blanket block so as to obtain a proper tritium breeding ratio. A pumped limiter is chosen for the plasma exhaust system. The study has shown the technical feasibility of a high power density fusion power reactor (330 kWe/tonne) with solid breeder blanket and many key physics and engineering issues are also clarified. (orig.)

  20. What fusion means to Canada

    International Nuclear Information System (INIS)

    Bolton, R.A.

    1983-06-01

    Fusion can and will play an ever-increasing role in the energy balance once it has been brought on line. Taming of this technology and the maturing processes of engineering and economic feasibility will proceed at a rate which depends very strongly upon international and collective national wills to see it through. Large experimental devices, particularly of the tokamak type, are now being completed; their performance should give a very good idea of the scientific feasibility. The next-stage devices are at the pre-proposal and proposal stages but are not yet approved, even in principle. An improved general economic climate sustained for a few years would certainly help re-establish the momentum of world international efforts in fusion. This paper gives an overview of fusion research on a world scale and details of the particular aspects that Canada has chosen to pursue

  1. Nuclear fusion: power for the next century

    International Nuclear Information System (INIS)

    1980-05-01

    The basis of fusion reactions is outlined, with special reference to deuterium and tritium (from lithium, by neutron reaction) as reactants, and the state of research worldwide is indicated. The problems inherent in fusion reactions are discussed, plasma is defined, and the steps to be taken to generate electricity from controlled nuclear fusion are stated. Methods of plasma heating and plasma confinement are considered, leading to a description of the tokamak plasma confinement system. Devices under construction include the JET (Joint European Torus) Undertaking in the UK. Plans and possibilities for fusion reactors are discussed. (U.K.)

  2. Fusion Concept Exploration Experiments at PPPL

    International Nuclear Information System (INIS)

    Stewart Zweben; Samuel Cohen; Hantao Ji; Robert Kaita; Richard Majeski; Masaaki Yamada

    1999-01-01

    Small ''concept exploration'' experiments have for many years been an important part of the fusion research program at the Princeton Plasma Physics Laboratory (PPPL). this paper describes some of the present and planned fusion concept exploration experiments at PPPL. These experiments are a University-scale research level, in contrast with the larger fusion devices at PPPL such as the National Spherical Torus Experiment (NSTX) and the Tokamak Fusion Test Reactor (TFTR), which are at ''proof-of-principle'' and ''proof-of-performance'' levels, respectively

  3. Nuclear fusion

    International Nuclear Information System (INIS)

    Huber, H.

    1978-01-01

    A comprehensive survey is presented of the present state of knowledge in nuclear fusion research. In the first part, potential thermonuclear reactions, basic energy balances of the plasma (Lawson criterion), and the main criteria to be observed in the selection of appropriate thermonuclear reactions are dealt with. This is followed by a discussion of the problems encountered in plasma physics (plasma confinement and heating, transport processes, plasma impurities, plasma instabilities and plasma diagnostics) and by a consideration of the materials problems involved, such as material of the first wall, fuel inlet and outlet, magnetic field generation, as well as repair work and in-service inspections. Two main methods have been developed to tackle these problems: reactor concepts using the magnetic pinch (stellarator, Tokamak, High-Beta reactors, mirror machines) on the one hand, and the other concept using the inertial confinement (laser fusion reactor). These two approaches and their specific problems as well as past, present and future fusion experiments are treated in detail. The last part of the work is devoted to safety and environmental aspects of the potential thermonuclear aspects of the potential thermonuclear reactor, discussing such problems as fusion-specific hazards, normal operation and potential hazards, reactor incidents, environmental pollution by thermal effluents, radiological pollution, radioactive wastes and their disposal, and siting problems. (orig./GG) [de

  4. Short fusion

    CERN Multimedia

    2002-01-01

    French and UK researchers are perfecting a particle accelerator technique that could aid the quest for fusion energy or make X-rays that are safer and produce higher-resolution images. Led by Dr Victor Malka from the Ecole Nationale Superieure des Techniques Avancees in Paris, the team has developed a better way of accelerating electrons over short distances (1 page).

  5. Magnetic fusion

    International Nuclear Information System (INIS)

    2002-01-01

    This document is a detailed lecture on thermonuclear fusion. The basic physics principles are recalled and the technological choices that have led to tokamaks or stellarators are exposed. Different aspects concerning thermonuclear reactors such as safety, economy and feasibility are discussed. Tore-supra is described in details as well as the ITER project

  6. Cold fusion

    International Nuclear Information System (INIS)

    Seo, Suk Yong; You, Jae Jun

    1996-01-01

    Nearly every technical information is chased in the world. All of them are reviewed and analyzed. Some of them are chosen to study further more to review every related documents. And a probable suggestion about the excitonic process in deuteron absorbed condensed matter is proposed a way to cold fusion. 8 refs. (Author)

  7. Collaborations in fusion research

    International Nuclear Information System (INIS)

    Barnes, D.; Davis, S.; Roney, P.

    1995-01-01

    This paper reviews current experimental collaborative efforts in the fusion community and extrapolates to operational scenarios for the Tokamak Physics Experiment (TPX) and the International Thermonuclear Experimental Reactor (ITER). Current requirements, available technologies and tools, and problems, issues and concerns are discussed. This paper specifically focuses on the issues that apply to experimental operational collaborations. Special requirements for other types of collaborations, such as theoretical or design and construction efforts, will not be addressed. Our current collaborative efforts have been highly successful, even though the tools in use will be viewed as primitive by tomorrow's standards. An overview of the tools and technologies in today's collaborations can be found in the first section of this paper. The next generation of fusion devices will not be primarily institutionally based, but will be national (TPX) and international (ITER) in funding, management, operation and in ownership of scientific results. The TPX will present the initial challenge of real-time remotely distributed experimental data analysis for a steady state device. The ITER will present new challenges with the possibility of several remote control rooms all participating in the real-time operation of the experimental device. A view to the future of remote collaborations is provided in the second section of this paper

  8. Feasibility of a laser or charged-particle-beam fusion-reactor concept with direct electric generation by magnetic-flux compression

    International Nuclear Information System (INIS)

    Lasche, G.P.

    1983-06-01

    A new concept for an inertial-confinement fusion reactor is described which, because of its fundamentally different approach to blanket geometry and energy conversion, makes possible a unique combination of high efficiency, high power density, and low radioactivity. The conventional blanket is replaced with a liquid-density mass of lithium contiguously surrounding the fusion yield. This compact blanket configuration produces the maximum shock-induced kinetic energy in liquid metal and the maximum neutron absorption per unit mass. The shock-induced kinetic energy of the liquid lithium is converted directly to electricity with high efficiency by work done against a pulsed normal-conducting magnetic field applied to the exterior of the lithium

  9. Cold fusion, Alchemist's dream

    International Nuclear Information System (INIS)

    Clayton, E.D.

    1989-09-01

    In this report the following topics relating to cold fusion are discussed: muon catalysed cold fusion; piezonuclear fusion; sundry explanations pertaining to cold fusion; cosmic ray muon catalysed cold fusion; vibrational mechanisms in excited states of D 2 molecules; barrier penetration probabilities within the hydrogenated metal lattice/piezonuclear fusion; branching ratios of D 2 fusion at low energies; fusion of deuterons into 4 He; secondary D+T fusion within the hydrogenated metal lattice; 3 He to 4 He ratio within the metal lattice; shock induced fusion; and anomalously high isotopic ratios of 3 He/ 4 He

  10. Magnetic fusion; La fusion magnetique

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

    This document is a detailed lecture on thermonuclear fusion. The basic physics principles are recalled and the technological choices that have led to tokamaks or stellarators are exposed. Different aspects concerning thermonuclear reactors such as safety, economy and feasibility are discussed. Tore-supra is described in details as well as the ITER project.

  11. Process and device for decontamination of the waste gas of the fuel circuit of a fusion reactor from tritium and/or deuterium in waste gas containing them in chemically bound form

    International Nuclear Information System (INIS)

    Penzhorn, R.D.; Glugla, M.

    1987-01-01

    The invention concerns a process and a device for the decontamination of the wate gases of the fuel circuit of a fusion reactor from tritum and/or deuterium in waste gas containing them in chemically bound form, in which the waste gas is taken over an oxidation catalyst and then over a hot metal bed, tritium and/or deuterium is released from its compounds, separated from the waste gas and is returned to the fuel circuit. The process is intended to prevent losses of tritum and/or deuterium by permeation and the high loading of the hot metal getter materials, as occurs in the previously known corresponding process, and to avoid the formation of nitrogen oxides. This is achieved by: a) The catalytic oxidation reaction being carried out at a temperature of 200 0 C to 300 0 C. b) The gas mixture then being brought into contact with a hot metal bed at 200 0 C to 300 0 C to remove the remaining O 2 and for the selective conversion of the proportion of water into the hydrogen isotope. c) The gas mixture being brought into contact with a diaphragm made of palladium or a palladium-silver alloy at 400 0 C to 450 0 C to decompose the ammonia, all the released hydrogen isotope being passed through the diaphragm, separated from the remaining waste gas flow and removed. (orig.) [de

  12. Splenogonadal Fusion

    Directory of Open Access Journals (Sweden)

    Sung-Lang Chen

    2008-11-01

    Full Text Available Splenogonadal fusion (SGF is a rare congenital non-malignant anomaly characterized by fusion of splenic tissue to the gonad, and can be continuous or discontinuous. Very few cases have been diagnosed preoperatively, and many patients who present with testicular swelling undergo unnecessary orchiectomy under the suspicion of testicular neoplasm. A 16-year-old boy presented with a left scrotal mass and underwent total excision of a 1.6-cm tumor without damaging the testis, epididymis or its accompanying vessels. Pathologic examination revealed SFG (discontinuous type. If clinically suspected before surgery, the diagnosis may be confirmed by Tc-99m sulfur colloid imaging, which shows uptake in both the spleen and accessory splenic tissue within the scrotum. Frozen section should be considered if there remains any doubt regarding the diagnosis during operation.

  13. Line voltage distortions due to operation of the power supply devices required for plasma heating and magnetic field generation in the W7X thermonuclear fusion experiment

    International Nuclear Information System (INIS)

    Werner, F.

    1997-03-01

    The operation of the W7-X plasma heating devices requires high voltage DC power supplies with a total electrical power of 40 MVA. For this purpose twelve-pulse AC/DC converters are projected. These converters enforce a non sinusoidal line current, whose harmonics are causing corresponding line voltage distortions. To evaluate the extent of these distortions, the reaction of the harmonic currents on the AC line, is investigated by numerical network analysis. This is done for both, the 20 kV-junction point of the converters and the 110 kV-line terminal of the electricity supply company. Furthermore the design of LC series-resonant circuits, projected for power factor correction and damping of the harmonic content of the line voltage, has been verified. The additional operation of the 1.5 MVA magnet power supplies also contributes, even though to a much smaller extent, to the line voltage distortion. The influence of these twelve-pulse AC/DC converters was investigated too. The numerical calculations have been done with the aid of the network simulation program 'Pspice'. In an equivalent circuit the transmission line network and the transformers are represented by their inductances respectively equivalent inductances. The rectifier units are simulated by a number of current sources, producing the current harmonics in amplitude, frequency and phase. The harmonics amplitudes of the plasma heating power supplies are frequency and phase. The harmonics amplitudes of the plasma heating power supplies are measured values given by the manufacturer. For the magnet power supplies, the harmonics are derived from the theoretical step like I(t) current shape by Fourier series decomposition. Due to the action of the LC circuits the achieved characteristic voltage quality values are far below the permissible values corresponding to the recommendations of VDE 0160. (orig.) [de

  14. Laser fusion

    International Nuclear Information System (INIS)

    Eliezer, S.

    1982-02-01

    In this paper, the physics of laser fusion is described on an elementary level. The irradiated matter consists of a dense inner core surrounded by a less dense plasma corona. The laser radiation is mainly absorbed in the outer periphery of the plasma. The absorbed energy is transported inward to the ablation surface where plasma flow is created. Due to this plasma flow, a sequence of inward going shock waves and heat waves are created, resulting in the compression and heating of the core to high density and temperature. The interaction physics between laser and matter leading to thermonuclear burn is summarized by the following sequence of events: Laser absorption → Energy transport → Compression → Nuclear Fusion. This scenario is shown in particular for a Nd:laser with a wavelength of 1 μm. The wavelength scaling of the physical processes is also discussed. In addition to the laser-plasma physics, the Nd high power pulsed laser is described. We give a very brief description of the oscillator, the amplifiers, the spatial filters, the isolators and the diagnostics involved. Last, but not least, the concept of reactors for laser fusion and the necessary laser system are discussed. (author)

  15. Fusion spectroscopy

    International Nuclear Information System (INIS)

    Peacock, N.J.

    1995-09-01

    This article traces developments in the spectroscopy of high temperature laboratory plasma used in controlled fusion research from the early 1960's until the present. These three and a half decades have witnessed many orders of magnitude increase in accessible plasma parameters such as density and temperature as well as particle and energy confinement timescales. Driven by the need to interpret the radiation in terms of the local plasma parameters, the thrust of fusion spectroscopy has been to develop our understanding of (i) the atomic structure of highly ionised atoms, usually of impurities in the hydrogen isotope fuel; (ii) the atomic collision rates and their incorporation into ionization structure and emissivity models that take into account plasma phenomena like plasma-wall interactions, particle transport and radiation patterns; (iii) the diagnostic applications of spectroscopy aided by increasingly sophisticated characterisation of the electron fluid. These topics are discussed in relation to toroidal magnetically confined plasmas, particularly the Tokamak which appears to be the most promising approach to controlled fusion to date. (author)

  16. Fusion engineering. Vol. 2

    International Nuclear Information System (INIS)

    Young, N.E.; Pinter, G.R.; Spinos, F.R.

    1983-01-01

    An x-ray crystal spectrometer is scheduled for installation in the Tokamak Fusion Test Reactor in 1984 coinciding with the TFTR deuterium operation phase. This spectrometer is designed to measure the spectra of hydrogen-like and helium-like impurities in the plasma. Ion temperatures electron temperatures, electron density and the distribution of impurity charge states are obtained from the x-ray detector count ratios. The velocity of the toroidal rotation of the plasma is also discerned using this device. The x-ray crystal spectrometer is based on the Bragg diffraction of x-rays from a curved crystal impinging on a multiwire proportional counter. Only those x-rays that satisfy the Bragg relationship (lambda =2d sin theta) will be diffracted and strike the proportional counter. This paper is limited to a discussion of the physical characteristics of the spectrometer and the methods devised to satisfy the operational aspects of such a device

  17. Liquid metal technology in fusion

    International Nuclear Information System (INIS)

    Torre Cabezas, M. de la; Martin Espigares, M.; Lapena, J.

    1985-01-01

    Lithium (or Li-Pb) is one of the several possible coolants being considered for the blanket of magnetic toroidal fusion reactor, not only because of its good thermal and neutron properties, but also because the tritium required to fuel the reactor can be produced by neutron reactions in the lithium. In this paper two main technology tasks to be proposed in our fusion programme have been identified: 1) the development of impurity monitoring devices for use in lithium and Li-Pb environments; 2) effects of Li and Li-Pb environments on the low cycle fatigue properties of different steels. (author)

  18. Design of robust hollow fiber membranes with high power density for osmotic energy production

    KAUST Repository

    Zhang, Sui; Sukitpaneenit, Panu; Chung, Neal Tai-Shung

    2014-01-01

    This study highlights the design strategy of highly asymmetric hollow fiber membranes that possess both characteristics of high flux and high mechanical strength to effectively reap the osmotic energy from seawater brine with an ultrahigh power density. An advanced co-extrusion technology was employed to fabricate the polyethersulfone (PES) hollow fiber supports with diversified structures from macrovoid to sponge-like. The microstructure of the supports is found critical for the stability and water permeability of the thin film composite (TFC) membranes. A high porosity in the porous layer is needed to reduce internal concentration polarization, while a thick and relatively dense skin layer underneath the TFC layer is required to maintain good mechanical stability and stress dissipation. The pore size of the supporting layer underneath the TFC layer must be small with a narrow pore size distribution to ensure the formation of a less-defective, highly permeable and mechanically stable TFC layer. The newly developed hollow fiber comprising high asymmetry, high porosity, and a thick skin layer with a small and narrow pore size distribution underneath the TFC layer produces a maximum power density of 24.3W/m2 at 20.0bar by using 1M NaCl as the concentrated brine and deionized (DI) water as the feed. The proposed design strategy for ultrahigh power density membranes clearly advances the osmotic energy production close to commercialization with a quite cost-effective and practicable approach. © 2013 Elsevier B.V.

  19. High power density superconducting rotating machines—development status and technology roadmap

    Science.gov (United States)

    Haran, Kiruba S.; Kalsi, Swarn; Arndt, Tabea; Karmaker, Haran; Badcock, Rod; Buckley, Bob; Haugan, Timothy; Izumi, Mitsuru; Loder, David; Bray, James W.; Masson, Philippe; Stautner, Ernst Wolfgang

    2017-12-01

    Superconducting technology applications in electric machines have long been pursued due to their significant advantages of higher efficiency and power density over conventional technology. However, in spite of many successful technology demonstrations, commercial adoption has been slow, presumably because the threshold for value versus cost and technology risk has not yet been crossed. One likely path for disruptive superconducting technology in commercial products could be in applications where its advantages become key enablers for systems which are not practical with conventional technology. To help systems engineers assess the viability of such future solutions, we present a technology roadmap for superconducting machines. The timeline considered was ten years to attain a Technology Readiness Level of 6+, with systems demonstrated in a relevant environment. Future projections, by definition, are based on the judgment of specialists, and can be subjective. Attempts have been made to obtain input from a broad set of organizations for an inclusive opinion. This document was generated through a series of teleconferences and in-person meetings, including meetings at the 2015 IEEE PES General meeting in Denver, CO, the 2015 ECCE in Montreal, Canada, and a final workshop in April 2016 at the University of Illinois, Urbana-Champaign that brought together a broad group of technical experts spanning the industry, government and academia.

  20. In-wheel PM motor : compromise between high power density and extended speed capability

    NARCIS (Netherlands)

    Lomonova, E.; Kazmin, Evgeny; Tang, Y.; Paulides, J.J.H.

    2011-01-01

    Purpose – Today's brushless permanent magnet (PM) drive systems usually adopt motors including the advancements in magnet technology, e.g. better thermal characteristics and higher magnetic strength. By this means, they become capable in the roughest applications yet maintain a high accuracy at

  1. Thermal-Hydraulic Performance of Cross-Shaped Spiral Fuel in High-Power-Density BWRs

    International Nuclear Information System (INIS)

    Conboy, Thomas; Hejzlar, Pavel

    2006-01-01

    Power up-rating of existing nuclear reactors promises to be an area of great study for years to come. One of the major approaches to efficiently increasing power density is by way of advanced fuel design, and cross-shaped spiral-fuel has shown such potential in previous studies. Our work aims to model the thermal-hydraulic consequences of filling a BWR core with these spiral-shaped pins. The helically-wound pins have a cross-section resembling a 4-petaled flower. They fill an assembly in a tight bundle, their dimensions chosen carefully such that the petals of neighboring pins contact each other at their outer-most extent in a self-supporting lattice, absent of grid spacers. Potential advantages of this design raise much optimism from a thermal-hydraulic perspective. These spiral rods possess about 40% larger surface area than traditional rods, resulting in increased cooling and a proportional reduction in average surface heat flux. The thin petal-like extensions help by lowering thermal resistance between the hot central region of the pin and the bulk coolant flow, decreasing the maximum fuel temperature by 200 deg. C according to Finite Element (COSMOS) models. However, COSMOS models also predict a potential problem area at the 'elbow' region of two adjoining petals, where heat flux peaking is twice that along the extensions. Preliminary VIPRE models, which account only for the surface area increase, predict a 22% increase in critical power. It is also anticipated that the spiral twist would provide the flowing coolant with an additional radial velocity component, and likely promote turbulence and mixing within an assembly. These factors are expected to provide further margin for increased power density, and are currently being incorporated into the VIPRE model. The reduction in pressure drop inherent in any core without grid-spacers is also expected to be significant in aiding core stability, though this has not yet been quantified. Spiral-fuel seems to be a favorable alternative to traditional pins from a thermal-hydraulic standpoint, though further study of the trends shown in this paper are required. (authors)

  2. Affordable High Power Density Engine Designs for Personal Air Vehicles, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Next generation General Aviation (GA) Sport Class air vehicles limited to 1200lbs, represent the first opportunity to overhaul the FAA certification process...

  3. Affordable High Power Density Engine Designs for Personal Air Vehicles, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — Next generation General Aviation (GA) Sport Class air vehicles limited to 1200lbs, represent the first opportunity to overhaul the FAA certification process...

  4. Biopolymer-nanocarbon composite electrodes for use as high-energy high-power density electrodes

    Science.gov (United States)

    Karakaya, Mehmet; Roberts, Mark; Arcilla-Velez, Margarita; Zhu, Jingyi; Podila, Ramakrishna; Rao, Apparao

    2014-03-01

    Supercapacitors (SCs) address our current energy storage and delivery needs by combining the high power, rapid switching, and exceptional cycle life of a capacitor with the high energy density of a battery. Although activated carbon is extensively used as a supercapacitor electrode due to its inexpensive nature, its low specific capacitance (100-120 F/g) fundamentally limits the energy density of SCs. We demonstrate that a nano-carbon based mechanically robust, electrically conducting, free-standing buckypaper electrode modified with an inexpensive biorenewable polymer, viz., lignin increases the electrode's specific capacitance (~ 600-700 F/g) while maintaining rapid discharge rates. In these systems, the carbon nanomaterials provide the high surface area, electrical conductivity and porosity, while the redox polymers provide a mechanism for charge storage through Faradaic charge transfer. The design of redox polymers and their incorporation into nanomaterial electrodes will be discussed with a focus on enabling high power and high energy density electrodes. Research supported by US NSF CMMI Grant 1246800.

  5. High power density supercapacitor electrodes of carbon nanotube films by electrophoretic deposition

    International Nuclear Information System (INIS)

    Du Chunsheng; Pan Ning

    2006-01-01

    Carbon nanotube thin films have been successfully fabricated by the electrophoretic deposition technique. The supercapacitors built from such thin film electrodes have a very small equivalent series resistance, and a high specific power density over 20 kW kg -1 was thus obtained. More importantly, the supercapacitors showed superior frequency response. Our study also demonstrated that these carbon nanotube thin films can serve as coating layers over ordinary current collectors to drastically enhance the electrode performance, indicating a huge potential in supercapacitor and battery manufacturing

  6. Design of robust hollow fiber membranes with high power density for osmotic energy production

    KAUST Repository

    Zhang, Sui

    2014-04-01

    This study highlights the design strategy of highly asymmetric hollow fiber membranes that possess both characteristics of high flux and high mechanical strength to effectively reap the osmotic energy from seawater brine with an ultrahigh power density. An advanced co-extrusion technology was employed to fabricate the polyethersulfone (PES) hollow fiber supports with diversified structures from macrovoid to sponge-like. The microstructure of the supports is found critical for the stability and water permeability of the thin film composite (TFC) membranes. A high porosity in the porous layer is needed to reduce internal concentration polarization, while a thick and relatively dense skin layer underneath the TFC layer is required to maintain good mechanical stability and stress dissipation. The pore size of the supporting layer underneath the TFC layer must be small with a narrow pore size distribution to ensure the formation of a less-defective, highly permeable and mechanically stable TFC layer. The newly developed hollow fiber comprising high asymmetry, high porosity, and a thick skin layer with a small and narrow pore size distribution underneath the TFC layer produces a maximum power density of 24.3W/m2 at 20.0bar by using 1M NaCl as the concentrated brine and deionized (DI) water as the feed. The proposed design strategy for ultrahigh power density membranes clearly advances the osmotic energy production close to commercialization with a quite cost-effective and practicable approach. © 2013 Elsevier B.V.

  7. Core design and fuel rod analyses of a super fast reactor with high power density

    International Nuclear Information System (INIS)

    Ju, Haitao; Cao, Liangzhi; Lu, Haoliang; Oka, Yoshiaki; Ikejiri, Satoshi; Ishiwatari, Yuki

    2009-01-01

    A Super Fast Reactor is a pressure-vessel type, fast spectrum SuperCritical Water Reactor (SCWR) that is presently researched in a Japanese project. One of the most important advantages of the Super Fast Reactor is the higher power density compared to the thermal spectrum SCWR, which reduces the capital cost. A preliminary core has an average power density of 158.8W/cc. In this paper, the principle of improving the average power density is studied and the core design is improved. After the sensitivity analyses on the fuel rod configurations, the fuel assembly configurations and the core configurations, an improved core with an average power density of 294.8W/cc is designed by 3-D neutronic/thermal-hydraulic coupled calculations. This power density is competitive with that of typical Liquid Metal Fast Breeder Reactors (LMFBR). In order to ensure the fuel rod integrity of this core design, the fuel rod behaviors on the normal operating condition are analyzed using FEMAXI-6 code. The power histories of each fuel rod are taken from the neutronics calculation results in the core design. The cladding surface temperature histories are taken from the thermal-hydraulic calculation results in the core design. Four types of the limiting fuel rods, with the Maximum Cladding Surface Temperature (MCST), Maximum Power Peak(MPP), Maximum Discharge Burnup(MDB) and Different Coolant Flow Pattern (DCFP), are chosen to cover all the fuel rods in the core. The available design range of the fuel rod design parameters, such as initial gas plenum pressure, gas plenum position, gas plenum length, grain size and gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900degC. (2) Maximum cladding stress in circumstance direction should be less than 100MPa. (3) Pressure difference on the cladding should be less than 1/3 of buckling collapse pressure. (4) Cumulative damage faction (CDF) of the cladding should be less than 1.0. Finally the improved fuel rod design of the new core is proposed. (author)

  8. High power density thin film SOFCs with YSZ/GDC bilayer electrolyte

    International Nuclear Information System (INIS)

    Cho, Sungmee; Kim, YoungNam; Kim, Jung-Hyun; Manthiram, Arumugam; Wang Haiyan

    2011-01-01

    Graphical abstract: . A: Cross-sectional TEM images show a GDC single layer and YSZ/GDC bilayer electrolyte structures. As clearly observed from TEM images, the YSZ interlayer thickness varies from ∼330 nm to ∼1 μm. B: The cell with the bilayer electrolyte (YSZ ∼330 nm) doubles the overall power output at 750 deg. C compared to that achieved in the GDC single layer cell. Display Omitted Highlights: → YSZ/ GDC bilayer thin film electrolytes were deposited by a pulsed laser deposition (PLD) technique. → Thin YSZ film as a blocking layer effectively suppresses the cell voltage drop without reducing the ionic conductivity of the electrolyte layer. → The YSZ/ GDC bilayer structure presents a feasible architecture for enhancing the overall power density and enabling chemical, mechanical, and structural stability in the cells. - Abstract: Bilayer electrolytes composed of a gadolinium-doped CeO 2 (GDC) layer (∼6 μm thickness) and an yttria-stabilized ZrO 2 (YSZ) layer with various thicknesses (∼330 nm, ∼440 nm, and ∼1 μm) were deposited by a pulsed laser deposition (PLD) technique for thin film solid oxide fuel cells (TFSOFCs). The bilayer electrolytes were prepared between a NiO-YSZ (60:40 wt.% with 7.5 wt.% carbon) anode and La 0.5 Sr 0.5 CoO 3 -Ce 0.9 Gd 0.1 O 1.95 (50:50 wt.%) composite cathode for anode-supported single cells. Significantly enhanced maximum power density was achieved, i.e., a maximum power density of 188, 430, and 587 mW cm -2 was measured in a bilayer electrolyte single cell with ∼330 nm thin YSZ at 650, 700, and 750 deg. C, respectively. The cell with the bilayer electrolyte (YSZ ∼330 nm) doubles the overall power output at 750 deg. C compared to that achieved in the GDC single layer cell. This signifies that the YSZ thin film serves as a blocking layer for preventing electrical current leakage in the GDC layer and also provides chemical, mechanical, and structural integrity in the cell, which leads to the overall enhanced performance.

  9. Towards High Power Density Metal Supported Solid Oxide Fuel Cell for Mobile Applications

    DEFF Research Database (Denmark)

    Nielsen, Jimmi; Persson, Åsa H.; Muhl, Thuy Thanh

    2018-01-01

    For use of metal supported solid oxide fuel cell (MS-SOFC) in mobile applications it is important to reduce the thermal mass to enable fast startup, increase stack power density in terms of weight and volume and reduce costs. In the present study, we report on the effect of reducing the Technical...

  10. Towards High Power Density Metal Supported Solid Oxide Fuel Cell for Mobile Applications

    DEFF Research Database (Denmark)

    Nielsen, Jimmi; Persson, Åsa Helen; Muhl, Thuy

    2017-01-01

    For use of metal supported SOFC in mobile applications it is important to reduce the thermal mass to enable fast start up, increase stack power density in terms of weight and volume and reduce costs. In the present study, we report on the effect of reducing the support layer thickness of 313 μm...

  11. Transport dynamics of a high-power-density matrix-type hydrogen-oxygen fuel cell

    Science.gov (United States)

    Prokopius, P. R.; Hagedorn, N. H.

    1974-01-01

    Experimental transport dynamics tests were made on a space power fuel cell of current design. Various operating transients were introduced and transport-related response data were recorded with fluidic humidity sensing instruments. Also, sampled data techniques were developed for measuring the cathode-side electrolyte concentration during transient operation.

  12. A Study on the Development of BLDC Motor with High Power Density

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun Cheol; Kong, Yeong Kyung; Choi, Tae In [Agency for Defense Development (Korea); Song, Jong Hwan [Hyosung Ltd., (Korea)

    2000-05-01

    The motor for torpedo propulsion is needed the compact and short rating high power characteristics. This paper describes the development of the motor through the theory and Finite Element Method(FEM) analysis for Brushless Direct Current Motor(BLDCM) of 7 phase 6 poles. Back EMF, inductance and eddy current loss were analyzed. The proposed methods like magnetic wedge acquired by these FEM analysis were introduced. Phase-leading angle using encoder was used. Test results on the motor of 7 phases 6 poles were showed the validity of proposed methods and phase-leading angle. (author). 9 refs., 12 figs., 5 tabs.

  13. Fusion at counterstreaming ion beams - ion optic fusion (IOF)

    International Nuclear Information System (INIS)

    Gryzinski, M.

    1981-01-01

    The results of investigation are briefly reviewed in the field of ion optic fusion performed at the Institute of Nuclear Research in Swierk. The ion optic fusion concept is based on the possibility of obtaining fusion energy at highly ordered motion of ions in counterstreaming ion beams. For this purpose TW ion beams must be produced and focused. To produce dense and charge-neutralized ion beams the selective conductivity and ballistic focusing ideas were formulated and used in a series of RPI devices with low-pressure cylindrical discharge between grid-type electrodes. 100 kA, 30 keV deuteron beams were successfully produced and focused into the volume of 1 cm 3 , yielding 10 9 neutrons per 200 ns shot on a heavy ice target. Cylindrically convergent ion beams with magnetic anti-defocusing were proposed in order to reach a positive energy gain at reasonable energy level. (J.U.)

  14. Thermostructural design of the first wall/blanket for the TITAN-RFP fusion reactor

    International Nuclear Information System (INIS)

    Orient, G.E.; Blanchard, J.P.; Ghoniem, N.M.

    1987-01-01

    The mass power density, which is defined as the average power per unit mass within the magnet boundary, is a rough and general measure of economic competitiveness. Conn et al. (1985) have identified a target value of 100 kW(e)/tonne as a reasonable threshold for 'compact' commercial fusion systems. In pursuit of this goal, Hagenson et al. (1984) and Najmabadi et al. (1987) have pointed out the inherent characteristics of the RFP toroidal confinement concept which allow it to exceed this target value. It is inevitable that the compactness of the fusion power core will introduce a unique set of design issues. The special design concerns stem from high thermal surface fluxes, high bulk energy deposition by neutrons, and a relatively short blanket structural lifetime. In the TITAN-RFP, study Najmabadi et al. (1987) investigate a number of blanket (B) and first wall (FW) options suitable for high power density fusion reactors. Final choices were made for two designs: A high pressure aqueous blanket and a vanadium/lithium self-cooled blanket. The first design utilizes a pressurized aqueous loop containing a lithium compound dissolved in water, while the second design is based upon a self-cooled lithium-vanadium blanket. In this paper, we consider the beginning-of-life (BOL) thermostructural design and analysis of only the second concept. (orig./GL)

  15. Evaluation of compatibility of flowing liquid lithium curtain for blanket with core plasma in fusion reactors

    International Nuclear Information System (INIS)

    Deng Baiquan; Huang Jinhua; Peng Lilin; Yan Jiancheng

    2003-01-01

    A global model analysis of the compatibility of flowing liquid lithium curtain for blanket with core plasma has been performed. The relationships between the surface temperature of lithium curtain and mean effective plasma charges, fuel dilution and produced fusion power have been obtained. Results show that under normal circumstances, the evaporation of liquid lithium does not affect Z eff seriously, but affects fuel dilution and fusion power sensitively. The authors have investigated the relationships between the flow velocity of liquid lithium and its surface temperature rise based on the conditions of the option II of the fusion experimental breeder (FEB-E) design with reversed shear configuration and fairly high power density. The authors concluded that the effects of evaporation from liquid lithium curtain for FEB-E on plasma are negligible even if the flow velocity of liquid lithium is as low as 0.5 m·s -1 . Finally, the sputtering yield of liquid lithium saturated by hydrogen isotopes is briefly discussed

  16. Economic effect of fusion in energy market. Economic impact of fusion deployment in energy market

    International Nuclear Information System (INIS)

    Konishi, Satoshi

    2002-01-01

    Energy model analysis estimates the significant contribution of fusion in the latter half of the century under the global environment constraints if it will be successfully developed and introduced into the market. The total possible economical impact of fusion is investigated from the aspect of energy cost savings, sales, and its effects on Gross Domestic Products. Considerable economical possibility will be found in the markets for fusion related devices, of currently developing countries, and for synthesized fuel. The value of fusion development could be evaluated from these possible economic impact in comparison with its necessary investment. (author)

  17. Fusion Machines

    International Nuclear Information System (INIS)

    Weynants, R.R.

    2004-01-01

    A concise overview is given of the principles of inertial and magnetic fusion, with an emphasis on the latter in view of the aim of this summer school. The basis of magnetic confinement in mirror and toroidal geometry is discussed and applied to the tokamak concept. A brief discussion of the reactor prospects of this configuration identifies which future developments are crucial and where alternative concepts might help in optimising the reactor design. The text also aims at introducing the main concepts encountered in tokamak research that will be studied and used in the subsequent lectures

  18. Thermonuclear fusion: Current status and future prospects

    International Nuclear Information System (INIS)

    Bruhns, H.; Maisonnier, Ch.

    1992-01-01

    Thermonuclear Fusion holds great promises for becoming an important energy source for the future. Fusion research and development is undertaken in al major countries of the world. The European Community pursues fusion in a large programme which embraces all R and D in the field of magnetic confinement fusion in the Member States, and to which Sweden and Switzerland are fully associated. The long-term objective of the programme is the joint creation of safe, environmentally sound prototype reactors. The main R and D line of the Community Fusion Programme is fusion by toroidal magnetic confinement on the basis of the Tokamak concept. Some related concepts are also studied which possibly could offer advantages for a reactor, and keep-in-touch activities exist for other approaches. Several small and medium sized specialised devices in Associated Laboratories have been built by the Community Fusion Programme as well as the Joint European Torus (JET Joint Undertaking) which is the largest and the most successful fusion device in the world. Recently, fusion power in the megawatt range has been achieved in JET. The long timescale and the large effort needed for the development of fusion as an energy source have been important elements to foster international collaboration. Engineering Design Activities for an International Thermonuclear Experimental Reactor (ITER) are undertaken, under the auspices of the IAEA, by the European Community, Japan, the Russian Federation and the United States of America. The objective of ITER is to achieve self-sustained thermonuclear burn and its control under long-pulse operation and to provide basic data for the engineering of a demonstration fusion reactor. (author)

  19. Congress turns cold on fusion

    International Nuclear Information System (INIS)

    Marshall, E.

    1984-01-01

    A 5% cut in fusion research budgets will force some programs to be dropped in order to keep the large machinery running unless US and European scientists collaborate instead of competing. Legislators became uneasy about the escalating costs of the new devices. The 1984 budget of $470 million for magnetic fusion research is only half the projected cost of the Tokomak Fusion Core Experiment (TFCX) planned to ignite, for the first time, a self-sustaining burn. Planning for the TCFX continued despite the message from Congress. Work at the large institutions at Princeton, MIT, etc. may survive at the expense of other programs, some of which will lose academic programs as well. Scientists point to the loss of new ideas and approaches when projects are cancelled. Enthusiasm is growing for international collaboration

  20. Fusion Canada issue 10

    International Nuclear Information System (INIS)

    1990-02-01

    A short bulletin from the National Fusion Program. Included in this issue is a report on Fusion Materials Research, ITER physics research, fusion performance record at JET, and design options for reactor building. 4 figs

  1. Contributions to the 20. EPS conference on controlled fusion and plasma physics

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-07-15

    The Conference covers research on different aspects of plasma physics and fusion technology, like technical aspects of Tokamak devices; plasma instabilities and impurities, development and testing of materials for fusion reactors etc.

  2. The present role of superconductivity in fusion

    International Nuclear Information System (INIS)

    Shimamoto, S.

    1986-01-01

    After completion of large fusion devices in the world, such as JT-60, JET and TFTR, high temperature plasma is proceeding to critical condition for fusion. The devices up to now use mainly conventional magnet. However, for the next generation machine which demonstrates fusion reaction, deuterium-tritium burning, superconducting magnet system is indispensable from view point of both net energy extraction and capacity limitation of power supply. In order to realize such a large and complicated system, a lot of development works is being carried out. This paper describes required parameters of superconducting magnet and helium refrigerator, the state of plasma condition and superconducting magnet. It is shown that the present technology of superconducting magnet is not so far from realization of fusion experimental reactor

  3. View of fusion from Capitol Hill

    International Nuclear Information System (INIS)

    Mense, A.T.

    1981-01-01

    On October 7, 1980, the Magnetic Fusion Energy Engineering Act of 1980 (nicknamed the 'McCormack Fusion Bill') was signed into Public Law (P.L. 96-386) by President Carter. This new law if carried through, would result in an accelerated program leading in the near term to: (1) the establishment of a national center for fusion engineering; and (2) the design, construction and operation of a multi-billion dollar fusion reactor called the Fusion Engineering Device (FED). It is the purpose of this paper to briefly outline some of the legislative history that led up to the passage of P.L. 96-386, and finally, to present some thought on the legislative climate with regard to the FY '82 Department of Energy budget

  4. Radiolytic production of chemical fuels in fusion reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Fish, J D

    1977-06-01

    Miley's energy flow diagram for fusion reactor systems is extended to include radiolytic production of chemical fuel. Systematic study of the economics and the overall efficiencies of fusion reactor systems leads to a criterion for evaluating the potential of radiolytic production of chemical fuel as a means of enhancing the performance of a fusion reactor system. The ecumenicity of the schema is demonstrated by application to (1) tokamaks, (2) mirror machines, (3) theta-pinch reactors, (4) laser-heated solenoids, and (5) inertially confined, laser-pellet devices. Pure fusion reactors as well as fusion-fission hybrids are considered.

  5. Radiolytic production of chemical fuels in fusion reactor systems

    International Nuclear Information System (INIS)

    Fish, J.D.

    1977-06-01

    Miley's energy flow diagram for fusion reactor systems is extended to include radiolytic production of chemical fuel. Systematic study of the economics and the overall efficiencies of fusion reactor systems leads to a criterion for evaluating the potential of radiolytic production of chemical fuel as a means of enhancing the performance of a fusion reactor system. The ecumenicity of the schema is demonstrated by application to (1) tokamaks, (2) mirror machines, (3) theta-pinch reactors, (4) laser-heated solenoids, and (5) inertially confined, laser-pellet devices. Pure fusion reactors as well as fusion-fission hybrids are considered

  6. Catalysed fusion

    CERN Document Server

    Farley, Francis

    2012-01-01

    A sizzling romance and a romp with subatomic particles at CERN. Love, discovery and adventure in the city where nations meet and beams collide. Life in a large laboratory. As always, the challenges are the same. Who leads? Who follows? Who succeeds? Who gets the credit? Who gets the women or the men? Young Jeremy arrives in CERN and joins the quest for green energy. Coping with baffling jargon and manifold dangers, he is distracted by radioactive rats, lovely ladies and an unscrupulous rival. Full of doubts and hesitations, he falls for a dazzling Danish girl, who leads him astray. His brilliant idea leads to a discovery and a new route to cold fusion. But his personal life is scrambled. Does it bring fame or failure? Tragedy or triumph?

  7. Fusion cuisine

    DEFF Research Database (Denmark)

    Peters, Chris; Broersma, Marcel

    2018-01-01

    JJournalism studies as an academic field is characterized by multidisciplinarity. Focusing on one object of study, journalism and the news, it established itself by integrating and synthesizing approaches from established disciplines – a tendency that lives on today. This constant gaze to the out......JJournalism studies as an academic field is characterized by multidisciplinarity. Focusing on one object of study, journalism and the news, it established itself by integrating and synthesizing approaches from established disciplines – a tendency that lives on today. This constant gaze...... to the outside for conceptual inspiration and methodological tools lends itself to a journalism studies that is a fusion cuisine of media, communication, and related scholarship. However, what happens when this object becomes as fragmented and multifaceted as the ways we study it? This essay addresses...

  8. The ''ATOS'' experimental device

    International Nuclear Information System (INIS)

    Belyaev, V.A.; Dorovskij, A.P.; Dubrovin, M.M.; Khlopkin, A.N.

    1980-08-01

    This paper contains a brief description of the ATOS experimental device at the I.V. Kurchatov Institute, Moscow, USSR, which has been designed in accordance with the merged beam principle to investigate collisions between heavy atomic particles and multiply-charged ions of impurity elements - following the programme of the Joint IFRC/INDC Subcommittee on Atomic and Molecular Data for Fusion

  9. Mirror Fusion Test Facility (MFTF)

    International Nuclear Information System (INIS)

    Thomassen, K.I.

    1978-01-01

    A large, new Mirror Fusion Test Facility is under construction at LLL. Begun in FY78 it will be completed at the end of FY78 at a cost of $94.2M. This facility gives the mirror program the flexibility to explore mirror confinement principles at a signficant scale and advances the technology of large reactor-like devices. The role of MFTF in the LLL program is described here

  10. History of controlled nuclear fusion in Japan

    International Nuclear Information System (INIS)

    Uematsu, Eisui; Nishio, Shigeko; Takeda, Tatsuoki

    2001-01-01

    A research development of nuclear fusion was divided four periods: the first period as prehistory (until about 1955), the second period as begin of research (1955 to 1969), the third as the growth period (1970 to 1985) and the forth as the large tokamak age. In this paper I explained the second period, because general physicists and young plasma and controlled nuclear fusion researcher did not know about this period. The controlled nuclear fusion research was begun by the experiment of hydrogen bomb by USA and USSR in 1952 and 1953. In Japan, on the basis of many societies, 'The Controlled Nuclear Fusion Meeting' was established as an independent system and KAKEA (Journal of Fusion Research) was published in 1958. Japan government began to make the system by the Nuclear Commission in 1957. The main research devices in 1962 were linear pinch, mirror device, toroidal pinch, helical system, plasma gun and plasma measurement. USSR showed the excellent results of tokamak device in 1968. Ookawa spoke the effect of the average minimum-B, the best report in this period, at the second IAEA meeting, 1965. JAERI constructed JFT-1 and JFT-2, the latter was the first class device in the world and made the first step of Japanese research into the world, for examples, to attain the equilibrium of divertor plasma and to control impurity. Many research centers of controlled fusion were established in many universities in Japan from 1966 to 1980. Cooperation researchs between Japan and USA, USSR and many countries has been carried out after 1978: JIFT (Joint Institute for Fusion Theory) and FPPC (Fusion Power Coordinating Committee). The important results increased in this period. After 1985, the research activities are processing and data increased very fast depend on the larger devices and system, good measurement system and development of information system. JT-60 in JAERI opened to the large tokamak period. It led controlled fusion researchs in the world the same as TFTR (US

  11. Beam limiter for thermonuclear fusion devices

    International Nuclear Information System (INIS)

    Kaminsky, M.S.

    1976-01-01

    A beam limiter circumscribes the interior surface of a vacuum vessel to inhibit collisions of contained plasma and the vessel walls. The cross section of the material making up the limiter has a flatsided or slightly concave portion of increased width towards the plasma and portions of decreased width towards the interior surface of the vessel. This configuration is designed to prevent a major fraction of the material sputtered, vaporized and blistered from the limiter from reaching the plasma. It also allows adequate heat transfer from the wider to the narrower portions. The preferred materials for the beam limiter are solids of sintered, particulate materials of low atomic number with low vapor pressure and low sputtering and blistering yields. 7 claims, 3 figures

  12. Leak testing and repair of fusion devices

    International Nuclear Information System (INIS)

    Kozman, T.A.

    1983-01-01

    The leak testing, reporting and vacuum leak repair techniques of the MFTF yin-yang number one magnet system, the world's largest superconducting magnet system, are discussed. Based on this experience, techniques will be developed for testing and repairing leaks on the 42 MFTF-B magnets. The leak-hunting techniques for the yin-yang magnet systems were applied to two helium circuits (the coil bundle and guard vacuum; both require helium flow for magnet cooldown), their associated piping, liquid nitrogen radiation shields, and piping. Additionally, during MFTF-B operation there will be warm water plasma shields and piping that require leak checking

  13. Aerodynamic window for a laser fusion device

    International Nuclear Information System (INIS)

    Masuda, Wataru

    1983-01-01

    Since the window of a laser system absorbs a part of the laser energy, the output power is determined by the characteristics of the window. The use of an aerodynamic window has been studied. The required characteristics are to keep the large pressure difference. An equation of motion of a vortex was presented and analyzed. The operation power of the system was studied. A multi-stage aerodynamic window was proposed to reduce the power. When the jet flow of 0.3 of the Mach number is used, the operation power will be several Megawatt, and the length of an optical path will be about 100 m. (Kato, T.)

  14. Coil supporting device in nuclear fusion apparatus

    International Nuclear Information System (INIS)

    Hoshi, Ryo; Imura, Yasuya.

    1974-01-01

    Object: To secure intermediate fittings with a coil fixed thereon by an insulating tape to a fixed body by means of fittings, thereby supporting the coil in a narrow space. Structure: A coil is secured to intermediate fittings by means of an insulating tape, after which the intermediate fittings is mounted on a fixed body through fittings to support the coil in a narrow clearance portion between a plasma sealed vessel and a main coil. (Kamimura, M.)

  15. Beam limiter for thermonuclear fusion devices

    International Nuclear Information System (INIS)

    Kaminsky, M.S.

    1977-01-01

    The invention pertains to a beam limiter to prevent collisions between a plasma and the inner surface of a hollow body in which the plasma is confined. The patent claims pertain to suitable geometrical shapes of the beam limiter. (GG) [de

  16. Compact imaging Bragg spectrometer for fusion devices

    International Nuclear Information System (INIS)

    Bertschinger, G.; Biel, W.; Jaegers, H.; Marchuk, O.

    2004-01-01

    A compact imaging x-ray spectrometer has been designed for tokamaks and stellarators to measure the plasma parameters at different spatial chords. It has been optimized for high spectral resolution and high sensitivity. High spectral resolution is obtained by using solid state detectors and minimizing the imaging errors of the spherical crystals. It is shown, that using spherical crystals the solid angle and hence the throughput can be increased significantly, without compromising the spectral resolution. The design is useful for the measurement of the spectra of He- and H-like ions from Si to Kr. The spectral resolution is sufficient for the measurement of plasma parameters. The temporal resolution is high enough for transport studies by gas puff and laser ablation experiments. The design is based on a modified Johann spectrometer mount, utilizing a spherically bent crystal instead of the cylindrically bent crystal in the traditional Johann mount. The astigmatism of the wavelength selective reflection on the spherical crystal is applied to obtain imaging of an extended plasma source on a two-dimensional detector. For each element, a separate crystal is required, only in few cases, a crystal can be used for the spectra of two elements. For the spectra of most of the He-like ions from Si up to Kr, suitable crystal cuts have been found on quartz, silicon and germanium crystals with Bragg angles in a small interval around the design value of 53.5 deg. All of the crystals have the same radius. They are fixed on a rotational table. The distance to the detector is adjusted by an x-y table to fit to the Rowland circle

  17. Status report on controlled thermonuclear fusion

    International Nuclear Information System (INIS)

    1990-06-01

    The International Fusion Research Council (IFRC), an advisory body to the International Atomic Energy Agency, reports on the current status of fusion; this report updates its 1978 status report. This report contains a General Overview and Executive Summary, and reports on all current approaches to fusion throughout the world; a series of technical reports is to be published elsewhere. This report is timely in that it not only shows progress which has occurred over the past, but interfaces with possible future devices, in particular the International Thermonuclear Experimental Reactor (ITER), whose conceptual design phase is nearing completion. 5 refs, 6 figs

  18. Culham: fusion and commerce

    International Nuclear Information System (INIS)

    Herman, R.

    1976-01-01

    An overview is given of present day work at the UKAEA research establishment at Culham. This consists not only of research into the practical and theoretical problems of nuclear fusion but also contract work for commercial companies and government agencies. This latter type of work includes studies on the effect of lightening striking aircraft, work on electrostatic hazards in oil tankers and potential uses of lasers in industrial processes such as cutting, drilling and welding. The six toroidal confinement devices at present at Culham are described. Reference is made to the superconducting levitron which is of the toroidal multipole type, two stellarators TORSO and CLEO, the reversed field pinch configuration Zeta experiment and the two tokamaks TOSCA and DITE. Work on such complex machines needs substantial support in engineering and computational skills, and these are also provided by groups at Culham. Research is already under way into overcoming the difficulties of an actual reactor system such as extracting energy from the neutrons coming out of the plasma, and the long term effects of neutron irradiation on cell modules for fusion reactor blankets. (U.K.)

  19. Fusion reactor remote maintenance study. Final report

    International Nuclear Information System (INIS)

    Sniderman, M.

    1979-04-01

    An analysis of a major maintenance operation, the remote replacement of a modular sector of a tokamak reactor, was performed in substantial detail. Specific assumptions were developed which included concepts from various existing designs so that the operation which was studied includes some design features generic to any fusion reactor design. Based on the work performed in this study, the principal conclusions are: (1) It appears feasible to design a tokamak fusion reactor plant with availability comparable to existing fossil and fission plants, but this will require diligence and comprehensive planning during the complete design phase. (2) Since the total fusion program is paced by the success of each device, maintenance considerations must be incorporated into each device during design, even if the device is an experimental unit. (3) Innovative approaches, such as automatic computer controlled operations, should be developed so that large step reductions in planned maintenance times can be achieved

  20. Towards nuclear fusion reactors

    International Nuclear Information System (INIS)

    1993-11-01

    The results of nuclear fusion researches in JAERI are summarized. In this report, following themes are collected: the concept of fusion reactor (including ITER), fusion reactor safety, plasma confinement, fusion reactor equipment, and so on. Includes glossary. (J.P.N.)