WorldWideScience

Sample records for bundle divertors

  1. Hybrid bundle divertor design

    International Nuclear Information System (INIS)

    A hybrid bundle divertor design is presented that produces <0.3% magnetic ripple at the center of the plasma while providing adequate space for the coil shielding and structure for a tokamak fusion test reactor similar to the International Tokamak Reactor and the Engineering Test Facility (with R = 5 m, B = 5 T, and a /SUB wall/ = 1.5 m, in particular). This hybrid divertor consists of a set of quadrupole ''wing'' coils running tangent to the tokamak plasma on either side of a bundle divertor. The wing coils by themselves pull the edge of the plasma out 1.5 m and spread the thickness of the scrape-off layer from 0.1 to 0.7 m at the midplane. The clear aperture of the bundle divertor throat is 1.0 m high and 1.8 m wide. For maintenance or replacement, the hybrid divertor can be disassembled into three parts, with the bundle divertor part pulling straight out between toroidal field coils and the wing coils then sliding out through the same opening

  2. Twin tori for a new bundle divertor

    International Nuclear Information System (INIS)

    A new bundle divertor system using the straight stagnation axis in toroidal field together with the uniform field along the axis is discussed in detail. We call this type of divertor as the ''muffler divertor'' because of its shape. (author)

  3. Optimization of a bundle divertor for FED

    International Nuclear Information System (INIS)

    Optimal double-T bundle divertor configurations have been obtained for the Fusion Engineering Device (FED). On-axis ripple is minimized, while satisfying a series of engineering constraints. The ensuing non-linear optimization problem is solved via a sequence of quadratic programming subproblems, using the VMCON algorithm. The resulting divertor designs are substantially improved over previous configurations

  4. Reactor application of an improved bundle divertor

    International Nuclear Information System (INIS)

    A Bundle Divertor was chosen as the impurity control and plasma exhaust system for the beam driven Demonstration Tokamak Hybrid Reactor - DTHR. In the context of a preconceptual design study of the reactor and associated facility a bundle divertor concept was developed and integrated into the reactor system. The overall system was found feasible and scalable for reactors with intermediate torodial field strengths on axis. The important design characteristics are: the overall average current density of the divertor coils is 0.73 kA for each tesla of toroidal field on axis; the divertor windings are made from super-conducting cables supported by steel structures and are designed to be maintainable; the particle collection assembly and auxiliary cryosorption vacuum pump are dual systems designed such that they can be reactivated alterntively to allow for continuous reactor operation; and the power requirement for energizing and operating the divertor is about 5 MW

  5. FLP: a field line plotting code for bundle divertor design

    International Nuclear Information System (INIS)

    A computer code was developed to aid in the design of bundle divertors. The code can handle discrete toroidal field coils and various divertor coil configurations. All coils must be composed of straight line segments. The code runs on the PDP-10 and displays plots of the configuration, field lines, and field ripple. It automatically chooses the coil currents to connect the separatrix produced by the divertor to the outer edge of the plasma and calculates the required coil cross sections. Several divertor designs are illustrated to show how the code works

  6. Structural evaluation of a DTHR bundle divertor particle collector

    International Nuclear Information System (INIS)

    The purpose of this report is to present a structural evaluation of the current bundle divertor particle collector BDPC design under a peak heat flux in relation to criteria that protect against coolant leakage into the plasma over replacement schedules planned during DTHR operation. In addition, an assessment of the BDPC structural integrity at higher heat fluxes is presented. Further, recommendations for modifications in the current BDPC design that would improve design reliability to be considered in future design studies are described. Finally, experimental test programs directed to establishing materials data necessary in providing greater confidence in subsequent structural evaluations of BDPC designs in relation to coolant leakage over planned replacement schedules are identified

  7. Study of the feasibility of installing a toroidal or bundle divertor in EBT-S. Phase I: EBT-S divertor project. Final report

    International Nuclear Information System (INIS)

    The following chapters are included: (1) magnetic field analysis of the basic EBT-S geometry with and without aspect ratio enhancement coils; (2) analyses of a toroidal divertor for EBT-S; (3) analysis of a bundle divertor for EBT-S; (4) engineering; and (5) divertor vacuum pumping

  8. Sizing of the thermal and electrical systems for an FED bundle divertor design with MgO insulation

    International Nuclear Information System (INIS)

    The high-order dependence of toroidal ripple from a bundle divertor on the magnet shield thickness increases the desirability of a magnet technology with minimal shielding requirements. A jacketed conductor with MgO powder insulation has been used successfully in highly irradiated environments. Its properties and limitations are described. A thermal and electrical sizing code has been developed for magnet design with this technology. Two design examples for ETF and FED missions show reduced recirculating power from previously reported designs

  9. Divertor detachment

    Science.gov (United States)

    Krasheninnikov, Sergei

    2015-11-01

    The heat exhaust is one of the main conceptual issues of magnetic fusion reactor. In a standard operational regime the large heat flux onto divertor target reaches unacceptable level in any foreseeable reactor design. However, about two decades ago so-called ``detached divertor'' regimes were found. They are characterized by reduced power and plasma flux on divertor targets and look as a promising solution for heat exhaust in future reactors. In particular, it is envisioned that ITER will operate in a partly detached divertor regime. However, even though divertor detachment was studied extensively for two decades, still there are some issues requiring a new look. Among them is the compatibility of detached divertor regime with a good core confinement. For example, ELMy H-mode exhibits a very good core confinement, but large ELMs can ``burn through'' detached divertor and release large amounts of energy on the targets. In addition, detached divertor regimes can be subject to thermal instabilities resulting in the MARFE formation, which, potentially, can cause disruption of the discharge. Finally, often inner and outer divertors detach at different plasma conditions, which can lead to core confinement degradation. Here we discuss basic physics of divertor detachment including different mechanisms of power and momentum loss (ionization, impurity and hydrogen radiation loss, ion-neutral collisions, recombination, and their synergistic effects) and evaluate the roles of different plasma processes in the reduction of the plasma flux; detachment stability; and an impact of ELMs on detachment. We also evaluate an impact of different magnetic and divertor geometries on detachment onset, stability, in- out- asymmetry, and tolerance to the ELMs. Supported by the U.S. Department of Energy Office of Science, Office of Fusion Energy Sciences under Award Number DE-DE-FG02-04ER54739 at UCSD.

  10. Divertor parameters and divertor operation in ASDEX

    Science.gov (United States)

    Fussmann, G.; Ditte, U.; Eckstein, W.; Grave, T.; Keilhacker, M.; McCormick, K.; Murmann, H.; Röhr, H.; Elshaer, M.; Steuer, K.-H.; Szymanski, Z.; Wagner, F.; Becker, G.; Bernhardi, K.; Eberhagen, A.; Gehre, O.; Gernhardt, J.; Gierke, G. V.; Glock, E.; Gruber, O.; Haas, G.; Hesse, M.; Janeschitz, G.; Karger, F.; Kissel, S.; Klüber, O.; Kornherr, M.; Lisitano, G.; Mayer, H. M.; Meisel, D.; Müller, E. R.; Poschenrieder, W.; Ryter, F.; Rapp, H.; Schneider, F.; Siller, G.; Smeulders, P.; Söldner, F.; Speth, E.; Stäbler, A.; Vollmer, O.

    1984-12-01

    Recent measurements of plasma boundary and divertor scrape-off parameters for ohmically and neutral injection heated plasmas are presented. For these data the power flow onto the divertor plates and the sputtering rates at the plates are calculated and compared with separate measurements. The impurity behaviour in front of the plates is also discussed.

  11. Bundled procurement

    OpenAIRE

    Chen, Yongmin; Li, Jianpei

    2015-01-01

    When procuring multiple products from competing firms, a buyer may choose separate purchase, pure bundling, or mixed bundling. We show that pure bundling will generate higher buyer surplus than both separate purchase and mixed bundling, provided that trade for each good is likely to be efficient. Pure bundling is superior because it intensifies the competition between firms by reducing their cost asymmetry. Mixed bundling is inferior because it allows firms to coordinate to ...

  12. The jet divertor coils

    International Nuclear Information System (INIS)

    This paper reports on the JET Tokamak which is to be modified to incorporate a divertor. A coil system in the vacuum vessel has been developed, which can produce a range of different divertor plasmas. The divertor coils are of conventional construction and are contained in this Inconel cases. They will be assembled in the vacuum vessel, welded into their cases and impregnated with epoxy resin

  13. Divertor efficiency in ASDEX

    Science.gov (United States)

    Engelhardt, W.; Becker, G.; Behringer, K.; Campbell, D.; Eberhagen, A.; Fussmann, G.; Gehre, O.; Gierke, G. V.; Glock, E.; Haas, G.; Huang, M.; Karger, F.; Keilhacker, M.; KlÜber, O.; Kornherr, M.; Lisitano, G.; Mayer, H.-M.; Meisel, D.; Müller, E. R.; Murmann, H.; Niedermeyer, H.; Poschenrieder, W.; Rapp, H.; Schneider, F.; Siller, G.; Steuer, K.-H.; Venus, G.; Vernickel, H.; Wagner, F.

    1982-12-01

    The divertor efficiency in ASDEX is discussed for ohmically heated plasmas. The parameters of the boundary layer both in the torus midplane and the divertor chamber have been measured. The results are reasonably well understood in terms of parallel and perpendicular transport. A high pressure of neutral hydrogen builds up in the divertor chamber and Franck-Condon particles recycle back through the divertor throat. Due to dissociation processes the boundary plasma is effectively cooled before it reaches the neutralizer plates. The shielding property of the boundary layer against impurity influx is comparable to that of a limiter plasma. The transport of iron is numerically simulated for an iron influx produced by sputtering of charge exchange neutrals at the wall. The results are consistent with the measured iron concentration. First results from a comparison of the poloidal divertor with toroidally closed limiters (stainless steel, carbon) are given. Diverted discharges are considerably cleaner and easier to create.

  14. 'EU divertor celebration day'

    International Nuclear Information System (INIS)

    The meeting 'EU divertor celebration day' organized on 16 January 2002 at Plansee AG, Reutte, Austria was held on the occasion of the completion of manufacturing activities of a complete set of near full-scale prototypes of divertor components including the vertical target, the dome liner and the cassette body. About 30 participants attended the meeting including Dr. Robert Aymar, ITER Director, representatives from EFDA, CEA, ENEA, IPP and others

  15. Diagnostics for the DIII-D radiative divertor

    Energy Technology Data Exchange (ETDEWEB)

    Nilson, D.G. [Lawrence Livermore National Lab., CA (United States); Brooks, N.H.; Smith, J.P.; Snider, R.T.

    1995-10-01

    This paper reviews the design of new diagnostics and the modifications to existing diagnostics needed to carry out radiative divertor experiments in DIII-D following installation in late 1996 of a set of baffle structures that will restrict the backflow to the core plasma of neutral deuterium atoms and impurity gases. The divertor slots formed by the new baffle structures will inhibit the easy view of the divertor legs and target plates that the open divertor geometry in DIII-D currently affords. We review a basic set of diagnostics that are needed to demonstrate the reduction of divertor heat loading and radiative dissipation of energy within the divertor. This will include IR cameras, bolometry, foil bolometers, and Langmuir probes. Within the limits of available funding, we will implement a supplemental set of instruments which provide a more detailed understanding of the underlying physical processes. Many existing diagnostics require only re-aiming to provide proper coverage of the initial 23 cm long divertor plasma configuration (X- point to floor distance). Other diagnostics need extensive reconfiguration using in-vessel fiber-optic bundles or high power laser mirrors. The new divertor baffle panels provide a protective shelf for diagnostic hardware mounted underneath them, but the water cooling channels in the panels limit the permissible size of through holes and, thereby, restrict the available views of under-the- baffle diagnostics. The successful resolution of the design and implementation of these diagnostic modifications is dependent on a strong coordination between GA and its many diagnostic collaborators.

  16. Diagnostics for the DIII-D radiative divertor

    International Nuclear Information System (INIS)

    This paper reviews the design of new diagnostics and the modifications to existing diagnostics needed to carry out radiative divertor experiments in DIII-D following installation in late 1996 of a set of baffle structures that will restrict the backflow to the core plasma of neutral deuterium atoms and impurity gases. The divertor slots formed by the new baffle structures will inhibit the easy view of the divertor legs and target plates that the open divertor geometry in DIII-D currently affords. We review a basic set of diagnostics that are needed to demonstrate the reduction of divertor heat loading and radiative dissipation of energy within the divertor. This will include IR cameras, bolometry, foil bolometers, and Langmuir probes. Within the limits of available funding, we will implement a supplemental set of instruments which provide a more detailed understanding of the underlying physical processes. Many existing diagnostics require only re-aiming to provide proper coverage of the initial 23 cm long divertor plasma configuration (X- point to floor distance). Other diagnostics need extensive reconfiguration using in-vessel fiber-optic bundles or high power laser mirrors. The new divertor baffle panels provide a protective shelf for diagnostic hardware mounted underneath them, but the water cooling channels in the panels limit the permissible size of through holes and, thereby, restrict the available views of under-the- baffle diagnostics. The successful resolution of the design and implementation of these diagnostic modifications is dependent on a strong coordination between GA and its many diagnostic collaborators

  17. Kinetic divertor modeling

    International Nuclear Information System (INIS)

    Highlights: ► We have studied the coupling among gas, plasma and surface in the divertor region. ► A one-dimensional PIC-DSMC model has been developed. ► Profiles of density and temperature of all the species involved have been provided. ► MAR processes are effective in a region smaller than 1.5 mm from the divertor plate. ► For regions more distant, the ionization of atoms, produced by MAR, starts to occur. - Abstract: The coupled dynamics and kinetics between gas and plasma in the divertor region is studied by means of a one-dimensional Particle in Cell-Direct Simulation Monte Carlo (PIC-DSMC) model. In particular, the collision-induced vibrational excitation/relaxation of H2 molecules and particle–surface interaction (vibrational relaxation and recombinative desorption) have been considered in detail to estimate the importance of plasma volumetric recombination by molecular assisted reaction (MAR). Spatially resolved results show that MAR processes are effective very close to the divertor plate in a region smaller than 1.5 mm from the divertor plate. For regions more distant the ionization of atoms, produced by MAR, starts to make molecular assisted recombination an ineffective reaction.

  18. Innovative divertor concepts for LHD

    International Nuclear Information System (INIS)

    We are developing various innovative divertor concepts which improve the LHD plasma performance. These are two divertor magnetic geometries (helical and local island divertors), three operational scenarios (radiative cooling in the high density, cold boundary, confinement improvement by generating high temperature divertor plasma and simultaneous achievement of radiative cooling and H-mode like confinement improvement) and technological development of new efficient hydrogen pumping schemes. (author)

  19. Magnetic divertor design for the compact reversed-field pinch reactor

    International Nuclear Information System (INIS)

    A recently completed design of a pumped-limiter-based Compact Reversed-Field Pinch Reactor is used to estimate for the first time the impact of magnetic divertors. A range of divertor options for the low-toroidal-field RFP is examined, and a design selection is made constrained by consideration of field ripple (magnetic island), blanket displacement, recirculating power, cost, heat flux, and access. Design choices based on diversion of minority (toroidal) field lead to a preference for (poloidally) symmetric or bundle divertor geometries

  20. Divertor for a torsatron

    International Nuclear Information System (INIS)

    The divertor for a torsatron comprising a toroidal vacuum chamber embracing the toroidal chamber of torsatron trap and communicating with it through the gaps between helical conductors of the system for creation of the trap magnetic field is described. The divertor comprises also a collector realized in a form of plates crossing magnetic field force lines. With the purpose of decreasing the plasma contamination level the collector plates realized curvilinear and embrace conductors at full their length and have the curvature less than that of the magnetic field force lines in the plate mounting point. The invention permits to decrease the plasma contamination by decreasing the particles flux formed as a result of collector plates errosion and accordingly increase plasma temperature in the trap

  1. Divertor plasma detachment

    Science.gov (United States)

    Krasheninnikov, S. I.; Kukushkin, A. S.; Pshenov, A. A.

    2016-05-01

    Regime with the plasma detached from the divertor targets (detached divertor regime) is a natural continuation of the high recycling conditions to higher density and stronger impurity radiation loss. Both the theoretical considerations and experimental data show clearly that the increase of the impurity radiation loss and volumetric plasma recombination causes the rollover of the plasma flux to the target when the density increases, which is the manifestation of detachment. Plasma-neutral friction (neutral viscosity effects), although important for the sustainment of high density/pressure plasma upstream and providing the conditions for efficient recombination and power loss, is not directly involved in the reduction of the plasma flux to the targets. The stability of detachment is also discussed.

  2. The JET divertor coil

    International Nuclear Information System (INIS)

    The divertor coil is mounted inside the Jet vacuum vessel and is able to carry 1 MA turns. It is of conventional construction - water cooled copper, epoxy glass insulation -and is contained in a thin stainless steel case. The coil has to be assembled, insulated and encased inside the Jet vacuum vessel. A description of the coil is given, together with technical information (including mechanical effects on the vacuum vessel), an outline of the manufacture process and a time schedule. (author)

  3. Bundling biodiversity

    OpenAIRE

    Heal, Geoffrey

    2002-01-01

    Biodiversity provides essential services to human societies. Many of these services are provided as public goods, so that they will typically be underprovided both by market mechanisms (because of the impossibility of excluding non-payers from using the services) and by government-run systems (because of the free rider problem). I suggest here that in some cases the public goods provided by biodiversity conservation can be bundled with private goods and their value to consumers captured in th...

  4. Numerical studies on divertor experiments

    International Nuclear Information System (INIS)

    Numerical analysis on the divertor experiments such as JFT-2M tokamak is made by use of the two-dimensional time-dependent simulation code. The plasma in the scrape-off layer (SOL) and divertor region is solved for the given particle and heat sources from the main plasma, Γp and QT. Effect of the direction of the toroidal magnetic field is studied. It is found that the heat flux which is proportional to b vector x ∇Ti has influences on the divertor plasmas, but has a small effect on the parameters on the midplane in the framework of the fluid model. Parameter survey on Γp and QT is made. The transient response of the SOL/divertor plasma to the sudden change of Γp and QT is studied. Time delay in the SOL and divertor region is calculated. (author)

  5. PDX divertor operation

    International Nuclear Information System (INIS)

    PDX was brought into operation in January 1980 as a diverted tokamak with typical parameters of Bsub(T) = 15-20 kG, a = 38 cm, R0 = 123-159 cm, Isub(p) = 180-300 kA, q approx. equal to 3.7, anti nsub(e) = 1-3.8 x 1013 cm-3, anti Z = 1.1-3, tausub(E)sub(e) approx. equal to 25 ms, and pulse lengths up to 0.7 s. Internal vacuum components that were exposed to the plasma (such as limiters, shields, microwave horns, etc.) were fabricated from 99% pure titanium. Glow discharge cleaning with 3 x 10-2 Torr H2 and pulse discharge cleaning were used to condition the vessel for high power discharges. For the divertor studies, work has concentrated on obtaining long, high current stable discharges. Radial position, plasma current, and gas injection control systems have been used to facilitate this effort. Discharges of inside-D, square, and inverse-D cross-section have been produced. Microwave interferometers, spectroscopy, an X-ray pulse height analyzer system, scanning and fixed bolometers, and thermocouple array have been used to determine plasma and impurity densities, temperature, radiation, and power loss to the divertor. A comparison of diverted and undiverted discharges is presented. (orig.)

  6. Investigation of tokamak solid-divertor target options

    International Nuclear Information System (INIS)

    Analysis of survival constraints on the design of solid targets for tokamak bundle divertors is presented. Previous target design efforts are reviewed. Considerations of heat removal, surface erosion, and fatigue life are included in a generalized design window methodology which facilitates target selection. Using subcooled water as coolant, eight possible target materials are evaluated for use in tubular and plate targets as substrates, coatings, and claddings. Subject to the severe environment of the tokamak plasma, the most promising conventional designs are identified. A thermally bonded, mechanically unbonded laminated design is proposed and evaluated as a target design well suited to the divertor target environment. Due to fatigue and sputtering erosion this configuration has limited life, but appears to constitute an upper bound for the capabilities of a solid target design. Needs for experimental work are identified

  7. Understanding impurity retention by divertors

    International Nuclear Information System (INIS)

    Simple, 1-D fluid model prescriptions are developed to predict under what circumstances impurities released at divertor targets would be expected to leak to the main plasma. The prescriptions are tested by comparison with results using the DIVIMP (divertor impurity) Monte Carlo code and are found to be well satisfied under strongly collisional conditions. The transition to collisionlessness degrades the agreement with the simple model. Usually, the simple model predicts a more-or-less catastrophic buildup of impurities outside the divertor. This, however, is an artificial result arising from the assumption of strictly one-dimensional, along B, motion; even weak cross-field transport can stop such impurity accumulation. ((orig.))

  8. High temperature divertor plasma operation

    International Nuclear Information System (INIS)

    High temperature divertor plasma operation has been proposed, which is expected to enhance the core energy confinement and eliminates the heat removal problem. In this approach, the heat flux is guided through divertor channel to a remote area with a large target surface, resulting in low heat load on the target plate. This allows pumping of the particles escaping from the core and hence maintaining of the high divertor temperature, which is comparable to the core temperature. The energy confinement is then determined by the diffusion coefficient of the core plasma, which has been observed to be much lower than the thermal diffusivity. (author)

  9. Fuel bundle

    International Nuclear Information System (INIS)

    This patent describes a method of forming a fuel bundle of a nuclear reactor. The method consists of positioning the fuel rods in the bottom plate, positioning the tie rod in the bottom plate with the key passed through the receptacle to the underside of the bottom plate and, after the tie rod is so positioned, turning the tie rod so that the key is in engagement with the underside of the bottom plate. Thereafter mounting the top plate is mounted in engagement with the fuel rods with the upper end of the tie rod extending through the opening in the top plate and extending above the top plate, and the tie rod is secured to the upper side of sid top plate thus simultaneously securing the key to the underside of the bottom plate

  10. The divertor remote maintenance project

    International Nuclear Information System (INIS)

    Remote replacement of the ITER divertor will be required several times during the life of ITER. To facilitate its regular exchange, the divertor is assembled in the ITER vacuum vessel from 60 cassettes. Radial movers transport each cassette along radial rails through the handling ports and into the vessel where a toroidal mover lifts and transports the cassette around a pair of toroidal rails. Once at its final position the cassette is locked to the toroidal rails and is accurately aligned in both poloidal and toroidal directions. A further requirement on the divertor is to minimise the amount of activated waste to be sent to a repository. To this end the cassettes have been designed to allow the remote replacement, in a hot cell, of their plasma facing components. The paper describes the two facilities built at ENEA Brasimone, Italy, whose aim is to demonstrate the reliable remote maintenance of the divertor cassettes. (author)

  11. R.H. divertor maintenance-the divertor refurbishment platform

    International Nuclear Information System (INIS)

    The ITER divertor assembly consists in 60 cassettes located in the bottom region of the vacuum vessel. Because of erosion and damage during, reactor operations, their replacement is expected to be required eight times during the machine lifetime. The cassettes will be withdrawn from the vessel through dedicated ducts and they will be transported to a hot cell for refurbishment. The divertor refurbishment platform (DRP) simulates the arrangement in the divertor hot cell for cassette inspection, component replacement and repair, measuring, and testing. The DRP had to demonstrate the feasibility of divertor cassette refurbishment, procedures, and the use of conventional remote handling equipment in a hot cell, for the refurbishment of high heat flux components (also called plasma facing components PFC), cassette locking systems, water feeds and post-repair, integrity testing. The true environmental conditions (temperature, atmosphere, radiation, contamination) have not been replicated in the DRP, but they were taken into account in the development of the mock ups, the remote handling equipment, and the operating procedures. The results permit to validate the hot cell operations for the cassette refurbishment and to specify the hot cell requirements. This paper describes the objectives, lay-out, test programme, test results, and future activities of the divertor refurbishment platform

  12. Edge plasma in snowflake divertor

    International Nuclear Information System (INIS)

    The snowflake divertor (Ryutov 2007, Phys. Plasmas 14, 064502) uses a 2nd order null of the poloidal magnetic field instead of the 1st order null used in the standard divertor. This leads to a number of interesting geometric properties such as stronger fanning of the poloidal flux, stronger magnetic shear in the edge region, larger radiating volume, and larger connection length in the scrape-off layer. These can potentially lead to new ways for alleviating heat loads on the divertor target plates. Discussion of properties of snowflake is presented, along with results of numerical modeling. Divertor leg volume is larger in snowflake than in the standard x-point configuration, which leads to larger fraction of radiated power in the divertor. This allows the snowflake to transition to a strongly detached plasma regime more easily than for the standard x-point. Besides, stronger shearing of the magnetic field in snowflake may be beneficial for controlling magneto-hydrodynamic instabilities in the edge (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  13. Detached divertor plasmas in JET

    Energy Technology Data Exchange (ETDEWEB)

    Horton, L.D.; Borrass, K.; Corrigan, G.; Gottardi, N.; Lingertat, J.; Loarte, A.; Simonini, R.; Stamp, M.F.; Taroni, A. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking; Stangeby, P.C. [Toronto Univ., ON (Canada). Inst. for Aerospace Studies

    1994-07-01

    In simulations with high radiated power fractions, it is possible to produce the drop in ion current to the divertor targets typical of detached plasmas. Despite the fact that these experiments are performed on beryllium target tiles, radiation from deuterium and beryllium cannot account for the measured power losses. The neutral deuterium levels in the SOL in these plasmas are higher than the model predicts. This may be due to leakage from the divertor or to additional wall sources related to the non-steady nature of these plasmas. In contrast, a surprisingly high level of carbon is present in these discharges; higher even than would be predicted are the divertor target tiles pure carbon. This level may well be large enough to produce the measured radiation. (authors). 6 refs., 2 figs., 1 tab.

  14. Actively convected liquid metal divertor

    Science.gov (United States)

    Shimada, Michiya; Hirooka, Yoshi

    2014-12-01

    The use of actively convected liquid metals with j × B force is proposed to facilitate heat handling by the divertor, a challenging issue associated with magnetic fusion experiments such as ITER. This issue will be aggravated even more for DEMO and power reactors because the divertor heat load will be significantly higher and yet the use of copper would not be allowed as the heat sink material. Instead, reduced activation ferritic/martensitic steel alloys with heat conductivities substantially lower than that of copper, will be used as the structural materials. The present proposal is to fill the lower part of the vacuum vessel with liquid metals with relatively low melting points and low chemical activities including Ga and Sn. The divertor modules, equipped with electrodes and cooling tubes, are immersed in the liquid metal. The electrode, placed in the middle of the liquid metal, can be biased positively or negatively with respect to the module. The j × B force due to the current between the electrode and the module provides a rotating motion for the liquid metal around the electrodes. The rise in liquid temperature at the separatrix hit point can be maintained at acceptable levels from the operation point of view. As the rotation speed increases, the current in the liquid metal is expected to decrease due to the v × B electromotive force. This rotating motion in the poloidal plane will reduce the divertor heat load significantly. Another important benefit of the convected liquid metal divertor is the fast recovery from unmitigated disruptions. Also, the liquid metal divertor concept eliminates the erosion problem.

  15. Strategic Aspects of Bundling

    International Nuclear Information System (INIS)

    The increase of bundle supply has become widespread in several sectors (for instance in telecommunications and energy fields). This paper review relates strategic aspects of bundling. The main purpose of this paper is to analyze profitability of bundling strategies according to the degree of competition and the characteristics of goods. Moreover, bundling can be used as price discrimination tool, screening device or entry barriers. In monopoly case bundling strategy is efficient to sort consumers in different categories in order to capture a maximum of surplus. However, when competition increases, the profitability on bundling strategies depends on correlation of consumers' reservations values. (author)

  16. Equivariant bundle gerbes

    CERN Document Server

    Murray, Michael K; Stevenson, Danny; Vozzo, Raymond F

    2015-01-01

    We develop the theory of simplicial extensions for bundle gerbes and their characteristic classes. This formalism is used to study descent problems and equivariance for bundle gerbes. We consider in detail two examples: the basic bundle gerbe on a unitary group and a string structure for a principal bundle. We show that the basic bundle gerbe is equivariant for the conjugation action and calculate its characteristic class and that a string structure gives rise to a bundle gerbe which is equivariant for a natural action of the String 2-group.

  17. Tokamak Physics Experiment divertor design

    International Nuclear Information System (INIS)

    The Tokamak Physics Experiment (TPX) tokamak requires a symmetric up/down double-null divertor capable of operation with steady-state heat flux as high as 7.5 MW/m2. The divertor is designed to operate in the radiative mode and employs a deep slot configuration with gas puffing lines to enhance radiative divertor operation. Pumping is provided by cryopumps that pump through eight vertical ports in the floor and ceiling of the vessel. The plasma facing surface is made of carbon-carbon composite blocks (macroblocks) bonded to multiple parallel copper tubes oriented vertically. Water flowing at 6 m/s is used, with the critical heat flux (CHF) margin improved by the use of enhanced heat transfer surfaces. In order to extend the operating period where hands on maintenance is allowed and to also reduce dismantling and disposal costs, the TPX design emphasizes the use of low activation materials. The primary materials used in the divertor are titanium, copper, and carbon-carbon composite. The low activation material selection and the planned physics operation will allow personnel access into the vacuum vessel for the first 2 years of operation. The remote handling system requires that all plasma facing components (PFCs) are configured as modular components of restricted dimensions with special provisions for lifting, alignment, mounting, attachment, and connection of cooling lines, and instrumentation and diagnostics services

  18. Bundling in Telecommunications

    OpenAIRE

    Begoña García-Mariñoso; Xavier Martinez-Giralt; Pau Olivella

    2008-01-01

    The paper offers an overview of the literature on bundling in the telecommunications sector and its application in the Spanish market. We argue that the use of bundling in the provision of services is associated to technological reasons. Therefore, there appears no need to regulate bundling activities. However, this is not to say that other related telecom markets should not be scrutinized and regulated, or that the regulator should not pay attention to other bundling-related anticompetitive ...

  19. R.H. divertor maintenance -- the divertor test platform

    International Nuclear Information System (INIS)

    The ITER divertor assembly consists in 60 cassettes located in the bottom region of the vacuum vessel. Because of erosion and damage, their replacement is expected to be required eight times during the machine lifetime. The cassettes will be remotely withdrawn from the vessel through dedicated ducts and they will be transported to a hot cell for refurbishment. To demonstrate the feasibility of the withdrawal operations, and to optimise the maintenance scenario and the handling equipment design, a test facility has been set-up at the ENEA Research Centre of Brasimone (Italy), i.e. the divertor test platform (DTP) that allows to simulate, in full scale, all handling operations inside the vacuum vessel. This paper describes the objectives, test programme, layout, test results and future activities of the DTP

  20. Control of divertor geometry and performance of the ergodic divertor of Tore Supra

    International Nuclear Information System (INIS)

    Experimental evidence of the location of the ergodic divertor separatrix is shown to agree with the predicted value given by codes. Variation of this position modifies the divertor tightness, defined as the ratio of the divertor to core density. This effect is governed by laminar transport, i.e., transport proportional to the magnitude of the perturbation. Operation with feedback control of the divertor temperature allows one to optimise the choice of injected impurity species. At 10 eV divertor temperature, nitrogen is shown to lead to the largest decrease in energy flux to the divertor at lowest contribution to Zeff. Parallel energy fluxes as low as 2 MW m-2 are thus achieved on the target plates. For this impurity, radiation is localised in the divertor volume thus leading to radiation compression close to 10. The ergodic divertor appears as a powerful tool to control plasma-wall interaction with no loss of core confinement or plasma current

  1. Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake

    Energy Technology Data Exchange (ETDEWEB)

    Kotschenreuther, Mike; Valanju, Prashant; Covele, Brent; Mahajan, Swadesh [Institute for Fusion Studies, The University of Texas at Austin, Austin, Texas 78712 (United States)

    2013-10-15

    Advanced divertors are magnetic geometries where a second X-point is added in the divertor region to address the serious challenges of burning plasma power exhaust. Invoking physical arguments, numerical work, and detailed model magnetic field analysis, we investigate the magnetic field structure of advanced divertors in the physically relevant region for power exhaust—the scrape-off layer. A primary result of our analysis is the emergence of a physical “metric,” the Divertor Index DI, which quantifies the flux expansion increase as one goes from the main X-point to the strike point. It clearly separates three geometries with distinct consequences for divertor physics—the Standard Divertor (DI = 1), and two advanced geometries—the X-Divertor (XD, DI > 1) and the Snowflake (DI < 1). The XD, therefore, cannot be classified as one variant of the Snowflake. By this measure, recent National Spherical Torus Experiment and DIIID experiments are X-Divertors, not Snowflakes.

  2. Kinetic Modeling of Divertor Plasma

    Science.gov (United States)

    Ishiguro, Seiji; Hasegawa, Hiroki; Pianpanit, Theerasarn

    2015-11-01

    Particle-in-Cell (PIC) simulation with the Monte Carlo collisions and the cumulative scattering angle coulomb collision can present kinetic dynamics of divertor plasmas. We are developing two types of PIC codes. The first one is the three dimensional bounded PIC code where three dimensional kinetic dynamics of blob is studied and current flow structures related to sheath formation are unveiled. The second one is the one spatial three velocity space dimensional (1D3V) PIC code with the Monte Carlo collisions where formation of detach plasma is studied. First target of our research is to construct self-consistent full kinetic simulation modeling of the linear divertor simulation experiments. This work is performed with the support and under the auspices of NIFS Collaboration Research program (NIFS15KNSS059, NIFS14KNXN279, and NIFS13KNSS038) and the Research Cooperation Program on Hierarchy and Holism in Natural Science at NINS.

  3. Development of divertor remote maintenance system

    International Nuclear Information System (INIS)

    The ITER divertor is categorized as a scheduled maintenance component because of extreme heat and particle loads it is exposed to by plasma. It is also highly activated by 14 MeV neutrons. Reliable remote handling equipment and tools are required for divertor maintenance under intense gamma radiation. To facilitate remote maintenance, the divertor is segmented into 60 cassettes, and each cassette weighing about 25 tons and maintained and replaced through four maintenance ports each 90 degrees. Divertor cassettes must be transported toroidally and radially for replacement through maintenance ports. Remote handling involving cassette movers and carriers for toroidal and radial transport has been developed. Under the ITER R and D program, technology critical to divertor cassette maintenance is being developed jointly by Japan, E.U., and U.S. home teams. This paper summarizes divertor remote maintenance design and the status of technology development by the Japan Home Team. (author)

  4. Simulation Analysis of Divertor Performance in EAST

    Institute of Scientific and Technical Information of China (English)

    Zhu Sizheng; Zha Xuejun

    2005-01-01

    A detailed study of the divertor performance in the EAST has been conducted for both its double null and single null configurations. The results of the application of the SOLPS (B2/Eirene) code package to the analysis of the EAST divertor are summarized. Here we concentrate on the effects of the increased geometrical closure and variation in the magnetic topology on the behavior of divertor plasmas. The results of numerical predictions for the EAST divertor's operational window are also described in this paper.

  5. Bundling and Tying

    OpenAIRE

    Nicholas Economides

    2014-01-01

    We discuss strategic ways that sellers can use tying and bundling with requirement conditions to extract consumer surplus. We analyze different types of tying and bundling creating (i) intra-product price discrimination; (ii) intra-consumer price discrimination; and (iii) inter-product price discrimination, and assess the antitrust liability that these practices may entail. We also discuss the impact on consumers and competition, as well as potential antitrust liability of bundling “incontest...

  6. SOL–divertor plasma simulations introducing anisotropic temperature with virtual divertor model

    International Nuclear Information System (INIS)

    A 1D SOL–divertor plasma simulation code by introducing the anisotropic ion temperature with virtual divertor model has been developed. By introducing the anisotropic ion temperature directly, the second-order derivative parallel ion viscosity term in the momentum transport equation can be excluded and the boundary condition at the divertor plate will not be required in the simulation. In order to express the effects of the divertor plate and accompanying sheath implicitly, a virtual divertor model which has artificial sinks for the particle, momentum and energy has been introduced. Periodic boundary condition becomes available by the use of the virtual divertor model. By using this model, SOL–divertor plasmas which satisfy the Bohm condition has been successfully obtained. The dependence of the ion temperature anisotropy on the normalized mean free path of ion and the validity of the parallel ion viscous flux for the Braginskii expression and the limited one are also investigated

  7. A simple model for biased divertors

    Energy Technology Data Exchange (ETDEWEB)

    Lachambre, J.-L.; Quirion, B.; Gunn, J.; Boucher, C.; Stansfield, B.; Gauvreau, J.-L. [Centre canadien de fusion magnetique, 1804, boulevard Lionel-Boulet, Varennes, Quebec, J3X 1S1 (Canada)

    1997-12-01

    Ionization near the target plate is shown to play an important role in biasing experiments. Our previous SOL model, which calculates the induced radial electric field, is found to be inadequate to treat the new divertor geometry of TdeV. When recycling is included via the measured D{sub {alpha}} emission near the plate, the upgraded model correctly reproduces all the observed electric currents and fields during biasing in the new divertor configuration. A simple divertor model using this calculated field has been developed to simulate the evolution of the divertor ion and neutral parameters under the action of neutralization plate biasing. Using a 1D adiabatic fluid model for the divertor ions, a 1D convective representation for the SOL neutrals and a 0D calculation for the plenum pressure, this divertor model satisfactorily simulates most of the TdeV biasing experiments at all biasing voltages and all toroidal field directions at low line-averaged densities. The weaker agreement at high densities is largely a consequence of the crudeness of the general divertor physics rather than of the deficiency of the biasing physics implemented in the model. The model is finally used to explain the polarity asymmetries observed in divertor efficiencies during biasing, and to demonstrate that no mechanism other than plate current saturation is required to interpret the saturation of toroidal rotation observed in the SOL at large biasing voltages of either polarity. (author)

  8. Moving Divertor Plates in a Tokamak

    International Nuclear Information System (INIS)

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions

  9. ITER-FEAT divertor maintenance and integration

    International Nuclear Information System (INIS)

    This paper presents the design status of the maintenance and integration of the ITER-FEAT divertor. It also includes the first results of a study showing how the in-vessel viewing system could be integrated at the divertor level. The studies are on-going, but already preliminary practical layouts have been produced

  10. Moving Divertor Plates in a Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S.J. Zweben, H. Zhang

    2009-02-12

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions.

  11. Contact fiber bundles

    OpenAIRE

    Lerman, Eugene

    2003-01-01

    We define contact fiber bundles and investigate conditions for the existence of contact structures on the total space of such a bundle. The results are analogous to minimal coupling in symplectic geometry. The two applications are construction of K-contact manifolds generalizing Yamazaki's fiber join construction and a cross-section theorem for contact moment maps

  12. Principal noncommutative torus bundles

    DEFF Research Database (Denmark)

    Echterhoff, Siegfried; Nest, Ryszard; Oyono-Oyono, Herve

    2008-01-01

    of bivariant K-theory (denoted RKK-theory) due to Kasparov. Using earlier results of Echterhoff and Williams, we shall give a complete classification of principal non-commutative torus bundles up to equivariant Morita equivalence. We then study these bundles as topological fibrations (forgetting the...

  13. Rapidly Moving Divertor Plates In A Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S. Zweben

    2011-05-16

    It may be possible to replace conventional actively cooled tokamak divertor plates with a set of rapidly moving, passively cooled divertor plates on rails. These plates would absorb the plasma heat flux with their thermal inertia for ~10-30 sec, and would then be removed from the vessel for processing. When outside the tokamak, these plates could be cooled, cleaned, recoated, inspected, and then returned to the vessel in an automated loop. This scheme could provide nearoptimal divertor surfaces at all times, and avoid the need to stop machine operation for repair of damaged or eroded plates. We describe various possible divertor plate designs and access geometries, and discuss an initial design for a movable and removable divertor module for NSTX-U.

  14. Rapidly Moving Divertor Plates In A Tokamak

    International Nuclear Information System (INIS)

    It may be possible to replace conventional actively cooled tokamak divertor plates with a set of rapidly moving, passively cooled divertor plates on rails. These plates would absorb the plasma heat flux with their thermal inertia for ∼10-30 sec, and would then be removed from the vessel for processing. When outside the tokamak, these plates could be cooled, cleaned, recoated, inspected, and then returned to the vessel in an automated loop. This scheme could provide nearoptimal divertor surfaces at all times, and avoid the need to stop machine operation for repair of damaged or eroded plates. We describe various possible divertor plate designs and access geometries, and discuss an initial design for a movable and removable divertor module for NSTX-U.

  15. Divertor and gas blanket impurity control study

    International Nuclear Information System (INIS)

    A simple calculational model for the transport of particles across the scrap off region between the plasma and the wall in the presence of a divertor or a gas blanket has been developed. The model departs from previous work in including: (a) the entire impurity transport as well as its effect on the energy balance equations; (b) the recycling neutrals from the divertor, and (c) the reflected neutrals from the wall. Results obtained with this model show how the steady state impurity level in the plasma depends on the divertor parameters such as the neutral backflow from the divertor, the particle residence time and the scrape off thickness; and on the gas blanket parameters such as the neutral source strength and the gas blanket thickness. The variation of the divertor or gas blanket performance as a function of the heat and particle fluxes escaping from the plasma, the wall material and the cross field diffusion is examined and numerical examples are given

  16. Scrape-off layer and divertor theory meeting: Proceedings

    International Nuclear Information System (INIS)

    This report contains viewgraphs on the following topics: fluid modelling of neutrals in the SOL and divertor; instabilities of gas-fueled divertors: theory and adaptive simulations; stability of ionization fronts of gaseous divertor plasmas; monte carlo calculation of heat transport; reduced charge model for edge impurity flows; thermally collapsed solutions for gaseous/radiative divertors; adaptive grid methods in transport simulation; advanced numerical solution algorithms applied to the multispecies edge plasma equations; two-dimensional edge plasma simulation using the multigrid method; neutral behavior and the effects of neutral-neutral and neutral-ion elastic scattering in the ITER gaseous divertor; particle throughput in the TPX divertor; marfes in tokamaks; a comparative study of the limiter and divertor edge plasmas in TEXT-U; issues of toroidal tokamak-type divertor simulators; ASDEX upgrade; the ITER divertor; the DIII-D divertor program and TPX divertor; DEGAS 2: a transmission/escape probabilities model for neutral particle transport: comparison with DEGAS 2; a collisional radiative model of hydrogen for high recycling divertors; comparison of fluid and non- fluid neutral models in B2.5; DIII-D radiative divertor simulations; 3-D fluid simulations of turbulence from conducting wall mode; turbulence and drifts in SOL plasmas; recent results for 1 1/2-D ITER gas target divertor modelling; evaluation of pumping and fueling in coupled core, SOL, and divertor chamber calculations; and ITER gas target divertors: comparison of volume recombination and large radial transport scenarios using DEGAS

  17. Magnetic divertors for experimental Tokamaks and fusion reactors

    International Nuclear Information System (INIS)

    Brief reports of working group discussions. These covered the requirements for a divertor in a fusion reactor including reducing impurities, exhausting the plasma and controlling the plasma-wall interactions. Divertor configurations were also reviewed and their merits and disadvantages compared. Existing divertor experiments were summarised and recommendations for further work made. Then the problems anticipated in designing a divertor for a conceptual reactor were considered. The physics of divertors and the scrape-off layer was discussed with reference to present models of plasma in divertors. Finally, experiments needed to demonstrate the feasibility of divertors for reactors and the development of specialised diagnostics for such experiments were considered. (U.K.)

  18. Dust divertor for a tokamak fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tang, X Z [Los Alamos National Laboratory; Delzanno, G L [Los Alamos National Laboratory

    2009-01-01

    Micron-size tungsten particulates find equilibrium position in the magnetized plasma sheath in the normal direction of the divertor surface, but are convected poloidally and toroidally by the sonic-ion-flow drag parallel to the divertor surface. The natural circulation of dust particles in the magnetized plasma sheath can be used to set up a flowing dust shield that absorbs and exhausts most of the tokamak heat flux to the divertor. The size of the particulates and the choice of materials offer substantial room for optimization.

  19. Engineering design of a radiative divertor for DIII-D

    International Nuclear Information System (INIS)

    A new divertor configuration is being developed for the DIII-D tokamak. This divertor will operate in the radiative mode. Experiments and modeling form the basis for the new design. The Radiative Divertor reduces the heat flux on the divertor plates by dispersing the power with radiation in the divertor region. In addition, the Radiative Divertor structure will allow density control in plasma shapes required for advanced tokamak operation. The divertor structure allows for operation in either double-null or single-null plasma configurations. Four independently controlled divertor cryopumps will enable pumping at either the inboard (upper and lower) or the outboard (upper and lower) divertor plates. An upgrade to the DIII-D cryogenic system is part of this project. The increased capabilities of the cryogenic system will allow delivery of liquid helium and nitrogen to the three new cryopumps. The Radiative Divertor design is very flexible, and will allow physics studies of the effects of slot width and length. Radiative Divertor diagnostics are being designed in parallel to provide comprehensive measurements for diagnosing the divertor. The Radiative divertor installation is scheduled for late 1996. Engineering experience gained in the DIII-D Advanced Divertor program form a foundation for the design work on the Radiative Divertor

  20. Restrictions of stable bundles

    CERN Document Server

    Balaji, V

    2011-01-01

    The Mehta-Ramanathan theorem ensures that the restriction of a stable vector bundle to a sufficiently high degree complete intersection curve is again stable. We improve the bounds for the "sufficiently high degree" and propose a possibly optimal conjecture.

  1. candu fuel bundle fabrication

    International Nuclear Information System (INIS)

    This paper describes works on CANDU fuel bundle fabrication in the Fuel Fabrication Development and Testing Section (FFDT) of AECL's Chalk River Laboratories. This work does not cover fuel design, pellet manufacturing, Zircaloy material manufacturing, but cover the joining of appendages to sheath tube, endcap preparation and welding, UO2 loading, end plate preparation and welding, and all inspections required in these steps. Materials used in the fabrication of CANDU fuel bundle are: 1)Ceramic UO2 Pellet 2)Zircaloy -4. Fuel Bundle Structural Material 3) Others (Zinc stearate, Colloidal graphite, Beryllium and Heium). Th fabrication of fuel element consist of three process: 1)pellet loading into the sheats, 2) endcap welding, and 3) the element profiling. Endcap welds is tested by metallography and He leak test. The endcaps of the elements are welded to the end plates to form the 37- element bundle assembly

  2. MAST-Upgrade Divertor Facility and Assessing Performance of Long-Legged Divertors

    CERN Document Server

    Fishpool, G; Cunningham, G; Harrison, J; Katramados, I; Kirk, A; Kovari, M; Meyer, H; Scannell, R

    2013-01-01

    A potentially important feature in a divertor design for a high-power tokamak is an extended and expanded divertor leg. The upgrade to MAST will allow a wide range of such divertor leg geometries to be produced, and hence will allow the roles of greatly increased connection length and flux expansion to be experimentally tested. This will include testing the potential of the Super-X configuration [1]. The design process for the upgrade has required analysis of producing and controlling the magnetic configurations, and has included consideration of the roles that divertor closure and increasing magnetic connection length will play.

  3. Stability, divertors and innovative concepts

    International Nuclear Information System (INIS)

    This paper contains a short resume of the sections on 'Stability, Divertors and Innovative Concepts' presented at the 19th IAEA Fusion Energy Conference. The main conclusions are: (1) the problem of type I ELMs in tokamaks seems to be not so dramatic; (2) it was demonstrated that the working pulse length of large thermonuclear devices can achieve 100 s and more; (3) the problem of tritium retention seems to be not so dramatic now; probable approaches of its solution are visible; (4) active methods of plasma instabilities suppression (NTM, RWM, sawteeth, external MHD) work successfully; (5) new methods of mitigation of the disruption consequences were offered. New technological ideas and new ideas on magnetic confinement were presented. (author)

  4. Impurity radiation modulations in an ergodic divertor

    Energy Technology Data Exchange (ETDEWEB)

    Laugier, F. E-mail: laugier@pegase.cad.cea.fr; Becoulet, M.; De Michelis, C.; Ghendrih, Ph.; Gunn, J.P.; Monier-Garbet, P.; Reichle, R.; Vallet, J.C

    2001-03-01

    The 3-D geometry of radiation losses is investigated in the Tore Supra ergodic divertor. Measurements from passive bolometers located on the divertor coils show evidence of toroidal and poloidal radiation modulations. They were interpreted using a 3-D code solving heat transport equation that gives the whole geometry of plasma radiation in a divertor configuration close to Tore Supra. The results of the code are in qualitative agreement with the measurements and they show that the total radiated power is underestimated when inferred from standard bolometers located between divertor modules. Maximum of radiation in front of the modules is explained by the multiplication of radiative zones at this place due to the intersection of field lines with the vessel wall. This effect leads to non-monotonic temperature profiles along field lines in the boundary plasma.

  5. Impurity radiation modulations in an ergodic divertor

    International Nuclear Information System (INIS)

    The 3-D geometry of radiation losses is investigated in the Tore Supra ergodic divertor. Measurements from passive bolometers located on the divertor coils show evidence of toroidal and poloidal radiation modulations. They were interpreted using a 3-D code solving heat transport equation that gives the whole geometry of plasma radiation in a divertor configuration close to Tore Supra. The results of the code are in qualitative agreement with the measurements and they show that the total radiated power is underestimated when inferred from standard bolometers located between divertor modules. Maximum of radiation in front of the modules is explained by the multiplication of radiative zones at this place due to the intersection of field lines with the vessel wall. This effect leads to non-monotonic temperature profiles along field lines in the boundary plasma

  6. High flux expansion divertor studies in NSTX

    CERN Document Server

    Soukhanovskii, V A; Bell, R E; Gates, D A; Kaita, R; Kugel, H W; LeBlanc, B P; Maqueda, R; Menard, J E; Mueller, D; Paul, S F; Raman, R; Roquemore, A L

    2009-01-01

    High flux expansion divertor studies have been carried out in the National Spherical Torus Experiment using steady-state X-point height variations from 22 to 5-6 cm. Small-ELM H-mode confinement was maintained at all X-point heights. Divertor flux expansions from 6 to 26-28 were obtained, with associated reduction in X-point connection length from 5-6 m to 2 m. Peak divertor heat flux was reduced from 7-8 MW/m$^2$ to 1-2 MW/m$^2$. In low X-point configuration, outer strike point became nearly detached. Among factors affecting deposition of parallel heat flux in the divertor, the flux expansion factor appeared to be dominant

  7. Subtleties Concerning Conformal Tractor Bundles

    CERN Document Server

    Graham, C Robin

    2012-01-01

    The realization of tractor bundles as associated bundles in conformal geometry is studied. It is shown that different natural choices of principal bundle with normal Cartan connection corresponding to a given conformal manifold can give rise to topologically distinct associated tractor bundles for the same inducing representation. Consequences for homogeneous models and conformal holonomy are described. A careful presentation is made of background material concerning standard tractor bundles and equivalence between parabolic geometries and underlying structures.

  8. Radiative divertor and SOL experiments in open and baffled divertors on DIII-D

    International Nuclear Information System (INIS)

    We present recent progress towards an understanding of the physical processes in the divertor and scrape-off-layer (SOL) plasmas in DIII-D. This has been made possible by a combination of new diagnostics, improved computational models, and changes in divertor geometry. We have focused primarily on ELMing H-mode discharges. The physics of Partially Detached Divertor (PDD) plasmas, with divertor heat flux reduction by divertor radiation enhancement using D2 puffing, has been studied in 2-D, and a model of the heat and particle transport has been developed that includes conduction, convection, ionization, recombination, and flows. Plasma and impurity particle flows have been measured with Mach probes and spectroscopy and these flows have been compared with the UEDGE model. The model now includes self-consistent calculations of carbon impurities. Impurity radiation has been increased in the divertor and SOL with 'puff and pump' techniques using SOL D2 puffing, divertor cryopumping, and argon puffing. The important physical processes in plasma-wall interactions have been examined with a DiMES probe, plasma characterization near the divertor plate, and the REDEP code. Experiments comparing single-null (SN) plasma operation in baffled and open divertors have demonstrated a change in the edge plasma profiles. These results are consistent with a reduction in the core ionization source calculated with UEDGE. Divertor particle control in ELMing H-mode with pumping and baffling has resulted in reduction in H-mode core densities to ne/ngw=0.25. Divertor particle exhaust and heat flux has been studied as the plasma shape was varied from a lower SN, to a balanced double null (DN), and finally to an upper SN. (author)

  9. Conceptual design of pebble drop divertor

    International Nuclear Information System (INIS)

    A pebble drop divertor concept is proposed for future fusion reactor. The marked feature of this system is the use of multi-layer pebbles that consists of a central kernel and some coating layers, as a divertor surface component. By using multi-layer pebbles, pebble drop divertor have the advantages such as steady state wall pumping with low bulk tritium retention. The performance of whole divertor system depends on the characteristics of the multi-layer pebble. Particularly the maximum heat load of the system is determined by the dimensions, the layer structure and the material of a kernel. A kernel also has an important role to determine surface temperature, which affects the wall pumping efficiency. This paper presents the numerical results of the maximum allowable heat load and the surface temperature of the divertor pebble. From the numerical estimation of thermal stress and surface temperature, it is found that the radius of divertor pebble with ceramic kernel should be 0.5 - 1 mm. (author)

  10. Conceptual design of pebble drop divertor

    International Nuclear Information System (INIS)

    A pebble drop divertor concept is proposed for future fusion reactor. The marked feature of this system is the use of multi-layer pebbles that consists of a central kernel and some coating layers, as a divertor surface component. By using multi-layer pebbles, pebble drop divertor have the advantages such as steady state wall pumping with low bulk tritium retention. The performance of whole divertor system depends on the characteristics of the multi-layer pebble. Particularly the maximum heat load of the system is determined by the dimensions, the layer structure and the material of a kernel. A kernel also has an important role to determine surface temperature, which affects the wall pumping efficiency. This paper presents the numerical results of the maximum allowable heat load and the surface temperature of the divertor pebble. From the numerical estimation of thermal stress and surface temperature, it is found that the radius of divertor pebble with ceramic kernel should be 0.5-1 mm. (author)

  11. Divertor Thomson scattering on DIII-D

    International Nuclear Information System (INIS)

    In this paper we describe the newly installed divertor Thomson scattering system for the DIII-D tokamak and present initial results from plasma discharges. Measured plasma densities have ranged from 5 x 1018 to 5 x 1020 m-3 and divertor plasma temperatures from 1 to 500 eV. These data are compared with earlier Langmuir probe data and qualitatively compared with UEDGE computer simulations. The divertor Thomson system uses one of the eight existing core Thomson scattering lasers (1 J, 20 Hz) which has been re-directed to probe the divertor region of the DIII-D vessel. Scattered light from this multipulse Nd:Yag laser is viewed with an f/6.8 collection optics system which provides eight spatial channels from 1-21 cm above the vessel floor (divertor target), each with 1.5 cm vertical resolution. Translating the plasma across the vessel floor using position controls provides a full scan of the divertor plasma. (orig.)

  12. Radiative power loading in the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Guillemaut, C., E-mail: christophe.guillemaut@cea.fr [ITER Organization, Route de Vinon CS 90 046, 13067 Saint-Paul-Lez-Durance (France); Pitts, R.A.; Kukushkin, A.S. [ITER Organization, Route de Vinon CS 90 046, 13067 Saint-Paul-Lez-Durance (France); O' Mullane, M. [Department of Physics, University of Strathclyde, Glasgow G4 0NG (United Kingdom)

    2011-12-15

    In ITER, steady state burning plasma operation will require a partially detached divertor state in order to reduce the peak power flux density to technologically achievable values at the actively cooled target plates ({approx}10 MW m{sup -2}). Such partially detached solutions require high radiative power dissipation in the divertor volume, with 60-70 MW expected in the baseline H-mode operating scenario. Power levels of this magnitude pose potential difficulties for divertor substructures, which, although also actively cooled, are not designed to withstand very high heat fluxes. This paper estimates the radiative power flux densities falling on critical divertor substructures during ITER burning plasma operation using commercial optical ray-tracing software to project radiation distributions simulated with the SOLPS plasma boundary simulation code onto a full 3D description of the divertor. The results indicate that inclusion of the real geometry provides heat flux densities due to photon illumination not higher than quasi-analytic estimates used in the original divertor design stages, and in some cases lower. When applied to the specific simple geometries used to develop the analytic expressions, the raytracing fully validates the analytic approach.

  13. Recent DIII-D divertor research

    International Nuclear Information System (INIS)

    DIII-D currently operates with a single- or double-null open divertor and graphite walls. Active particle control with a divertor cryopump has demonstrated density control, efficient helium exhaust, and reduction of the inventory of particles in the wall. Gas puffing of D2 and impurities has demonstrated reduction of the peak divertor beat flux by factors of 3--5 by radiation. A combination of active cryopumping and feedback-controlled D2 gas puffing has produced similar divertor heat flux reduction with density control. Experiments with neon puffing have shown that the radiation is equally-divided between a localized zone near the X-point and a mantle around the plasma core. The density in these experiments has also been controlled with cryopumping. These experimental results combined with modeling were used to develop the new Radiative Divertor for DIII-D. This is a double-null slot divertor with four cryopumps to provide particle control and neutral shielding for high-triangularity advanced tokamak discharges. UEDGE and DEGAS simulations, benchmarked to experimental data, have been used to optimize the design

  14. Control of divertor geometry and performance of the ergodic divertor of Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Ghendrih, Ph. E-mail: ghendrih@drfc.cad.cea.fr; Becoulet, M.; Costanzo, L.; Corre, Y.; Grisolia, C.; Grosman, A.; Guirlet, R.; Gunn, J.; Loarer, T.; Monier-Garbet, P.; Mank, G.; Reichle, R.; Vallet, J.-C.; Zabiego, M.; Azeroual, A.; Bucalossi, J.; Devynck, P.; De Michelis, C; Finken, K.H.; Hogan, J.; Laugier, F.; Nguyen, F.; Pegourie, B.; Saint-Laurent, F.; Schunke, B

    2001-03-01

    Experimental evidence of the location of the ergodic divertor separatrix is shown to agree with the predicted value given by codes. Variation of this position modifies the divertor tightness, defined as the ratio of the divertor to core density. This effect is governed by laminar transport, i.e., transport proportional to the magnitude of the perturbation. Operation with feedback control of the divertor temperature allows one to optimise the choice of injected impurity species. At 10 eV divertor temperature, nitrogen is shown to lead to the largest decrease in energy flux to the divertor at lowest contribution to Z{sub eff}. Parallel energy fluxes as low as 2 MW m{sup -2} are thus achieved on the target plates. For this impurity, radiation is localised in the divertor volume thus leading to radiation compression close to 10. The ergodic divertor appears as a powerful tool to control plasma-wall interaction with no loss of core confinement or plasma current.

  15. Application of the radiating divertor approach to innovative tokamak divertor concepts

    International Nuclear Information System (INIS)

    We survey the results of recent DIII-D experiments that tested the effectiveness of three innovative tokamak divertor concepts in reducing divertor heat flux while still maintaining acceptable energy confinement under neon/deuterium-based radiating divertor (RD) conditions: (1) magnetically unbalanced high performance double-null divertor (DND) plasmas, (2) high performance double-null “Snowflake” (SF-DN) plasmas, and (3) single-null H-mode plasmas having different isolation from their divertor targets. In general, all three concepts adapt well to RD conditions, achieving significant reduction in divertor heat flux (q⊥p) and maintaining high performance metrics, e.g., 50–70% reduction in peak divertor heat flux for DND and SF-DN plasmas that are characterized by βN ≅ 3.0 and H98(y,2) ≈ 1.35. It is also demonstrated that q⊥p could be reduced ≈50% by extending the parallel connection length (L||-XPT) in the scrape-off layer between the X-point and divertor targets over a variety of the RD and non-RD environments tested

  16. Application of the radiating divertor approach to innovative tokamak divertor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Petrie, T.W., E-mail: petrie@fusion.gat.com [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Allen, S.L.; Fenstermacher, M.E. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Groebner, R.J. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Holcomb, C.T. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Kolemen, E. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); La Haye, R.J. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Lasnier, C.J. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Leonard, A.W.; Luce, T.C. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Maingi, R. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); Moyer, R.A. [University of California San Diego, 9500 Gilman Dr., La Jolla, CA 92093-0417 (United States); Solomon, W.M. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Turco, F. [Columbia University, 2960 Broadway, New York, NY 10027 (United States); Watkins, J.G. [Sandia National Laboratory, PO Box 5800, Albuquerque, NM 87185 (United States)

    2015-08-15

    We survey the results of recent DIII-D experiments that tested the effectiveness of three innovative tokamak divertor concepts in reducing divertor heat flux while still maintaining acceptable energy confinement under neon/deuterium-based radiating divertor (RD) conditions: (1) magnetically unbalanced high performance double-null divertor (DND) plasmas, (2) high performance double-null “Snowflake” (SF-DN) plasmas, and (3) single-null H-mode plasmas having different isolation from their divertor targets. In general, all three concepts adapt well to RD conditions, achieving significant reduction in divertor heat flux (q{sub ⊥p}) and maintaining high performance metrics, e.g., 50–70% reduction in peak divertor heat flux for DND and SF-DN plasmas that are characterized by β{sub N} ≅ 3.0 and H{sub 98(y,2)} ≈ 1.35. It is also demonstrated that q{sub ⊥p} could be reduced ≈50% by extending the parallel connection length (L{sub ||-XPT}) in the scrape-off layer between the X-point and divertor targets over a variety of the RD and non-RD environments tested.

  17. FINAL REPORT FOR THE DIII-D RADIATIVE DIVERTOR PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    O' NEIL, RC; STAMBAUGH, RD

    2002-06-01

    OAK A271 FINAL REPORT FOR THE DIII-D RADIATIVE DIVERTOR PROJECT. The Radiative Divertor Project originated in 1993 when the DIII-D Five Year Plan for the period 1994--1998 was prepared. The Project Information Sheet described the objective of the project as ''to demonstrate dispersal of divertor power by a factor of then with sufficient diagnostics and modeling to extend the results to ITER and TPX''. Key divertor components identified were: (1) Carbon-carbon and graphite armor tiles; (2) The divertor structure providing a gas baffle and cooling; and (3) The divertor cryopumps to pump fuel and impurities.

  18. Damage structure in divertor armor materials exposed to multiple ITER relevant ELM loads

    International Nuclear Information System (INIS)

    The damage threshold and damage mechanisms of divertor armor materials, i.e. CFC and tungsten, were studied under the impact of ITER relevant ELM-like loads. These experiments were carried out in a Quasi-Stationary Plasma Accelerator applying repetitive pulses of 500 μs up to 100 cycles. CFC showed preferential erosion of the PAN fiber-bundles above 0.6 MJ/m2 and cracking of pitch fiber-bundles. Tungsten showed cracking already at 0.2 MJ/m2 and melting at flat surfaces above 1 MJ/m2. Cracks in tungsten were identified as primary and secondary cracks which all propagated in the vertical direction, which was considered to be less critical. At an energy density of 1.5 MJ/m2, the melt-layer completely covered the surface and bridged the castellation slots.

  19. Right bundle branch block

    DEFF Research Database (Denmark)

    Bussink, Barbara E; Holst, Anders Gaarsdal; Jespersen, Lasse;

    2013-01-01

    AimsTo determine the prevalence, predictors of newly acquired, and the prognostic value of right bundle branch block (RBBB) and incomplete RBBB (IRBBB) on a resting 12-lead electrocardiogram in men and women from the general population.Methods and resultsWe followed 18 441 participants included in...... men vs. 0.5%/2.3% in women, P <0.001). Significant predictors of newly acquired RBBB were male gender, increasing age, high systolic blood pressure, and presence of IRBBB, whereas predictors of newly acquired IRBBB were male gender, increasing age, and low BMI. Right bundle branch block was associated...... with significantly increased all-cause and cardiovascular mortality in both genders with age-adjusted hazard ratios (HR) of 1.31 [95% confidence interval (CI), 1.11-1.54] and 1.87 (95% CI, 1.48-2.36) in the gender pooled analysis with little attenuation after multiple adjustment. Right bundle branch...

  20. Principal -bundles on Nodal Curves

    Indian Academy of Sciences (India)

    Usha N Bhosle

    2001-08-01

    Let be a connected semisimple affine algebraic group defined over . We study the relation between stable, semistable -bundles on a nodal curve and representations of the fundamental group of . This study is done by extending the notion of (generalized) parabolic vector bundles to principal -bundles on the desingularization of and using the correspondence between them and principal -bundles on . We give an isomorphism of the stack of generalized parabolic bundles on with a quotient stack associated to loop groups. We show that if is simple and simply connected then the Picard group of the stack of principal -bundles on is isomorphic to ⊕ , being the number of components of .

  1. Integrated core-edge-divertor modeling studies

    International Nuclear Information System (INIS)

    An integrated calculation model for simulating the interaction of physics phenomena taking place in the plasma core, in the plasma edge and in the SOL and divertor of tokamaks has been developed and applied to study such interactions. The model synthesises a combination of numerical calculations (1) the power and particle balances for the core plasma, using empirical confinement scaling laws and taking into account radiation losses (2), the particle, momentum and power balances in the SOL and divertor, taking into account the effects of radiation and recycling neutrals, (3) the transport of feeling and recycling neutrals, explicitly representing divertor and pumping geometry, and (4) edge pedestal gradient scale lengths and widths, evaluation of theoretical predictions (5) confinement degradation due to thermal instabilities in the edge pedestals, (6) detachment and divertor MARFE onset, (7) core MARFE onsets leading to a H-L transition, and (8) radiative collapse leading to a disruption and evaluation of empirical fits (9) power thresholds for the L-H and H-L transitions and (10) the width of the edge pedestals. The various components of the calculation model are coupled and must be iterated to a self-consistent convergence. The model was developed over several years for the purpose of interpreting various edge phenomena observed in DIII-D experiments and thereby, to some extent, has been benchmarked against experiment. Because the model treats the interactions of various phenomena in the core, edge and divertor, yet is computationally efficient, it lends itself to the investigation of the effects of different choices of various edge plasma operating conditions on overall divertor and core plasma performance. Studies of the effect of feeling location and rate, divertor geometry, plasma shape, pumping and over 'edge parameters' on core plasma properties (line average density, confinement, density limit, etc.) have been performed for DIII-D model problems. A

  2. Atomic and molecular processes in JT-60U divertor plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Takenaga, H.; Shimizu, K.; Itami, K. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    1997-01-01

    Atomic and molecular data are indispensable for the understanding of the divertor characteristics, because behavior of particles in the divertor plasma is closely related to the atomic and molecular processes. In the divertor configuration, heat and particles escaping from the main plasma flow onto the divertor plate along the magnetic field lines. In the divertor region, helium ash must be effectively exhausted, and radiation must be enhanced for the reduction of the heat load onto the divertor plate. In order to exhaust helium ash effectively, the difference between behavior of neutral hydrogen (including deuterium and tritium) and helium in the divertor plasma should be understood. Radiation from the divertor plasma generally caused by the impurities which produced by the erosion of the divertor plate and/or injected by gas-puffing. Therefore, it is important to understand impurity behavior in the divertor plasma. The ions hitting the divertor plate recycle through the processes of neutralization, reflection, absorption and desorption at the divertor plates and molecular dissociation, charge-exchange reaction and ionization in the divertor plasma. Behavior of hydrogen, helium and impurities in the divertor plasmas can not be understood without the atomic and molecular data. In this report, recent results of the divertor study related to the atomic and molecular processes in JT-60U were summarized. Behavior of neural deuterium and helium was discussed in section 2. In section 3, the comparisons between the modelling of the carbon impurity transport and the measurements of C II and C IV were discussed. In section 4, characteristics of the radiative divertor using Ne puffing were reported. The new diagnostic method for the electron density and temperature in the divertor plasmas using the intensity ratios of He I lines was described in section 5. (author)

  3. Snowflake divertor configuration studies in National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V. A.; McLean, A. G.; Rognlien, T. D.; Ryutov, D. D.; Umansky, M. V. [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States); Bell, R. E.; Diallo, A.; Gerhardt, S.; Kaye, S.; Kolemen, E.; LeBlanc, B. P.; Menard, J. E.; Paul, S. F.; Podesta, M.; Roquemore, A. L.; Scotti, F.; Battaglia, D.; Bell, M. G.; Gates, D. A.; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); and others

    2012-08-15

    Experimental results from NSTX indicate that the snowflake divertor (D. Ryutov, Phys. Plasmas 14, 064502 (2007)) may be a viable solution for outstanding tokamak plasma-material interface issues. Steady-state handling of divertor heat flux and divertor plate erosion remains to be critical issues for ITER and future concept devices based on conventional and spherical tokamak geometry with high power density divertors. Experiments conducted in 4-6 MW NBI-heated H-mode plasmas in NSTX demonstrated that the snowflake divertor is compatible with high-confinement core plasma operation, while being very effective in steady-state divertor heat flux mitigation and impurity reduction. A steady-state snowflake divertor was obtained in recent NSTX experiments for up to 600 ms using three divertor magnetic coils. The high magnetic flux expansion region of the scrape-off layer (SOL) spanning up to 50% of the SOL width {lambda}{sub q} was partially detached in the snowflake divertor. In the detached zone, the heat flux profile flattened and decreased to 0.5-1 MW/m{sup 2} (from 4-7 MW/m{sup 2} in the standard divertor) indicative of radiative heating. An up to 50% increase in divertor, P{sub rad} in the snowflake divertor was accompanied by broadening of the intrinsic C III and C IV radiation zones, and a nearly order of magnitude increase in divertor high-n Balmer line emission indicative of volumetric recombination onset. Magnetic reconstructions showed that the x-point connection length, divertor plasma-wetted area and divertor volume, all critical parameters for geometric reduction of deposited heat flux, and increased volumetric divertor losses were significantly increased in the snowflake divertor, as expected from theory.

  4. Microturbulence measurements during divertor biasing

    International Nuclear Information System (INIS)

    The application of a bias voltage to a neutralization plate of the upper divertor with respect to the vacuum chamber in the Tokamak de Varennes (TdeV) influences the plasma well inside the separatrix. In particular, the unbiased Ohmic poloidal rotation edge velocity measured by visible spectroscopy is found to be in the electron diamagnetic drift direction (2-3 km/s) and increases by a factor of two for Vbias = 100 V. This coincides with a major reduction of the microturbulence signal at low frequencies (50 kHz -1 -1), as determined from coherent laser scattering measurements. One possible explanation is that the turbulence signal is simply Doppler shifted to frequencies outside the accessible range. This scenario is, however, difficult to reconcile with some observations. Another explanation invokes a reduction of the turbulence level. The variation of the turbulence signal as a function of the applied bias voltage can indeed be reproduced with a theoretical model based on radial and poloidal decorrelation mechanisms, the latter corresponding to poloidal velocity shear stabilization. This model also explains the observed steepening of the k-spectrum decay during biasing. Biasing also modifies the electron density profile inside the separatrix. These changes of nabla ne cannot explain the behaviour of microturbulence behaviour, when explained in terms of stabilization, would agree with the plasma maintaining a steeper electron density gradient. (author). 17 refs, 9 figs

  5. EU R and D on divertor components

    International Nuclear Information System (INIS)

    Since the last SOFT conference held in Helsinki in 2002, substantial progress has been made in the EU R and D on the divertor components. A number of activities have been completed and new ones have been launched. The present paper gives an update of the works carried out by the EU Participating Team in support of the development of the divertor, which is one of the most challenging components of the next-step ITER machine. The following topics are covered: (1) the further development and consolidation of suitable technologies for the production of high heat-flux components, which culminated with the successful manufacturing and testing of a full-scale vertical target prototype; (2) the completion of the post-irradiation testing of divertor mock-ups and samples; (3) the preparation for the hydraulic and assembly tests of a complete set of full-scale divertor components; (4) the on-going R and D on the definition of workable acceptance criteria for the procurement of ITER high heat-flux components; (5) the activities in support of the divertor design

  6. Impurity-induced divertor plasma oscillations

    International Nuclear Information System (INIS)

    Two different oscillatory plasma regimes induced by seeding the plasma with high- and low-Z impurities are found for ITER-like divertor plasmas, using computer modeling with the DUSTT/UEDGE and SOLPS4.3 plasma-impurity transport codes. The oscillations are characterized by significant variations of the impurity-radiated power and of the peak heat load on the divertor targets. Qualitative analysis of the divertor plasma oscillations reveals different mechanisms driving the oscillations in the cases of high- and low-Z impurity seeding. The oscillations caused by the high-Z impurities are excited near the X-point by an impurity-related instability of the radiation-condensation type, accompanied by parallel impurity ion transport affected by the thermal and plasma friction forces. The driving mechanism of the oscillations induced by the low-Z impurities is related to the cross-field transport of the impurity atoms, causing alteration between the high and low plasma temperature regimes in the plasma recycling region near the divertor targets. The implications of the impurity-induced plasma oscillations for divertor operation in the next generation tokamaks are also discussed

  7. Impurity-induced divertor plasma oscillations

    Energy Technology Data Exchange (ETDEWEB)

    Smirnov, R. D., E-mail: rsmirnov@ucsd.edu; Krasheninnikov, S. I.; Pigarov, A. Yu. [University of California, San Diego, La Jolla, California 92093 (United States); Kukushkin, A. S. [NRC “Kurchatov Institute”, Moscow 123182 (Russian Federation); National Research Nuclear University MEPhI, Moscow 115409 (Russian Federation); Rognlien, T. D. [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States)

    2016-01-15

    Two different oscillatory plasma regimes induced by seeding the plasma with high- and low-Z impurities are found for ITER-like divertor plasmas, using computer modeling with the DUSTT/UEDGE and SOLPS4.3 plasma-impurity transport codes. The oscillations are characterized by significant variations of the impurity-radiated power and of the peak heat load on the divertor targets. Qualitative analysis of the divertor plasma oscillations reveals different mechanisms driving the oscillations in the cases of high- and low-Z impurity seeding. The oscillations caused by the high-Z impurities are excited near the X-point by an impurity-related instability of the radiation-condensation type, accompanied by parallel impurity ion transport affected by the thermal and plasma friction forces. The driving mechanism of the oscillations induced by the low-Z impurities is related to the cross-field transport of the impurity atoms, causing alteration between the high and low plasma temperature regimes in the plasma recycling region near the divertor targets. The implications of the impurity-induced plasma oscillations for divertor operation in the next generation tokamaks are also discussed.

  8. Dynamics of flagellar bundling

    Science.gov (United States)

    Janssen, Pieter; Graham, Michael

    2010-11-01

    Flagella are long thin appendages of microscopic organisms used for propulsion in low-Reynolds environments. For E. coli the flagella are driven by a molecular motor, which rotates the flagella in a counter-clockwise motion (CCM). When in a forward swimming motion, all flagella bundle up. If a motor reverses rotation direction, the flagella unbundle and the cell makes a tumbling motion. When all motors turn in the same CC direction again, the flagella bundle up, and forward swimming continues. To investigate the bundling, we consider two flexible helices next to each other, as well as several flagella attached to a spherical body. Each helix is modeled as several prolate spheroids connected at the tips by springs. For hydrodynamic interactions, we consider the flagella to made up of point forces, while the finite size of the body is incorporated via Fax'en's laws. We show that synchronization occurs quickly relative to the bundling process. For flagella next to each other, the initial deflection is generated by rotlet interactions generated by the rotating helices. At longer times, simulations show the flagella only wrap once around each other, but only for flagella that are closer than about 4 helix radii. Finally, we show a run-and-tumble motion of the body with attached flagella.

  9. First experiments on the TO-2 tokamak with a divertor

    International Nuclear Information System (INIS)

    Long stable discharges have been obtained in a recetrack tokamak with toroidal divertors in low plasma density regime. Divertors sharply limit plasma filament cross section, plasma density decreasing by an order at 1 cm length near the separatrix. 8 mm thick well formed flux of plasma appears at the divertor plate. Divertor power efficiency at different modes of operation is 50- 70 %. As compared to the TO-1 nondivertor tokamak some plasma filament hot zone expansion is recorded in the TO-2 tokamak

  10. Divertor E X B Plasma Convection in DIII-D

    International Nuclear Information System (INIS)

    Extensive two-dimensional measurements of plasma potential in the DIII-D tokamak divertor region are reported for standard (ion VBT drift toward divertor X-point) and reversed BT directions; for low (L) and high (H) confinement modes; and for partially detached divertor mode. The data are consistent with recent computational modeling identifying E x BT circulation, due to potentials sustained by plasma gradients, as the main cause of divertor plasma sensitivity to BT direction

  11. Impact of divertor geometry on radiative divertor performance in JET H-mode plasmas

    Science.gov (United States)

    Jaervinen, A. E.; Brezinsek, S.; Giroud, C.; Groth, M.; Guillemaut, C.; Belo, P.; Brix, M.; Corrigan, G.; Drewelow, P.; Harting, D.; Huber, A.; Lawson, K. D.; Lipschultz, B.; Maggi, C. F.; Matthews, G. F.; Meigs, A. G.; Moulton, D.; Stamp, M. F.; Wiesen, S.; Contributors, JET

    2016-04-01

    Radiative divertor operation in JET high confinement mode plasmas with the ITER-like wall has been experimentally investigated and simulated with EDGE2D-EIRENE in horizontal and vertical low field side (LFS) divertor configurations. The simulations show that the LFS divertor heat fluxes are reduced with N2-injection in similar fashion in both configurations, qualitatively consistent with experimental observations. The simulations show no substantial difference between the two configurations in the reduction of the peak LFS heat flux as a function of divertor radiation, nitrogen concentration, or pedestal Zeff. Consistently, experiments show similar divertor radiation and nitrogen injection levels for similar LFS peak heat flux reduction in both configurations. Nevertheless, the LFS strike point is predicted to detach at 20% lower separatrix density in the vertical than in the horizontal configuration. However, since the peak LFS heat flux in partial detachment in the vertical configurations is shifted towards the far scrape-off layer (SOL), the simulations predict no benefit in the reduction of LFS peak heat flux for a given upstream density in the vertical configuration relative to a horizontal one. A factor of 2 reduction of deuterium ionization source inside the separatrix is observed in the simulations when changing to the vertical configuration. The simulations capture the experimentally observed particle and heat flux reduction at the LFS divertor plate in both configurations, when adjusting the impurity injection rate to reproduce the measured divertor radiation. However, the divertor D α -emissions are underestimated by a factor of 2-5, indicating a short-fall in radiation by the fuel species. In the vertical configuration, detachment is experimentally measured and predicted to start next to the strike point, extending towards the far SOL with increasing degree of detachment. In contrast, in the horizontal configuration, the entire divertor particle flux

  12. On framed quantum principal bundles

    CERN Document Server

    Durdevic, M

    1995-01-01

    A noncommutative-geometric formalism of framed principal bundles is sketched, in a special case of quantum bundles (over quantum spaces) possessing classical structure groups. Quantum counterparts of torsion operators and Levi-Civita type connections are analyzed. A construction of a natural differential calculus on framed bundles is described. Illustrative examples are presented.

  13. Analysis of particle transport in a gas target divertor

    Energy Technology Data Exchange (ETDEWEB)

    Ohtsu, Shigeki; Tanaka, Satoru [Tokyo Univ. (Japan). Faculty of Engineering

    1996-10-01

    2-dimensional modelling of divertor plasma was performed with three types of the divertor geometry configuration. Pumping is effective to reduce neutral recycling to core region in the configuration without baffle. In baffle configuration, a good shielding of neutrals in the divertor region can be achieved. The dome configuration reduces plasma density near the null region and flow shear near the separatrix. (author)

  14. Thermal effects of runaway electrons in an armoured divertor

    International Nuclear Information System (INIS)

    This report describes the results of a numerical thermal analysis of the heat deposition of runaway electrons accompanying plasma disruptions in a armoured divertor. The divertor concepts studied are carbon on molybdenum and beryllium on copper. The conclusion is that the runaway electrons can cause melting of the armour as well as melting of the structure and can damage the divertor severely. (orig.)

  15. Liquid metal cooled divertor for ARIES

    International Nuclear Information System (INIS)

    A liquid metal, Ga-cooled divertor design was completed for the double null ARIES-II divertor design. The design analysis indicated a surface heat flux removal capability of up to 15 MW/m2, and its relative easy maintenance. Design issues of configuration, thermal hydraulics, thermal stresses, liquid metal loop and safety effects were evaluated. For coolant flow control, it was found that it is necessary to use some part of the blanket cooling ducts for the draining of liquid metal from the top divertor. In order to minimize the inventory of Ga, it was recommended that the liquid metal loop equipment should be located as close to the torus as possible. More detailed analysis of transient conditions especially under accident conditions was identified as an issue that will need to be addressed

  16. Designing divertor targets for uniform power load

    Science.gov (United States)

    Dekeyser, W.; Reiter, D.; Baelmans, M.

    2015-08-01

    Divertor design for next step fusion reactors heavily relies on 2D edge plasma modeling with codes as e.g. B2-EIRENE. While these codes are typically used in a design-by-analysis approach, in previous work we have shown that divertor design can alternatively be posed as a mathematical optimization problem, and solved very efficiently using adjoint methods adapted from computational aerodynamics. This approach has been applied successfully to divertor target shape design for more uniform power load. In this paper, the concept is further extended to include all contributions to the target power load, with particular focus on radiation. In a simplified test problem, we show the potential benefits of fully including the radiation load in the design cycle as compared to only assessing this load in a post-processing step.

  17. MAST-Upgrade Divertor Facility and Assessing Performance of Long-Legged Divertors

    OpenAIRE

    Fishpool, G.; Canik, J.; Cunningham, G.; Harrison, J.; Katramados, I.; Kirk, A.; Kovari, M.; H. Meyer; Scannell, R.; Team, the MAST-Upgrade

    2013-01-01

    A potentially important feature in a divertor design for a high-power tokamak is an extended and expanded divertor leg. The upgrade to MAST will allow a wide range of such divertor leg geometries to be produced, and hence will allow the roles of greatly increased connection length and flux expansion to be experimentally tested. This will include testing the potential of the Super-X configuration [1]. The design process for the upgrade has required analysis of producing and controlling the mag...

  18. Design Integration of Liquid Surface Divertors

    Energy Technology Data Exchange (ETDEWEB)

    Nygren, R E; Cowgill, D F; Ulrickson, M A; Nelson, B E; Fogarty, P J; Rognlien, T D; Rensink, M E; Hassanein, A; Smolentsev, S S; Kotschenreuther, M

    2003-11-13

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

  19. Control of divertor configuration in JT-60

    International Nuclear Information System (INIS)

    The control algorithm of JT-60 divertor configuration is presented. JT-60 has five types of poloidal magnetic field coil with each power supply in order to regulate the control objectives mentioned above. However, if one controls each objective by each coil current independently, there must inevitably occur large interaction between control objectives. Because the relation between control objectives and coil currents is complicated. This situation may be the same with a fusion reactor device. For making it possible to control each objective independently without causing large interaction, the authors adopt the noninteracting control algorithm. Hence, this report demonstrates the availability of this method to the control of JT-60 divertor configuration

  20. Design of impurity influx monitor (divertor) for ITER

    International Nuclear Information System (INIS)

    Because of the changes of interfaces between ITER and the Impurity Influx Monitor (divertor) accompanied with the change of the ITER design to a reduced size machine, changes in the design of the monitor are required. In this design work, optics compatible with new interfaces, a calibration system and an alignment system of the optical axis and the focus were designed and investigated. The design of the optical systems was simplified to save the cost. To simplify the optics, the design of the collection optics was changed from an off-axis aspherical mirror system, which is a previous design, to a simple Cassegrain telescope system composed of simple spherical mirrors and lenses. In addition, a micro lens array is inserted just in front of the fiber bundle to increase the light detected. The ray-trace analysis shows that the spatial resolution of ITER requirement (50 mm) will be achieved by these optical systems designed here. The in-situ sensitivity calibration will be realized by applying the light on a micro retro-reflector array installed in front of the plasma facing first mirror from the outside the vacuum chamber through the same optics for plasma measurement and measuring the intensity of the reflected light from the array by using the same optics. In addition, optical design of an adjustment system and a focusing system, and a conceptual design of the shutter system of the monitor were carried out. Moreover, the detector system for the monitor was investigated and designed. (author)

  1. Bundles of Banach algebras

    Directory of Open Access Journals (Sweden)

    D. A. Robbins

    1994-12-01

    Full Text Available We study bundles of Banach algebras π:A→X, where each fiber Ax=π−1({x} is a Banach algebra and X is a compact Hausdorff space. In the case where all fibers are commutative, we investigate how the Gelfand representation of the section space algebra Γ(π relates to the Gelfand representation of the fibers. In the general case, we investigate how adjoining an identity to the bundle π:A→X relates to the standard adjunction of identities to the fibers.

  2. Helices and vector bundles

    CERN Document Server

    Rudakov, A N

    1990-01-01

    This volume is devoted to the use of helices as a method for studying exceptional vector bundles, an important and natural concept in algebraic geometry. The work arises out of a series of seminars organised in Moscow by A. N. Rudakov. The first article sets up the general machinery, and later ones explore its use in various contexts. As to be expected, the approach is concrete; the theory is considered for quadrics, ruled surfaces, K3 surfaces and P3(C).

  3. Bundled monocapillary optics

    Science.gov (United States)

    Hirsch, Gregory

    2002-01-01

    A plurality of glass or metal wires are precisely etched to form the desired shape of the individual channels of the final polycapillary optic. This shape is created by carefully controlling the withdrawal speed of a group of wires from an etchant bath. The etched wires undergo a subsequent operation to create an extremely smooth surface. This surface is coated with a layer of material which is selected to maximize the reflectivity of the radiation being used. This reflective surface may be a single layer of material, or a multilayer coating for optimizing the reflectivity in a narrower wavelength interval. The collection of individual wires is assembled into a close-packed multi-wire bundle, and the wires are bonded together in a manner which preserves the close-pack configuration, irrespective of the local wire diameter. The initial wires are then removed by either a chemical etching procedure or mechanical force. In the case of chemical etching, the bundle is generally segmented by cutting a series of etching slots. Prior to removing the wire, the capillary array is typically bonded to a support substrate. The result of the process is a bundle of precisely oriented radiation-reflecting hollow channels. The capillary optic is used for efficiently collecting and redirecting the radiation from a source of radiation which could be the anode of an x-ray tube, a plasma source, the fluorescent radiation from an electron microprobe, a synchrotron radiation source, a reactor or spallation source of neutrons, or some other source.

  4. Bundling harvester; Nippukorjausharvesteri

    Energy Technology Data Exchange (ETDEWEB)

    Koponen, K. [Eko-Log Oy, Kuopio (Finland)

    1996-12-31

    The staring point of the project was to design and construct, by taking the silvicultural point of view into account, a harvesting and processing system especially for energy-wood, containing manually driven bundling harvester, automatizing of the harvester, and automatized loading. The equipment forms an ideal method for entrepreneur`s-line harvesting. The target is to apply the system also for owner`s-line harvesting. The profitability of the system promotes the utilization of the system in both cases. The objectives of the project were: to construct a test equipment and prototypes for all the project stages, to carry out terrain and strain tests in order to examine the usability and durability, as well as the capacity of the machine, to test the applicability of the Eko-Log system in simultaneous harvesting of energy and pulp woods, and to start the marketing and manufacturing of the products. The basic problems of the construction of the bundling harvester have been solved using terrain-tests. The prototype machine has been shown to be operable. Loading of the bundles to form sufficiently economically transportable loads has been studied, and simultaneously, the branch-biomass has been tried to be utilized without loosing the profitability of transportation. The results have been promising, and will promote the profitable utilization of wood-energy

  5. Divertor detachment and exhaust on the TdeV tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Decoste, R.; Stansfield, B.L.; Gauvreau, J.L. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada)] [and others

    1996-12-01

    Experimental data, analysis and simulations are used to describe the physics of divertor detachment and He exhaust under detached conditions in TdeV. With increasingly density, the plasma is found to detach progressively from the outboard divertor plates with a marked reduction of the ion flux to the plates, the generation of a pressure gradient between an ionization front and the target plate, and strong cross-field transport in the divertor. Local interactions between the divertor plasma and the plates are described, with evidence for carbon sputtering and molecular processes near the divertor plates. Divertor exhaust and retention continue to increase through detachment and He exhaust is not affected although the divertor He enrichment remains low but constant at about 0.2. A moderate density of n-bar{sub e} {approx} 5 x 10{sup 19} m{sup -3} seems to be sufficient both for efficient peak power load reduction at the divertor plate and good He exhaust through the divertor. Simulation of the edge and divertor plasmas using the B2/EIRENE and DIVIMP codes give reasonable agreement with the measurements and indicate possible divertor geometry improvements. (author).

  6. Divertor detachment and exhaust on the TdeV tokamak

    International Nuclear Information System (INIS)

    Experimental data, analysis and simulations are used to describe the physics of divertor detachment and He exhaust under detached conditions in TdeV. With increasingly density, the plasma is found to detach progressively from the outboard divertor plates with a marked reduction of the ion flux to the plates, the generation of a pressure gradient between an ionization front and the target plate, and strong cross-field transport in the divertor. Local interactions between the divertor plasma and the plates are described, with evidence for carbon sputtering and molecular processes near the divertor plates. Divertor exhaust and retention continue to increase through detachment and He exhaust is not affected although the divertor He enrichment remains low but constant at about 0.2. A moderate density of n-bare ∼ 5 x 1019 m-3 seems to be sufficient both for efficient peak power load reduction at the divertor plate and good He exhaust through the divertor. Simulation of the edge and divertor plasmas using the B2/EIRENE and DIVIMP codes give reasonable agreement with the measurements and indicate possible divertor geometry improvements. (author)

  7. Detailed Radiative Transport Modeling of a Radiative Divertor

    CERN Document Server

    Wan, A S; Scott, H A; Post, D; Rognlien, T D

    1995-01-01

    An effective radiative divertor maximizes the utilization of atomic processes to spread out the energy deposition to the divertor chamber walls and to reduce the peak heat flux. Because the mixture of neutral atoms and ions in the divertor can be optically thick to a portion of radiated power, it is necessary to accurately model the magnitude and distribution of line radiation in this complex region. To assess their importance we calculate the effects of radiation transport using CRETIN, a multi-dimensional, non-local thermodynamic equilibrium simulation code that includes the atomic kinetics and radiative transport processes necessary to model the complex environment of a radiative divertor. We also include neutral transport to model radiation from recycling neutral atoms. This paper presents a case study of a high-recycling radiative divertor with a typical large neutral pressure at the divertor plate to estimate the impact of H line radiation on the overall power balance in the divertor region with conside...

  8. First study of EAST divertor by impurities puffing

    International Nuclear Information System (INIS)

    A series of experiments has recently been carried out in EAST under different plasma conditions to investigate the basic divertor performance, divertor and SOL screening efficiency and radiative divertor effect. Detached divertor plasmas have been achieved by density ramp-up. It is found that CIII emission from the low-field side (LFS) exhibits a strong dependence on poloidal locations and plasma operation regimes from methane (CH4) puffing experiments. In addition, the radiative divertor experiments by injection of mixed Ar (5.7% Ar in D2) into the outer divertor chamber reduce the peak heat flux by 50% at the outer target plate, which also reduce the divertor plasma temperature. The in-out heat flux distribution asymmetry is improved.

  9. The tungsten divertor experiment at ASDEX Upgrade

    Science.gov (United States)

    Neu, R.; Asmussen, K.; Krieger, K.; Thoma, A.; Bosch, H.-S.; Deschka, S.; Dux, R.; Engelhardt, W.; García-Rosales, C.; Gruber, O.; Herrmann, A.; Kallenbach, A.; Kaufmann, M.; Mertens, V.; Ryter, F.; Rohde, V.; Roth, J.; Sokoll, M.; Stäbler, A.; Suttrop, W.; Weinlich, M.; Zohm, H.; Alexander, M.; Becker, G.; Behler, K.; Behringer, K.; Behrisch, R.; Bergmann, A.; Bessenrodt-Weberpals, M.; Brambilla, M.; Brinkschulte, H.; Büchl, K.; Carlson, A.; Chodura, R.; Coster, D.; Cupido, L.; de Blank, H. J.; de Peña Hempel, S.; Drube, R.; Fahrbach, H.-U.; Feist, J.-H.; Feneberg, W.; Fiedler, S.; Franzen, P.; Fuchs, J. C.; Fußmann, G.; Gafert, J.; Gehre, O.; Gernhardt, J.; Haas, G.; Herppich, G.; Herrmann, W.; Hirsch, S.; Hoek, M.; Hoenen, F.; Hofmeister, F.; Hohenöcker, H.; Jacobi, D.; Junker, W.; Kardaun, O.; Kass, T.; Kollotzek, H.; Köppendörfer, W.; Kurzan, B.; Lackner, K.; Lang, P. T.; Lang, R. S.; Laux, M.; Lengyel, L. L.; Leuterer, F.; Manso, M. E.; Maraschek, M.; Mast, K.-F.; McCarthy, P.; Meisel, D.; Merkel, R.; Müller, H. W.; Münich, M.; Murmann, H.; Napiontek, B.; Neu, G.; Neuhauser, J.; Niethammer, M.; Noterdaeme, J.-M.; Pasch, E.; Pautasso, G.; Peeters, A. G.; Pereverzev, G.; Pitcher, C. S.; Poschenrieder, W.; Raupp, G.; Reinmüller, K.; Riedl, R.; Röhr, H.; Salzmann, H.; Sandmann, W.; Schilling, H.-B.; Schlögl, D.; Schneider, H.; Schneider, R.; Schneider, W.; Schramm, G.; Schweinzer, J.; Scott, B. D.; Seidel, U.; Serra, F.; Speth, E.; Silva, A.; Steuer, K.-H.; Stober, J.; Streibl, B.; Treutterer, W.; Troppmann, M.; Tsois, N.; Ulrich, M.; Varela, P.; Verbeek, H.; Verplancke, Ph; Vollmer, O.; Wedler, H.; Wenzel, U.; Wesner, F.; Wolf, R.; Wunderlich, R.; Zasche, D.; Zehetbauer, T.; Zehrfeld, H.-P.

    1996-12-01

    Tungsten-coated tiles, manufactured by plasma spray on graphite, were mounted in the divertor of the ASDEX Upgrade tokamak and cover almost 90% of the surface facing the plasma in the strike zone. Over 600 plasma discharges have been performed to date, around 300 of which were auxiliary heated with heating powers up to 10 MW. The production of tungsten in the divertor was monitored by a W I line at 400.8 nm. In the plasma centre an array of spectral lines at 5 nm emitted by ionization states around W XXX was measured. From the intensity of these lines the W content was derived. Under normal discharge conditions W-concentrations around 0741-3335/38/12A/013/img12 or even lower were found. The influence on the main plasma parameters was found to be negligible. The maximum concentrations observed decrease with increasing heating power. In several low power discharges accumulation of tungsten occurred and the temperature profile was flattened. The concentrations of the intrinsic impurities carbon and oxygen were comparable to the discharges with the graphite divertor. Furthermore, the density and the 0741-3335/38/12A/013/img13 limits remained unchanged and no negative influence on the energy confinement or on the H-mode threshold was found. Discharges with neon radiative cooling showed the same behaviour as in the graphite divertor case.

  10. ELM induced divertor heat loads on TCV

    Czech Academy of Sciences Publication Activity Database

    Marki, J.; Pitts, R. A.; Horáček, Jan; Turri, G.; Tskhakaya, D.; TCV, team.

    Basel: Swiss Physical Society, 2008. s. 75-75. ISBN N. [Annual meeting of the Swiss physical society 2008. 26.03.208-27.03.2008, Geneva] Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak TCV * divertor heat load * ELM Subject RIV: BL - Plasma and Gas Discharge Physics http://www.sps.ch/uploads/media/SPS2008_Plasma.pdf

  11. Divertor Coil Design and Implementation on Pegasus

    Science.gov (United States)

    Shriwise, P. C.; Bongard, M. W.; Cole, J. A.; Fonck, R. J.; Kujak-Ford, B. A.; Lewicki, B. T.; Winz, G. R.

    2012-10-01

    An upgraded divertor coil system is being commissioned on the Pegasus Toroidal Experiment in conjunction with power system upgrades in order to achieve higher β plasmas, reduce impurities, and possibly achieve H-mode operation. Design points for the divertor coil locations and estimates of their necessary current ratings were found using predictive equilibrium modeling based upon a 300 kA target plasma. This modeling represented existing Pegasus coil locations and current drive limits. The resultant design calls for 125 kA-turns from the divertor system to support the creation of a double null magnetic topology in plasmas with IpIGBT power supply modules to provide IDIV<=4 kA. The resulting 20 kA-turn capability of the existing divertor coil will be augmented by a new coil providing additional A-turns in series. Induced vessel wall current modeling indicates the time response of a 28 turn augmentation coil remains fast compared to the poloidal field penetration rate through the vessel. First results operating the augmented system are shown.

  12. Divertor erosion in DIII-D

    International Nuclear Information System (INIS)

    Net erosion rates of carbon target plates have been measured in situ for the DIII-D lower divertor. The principal method of obtaining this data is the DiMES sample probe. Recent experiments have focused on erosion at the outer strike-point of two divertor plasma conditions: (1) attached (Te > 40 eV) ELMing plasmas and (2) detached (Te 10 cm/year, even with incident heat flux 2. In this case, measurements and modeling agree for both gross and net carbon erosion, showing the near-surface transport and redeposition of the carbon is well understood and that effective sputtering yields are > 10%. In ELM-free discharges, this erosion rate can account for the rate of carbon accumulation in the core plasma. Divertor plasma detachment eliminates physical sputtering, while spectroscopically measured chemical erosion yields are also found to be low (Y(C/D+) ≤ 2.0 x 10-3). This leads to suppression of net erosion at the outer strike-point, which becomes a region of net redeposition (∼ 4 cm/year). The private flux wall is measured to be a region of net redeposition with dense, high neutral pressure, attached divertor plasmas. Leading edges intercepting parallel heat flux (∼ 50 MW/m2) have very high net erosion rates (∼ 10 microm/s) at the OSP of an attached plasma. Leading edge erosion, and subsequent carbon redeposition, caused by tile gaps can account for half of the deuterium codeposition in the DIII-D divertor

  13. Divertor Materials Evaluation System (DiMES)

    International Nuclear Information System (INIS)

    The mission of the Divertor Materials Evaluation System (DiMES) in DIII-D is to establish an integrated data base from measurements in the divertor of a tokamak in order to address some of the ITER and fusion power reactor plasma material interaction issues. Carbon and metal coatings of Be, W, V, and Mo were exposed to the steady-state outer strike point on DIII-D for 4-18 s. These short exposure times ensure controlled exposure conditions, and the extensive arrays of DIII-D divertor diagnostics provide a well-characterized plasma for modeling efforts. Postexposure analysis provides a direct measure of surface material erosion rates and the amount of retained deuterium. For carbon, these results match closely with the results of accumulated carbon deposition and erosion, and the corresponding deuterium retention of long term exposure tiles in DIII-D. Deuterium retention of different materials was measured using the 3He(d,p) 4He nuclear reaction. For carbon, these measurements showed peak deuterium areal density of about 8 x 10 18 D/cm2 in a co-deposited layer about 6 microm deep, mainly at the usually detached inboard divertor leg. That layer of carbon near the inner divertor strike point has an atomic saturation concentration of D/C ∼ 0.25, which is not significantly lower than the laboratory-measured saturation retention of 0.4. Under the carbon contaminated background plasma of DIII-D, metal coatings of Be, V, Mo, and W were exposed to the steady state outer strike point under ELMing and ELM-free H-mode discharges. The rate of material erosion and tritium retention were measured. As expected, W shows the lowest erosion rate at 0.1 nm/s and the lowest deuterium uptake

  14. Study of the radiation in divertor plasmas

    International Nuclear Information System (INIS)

    We have studied the cooling of the edge plasma by radiation in the divertor volume, in order to optimize the extraction of power in tokamaks and to limit the wall erosion. In attached divertor plasmas experiments, the concentration of intrinsic impurities at the edge is related to the response of the wall to the incident energy flow of plasma, depending on a phenomenological law. We carried out an analysis of the radiation according to this law and to the control parameters of the discharges. The largest radiated fraction and best synergy are obtained when the concentration of intrinsic impurities strongly increases with the energy of incident plasma. On the other hand, the erosion of the wall is stronger. In detached plasmas, we proved that the performances in terms of incident plasma energy loss and pressure loss are optimal when the density of the slowest neutrals is strong at the edge and when their radial penetration is small. On Tore Supra, we highlighted the correlations between the maximum Mach number of incident plasma flow, the radiation front and the penetration of the neutrals. A simple diagnostic based on the localization of the maximum Mach number proves that detached mode is not optimal on Tore Supra, because the radial penetration of the slowest neutrals is not sufficiently small. In the last part, we obtained the three-dimensional topology of the radiation in the ergodic divertor using a spectral analysis code and boundary conditions consistent with the temperature distribution on the wall. The radiation is maximum in front of the divertor modules. As a consequence, radiated power is underestimated by standards measurements of Tore Supra that are located between the modules. We finally showed that the profiles of temperature along the field lines are modulated, this is specific to the ergodic divertor. (author)

  15. Kernel Bundle EPDiff

    DEFF Research Database (Denmark)

    Sommer, Stefan Horst; Lauze, Francois Bernard; Nielsen, Mads; Pennec, Xavier

    information to be automatically incorporated in registrations and promises to improve the standard framework in several aspects. We present the mathematical foundations of LDDKBM and derive the KB-EPDiff evolution equations, which provide optimal warps in this new framework. To illustrate the resulting......In the LDDMM framework, optimal warps for image registration are found as end-points of critical paths for an energy functional, and the EPDiff equations describe the evolution along such paths. The Large Deformation Diffeomorphic Kernel Bundle Mapping (LDDKBM) extension of LDDMM allows scale space...... diffeomorphism paths, we give examples showing the decoupled evolution across scales and how the method automatically incorporates deforma- tion at appropriate scales....

  16. Twists of symmetric bundles

    OpenAIRE

    Cassou-Nogues, Ph.; Erez, B.; Taylor, M. J.

    2004-01-01

    We establish comparison results between the Hasse-Witt invariants w_t(E) of a symmetric bundle E over a scheme and the invariants of one of its twists E_{\\alpha}. For general twists we describe the difference between w_t(E) and w_t(E_{\\alpha}) up to terms of degree 3. Next we consider a special kind of twist, which has been studied by A. Fr\\"ohlich. This arises from twisting by a cocycle obtained from an orthogonal representation. We show how to explicitly describe the twist for representatio...

  17. REBEKA bundle experiments

    International Nuclear Information System (INIS)

    This report is a summary of experimental investigations describing the fuel rod behavior in the refilling and reflooding phase of a loss-of-coolant accident of a PWR. The experiments were performed with 5x5 and 7x7 rod bundles, using indirectly electrically heated fuel rod simulators of full length with original PWR-KWU-geometry, original grid spacers and Zircaloy-4-claddings (Type Biblis B). The fuel rod simulators showed a cosine shaped axial power profile in 7 steps and continuous, respectively. The results describe the influence of the different parameters such as bundle size on the maximum coolant channel blockage, that of the cooling on the size of the circumferential strain of the cladding (azimuthal temperature distribution) a cold control rod guide thimble and the flow direction (axial temperature distribution) on the resulting coolant channel blockage. The rewetting behavior of different fuel rod simulators including ballooned and burst Zircaloy claddings is discussed as well as the influence of thermocouples on the cladding temperature history and the rewetting behavior. All results prove the coolability of a PWR in the case of a LOCA. Therefore, it can be concluded that the ECC-criteria established by licensing authorities can be fulfilled. (orig./HP)

  18. Module of lithium divertor for KTM tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lyublinski, I., E-mail: yublinski@yandex.ru [FSUE ' Red Star' , Moscow (Russian Federation); Vertkov, A.; Evtikhin, V.; Balakirev, V.; Ionov, D.; Zharkov, M. [FSUE ' Red Star' , Moscow (Russian Federation); Tazhibayeva, I. [IAE NNC RK, Kurchatov (Kazakhstan); Mirnov, S. [TRINITI, Troitsk, Moscow Region (Russian Federation); Khomiakov, S.; Mitin, D. [OJSC Dollezhal Institute, Moscow (Russian Federation); Mazzitelli, G. [ENEA RC Frascati (Italy); Agostini, P. [ENEA RC Brasimone (Italy)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium-metal with low Z. Black-Right-Pointing-Pointer Capillary-porous system (CPS) - new material in which liquid lithium fill a solid matrix from porous material. Black-Right-Pointing-Pointer Lithium divertor module for KTM tokamak is under development. Black-Right-Pointing-Pointer Lithium filled tungsten felt is offered as the base plasma facing material of divertor. Black-Right-Pointing-Pointer Results of this project addresses to the progress in the field of fusion neutrons source and fusion energy source creation. - Abstract: Activity on projects of ITER and DEMO reactors has shown that solution of problems of divertor target plates and other plasma facing elements (PFEs) based on the solid plasma facing materials cause serious difficulties. Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium-metal with low Z. Application of lithium will allow to create a self-renewal and MHD stable liquid metal surface of the in-vessel devices possessing practically unlimited service life; to reduce power flux due to intensive re-irradiation on lithium atoms in plasma periphery that will essentially facilitate a problem of heat removal from PFE; to reduce Z{sub eff} of plasma to minimally possible level close to 1; to exclude tritium accumulation, that is provided with absence of dust products and an opportunity of the active control of the tritium contents in liquid lithium. Realization of these advantages is based on use of so-called lithium capillary-porous system (CPS) - new material in which liquid lithium fill a solid matrix from porous material. The progress in development of lithium technology and also activity in lithium experiments in the tokamaks TFTR, T-11M, T-10, FTU, NSTX, HT-7 and stellarator TJ II permits of solving the problems in development of

  19. NSTX Plasma Response to Lithium Coated Divertor

    Energy Technology Data Exchange (ETDEWEB)

    H.W. Kugel, M.G. Bell, J.P. Allain, R.E. Bell, S. Ding, S.P. Gerhardt, M.A. Jaworski, R. Kaita, J. Kallman, S.M. Kaye, B.P. LeBlanc, R. Maingi, R. Majeski, R. Maqueda, D.K. Mansfield, D. Mueller, R. Nygren, S.F. Paul, R. Raman, A.L. Roquemore, S.A. Sabbagh, H. Schneider, C.H. Skinner, V.A. Soukhanovskii, C.N. Taylor, J.R. Timberlak, W.R. Wampler, L.E. Zakharov, S.J. Zweben, and the NSTX Research Team

    2011-01-21

    NSTX experiments have explored lithium evaporated on a graphite divertor and other plasma facing components in both L- and H- mode confinement regimes heated by high-power neutral beams. Improvements in plasma performance have followed these lithium depositions, including a reduction and eventual elimination of the HeGDC time between discharges, reduced edge neutral density, reduced plasma density, particularly in the edge and the SOL, increased pedestal electron and ion temperature, improved energy confinement and the suppression of ELMs in the H-mode. However, with improvements in confinement and suppression of ELMs, there was a significant secular increase in the effective ion charge Zeff and the radiated power in H-mode plasmas as a result of increases in the carbon and medium-Z metallic impurities. Lithium itself remained at a very low level in the plasma core, <0.1%. Initial results are reported from operation with a Liquid Lithium Divertor (LLD) recently installed.

  20. ITER tungsten divertor design development and qualification program

    Energy Technology Data Exchange (ETDEWEB)

    Hirai, T., E-mail: takeshi.hirai@iter.org [ITER Organization, Route de Vinon sur Verdon, F-13115 Saint Paul lez Durance (France); Escourbiac, F.; Carpentier-Chouchana, S.; Fedosov, A.; Ferrand, L.; Jokinen, T.; Komarov, V.; Kukushkin, A.; Merola, M.; Mitteau, R.; Pitts, R.A.; Shu, W.; Sugihara, M. [ITER Organization, Route de Vinon sur Verdon, F-13115 Saint Paul lez Durance (France); Riccardi, B. [F4E, c/ Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Suzuki, S. [JAEA, Fusion Research and Development Directorate JAEA, 801-1 Mukouyama, Naka, Ibaragi 311-0193 (Japan); Villari, R. [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, I-00044 Frascati, Rome (Italy)

    2013-10-15

    Highlights: • Detailed design development plan for the ITER tungsten divertor. • Latest status of the ITER tungsten divertor design. • Brief overview of qualification program for the ITER tungsten divertor and status of R and D activity. -- Abstract: In November 2011, the ITER Council has endorsed the recommendation that a period of up to 2 years be set to develop a full-tungsten divertor design and accelerate technology qualification in view of a possible decision to start operation with a divertor having a full-tungsten plasma-facing surface. To ensure a solid foundation for such a decision, a full tungsten divertor design, together with a demonstration of the necessary high performance tungsten monoblock technology should be completed within the required timescale. The status of both the design and technology R and D activity is summarized in this paper.

  1. ITER tungsten divertor design development and qualification program

    International Nuclear Information System (INIS)

    Highlights: • Detailed design development plan for the ITER tungsten divertor. • Latest status of the ITER tungsten divertor design. • Brief overview of qualification program for the ITER tungsten divertor and status of R and D activity. -- Abstract: In November 2011, the ITER Council has endorsed the recommendation that a period of up to 2 years be set to develop a full-tungsten divertor design and accelerate technology qualification in view of a possible decision to start operation with a divertor having a full-tungsten plasma-facing surface. To ensure a solid foundation for such a decision, a full tungsten divertor design, together with a demonstration of the necessary high performance tungsten monoblock technology should be completed within the required timescale. The status of both the design and technology R and D activity is summarized in this paper

  2. Modeling detachment physics in the NSTX snowflake divertor

    Energy Technology Data Exchange (ETDEWEB)

    Meier, E.T., E-mail: emeier@wm.edu [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Bell, R.E.; Diallo, A.; Kaita, R.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, Princeton, NJ 08540 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Podestà, M. [Princeton Plasma Physics Laboratory, Princeton, NJ 08540 (United States); Rognlien, T.D.; Scotti, F. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States)

    2015-08-15

    The snowflake divertor is a proposed technique for coping with the tokamak power exhaust problem in next-step experiments and eventually reactors, where extreme power fluxes to material surfaces represent a leading technological and physics challenge. In lithium-conditioned National Spherical Torus Experiment (NSTX) discharges, application of the snowflake divertor typically induced partial outer divertor detachment and severalfold heat flux reduction. UEDGE is used to analyze and compare conventional and snowflake divertor configurations in NSTX. Matching experimental upstream profiles and divertor measurements in the snowflake requires target recycling of 0.97 vs. 0.91 in the conventional case, implying partial saturation of the lithium-based pumping mechanism. Density scans are performed to analyze the mechanisms that facilitate detachment in the snowflake, revealing that increased divertor volume provides most of the parallel heat flux reduction. Also, neutral gas power loss is magnified by the increased wetted area in the snowflake, and plays a key role in generating volumetric recombination.

  3. Development of a radiative divertor for DIII-D

    International Nuclear Information System (INIS)

    We have used experiments and modeling to develop a new radiative divertor configuration for DIII-D. Gas puffing experiments with the existing open divertor have shown the creation of a localized (∼10 cm diameter) radiation zone which results in substantial reduction (3--10) in the divertor heat flux while δE remains ∼2 times ITER-89P scaling. However, ne increases with D2 puffing, and Zeff increases with neon puffing. Divertor structures are required to minimize the effects on the core plasma. The UEDGE fluid code, benchmarked with DIII-D data, and the DEGAS neutrals transport code are used to estimate the effectiveness of divertor configurations; slots reduce the core ionization more than baffles. The overall divertor shape is set by confinement studies which indicate that high triangularity (δ ∼0.8) is important for high τE VH-modes. Results from engineering feasibility studies, including diagnostic access, will be presented

  4. Divertor and scoop limiter experiments on PDX

    Energy Technology Data Exchange (ETDEWEB)

    McGuire, K.; Beiersdorfer, P.; Bell, M.; Bol, K.; Boyd, D.; Buchenauer, D.; Budny, R.; Cavallo, A.; Couture, P.; Crowley, T.

    1985-01-01

    Routine operation in the enhanced energy confinement (or H-mode) regime during neutral beam injection was achieved by modifying the PDX divertor hardware to inhibit the influx of neutral gas from the divertor region to the main plasma chamber. A particle scoop limiter has been studied as a mechanical means of controlling particles at the plasma edge, and neutral beam heated discharges with this limiter show similar confinement times (normalized to tau/sub E//I/sub p/) to average H-mode plasmas. Two new instabilities are observed near the plasma edge in PDX during H-mode operation. The first, a quasicoherent fluctuation, occurred in bursts at well-defined frequencies (..delta omega../..omega.. less than or equal to 0.1) in the range 50 to 180 kHz, and had no obvious effects on confinement. The second instability, the edge relaxation phenomena (ERP), did cause deterioration in the global confinement time. The ERP's are characterized by sharp spikes in the divertor plasma density, H/sub ..cap alpha../ emission, and on the x-ray signals they appear as sawtoothlike relaxations at the plasma edge with an inversion radius near the separatrix. Attempts to obtain high ..beta../sub T/ in the H-mode discharges were hampered by a deterioration in the H-mode confinement and major disruptions which limited the achievable ..beta../sub T/. A study of the stability of both the limiter L-mode and divertor H-mode discharges close to the theoretical ..beta.. boundary, showed that the major disruptions observed there are sometimes caused by a fast growing m/n = 1/1 mode with no observable external precursor oscillations.

  5. Divertor and scoop limiter experiments on PDX

    International Nuclear Information System (INIS)

    Routine operation in the enhanced energy confinement (or H-mode) regime during neutral beam injection was achieved by modifying the PDX divertor hardware to inhibit the influx of neutral gas from the divertor region to the main plasma chamber. A particle scoop limiter has been studied as a mechanical means of controlling particles at the plasma edge, and neutral beam heated discharges with this limiter show similar confinement times (normalized to tau/sub E//I/sub p/) to average H-mode plasmas. Two new instabilities are observed near the plasma edge in PDX during H-mode operation. The first, a quasicoherent fluctuation, occurred in bursts at well-defined frequencies (Δω/ω less than or equal to 0.1) in the range 50 to 180 kHz, and had no obvious effects on confinement. The second instability, the edge relaxation phenomena (ERP), did cause deterioration in the global confinement time. The ERP's are characterized by sharp spikes in the divertor plasma density, H/sub α/ emission, and on the x-ray signals they appear as sawtoothlike relaxations at the plasma edge with an inversion radius near the separatrix. Attempts to obtain high β/sub T/ in the H-mode discharges were hampered by a deterioration in the H-mode confinement and major disruptions which limited the achievable β/sub T/. A study of the stability of both the limiter L-mode and divertor H-mode discharges close to the theoretical β boundary, showed that the major disruptions observed there are sometimes caused by a fast growing m/n = 1/1 mode with no observable external precursor oscillations

  6. ELM induced divertor heat loads on TCV

    Czech Academy of Sciences Publication Activity Database

    Marki, J.; Pitts, R. A.; Horáček, Jan; Turri, G.; Tskhakaya, D.; TCV Team, T.

    Madrid: Centro de Investigaciones Energética Medioambiental y Tecnológica (CIEMAT), 2008. P1-41-P1-41. ISBN N. [International Conference on Plasma Surface Interactions ,PSI 18/18th./. 26.05.2008-30.05.2008, Toledo] Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak TCV * divertor heat load * ELM Subject RIV: BL - Plasma and Gas Discharge Physics http://psi2008.ciemat.es/documents/Book_of_abstracts3.pdf

  7. ELM induced divertor heat loads on TCV

    Czech Academy of Sciences Publication Activity Database

    Marki, J.; Pitts, R. A.; Horáček, Jan; Tskhakaya, D.; TCV, team.

    309-391, - (2009), s. 801-805. ISSN 0022-3115. [International Conference on Plasma-Surface Interactions in Controlled Fusion Devices/18th./. Toledo, 26.05.2008-30.5.2008] Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak TCV * divertor heat load * ELM * EVOLUTION * JET Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.933, year: 2009 http://dx.doi.org/10.1016/j.jnucmat.2009.01.212

  8. Influence of helium puff on divertor asymmetry in experimental advanced superconducting tokamak

    DEFF Research Database (Denmark)

    Liu, S. C.; Guo, H. Y.; Xu, G. S.; Wang, L.; Wang, H. Q.; Ding, R.; Duan, Y. M.; Gan, K. F.; Shao, L. M.; Chen, L.; Yan, Ning; Zhang, W.; Chen, R.; Xiong, H.; Ding, S.; Hu, G. H.; Liu, Y. L.; Zhao, N.; Li, Y. L.; Gao, X.

    2014-01-01

    Divertor asymmetries with helium puffing are investigated in various divertor configurations on Experimental Advanced Superconducting Tokamak (EAST). The outer divertor electron temperature decreases significantly during the gas injection at the outer midplane. As soon as the gas is injected into...

  9. Divertor Development for a Future Fusion Power Plant

    OpenAIRE

    Norajitra, Prachai

    2011-01-01

    The thesis begins by describing the fusion process and operation of a fusion reactor, the approach in the conceptual development of a helium-cooled divertor, and leads to the KIT helium-cooled modular divertor design. Then the methods of verification and validation of the design by tests are described, results presented and discussed. The developed divertor concept has demonstrated its principal functionality and hence the used design process and tools can be conceived as verified and validated.

  10. Plasma diagnostics for the DIII-D divertor upgrade

    International Nuclear Information System (INIS)

    The DIII-D tokamak is being upgraded to allow for divertor biasing, baffling, and pumping experiments. This paper gives an overview of the new diagnostics added to DIII-D as part of this Advanced Divertor Program. They include tile current monitors, fast reciprocating Langmuir probes, a fixed probe array in the divertor, fast neutral pressure gauges, and Hα measurements with TV cameras and fiber optics coupled to a high resolution spectrometer. 9 refs

  11. ADX - Advanced Divertor and RF Tokamak Experiment

    Science.gov (United States)

    Greenwald, Martin; Labombard, Brian; Bonoli, Paul; Irby, Jim; Terry, Jim; Wallace, Greg; Vieira, Rui; Whyte, Dennis; Wolfe, Steve; Wukitch, Steve; Marmar, Earl

    2015-11-01

    The Advanced Divertor and RF Tokamak Experiment (ADX) is a design concept for a compact high-field tokamak that would address boundary plasma and plasma-material interaction physics challenges whose solution is critical for the viability of magnetic fusion energy. This device would have two crucial missions. First, it would serve as a Divertor Test Tokamak, developing divertor geometries, materials and operational scenarios that could meet the stringent requirements imposed in a fusion power plant. By operating at high field, ADX would address this problem at a level of power loading and other plasma conditions that are essentially identical to those expected in a future reactor. Secondly, ADX would investigate the physics and engineering of high-field-side launch of RF waves for current drive and heating. Efficient current drive is an essential element for achieving steady-state in a practical, power producing fusion device and high-field launch offers the prospect of higher efficiency, better control of the current profile and survivability of the launching structures. ADX would carry out this research in integrated scenarios that simultaneously demonstrate the required boundary regimes consistent with efficient current drive and core performance.

  12. Divertor materials evaluation system (DiMES)

    International Nuclear Information System (INIS)

    The mission of the Divertor Materials Evaluation System (DiMES) in DIII-D is to establish an integrated data base from measurements in the divertor of a tokamak in order to address some of the ITER and fusion power reactor plasma material interaction issues. Carbon and metal coatings of Be, W, V, and Mo were exposed to the steady-state outer strike point on DIII-D for 4--18 s. These short exposure times ensure controlled exposure conditions, and the extensive arrays of DIII-D divertor diagnostics provide a well-characterized plasma for modeling efforts. Post-exposure analysis provides a direct measure of surface material erosion rates and the amount of retained deuterium. For carbon, these results match closely with the results of accumulated carbon deposition and erosion, and the corresponding deuterium retention of long term exposure tiles in DIII-D. Under the carbon-contaminated background plasma of DIII-D, metal coatings of Be, V, Mo, and W were exposed to the steady-state outer strike point under ELMing and ELM-free H-mode discharges. The rate of material erosion and deuterium retention were measured. As expected, W shows the lowest erosion rate at 0.1 mm/s and the lowest deuterium uptake of 2 x 1020/m2

  13. Divertor geometry optimization for ASDEX Upgrade

    International Nuclear Information System (INIS)

    One of the critical questions to be solved for ITER (or any other reactor) is the power exhaust problem (compatible with particle exhaust). Optimized divertors have to be tested in existing geometries based mainly on the idea of closing them very efficiently to the main chamber and, by the choice of the plate and baffle geometry, positively influencing the flow pattern of hydrogen favoring good impurity entrainment. Also, for ASDEX Upgrade there is an experimental necessity for an improved divertor due to the increased heating power (24 MW will be available from 1997 on compared to the present 18 MW). We present the optimization strategy for the divertor II of ASDEX Upgrade, using elaborate numerical models and codes (B2-Eirene) as well as simple models. We start with the choice of a proper target plate geometry, and then further discuss how main chamber and private flux baffling will be done, and how this affects neutral recirculation pattern and pumping properties. For the final configuration the impurity entrainment properties are analyzed. (orig.)

  14. Bundle Security Protocol for ION

    Science.gov (United States)

    Burleigh, Scott C.; Birrane, Edward J.; Krupiarz, Christopher

    2011-01-01

    This software implements bundle authentication, conforming to the Delay-Tolerant Networking (DTN) Internet Draft on Bundle Security Protocol (BSP), for the Interplanetary Overlay Network (ION) implementation of DTN. This is the only implementation of BSP that is integrated with ION.

  15. CANFLEX fuel bundle impact test

    International Nuclear Information System (INIS)

    This document outlines the test results for the impact test of the CANFLEX fuel bundle. Impact test is performed to determine and verify the amount of general bundle shape distortion and defect of the pressure tube that may occur during refuelling. The test specification requires that the fuel bundles and the pressure tube retain their integrities after the impact test under the conservative conditions (10 stationary bundles with 31kg/s flow rate) considering the pressure tube creep. The refuelling simulator operating with pneumatic force and simulated shield plug were fabricated and the velocity/displacement transducer and the high speed camera were also used in this test. The characteristics of the moving bundle (velocity, displacement, impacting force) were measured and analyzed with the impact sensor and the high speed camera system. The important test procedures and measurement results were discussed as follows. 1) Test bundle measurements and the pressure tube inspections 2) Simulated shield plug, outlet flange installation and bundle loading 3) refuelling simulator, inlet flange installation and sensors, high speed camera installation 4) Perform the impact test with operating the refuelling simulator and measure the dynamic characteristics 5) Inspections of the fuel bundles and the pressure tube. (author). 8 refs., 23 tabs., 13 figs

  16. CANFLEX fuel bundle impact test

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Seok Kyu; Chung, C. H.; Park, J. S.; Hong, S. D.; Kim, B. D.

    1997-08-01

    This document outlines the test results for the impact test of the CANFLEX fuel bundle. Impact test is performed to determine and verify the amount of general bundle shape distortion and defect of the pressure tube that may occur during refuelling. The test specification requires that the fuel bundles and the pressure tube retain their integrities after the impact test under the conservative conditions (10 stationary bundles with 31kg/s flow rate) considering the pressure tube creep. The refuelling simulator operating with pneumatic force and simulated shield plug were fabricated and the velocity/displacement transducer and the high speed camera were also used in this test. The characteristics of the moving bundle (velocity, displacement, impacting force) were measured and analyzed with the impact sensor and the high speed camera system. The important test procedures and measurement results were discussed as follows. 1) Test bundle measurements and the pressure tube inspections 2) Simulated shield plug, outlet flange installation and bundle loading 3) refuelling simulator, inlet flange installation and sensors, high speed camera installation 4) Perform the impact test with operating the refuelling simulator and measure the dynamic characteristics 5) Inspections of the fuel bundles and the pressure tube. (author). 8 refs., 23 tabs., 13 figs.

  17. Fiber Bundles and Parseval Frames

    OpenAIRE

    Agrawal, Devanshu; Knisley, Jeff

    2015-01-01

    Continuous frames over a Hilbert space have a rich and sophisticated structure that can be represented in the form of a fiber bundle. The fiber bundle structure reveals the central importance of Parseval frames and the extent to which Parseval frames generalize the notion of an orthonormal basis.

  18. UEDGE modeling of the effect of divertor modifications on divertor performance

    International Nuclear Information System (INIS)

    The DIII'D upper divertor will be modified in late 1999 by installing a continuous dome in the private flux region with an independent pumping capability for the inner strike point. A ''bump'' on the inner cylinder also has been considered to enhance impurity and neutral control. Using the UEDGE code, we have examined the effect of this dome and the ''bump'' on core ionization and core impurity content. For typical parameters, results indicate that the planned divertor modifications enable detachment at higher heating power and the ''fueling efficiency'' (ratio of core neutral ionization rate to total divertor ion current) decreases, however, the core carbon content increases. The inner ''bump'' does enhance ''fueling efficiency'' compared to the private flux dome alone, but it does not reduce the increased core impurity content

  19. Fiber bundle phase conjugate mirror

    Science.gov (United States)

    Ward, Benjamin G.

    2012-05-01

    An improved method and apparatus for passively conjugating the phases of a distorted wavefronts resulting from optical phase mismatch between elements of a fiber laser array are disclosed. A method for passively conjugating a distorted wavefront comprises the steps of: multiplexing a plurality of probe fibers and a bundle pump fiber in a fiber bundle array; passing the multiplexed output from the fiber bundle array through a collimating lens and into one portion of a non-linear medium; passing the output from a pump collection fiber through a focusing lens and into another portion of the non-linear medium so that the output from the pump collection fiber mixes with the multiplexed output from the fiber bundle; adjusting one or more degrees of freedom of one or more of the fiber bundle array, the collimating lens, the focusing lens, the non-linear medium, or the pump collection fiber to produce a standing wave in the non-linear medium.

  20. Towards a physics-integrated view on divertor pumping

    International Nuclear Information System (INIS)

    Highlights: • Physics-integrated design approaches are to be preferred over approaches based on simple requirement lists. • A physics-integrated assessment is presented for the divertor vacuum pumping system based on detachment onset conditions for the divertor. • This approach considers density dependent pump albedo to reflect the effects of gas recycling at the divertor and the changes in flow regime with density. • A comparison with DEMO indicates that the divertor pumping system for a pulsed DEMO scales less than linearly with fusion power. - Abstract: One key requirement to design the inner fuel cycle of a divertor tokamak is defined by the torus vessel gas throughput and composition, and the sub-divertor neutral pressure at which the exhaust gas has to be pumped. This paper illustrates how divertor physics aspects can be translated to requirements on the divertor vacuum pumping system. An example workflow is presented that links the realization of detachment conditions with the sub-divertor neutral gas flow patterns in order to determine the appropriate number of torus vacuum pumps. For the example case of a fusion DEMO size machine, it was found that 7 actively pumping cryopumps (ITER-type) are necessary to handle the gas throughput that is needed to manage the heat flux and densities related to detachment onset

  1. Magnetic geometry and particle source drive of supersonic divertor regimes

    International Nuclear Information System (INIS)

    We present a comprehensive picture of the mechanisms driving the transition from subsonic to supersonic flows in tokamak plasmas. We demonstrate that supersonic parallel flows into the divertor volume are ubiquitous at low density and governed by the divertor magnetic geometry. As the density is increased, subsonic divertor plasmas are recovered. On detachment, we show the change in particle source can also drive the transition to a supersonic regime. The comprehensive theoretical analysis is completed by simulations in ITER geometry. Such results are essential in assessing the divertor performance and when interpreting measurements and experimental evidence. (technical note)

  2. Divertor bypass in the Alcator C-Mod tokamak

    Science.gov (United States)

    Pitcher, C. S.; LaBombard, B.; Danforth, R.; Pina, W.; Silveira, M.; Parkin, B.

    2001-01-01

    The Alcator C-Mod divertor bypass has for the first time allowed in situ variations to the mechanical baffle design in a tokamak. The design utilizes small coils which interact with the ambient magnetic field inside the vessel to provide the torque required to control small flaps of a Venetian blind geometry. Plasma physics experiments with the bypass have revealed the importance of the divertor baffling to maintain high divertor gas pressures. These experiments have also indicated that the divertor baffling has only a limited effect on the main chamber pressure in C-Mod.

  3. Comparative study of divertor and limiter concepts in FER

    International Nuclear Information System (INIS)

    Comparative study of engineering features is carried out for divertor and pumped limiter reactor concepts. These concepts are: double null divertor as a reference concept of FER, single null divertor, single pumped limiter with medium edge temperature and single pumped limiter with low edge temperature. Plasma parameters of these concepts are determined by maintaining plasma confinement performance. It is found that the double null divertor is the least favorable; and the medium edge temperature limiter is the most favorable from most of the engineering standpoints. (author)

  4. Divertor Heat Flux Mitigation in the National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V A; Maingi, R; Gates, D A; Menard, J E; Paul, S F; Raman, R; Roquemore, A L; Bell, M G; Bell, R E; Boedo, J A; Bush, C E; Kaita, R; Kugel, H W; LeBlanc, B P; Mueller, D

    2008-08-04

    Steady-state handling of divertor heat flux is a critical issue for both ITER and spherical torus-based devices with compact high power density divertors. Significant reduction of heat flux to the divertor plate has been achieved simultaneously with favorable core and pedestal confinement and stability properties in a highly-shaped lower single null configuration in the National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 2000] using high magnetic flux expansion at the divertor strike point and the radiative divertor technique. A partial detachment of the outer strike point was achieved with divertor deuterium injection leading to peak flux reduction from 4-6 MW m{sup -2} to 0.5-2 MW m{sup -2} in small-ELM 0.8-1.0 MA, 4-6 MW neutral beam injection-heated H-mode discharges. A self-consistent picture of outer strike point partial detachment was evident from divertor heat flux profiles and recombination, particle flux and neutral pressure measurements. Analytic scrape-off layer parallel transport models were used for interpretation of NSTX detachment experiments. The modeling showed that the observed peak heat flux reduction and detachment are possible with high radiated power and momentum loss fractions, achievable with divertor gas injection, and nearly impossible to achieve with main electron density, divertor neutral density or recombination increases alone.

  5. Twisted Vector Bundles on Pointed Nodal Curves

    Indian Academy of Sciences (India)

    Ivan Kausz

    2005-05-01

    Motivated by the quest for a good compactification of the moduli space of -bundles on a nodal curve we establish a striking relationship between Abramovich’s and Vistoli’s twisted bundles and Gieseker vector bundles.

  6. Plasma flow in the DIII-D divertor

    Energy Technology Data Exchange (ETDEWEB)

    Boedo, J.A. [Univ. of California, San Diego, CA (United States); Porter, G.D. [Lawrence Livermore National Lab., CA (United States); Schaffer, M.J. [General Atomics, San Diego, CA (United States)] [and others

    1998-07-01

    Indications that flows in the divertor can exhibit complex behavior have been obtained from 2-D modeling but so far remain mostly unconfirmed by experiment. An important feature of flow physics is that of flow reversal. Flow reversal has been predicted analytically and it is expected when the ionization source arising from neutral or impurity ionization in the divertor region is large, creating a high pressure zone. Plasma flows arise to equilibrate the pressure. A radiative divertor regime has been proposed in order to reduce the heat and particle fluxes to the divertor target plates. In this regime, the energy and momentum of the plasma are dissipated into neutral gas introduced in the divertor region, cooling the plasma by collisional, radiative and other atomic processes so that the plasma becomes detached from the target plates. These regimes have been the subject of extensive studies in DIII-D to evaluate their energy and particle transport properties, but only recently it has been proposed that the energy transport over large regions of the divertor must be dominated by convection instead of conduction. It is therefore important to understand the role of the plasma conditions and geometry on determining the region of convection-dominated plasma in order to properly control the heat and particle fluxes to the target plates and hence, divertor performance. The authors have observed complex structures in the deuterium ion flows in the DIII-D divertor. Features observed include reverse flow, convective flow over a large volume of the divertor and stagnant flow. They have measured large gradients in the plasma potential across the separatrix in the divertor and determined that these gradients induce poloidal flows that can potentially affect the particle balance in the divertor.

  7. Plasma flow in the DIII-D divertor

    International Nuclear Information System (INIS)

    Indications that flows in the divertor can exhibit complex behavior have been obtained from 2-D modeling but so far remain mostly unconfirmed by experiment. An important feature of flow physics is that of flow reversal. Flow reversal has been predicted analytically and it is expected when the ionization source arising from neutral or impurity ionization in the divertor region is large, creating a high pressure zone. Plasma flows arise to equilibrate the pressure. A radiative divertor regime has been proposed in order to reduce the heat and particle fluxes to the divertor target plates. In this regime, the energy and momentum of the plasma are dissipated into neutral gas introduced in the divertor region, cooling the plasma by collisional, radiative and other atomic processes so that the plasma becomes detached from the target plates. These regimes have been the subject of extensive studies in DIII-D to evaluate their energy and particle transport properties, but only recently it has been proposed that the energy transport over large regions of the divertor must be dominated by convection instead of conduction. It is therefore important to understand the role of the plasma conditions and geometry on determining the region of convection-dominated plasma in order to properly control the heat and particle fluxes to the target plates and hence, divertor performance. The authors have observed complex structures in the deuterium ion flows in the DIII-D divertor. Features observed include reverse flow, convective flow over a large volume of the divertor and stagnant flow. They have measured large gradients in the plasma potential across the separatrix in the divertor and determined that these gradients induce poloidal flows that can potentially affect the particle balance in the divertor

  8. The Atiyah Bundle and Connections on a Principal Bundle

    Indian Academy of Sciences (India)

    Indranil Biswas

    2010-06-01

    Let be a ∞ manifold and a Lie a group. Let $E_G$ be a ∞ principal -bundle over . There is a fiber bundle $\\mathcal{C}(E_G)$ over whose smooth sections correspond to the connections on $E_G$. The pull back of $E_G$ to $\\mathcal{C}(E_G)$ has a tautological connection. We investigate the curvature of this tautological connection.

  9. Divertor target shape optimization in realistic edge plasma geometry

    International Nuclear Information System (INIS)

    Tokamak divertor design for next-step fusion reactors heavily relies on numerical simulations of the plasma edge. Currently, the design process is mainly done in a forward approach, where the designer is strongly guided by his experience and physical intuition in proposing divertor shapes, which are then thoroughly assessed by numerical computations. On the other hand, automated design methods based on optimization have proven very successful in the related field of aerodynamic design. By recasting design objectives and constraints into the framework of a mathematical optimization problem, efficient forward-adjoint based algorithms can be used to automatically compute the divertor shape which performs the best with respect to the selected edge plasma model and design criteria. In the past years, we have extended these methods to automated divertor target shape design, using somewhat simplified edge plasma models and geometries. In this paper, we build on and extend previous work to apply these shape optimization methods for the first time in more realistic, single null edge plasma and divertor geometry, as commonly used in current divertor design studies. In a case study with JET-like parameters, we show that the so-called one-shot method is very effective is solving divertor target design problems. Furthermore, by detailed shape sensitivity analysis we demonstrate that the development of the method already at the present state provides physically plausible trends, allowing to achieve a divertor design with an almost perfectly uniform power load for our particular choice of edge plasma model and design criteria. (paper)

  10. 2D modelling and assessment of divertor performance for ITER

    International Nuclear Information System (INIS)

    The results of the ITER divertor modelling performed during the EDA are summarised in the paper. Studies on the operating window and optimisation of the divertor geometry are presented together with preliminary results on the start-up limiter performance. The issue of model validation against the experimental data which is crucial for extrapolation to ITER is also addressed. (author)

  11. Study of the radiation in divertor plasmas; Etude du rayonnement dans les plasmas de divertor

    Energy Technology Data Exchange (ETDEWEB)

    Laugier, F

    2000-10-19

    We have studied the cooling of the edge plasma by radiation in the divertor volume, in order to optimize the extraction of power in tokamaks and to limit the wall erosion. In attached divertor plasmas experiments, the concentration of intrinsic impurities at the edge is related to the response of the wall to the incident energy flow of plasma, depending on a phenomenological law. We carried out an analysis of the radiation according to this law and to the control parameters of the discharges. The largest radiated fraction and best synergy are obtained when the concentration of intrinsic impurities strongly increases with the energy of incident plasma. On the other hand, the erosion of the wall is stronger. In detached plasmas, we proved that the performances in terms of incident plasma energy loss and pressure loss are optimal when the density of the slowest neutrals is strong at the edge and when their radial penetration is small. On Tore Supra, we highlighted the correlations between the maximum Mach number of incident plasma flow, the radiation front and the penetration of the neutrals. A simple diagnostic based on the localization of the maximum Mach number proves that detached mode is not optimal on Tore Supra, because the radial penetration of the slowest neutrals is not sufficiently small. In the last part, we obtained the three-dimensional topology of the radiation in the ergodic divertor using a spectral analysis code and boundary conditions consistent with the temperature distribution on the wall. The radiation is maximum in front of the divertor modules. As a consequence, radiated power is underestimated by standards measurements of Tore Supra that are located between the modules. We finally showed that the profiles of temperature along the field lines are modulated, this is specific to the ergodic divertor. (author)

  12. Left bundle-branch block

    DEFF Research Database (Denmark)

    Risum, Niels; Strauss, David; Sogaard, Peter;

    2013-01-01

    The relationship between myocardial electrical activation by electrocardiogram (ECG) and mechanical contraction by echocardiography in left bundle-branch block (LBBB) has never been clearly demonstrated. New strict criteria for LBBB based on a fundamental understanding of physiology have recently...

  13. Bundling ecosystem services in Denmark

    DEFF Research Database (Denmark)

    Turner, Katrine Grace; Odgaard, Mette Vestergaard; Bøcher, Peder Klith; Dalgaard, Tommy; Svenning, J.-C.

    2014-01-01

    We made a spatial analysis of 11 ecosystem services at a 10 km × 10 km grid scale covering most of Denmark. Our objective was to describe their spatial distribution and interactions and also to analyze whether they formed specific bundle types on a regional scale in the Danish cultural landscape....... We found clustered distribution patterns of ecosystem services across the country. There was a significant tendency for trade-offs between on the one hand cultural and regulating services and on the other provisioning services, and we also found the potential of regulating and cultural services to...... form synergies. We identified six distinct ecosystem service bundle types, indicating multiple interactions at a landscape level. The bundle types showed specialized areas of agricultural production, high provision of cultural services at the coasts, multifunctional mixed-use bundle types around urban...

  14. Structure of the acrosomal bundle.

    Science.gov (United States)

    Schmid, Michael F; Sherman, Michael B; Matsudaira, Paul; Chiu, Wah

    2004-09-01

    In the unactivated Limulus sperm, a 60- micro m-long bundle of actin filaments crosslinked by the protein scruin is bent and twisted into a coil around the base of the nucleus. At fertilization, the bundle uncoils and fully extends in five seconds to support a finger of membrane known as the acrosomal process. This biological spring is powered by stored elastic energy and does not require the action of motor proteins or actin polymerization. In a 9.5-A electron cryomicroscopic structure of the extended bundle, we show that twist, tilt and rotation of actin-scruin subunits deviate widely from a 'standard' F-actin filament. This variability in structural organization allows filaments to pack into a highly ordered and rigid bundle in the extended state and suggests a mechanism for storing and releasing energy between coiled and extended states without disassembly. PMID:15343340

  15. Locking means for fuels bundles

    International Nuclear Information System (INIS)

    A nuclear power reactor fuel bundle is described which has a plurality of fuel rods disposed between two end plates positioned by tie rods extending therebetween. The assembled bundle is secured by one or more locking forks which pass through slots in the tie rod ends. Springs mounted on the fuel rods and tie rods are compressed by assembling the bundle and forcing one end plate against the locking fork to maintain the fuel rods and tie rods in position between the end plates. Downward pressure on the end plate permits removal of the locking fork so that the end plates may be removed, thus giving access to the fuel rods. This construction facilitates disassembly of an irradiated fuel bundle under water

  16. DIII-D radiative divertor project, status and plans

    International Nuclear Information System (INIS)

    New divertor hardware is being designed and fabricated for the Radiative Divertor modification of the DIII-D tokamak. The installation of the hardware has been separated into two phases, the first phase starting in October of 1996 and the and second final phase, in 1988. When completed, the Radiative Divertor Project hardware will provide pumping at all four strike points of a double-null, high triangularity discharge and provide baffling of the neutral particles from transport back to the core plasma. Radiative Divertor diagnostics are being designed to provide comprehensive measurements for diagnosing the divertor. Minimal modifications are required to diagnostics for the Phase I installation. More extensive diagnostic changes are planned for the Phase 2 installation. 3 refs., 6 figs

  17. Damage in Fiber Bundle Models

    OpenAIRE

    Kun, Ferenc; Zapperi, Stefano; Herrmann, Hans J.

    1999-01-01

    We introduce a continuous damage fiber bundle model that gives rise to macroscopic plasticity and compare its behavior with that of dry fiber bundles. Several interesting constitutive behaviors are found in this model depending on the value of the damage parameter and on the form of the disorder distribution. In addition, we compare the behavior of global load transfer models with local load transfer models and study in detail the damage evolution before failure. We emphasize the analogies be...

  18. Recent results from divertor and sol studies at JET

    International Nuclear Information System (INIS)

    Recent progress in the study of divertor and scrape-off layer plasma (SOL) phenomena in JET is reviewed. Up to the present time, three pumped divertors (Mark I, Mark IIA/AP and Mark IIGB) have been installed and exploited under reactor relevant conditions. With increased divertor closure, it is found that the particle exhaust rate has increased and neutral compression factors of >100 are obtained with the Mark IIGB divertor. Helium enrichment factors of >0.2 are measured under a wide range of conditions and satisfy the minimum requirements for ITER. Fast infrared camera measurements show broad deposition profiles during type I ELMs and energy densities of ∼0.12MJm-2. During the recent D-T experiments, the codeposition of tritium on cold shadowed surfaces in the inner divertor has been identified as an important form of long-term tritium retention. This has serious implications for the divertor design and tritium inventory in a next-step tokamak. Core plasma purity has not improved with enhanced divertor closure or decreased main chamber neutral pressure. Studies of the chemical sputtering yield have shown a dependence on surface temperature and hydrogen isotope. This accounts for the observation of increased impurity production and lower disruptive density limits in Mark II (at 500K) compared to Mark I (at 300K). Significant progress has been made in the study of divertor detachment, and volume recombination has been spectroscopically identified. With increasing isotope mass, detachment and the disruptive density limit occur at lower main plasma density as predicted by the EDGE2D/NIMBUS codes. Using differential gas fuelling in the Mark IIGB divertor, it has been possible to modify the in-out asymmetry of the divertor plasma for the first time. (author)

  19. Recent results from divertor and SOL studies at JET

    International Nuclear Information System (INIS)

    Recent progress in the study of divertor and scrape-off layer plasma (SOL) phenomena in JET is reviewed. Up to the present time, three pumped divertors (Mark I, Mark IIA/AP and Mark IIGB) have been installed and exploited under reactor relevant conditions. With increased divertor closure, it is found that the particle exhaust rate has increased and neutral compression factors of >100 are obtained with the Mark IIGB divertor. Helium enrichment factors of >0.2 are measured under a wide range of conditions and satisfy the minimum requirements for ITER. Fast infra-red camera measurements show broad deposition profiles during type I ELMs and energy densities of ∼0.12MJm-2. During the recent D-T experiments, the codeposition of tritium on cold shadowed surfaces in the inner divertor has been identified as an important form of long-term tritium retention. This has serious implications for the divertor design and tritium inventory in a next-step tokamak. Core plasma purity has not improved with enhanced divertor closure or decreased main chamber neutral pressure. Studies of the chemical sputtering yield have shown a dependence on surface temperature and hydrogen isotope. This accounts for the observation of increased impurity production and lower disruptive density limits in Mark II (at 500K) compared to Mark I (at 300K). Significant progress has been made in the study of divertor detachment, and volume recombination has been spectroscopically identified. With increasing isotope mass, detachment and the disruptive density limit occur at lower main plasma density as predicted by the EDGE2D/NIMBUS codes. Using differential gas fuelling in the Mark IIGB divertor, it has been possible to modify the in-out asymmetry of the divertor plasma for the first time. (author)

  20. Divertor asymmetry and scrape-off layer flow in various divertor configurations in Experimental Advanced Superconducting Tokamak

    DEFF Research Database (Denmark)

    Liu, S. C.; Guo, H. Y.; Xu, Guandong;

    2012-01-01

    plasmas exhibit the usual in-out asymmetry in particle and heat fluxes in LSN with the ion del B direction toward the lower X-point, favoring the outer divertor, especially at high density. The in-out asymmetry is reversed when changing the divertor configuration from LSN to USN, thus clearly...

  1. First measurements of electron temperature and density with divertor Thomson Scattering in radiative divertor discharges on DIII-D

    International Nuclear Information System (INIS)

    We have obtained the first measurements of ne and Te in the DIII-D divertor region with a multi-pulse (20 Hz) Divertor Thomson Scattering (DTS) system. Eight measurement locations are distributed vertically up to 21 cm above the divertor plate. Two-dimensional distributions have been obtained by sweeping the divertor plasma across the DTS measurement location. Several operating modes have been studied, including ohmic, L-mode, Elming H-mode, and Radiative Divertor operation with puffing of D2 and impurities. Mapping of the data to either the (Lpol, φ) or (R, Z) planes with the EFIT equilibrium is used to analyze the 2D profiles. We find that in ELMing H-mode: ne, Te, and Pe are relatively constant along field lines from the X-point to the divertor plate, especially near the separatrix field line. With D2 puffing, the DTS profiles indicate that Te in a large part of divertor region below the X-point is dramatically reduced from ∼30-40 eV in ELMing H-mode to 1-2 eV. This results in a fairly uniform low-Te divertor, with an increased electron density in the range of 2 to 4 x 1020 m-3. Detailed comparisons of the spatial profiles of ne, Te, and electron pressure Pe, are presented for several operating modes. In addition, these data are compared with initial calculations from the UEDGE fluid code

  2. Holomorphic bundles over elliptic manifolds

    International Nuclear Information System (INIS)

    In this lecture we shall examine holomorphic bundles over compact elliptically fibered manifolds. We shall examine constructions of such bundles as well as (duality) relations between such bundles and other geometric objects, namely K3-surfaces and del Pezzo surfaces. We shall be dealing throughout with holomorphic principal bundles with structure group GC where G is a compact, simple (usually simply connected) Lie group and GC is the associated complex simple algebraic group. Of course, in the special case G = SU(n) and hence GC = SLn(C), we are considering holomorphic vector bundles with trivial determinant. In the other cases of classical groups, G SO(n) or G = Sympl(2n) we are considering holomorphic vector bundles with trivial determinant equipped with a non-degenerate symmetric, or skew symmetric pairing. In addition to these classical cases there are the finite number of exceptional groups. Amazingly enough, motivated by questions in physics, much interest centres around the group E8 and its subgroups. For these applications it does not suffice to consider only the classical groups. Thus, while often first doing the case of SU(n) or more generally of the classical groups, we shall extend our discussions to the general semi-simple group. Also, we shall spend a good deal of time considering elliptically fibered manifolds of the simplest type, namely, elliptic curves

  3. Does size matter? : disentangling consumers' bundling preferences

    OpenAIRE

    Manoj K. Agarwal; Frambach, Ruud T.; Stremersch, Stefan

    2000-01-01

    Previous marketing literature has focused to a large extent on the effect of bundle characteristics on a consumer’s decision to buy a (fixed) bundle in a non-competitive setting. This study extends this narrow focus in four major ways. First, the authors address bundles that are customizable. Second, they distinguish between a consumer’s decision of whether to bundle (bundle choice) and the decision of how many goods or services to include in a bundle (bundle size). Third, they extend the foc...

  4. Results from Radiating Divertor Experiments with RMP ELM Suppression

    International Nuclear Information System (INIS)

    Full text: The successful integration of ELM suppression using resonant magnetic perturbations (RMPs) with puff-and-pump radiating divertor operation is demonstrated. In addition, because higher gas injection rates are needed to maintain plasma density after the RMP coils have been activated, a radiating divertor with resonant magnetic perturbation (RMP) produces considerably higher levels of radiated power from the divertor and scrape-off layer (SOL)/edge plasma regions than comparable non-RMP discharges at the same density. The radiating divertor has long been proposed as a reliable way to moderate steady heat flux at the divertor targets. Studies at DIII-D have demonstrated that RMPs are effective in suppressing ELMs and thus might be an attractive way to deal with the transient ELM-related heat flux problems expected for ITER. However, it was not clear as to whether RMP-based ELM suppression and radiating divertor scenarios were compatible. Recent DIII-D experiments comprise our first attempts to assess this compatibility by directly comparing the behaviors of injected 'seed' argon impurities in RMP and non-RMP puff-and-pump environments and by identifying issues that might limit the use of RMP ELM suppression with a radiating divertor approach. In the puff-and-pump scenarios used, argon was injected in the private flux region of a single-null magnetic configuration near the outer divertor target, while plasma flows into the divertor were enhanced by a combination of particle pumping near the outer divertor target and deuterium gas puffing upstream of the divertor targets. Differences in argon accumulation in the main plasma between RMP ELM-suppressed and similar non-RMP ELMing H-mode plasmas were relatively small, typically less than 20%. The core concentration of argon decreased as the deuterium gas puff rate was raised in RMP and non-RMP cases, suggesting that the detailed UEDGE analysis reported previously [1] for non-RMP divertor and SOL radiating divertor

  5. Researches on the Neutral Gas Pressure in the Divertor Chamber of the HL-2A Tokamak

    Institute of Scientific and Technical Information of China (English)

    WANGMingxu; LIBo; YANGZhigang; YANLongwen; HONGWenyu; YUANBaoshan; LIULi; CAOZeng; CUIChenghe; LIUYong; WANGEnyao; ZHANGNianman

    2003-01-01

    The neutral gas pressure in divertor chamber is a very basic and important physics parameter because it determines the temperature of charged particles, the thermal flux density onto divertor plates, the erosion of divertor plates, impurity retaining and exhausting, particle transportation and confinement performance of plasma in tokamaks. Therefore, the pressure measurement in divertor chamber is taken into account in many large tokamaks.

  6. Advanced Fuel Bundles for PHWRS

    International Nuclear Information System (INIS)

    The fuel used by NPCIL presently is natural uranium dioxide in the form of 19- element fuel bundles for 220 MWe PHWRs and 37-element fuel bundles for the TAPP-3&4 540 MWe units. The new 700 MWe PHWRs also use 37-element fuel bundles. These bundles are of short 0.5 m length of circular geometry. The cladding is of collapsible type made of Zircaloy-4 material. PHWRs containing a string of short length fuel bundles and the on-power refueling permit flexibility in using different advanced fuel designs and in core fuel management schemes. Using this flexibility, alternative fuel concepts are tried in Indian PHWRs. The advances in PHWR fuel designs are governed by the desire to use resources other than uranium, improve fuel economics by increasing fuel burnup and reduce overall spent nuclear fuel waste and improve reactor safety. The rising uranium prices are leading to a relook into the Thorium based fuel designs and reprocessed Uranium based and Plutonium based MOX designs and are expected to play a major role in future. The requirement of synergism between different type of reactors also plays a role. Increase in fuel burnup beyond 15 000 MW∙d/TeU in PHWRs, using higher fissile content materials like slightly enriched uranium, Mixed Oxide and Thorium Oxide in place of natural uranium in fuel elements, was studied many PHWR operating countries. The work includes reactor physics studies and test irradiation in research reactors and power reactors. Due to higher fissile content these bundles will be capable of delivering higher burnup than the natural uranium bundles. In India the fuel cycle flexibility of PHWRs is demonstrated by converting this type of technical flexibility to the real economy by irradiating these different types of advanced fuel materials namely Thorium, MOX, SEU, etc. The paper gives a review of the different advanced fuel design concepts studied for Indian PHWRs. (author)

  7. SOLPS Modeling of Slot Divertor Configuration on DIII-D

    Science.gov (United States)

    Sang, C. F.; Stangeby, P. C.; Guo, H. Y.; Lao, L. L.

    2015-11-01

    A major thrust of the DIII-D boundary/PMI initiative is to develop an advanced divertor configuration for next-step devices, such as FNSF and DEMO. We are adopting an integrated approach by optimizing both divertor structure and magnetic shape. Initial SOLPS modeling was carried out to optimize divertor structure shape to enhance divertor power dissipation, focusing on slot configurations. In particular, four different slot divertor structures, i.e., orthogonal-target slot, slanted-target slot, very narrow slot and v-shaped slot have been analyzed and comparisons made with an open divertor structure. It is found that the slot helps to trap recycling neutrals and impurities thus increasing radiative power dissipation in the divertor, reducing the electron temperature Te and the perpendicular heat flux q⊥ at the target plate. As expected, a narrower slot leads to lower Te and q⊥ than a less narrow one. The v-shaped slot appears to be especially effective at redirecting and concentrating recycling neutrals and impurities near the separatrix, thus promoting detachment at a lower upstream density than the other configurations. Work supported by US DOE under DE-FC02-04ER54698.

  8. A super-cusp divertor configuration for tokamaks

    International Nuclear Information System (INIS)

    Our study demonstrates a remarkable flexibility of advanced divertor configurations created with the remote poloidal field coils. The emphasis here is on the configurations with three poloidal field nulls in the divertor area. We are seeking the structures where all three nulls lie on the same separatrix, thereby creating two zones of a very strong flux expansion, as envisaged in the concept of Takase's cusp divertor. It turns out that the set of remote coils can produce a cusp divertor, with additional advantages of: (i) a large stand-off distance between the divertor and the coils and (ii) a thorough control that these coils exert over the fine features of the configuration. In reference to these additional favourable properties acquired by the cusp divertor, the resulting configuration could be called 'a super-cusp'. General geometrical features of the three-null configurations produced by remote coils are described. Furthermore, issues on the way to practical applications include the need for a more sophisticated control system and possible constraints related to excessively high currents in the divertor coils

  9. DIVERT: a divertor magnetic field line following code

    International Nuclear Information System (INIS)

    The computer code DIVERT has been written to trace magnetic field lines in the presence of a divertor. Its purpose is to allow a user to estimate the thickness of the plasma scrapeoff region and to provide a visual mapping of the magnetic field lines near the divertor. Included in the code is the capability to provide auxiliary graphics and compute the field ripple. The code can handle a divertor made up of any arrangement of straight line coil segments and will provide a graph of the field line configuration on output

  10. A review of ELMs in divertor tokamaks

    International Nuclear Information System (INIS)

    This paper reviews what is known about edge localized modes (ELMs), with an emphasis on their effect on the scrape-off layer and divertor plasmas. ELM effects have been measured in the ASDEX-U, C-Mod, COMPASS-D, DIII-D, JET, JFT-2M,JT-60U, and TCV tokamaks and are reported here. At least three types of ELMs have been identified and their salient features determined. Type-1 giant ELMs can cause the sudden loss of up to 10-15% of the plasma stored energy but their amplitude (ΔW/W) does not increase with increasing power. Type- 3 ELMs are observed near the H-mode power threshold and produce small energy dumps (1-3% of the stored energy). All ELMs increase the scrape- off layer plasma and produce particle fluxes on the divertor targets which are as much as ten times larger that the quiescent phase between ELMs. The divertor heat pulse is largest on the inner target, unlike that of L-Mode or quiescent H-mode; some tokamaks report radial structure in the heat flux profile which is suggestive of islands or helical structures. The power scaling of Type-1 ELM amplitude and frequency have been measured in several tokamaks and has recently been applied to predictions of the ELM Size in ITER. Concern over the expected ELM amplitude has led to a number of experiments aimed at demonstrating active control of ELMs. Impurity gas injection with feedback control on the radiation loss in ASDEX-U suggests that a promising mode of operation (the CDH-mode) with a very small type-3 ELMs can be maintained with heating power sell above the H-mode threshold, where giant type-1 ELMs can be maintained with heating power well above the H-mode threshold, where Giant type-1 ELMs are normally observed. While ELMs have many potential negative effects, the beneficial effect of ELMs in providing density control and limiting the core plasma impurity content in high confinement H- mode discharges should not be overlooked

  11. Cohomology of line bundles: Applications

    Science.gov (United States)

    Blumenhagen, Ralph; Jurke, Benjamin; Rahn, Thorsten; Roschy, Helmut

    2012-01-01

    Massless modes of both heterotic and Type II string compactifications on compact manifolds are determined by vector bundle valued cohomology classes. Various applications of our recent algorithm for the computation of line bundle valued cohomology classes over toric varieties are presented. For the heterotic string, the prime examples are so-called monad constructions on Calabi-Yau manifolds. In the context of Type II orientifolds, one often needs to compute cohomology for line bundles on finite group action coset spaces, necessitating us to generalize our algorithm to this case. Moreover, we exemplify that the different terms in Batyrev's formula and its generalizations can be given a one-to-one cohomological interpretation. Furthermore, we derive a combinatorial closed form expression for two Hodge numbers of a codimension two Calabi-Yau fourfold.

  12. Ergodic divertor impact on Tore Supra edge

    International Nuclear Information System (INIS)

    The present ergodic divertor experiments in Tore Supra have been devoted to benchmarking the operational regimes of the apparatus. Two major effects are reported: on the one hand, strong changes occur in the ergodized boundary layer (up to 20% of the minor radius), and on the other hand, the central plasma and especially the confinement is not directly affected, i.e. the observed modifications are induced by edge effects. The basic trends, which are recorded are a decrease of both the edge electronic temperature and the edge density gradient while the radiated power is increased at the very edge of the ergodic region. The latter feature is in agreement with the impurity line emission characterized by an increase of the peripheral lines with a strong decrease of the central lines. (orig.)

  13. Principal bundles the classical case

    CERN Document Server

    Sontz, Stephen Bruce

    2015-01-01

    This introductory graduate level text provides a relatively quick path to a special topic in classical differential geometry: principal bundles.  While the topic of principal bundles in differential geometry has become classic, even standard, material in the modern graduate mathematics curriculum, the unique approach taken in this text presents the material in a way that is intuitive for both students of mathematics and of physics. The goal of this book is to present important, modern geometric ideas in a form readily accessible to students and researchers in both the physics and mathematics communities, providing each with an understanding and appreciation of the language and ideas of the other.

  14. Divertor IR thermography on Alcator C-Moda)

    Science.gov (United States)

    Terry, J. L.; LaBombard, B.; Brunner, D.; Payne, J.; Wurden, G. A.

    2010-10-01

    Alcator C-Mod is a particularly challenging environment for thermography. It presents issues that will similarly face ITER, including low-emissivity metal targets, low-Z surface films, and closed divertor geometry. In order to make measurements of the incident divertor heat flux using IR thermography, the C-Mod divertor has been modified and instrumented. A 6° toroidal sector has been given a 2° toroidal ramp in order to eliminate magnetic field-line shadowing by imperfectly aligned divertor tiles. This sector is viewed from above by a toroidally displaced IR camera and is instrumented with thermocouples and calorimeters. The camera provides time histories of surface temperatures that are used to compute incident heat-flux profiles. The camera sensitivity is calibrated in situ using the embedded thermocouples, thus correcting for changes and nonuniformities in surface emissivity due to surface coatings.

  15. Impact of a poloidal divertor in ignition tokamak design

    International Nuclear Information System (INIS)

    System design studies were performed to assess the effect of assuming a poloidal divertor instead of a limiter as a means of impurity control for ignition tokamak configurations. Results show that for the nominal Tokamak Fusion Core Experiment (TFCX) device with superconducting TF coils, a feasible poloidal divertor configuration can be obtained without increasing the major radius. In the TFCX nominal copper TF coil device, however, field limits at the PF coils are exceeded when the effects of asymmetry associated with a poloidal divertor are included. It was found that a 12% increase in the major radius of this device is necessary to simultaneously satisfy the plasma-shaping requirements of a poloidal divertor and the magnetics constraints at the superconducting PF coils

  16. Two-dimensional divertor modeling and scaling laws

    International Nuclear Information System (INIS)

    Two-dimensional numerical models of divertors contain large numbers of dimensionless parameters that must be varied to investigate all operating regimes of interest. To simplify the task and gain insight into divertor operation, we employ similarity techniques to investigate whether model systems of equations plus boundary conditions in the steady state admit scaling transformations that lead to useful divertor similarity scaling laws. A short mean free path neutral-plasma model of the divertor region below the x-point is adopted in which all perpendicular transport is due to the neutrals. We illustrate how the results can be used to benchmark large computer simulations by employing a modified version of UEDGE which contains a neutral fluid model. (orig.)

  17. Neutral gas blanket effects in a gaseous divertor

    International Nuclear Information System (INIS)

    The gaseous divertor employs a neutral gas blanket to absorb the plasma heat flux in the divertor chamber. This novel method for resolving the heat loading problem in a conventional divertor system is simulated experimentally. In our operational range (nsub(e) 13 cm-3, Tsub(e) <= 5 eV) it is demonstrated that the localized plasma heat flux is scattered relatively uniformly with neutral pressures of a few microns. At large neutral pressures the plasma stream is neutralized without touching a material wall. Plasma pumping inhibits neutral backflow and can sustain a neutral pressure difference comparable to the plasma pressure. Effective divertor channel conductance is measured to be reduced by a factor of six. (orig.)

  18. Compatibility of detached divertor operation with robust edge pedestal performance

    Energy Technology Data Exchange (ETDEWEB)

    Leonard, A.W., E-mail: leonard@fusion.gat.com [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Makowski, M.A.; McLean, A.G. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Osborne, T.H.; Snyder, P.B. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States)

    2015-08-15

    The compatibility of detached radiative divertor operation with a robust H-mode pedestal is examined in DIII-D. A density scan produced low temperature plasmas at the divertor target, T{sub e} ⩽ 2 eV, with high radiation leading to a factor of ⩾4 drop in peak divertor heat flux. The cold radiative plasma was confined to the divertor and did not extend across the separatrix in X-point region. A robust H-mode pedestal was maintained with a small degradation in pedestal pressure at the highest densities. The response of the pedestal pressure to increasing density is reproduced by the EPED pedestal model. However, agreement of the EPED model with experiment at high density requires an assumption of reduced diamagnetic stabilization of edge Peeling–Ballooning modes.

  19. Numerical analysis of divertor plasma for demo-CREST

    International Nuclear Information System (INIS)

    The numerical analysis of the demonstration fusion reactor Demo-CREST has been carried out; this analysis focuses on impurity seeding. Several design activities for DEMO have been carried out; however, its detailed divertor plasma analysis remains to be carried out. Therefore, in this study, we discuss the possibility of neon puffing in demo-CREST to decrease the power load to the divertor plate by using the B2-EIRENE code. It has been shown that the radiation power loss by neon increases with upstream plasma density and that the peak power load to the divertor plate comes close to the allowable level by using the preliminary divertor configuration (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  20. Plasma detachment with molecular processes in divertor plasmas

    International Nuclear Information System (INIS)

    Molecular processes in detached recombining plasmas are briefly reviewed. Several reactions with vibrationally excited hydrogen molecule related to recombination processes are described. Experimental evidence of molecular activated recombination observed in a linear divertor plasma simulator is also shown. (author)

  1. Status of National Spherical Torus Experiment Liquid Lithium Divertor

    Science.gov (United States)

    Kugel, H. W.; Viola, M.; Ellis, R.; Bell, M.; Gerhardt, S.; Kaita, R.; Kallman, J.; Majeski, R.; Mansfield, D.; Roquemore, A. L.; Schneider, H.; Timberlake, J.; Zakharov, L.; Nygren, R. E.; Allain, J. P.; Maingi, R.; Soukhanovskii, V.

    2009-11-01

    Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. The next step in this work is the 2009 installation of a Liquid Lithium Divertor (LLD). The 20 cm wide LLD located on the lower outer divertor, consists of four, 80 degree sections; each section is separated by a row of graphite diagnostic tiles. The temperature controlled LLD structure consists of a 0.01cm layer of vacuum flame-sprayed, 50 percent porous molybdenum, on top of 0.02 cm, 316-SS brazed to a 1.9 cm Cu base. The physics design of the LLD encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

  2. Transport and divertor studies in the FM-1 spherator

    International Nuclear Information System (INIS)

    Fundamental problems of toroidal fusion devices have been investigated in the FM-1 Spherator. These subjects include the transport due to drift wave turbulence in the trapped electron regime, poloidal divertor and impurities, and lower hybrid heating. (auth)

  3. Thermal Fatigue Study on the Divertor Plate Materials

    Institute of Scientific and Technical Information of China (English)

    吴继红; 张斧; 许增裕; 严建成

    2002-01-01

    Thermal fatigue property of the divertor plate is one of the key issues that governs the lifetime of the divertor plate. Taking tungsten as surface material, a small-mock-up divertor plate was made by hot isostatic press welding (HIP). A thermal cycling experiment for divertor mock-up was carried out in the vacuum, where a high-heat-flux electronic gun was used as the thermal source. A cyclic heat flux of 9 MW/m2 was loaded onto the mock-up, a heating duration of 20 s was selected, the cooling water flow rate was 80 ml/s. After 1000 cycles, the surface and the W/Cu joint of the mock-up did not show any damage. The SEM was used to analyze the microstructure of the welding joint, where no cracks were found also.

  4. Non-ambipolar divertor flows in heliotron E

    International Nuclear Information System (INIS)

    The object of the work is to find out (1) the poloidal distributions of PEC in different poloidal cross-sections of the torus within one field period; (2) the link between PEC in the divertor flows (DF) and the characteristics of the divertor field lines; (3) the effect of different methods and regimes of heating on PEC. The data having been obtained enable us to understand at least partially the nature of PEC in the diverted plasma of H-E

  5. Design of divertor impurity monitoring system for ITER. 2

    International Nuclear Information System (INIS)

    The divertor impurity monitoring system of ITER has been designed. The main functions of this system are to identify impurity species and to measure the two-dimensional distributions of the particle influxes in the divertor plasmas. The wavelength range is 200 nm to 1000 nm. The viewing fans are realized by molybdenum mirrors located in the divertor cassette. With additional viewing fans seeing through the gap between the divertor cassettes, the region approximately from the divertor leg to the x-point will be observed. The light from the divertor region passes through the quartz windows on the divertor port plug and the cryostat, and goes through the dog-leg optics in the biological shield. Three different type of spectrometers: (i) survey spectrometers for impurity species monitoring, (ii) filter spectrometers for the particle influx measurement with the spatial resolution of 10 mm and the time resolution of 1 ms and (iii) high dispersion spectrometers for high resolution wavelength measurements are designed. These spectrometers are installed just behind the biological shield (for λ < 450 nm) to prevent the transmission loss in fiber and in the diagnostic room (for λ ≥ 450 nm) from the point of view of accessibility and flexibility. The optics have been optimized by a ray trace analysis. As a result, 10-15 mm spatial resolution will be achieved in all regions of the divertor. In addition, the measurable limit, the neutron and γ-ray irradiation effect on windows, a calibration method, an alignment method, a remote handling method and a data acquisition method are considered. (author)

  6. Design of divertor impurity monitoring system for ITER

    International Nuclear Information System (INIS)

    The divertor impurity monitoring system of ITER has been designed. The main objectives of this system are to identify impurity species and to measure two-dimensional distributions of particle influxes in the divertor plasma. This system, which is one of the most important diagnostic systems for plasma control of ITER, is nominated for the start-up set of ITER diagnostics. The conceptual design, the optical design and the mechanical design are mainly carried out. In order to satisfy the required measurements, three deferent type of spectral systems are selected corresponding to each objectives. First is the spectral system for impurity species monitoring. Second is the spectral system for particle influx measurement with spatial and time resolution. Third is the spectral system with high dispersion for particle energy distribution measurement in the divertor. The divertor impurity monitoring system is composed of these three systems. The two-dimensional measurement in the divertor is carried out with two viewing fans intersected each other. These viewing fans are realized by metallic mirrors (made of molybdenum or copper) sitting in the divertor cassette. In the optical design, the optimization of the optical system from the divertor to the spectrometer are carried out by using ray trace analysis. As the result, it is difficult to satisfy the spatial resolution of 3 mm in the divertor region. About 10 mm resolution will be reasonable. In addition, the measurable limit, the neutron and γ-ray irradiation effect on the optical fiber, the remote handling concept and the space requirement are considered preliminarily. The necessary design works during EDA, and necessary R and D are also listed. (author)

  7. Design of divertor impurity monitoring system for ITER. 2

    Energy Technology Data Exchange (ETDEWEB)

    Sugie, Tatsuo; Ogawa, Hiroaki; Ebisawa, Katsuyuki; Ando, Toshiro; Kasai, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Katsunuma, Atsushi; Maruo, Mitsumasa; Kita, Yoshio

    1998-11-01

    The divertor impurity monitoring system of ITER has been designed. The main functions of this system are to identify impurity species and to measure the two-dimensional distributions of the particle influxes in the divertor plasmas. The wavelength range is 200 nm to 1000 nm. The viewing fans are realized by molybdenum mirrors located in the divertor cassette. With additional viewing fans seeing through the gap between the divertor cassettes, the region approximately from the divertor leg to the x-point will be observed. The light from the divertor region passes through the quartz windows on the divertor port plug and the cryostat, and goes through the dog-leg optics in the biological shield. Three different type of spectrometers: (i) survey spectrometers for impurity species monitoring, (ii) filter spectrometers for the particle influx measurement with the spatial resolution of 10 mm and the time resolution of 1 ms and (iii) high dispersion spectrometers for high resolution wavelength measurements are designed. These spectrometers are installed just behind the biological shield (for {lambda} < 450 nm) to prevent the transmission loss in fiber and in the diagnostic room (for {lambda} {>=} 450 nm) from the point of view of accessibility and flexibility. The optics have been optimized by a ray trace analysis. As a result, 10-15 mm spatial resolution will be achieved in all regions of the divertor. In addition, the measurable limit, the neutron and {gamma}-ray irradiation effect on windows, a calibration method, an alignment method, a remote handling method and a data acquisition method are considered. (author)

  8. Nuclear analysis of the ITER full-tungsten divertor

    Energy Technology Data Exchange (ETDEWEB)

    Villari, R., E-mail: rosaria.villari@enea.it [ENEA Fusion Division, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Barabash, V.; Escourbiac, F.; Ferrand, L.; Hirai, T.; Komarov, V.; Loughlin, M.; Merola, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Moro, F. [ENEA Fusion Division, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Petrizzi, L. [IAEA Representative at OECD Nuclear Energy Agency, 92130 Issy-les-Moulinaux (France); Podda, S. [ENEA Fusion Division, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Polunovsky, E. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Brolatti, G. [ENEA Fusion Division, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy)

    2013-10-15

    Highlights: • 3D nuclear analysis of the last design of ITER full-W divertor. • Calculate nuclear heating, damage and he-production in divertor components. • Evaluate impact on design and maintenance of the system. -- Abstract: This paper presents the nuclear analysis performed for the ITER full-tungsten divertor using the MCNP-5 Monte Carlo Code in a 3-D geometry. A detailed model of the geometry of the divertor based on the last design specifications has been integrated into the latest ITER MCNP model. Nuclear heating, damage and helium production have been calculated. The presented results are consistent with recent analysis performed with ATTILA code and with the previous ones with MCNP. The shielding capabilities of the last design are reduced in comparison with the design of 2004 but a negligible impact is expected. The reweldability of radial pipes of the cassette body (CB) is not a concern even assuming irradiation during the whole ITER lifetime. The reweldability of pipes for the refurbishment of the CB with new plasma facing components depends on maintenance scenario. It is not expected that last 2012 divertor design leads to an appreciable increase of the nuclear loads on vacuum vessel and toroidal field coils, since the relative contribution of the divertor region remains low.

  9. Multi-fluid modeling of low-recycling divertor regimes

    International Nuclear Information System (INIS)

    The low-recycling regimes of divertor operation in a single-null NSTX magnetic configuration are studied using computer simulations with the edge plasma transport code UEDGE. The edge plasma transport properties pertinent to the low-recycling regimes are demonstrated. These include the flux-limited character of the parallel heat transport and the high plasma temperatures with the flattened profiles in the scrape-off-layer. It is shown that to maintain the balance of particle fluxes at the core interface the deuterium gas puffing rate should increase as the divertor recycling coefficient decreases. The radial profiles of the heat load to the outer divertor plate, the upstream radial plasma profiles, and the effects of the cross-field plasma transport in the low-recycling regimes are discussed. It is also shown that recycling of lithium impurities evaporating from the divertor plate at high surface temperatures can reverse the low-recycling divertor operational regime to the high-recycling one and may cause thermal instability of the divertor plate (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  10. Turbulent Simulations of Divertor Detachment Based On BOUT + + Framework

    Science.gov (United States)

    Chen, Bin; Xu, Xueqiao; Xia, Tianyang; Ye, Minyou

    2015-11-01

    China Fusion Engineering Testing Reactor is under conceptual design, acting as a bridge between ITER and DEMO. The detached divertor operation offers great promise for a reduction of heat flux onto divertor target plates for acceptable erosion. Therefore, a density scan is performed via an increase of D2 gas puffing rates in the range of 0 . 0 ~ 5 . 0 ×1023s-1 by using the B2-Eirene/SOLPS 5.0 code package to study the heat flux control and impurity screening property. As the density increases, it shows a gradually change of the divertor operation status, from low-recycling regime to high-recycling regime and finally to detachment. Significant radiation loss inside the confined plasma in the divertor region during detachment leads to strong parallel density and temperature gradients. Based on the SOLPS simulations, BOUT + + simulations will be presented to investigate the stability and turbulent transport under divertor plasma detachment, particularly the strong parallel gradient driven instabilities and enhanced plasma turbulence to spread heat flux over larger surface areas. The correlation between outer mid-plane and divertor turbulence and the related transport will be analyzed. Prepared by LLNL under Contract DE-AC52-07NA27344. LLNL-ABS-675075.

  11. Exploring Bundling Theory with Geometry

    Science.gov (United States)

    Eckalbar, John C.

    2006-01-01

    The author shows how instructors might successfully introduce students in principles and intermediate microeconomic theory classes to the topic of bundling (i.e., the selling of two or more goods as a package, rather than separately). It is surprising how much students can learn using only the tools of high school geometry. To be specific, one can…

  12. Bundled Discounts and EC Judicial Review

    OpenAIRE

    Christian Roques

    2008-01-01

    The Community Courts' case law is rich with cases relating to tying or bundling practices in their classical economic form. However, the same cannot be said for the second acceptance of bundled discounts.

  13. Failure properties of fiber bundle models

    OpenAIRE

    Pradhan, Srutarshi; Chakrabarti, Bikas K.

    2003-01-01

    We study the failure properties of fiber bundles when continuous rupture goes on due to the application of external load on the bundles. We take the two extreme models: equal load sharing model (democratic fiber bundles) and local load sharing model. The strength of the fibers are assumed to be distributed randomly within a finite interval. The democratic fiber bundles show a solvable phase transition at a critical stress (load per fiber). The dynamic critical behavior is obtained analyticall...

  14. Bundling Information Goods: Pricing, Profits, and Efficiency

    OpenAIRE

    Yannis Bakos; Erik Brynjolfsson

    1999-01-01

    We study the strategy of bundling a large number of information goods, such as those increasingly available on the Internet, and selling them for a fixed price. We analyze the optimal bundling strategies for a multiproduct monopolist, and we find that bundling very large numbers of unrelated information goods can be surprisingly profitable. The reason is that the law of large numbers makes it much easier to predict consumers' valuations for a bundle of goods than their valuations for the indi...

  15. Quantum principal bundles and corresponding gauge theories

    CERN Document Server

    Durdevic, M

    1995-01-01

    A generalization of classical gauge theory is presented, in the framework of a noncommutative-geometric formalism of quantum principal bundles over smooth manifolds. Quantum counterparts of classical gauge bundles, and classical gauge transformations, are introduced and investigated. A natural differential calculus on quantum gauge bundles is constructed and analyzed. Kinematical and dynamical properties of corresponding gauge theories are discussed.

  16. Strategic and welfare implications of bundling

    DEFF Research Database (Denmark)

    Martin, Stephen

    1999-01-01

    A standard oligopoly model of bundling shows that bundling by a firm with a monopoly over one product has a strategic effect because it changes the substitution relationships between the goods among which consumers choose. Bundling in appropriate proportions is privately profitable, reduces rivals...

  17. On Volumes of Arithmetic Line Bundles

    OpenAIRE

    Yuan, Xinyi

    2008-01-01

    We show an arithmetic generalization of the recent work of Lazarsfeld-Mustata which uses Okounkov bodies to study linear series of line bundles. As applications, we derive a log-concavity inequality on volumes of arithmetic line bundles and an arithmetic Fujita approximation theorem for big line bundles.

  18. Variation of divertor plasma parameters with divertor depth for H-mode discharges in DIII-D

    International Nuclear Information System (INIS)

    We report here the results of experiments aimed at quantifying the advantages of increasing the X-point to target-plate distance in a divertor tokamak operating with H-mode confinement. Larger distances should lower the peak electron temperature at the target plates, thereby reducing sputtering and lowering the impurity concentration in the core plasma. When gas puffing is used to reduce the divertor heat flux, extra field-line length may increase the volume available for radiation and increase gas isolation between the core and divertor regions. These experiments were carried out using a lower single-null open divertor configuration (IP = 1.4 MA, BT = 2.1 T) with neutral beam heating (PNBI = 4.8 and 6.8 MW) to produce ELMing H-mode discharges lasting about 3 s. The X-point height (zx) was varied from 1.5-32 cm above the target plates by changing the plasma elongation on a shot by shot basis; the X-point radius was also varied in order to keep the outer strike point aligned with divertor Langmuir probe tips. Though there was no gas fueling during the H-mode phase of the discharge, the plasma density remained constant for all Zx obtained. Additional D2 gas puffing for radiative divertor experiments was applied for the last 1.5 s of the H-mode period. (author) 5 refs., 4 figs

  19. Comment on "Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake" [Phys. Plasmas 20, 102507 (2013)

    Science.gov (United States)

    Ryutov, D. D.; Cohen, R. H.; Rognlien, T. D.; Soukhanovskii, V. A.; Umansky, M. V.

    2014-05-01

    In the recently published paper "Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake" [Phys. Plasmas 20, 102507 (2013)], the authors raise interesting and important issues concerning divertor physics and design. However, the paper contains significant errors: (a) The conceptual framework used in it for the evaluation of divertor "quality" is reduced to the assessment of the magnetic field structure in the outer Scrape-Off Layer. This framework is incorrect because processes affecting the pedestal, the private flux region and all of the divertor legs (four, in the case of a snowflake) are an inseparable part of divertor operation. (b) The concept of the divertor index focuses on only one feature of the magnetic field structure and can be quite misleading when applied to divertor design. (c) The suggestion to rename the divertor configurations experimentally realized on NSTX (National Spherical Torus Experiment) and DIII-D (Doublet III-D) from snowflakes to X-divertors is not justified: it is not based on comparison of these configurations with the prototypical X-divertor, and it ignores the fact that the NSTX and DIII-D poloidal magnetic field geometries fit very well into the snowflake "two-null" prescription.

  20. Simplicial principal bundles in parametrized spaces

    CERN Document Server

    Roberts, David M

    2012-01-01

    In this paper, motivated by recent interest in higher gauge theory, we prove that the fiberwise geometric realization functor takes a certain class of simplicial principal bundles in a suitable category of spaces over a fixed space $B$ to fiberwise principal bundles. As an application we show that the fiberwise geometric realization of the universal simplicial principal bundle for a simplicial group $G$ in the category of spaces over $B$ gives rise to a fiberwise principal bundle with structure group $|G|$. An application to classifying theory for fiberwise principal bundles is described.

  1. Multipath packet switch using packet bundling

    DEFF Research Database (Denmark)

    Berger, Michael Stubert

    The basic concept of packet bundling is to group smaller packets into larger packets based on, e.g., quality of service or destination within the packet switch. This paper presents novel applications of bundling in packet switching. The larger packets created by bundling are utilized to extend...... switching capacity by use of parallel switch planes. During the bundling operation, packets will experience a delay that depends on the actual implementation of the bundling and scheduling scheme. Analytical results for delay bounds and buffer size requirements are presented for a specific scheduling...

  2. Design, R&D and commissioning of EAST tungsten divertor

    Science.gov (United States)

    Yao, D. M.; Luo, G. N.; Zhou, Z. B.; Cao, L.; Li, Q.; Wang, W. J.; Li, L.; Qin, S. G.; Shi, Y. L.; Liu, G. H.; Li, J. G.

    2016-02-01

    After commissioning in 2005, the EAST superconducting tokamak had been operated with its water cooled divertors for eight campaigns up to 2012, employing graphite as plasma facing material. With increase in heating power over 20 MW in recent years, the heat flux going to the divertors rises rapidly over 10 MW m-2 for steady state operation. To accommodate the rapid increasing heat load in EAST, the bolting graphite tile divertor must be upgraded. An ITER-like tungsten (W) divertor has been designed and developed; and firstly used for the upper divertor of EAST. The EAST upper W divertor is modular structure with 80 modules in total. Eighty sets of W/Cu plasma-facing components (PFC) with each set consisting of an outer vertical target (OVT), an inner vertical target (IVT) and a DOME, are attached to 80 stainless steel cassette bodies (CB) by pins. The monoblock W/Cu-PFCs have been developed for the strike points of both OVT and IVT, and the flat type W/Cu-PFCs for the DOME and the baffle parts of both OVT and IVT, employing so-called hot isostatic pressing (HIP) technology for tungsten to CuCrZr heat sink bonding, and electron beam welding for CuCrZr to CuCrZr and CuCrZr to other material bonding. Both monoblock and flat type PFC mockups passed high heat flux (HHF) testing by means of electron beam facilities. The 80 divertor modules were installed in EAST in 2014 and results of the first commissioning are presented in this paper.

  3. Spectroscopic investigations of divertor detachment in TCV

    CERN Document Server

    Verhaegh, K; Duval, B P; Harrison, J R; Reimerdes, H; Theiler, C; Labit, B; Maurizio, R; Marini, C; Nespoli, F; Sheikh, U; Tsui, C K; Vianello, N; Vijvers, W A J

    2016-01-01

    The aim of this work is to provide an understanding of detachment at TCV with emphasis on analysis of the Balmer line emission. A new Divertor Spectroscopy System has been developed for this purpose. Further development of Balmer line analysis techniques has allowed detailed information to be extracted on free-free and three-body recombination. During density ramps, the plasma at the target detaches as inferred from a drop in density at, and ion current to, the target. At the same time the Balmer $6\\rightarrow2$ and $7\\rightarrow2$ line emission near the target is dominated by recombination, indicating that the ionization region has also detached from the target to be replaced by a recombining region with densities more than a factor 2 higher than at the target. As the core density increases further, the density and recombination rate are rising all along the outer leg to the x-point while remaining highest at the target. Even at the highest core densities accessed (Greenwald fraction 0.7) the peaks in recomb...

  4. Model of turbine blades bundles

    Czech Academy of Sciences Publication Activity Database

    Půst, Ladislav; Pešek, Luděk

    Prague : Institute of Thermomechanics, Academy of Sciences of the Czech Republic, v. v. i., 2013 - (Zolotarev, I.), s. 467-477 ISBN 978-80-87012-47-5. ISSN 1805-8256. [Engineering Mechanics 2013 /19./. Svratka (CZ), 13.05.2013-16.05.2013] R&D Projects: GA ČR GA101/09/1166 Institutional support: RVO:61388998 Keywords : free and forced vibrations * eigenmodes * mathematical model * bundle of blades Subject RIV: BI - Acoustics

  5. Model of turbine blades bundles

    Czech Academy of Sciences Publication Activity Database

    Půst, Ladislav; Pešek, Luděk

    Praha : Insitute of Thermomechanics ASCR, v. v. i., 2013 - (Zolotarev, I.). s. 125-126 ISBN 978-80-87012-46-8. ISSN 1805-8248. [Engineering Mechanics 2013 /19./. 13.05.2013-16.05.2013, Svratka] R&D Projects: GA ČR GA101/09/1166 Institutional support: RVO:61388998 Keywords : free and forced vibrations * eigenmodes * bundle of blades Subject RIV: BI - Acoustics

  6. Competitive nonlinear pricing and bundling

    OpenAIRE

    Armstrong, Mark; Vickers, John

    2006-01-01

    We examine the impact of multiproduct nonlinear pricing on profit, consumer surplus and welfare in a duopoly. When consumers buy all their products from one firm (the one-stop shopping model), nonlinear pricing leads to higher profit and welfare, but often lower consumer surplus, than linear pricing. By contrast, in a unit-demand model where consumers may buy one product from one firm and another product from another firm, bundling generally acts to reduce profit and welfare and to boost cons...

  7. Interpretation of the impurity distribution in the divertor during divertor plate biasing using the DIVIMP code

    Energy Technology Data Exchange (ETDEWEB)

    Haddad, E. E-mail: haddad@ccfm.ireq.ca; Meo, F.; Marchand, R.; Ratel, G.; Stansfield, B.L.; Gunn, J.; Stangeby, P.C.; Elder, J.D.; Lisgo, S.; Krieger, K

    2000-02-01

    Simulations of carbon transport using the DIVIMP code [P.C. Stangeby, J.D. Elder, J. Nucl. Mater. 196-198 (1992) 258] are compared with 2D toroidal images of CII and CIII radiation near the external divertor plates in TdeV ohmic plasmas (I{sub p}=170 kA, n{sub e}=3 x 10{sup 19} m{sup -3}, B{sub T}=1.4T). The main plasma parameters in the SOL and divertor are calculated by the onion skin model (OSM) [K. Shimizu et al., J. Nucl. Mater. 196-198 (1992) 476] included in DIVIMP, the neutrals being calculated by EIRENE [D. Reiter, Internal Report, KFA, Julich, 1947 (1984), 2599 (1992)] in an iterative loop. The results show that the carbon is mainly created by chemical sputtering, with a considerable fraction coming from the external oblique plate. By interpreting experimental CII and CIII distributions, it is found that carbon is affected by the biasing (-125 to +125 V) through a combination of at least three processes: the ion flux to the plates, the ExB drift velocity, and the cross field diffusion.

  8. Interpretation of the impurity distribution in the divertor during divertor plate biasing using the DIVIMP code

    International Nuclear Information System (INIS)

    Simulations of carbon transport using the DIVIMP code [P.C. Stangeby, J.D. Elder, J. Nucl. Mater. 196-198 (1992) 258] are compared with 2D toroidal images of CII and CIII radiation near the external divertor plates in TdeV ohmic plasmas (Ip=170 kA, ne=3 x 1019 m-3, BT=1.4T). The main plasma parameters in the SOL and divertor are calculated by the onion skin model (OSM) [K. Shimizu et al., J. Nucl. Mater. 196-198 (1992) 476] included in DIVIMP, the neutrals being calculated by EIRENE [D. Reiter, Internal Report, KFA, Julich, 1947 (1984), 2599 (1992)] in an iterative loop. The results show that the carbon is mainly created by chemical sputtering, with a considerable fraction coming from the external oblique plate. By interpreting experimental CII and CIII distributions, it is found that carbon is affected by the biasing (-125 to +125 V) through a combination of at least three processes: the ion flux to the plates, the ExB drift velocity, and the cross field diffusion

  9. Quantum bundles and their symmetries

    International Nuclear Information System (INIS)

    Wave functions in the domain of observables such as the Hamiltonian are not always smooth functions on the classical configuration space Q. Rather, they are often best regarded as functions on a G bundle EG over Q or as sections of an associated bundle. If H is a classical group which acts on Q, its quantum version HG, which acts on EG, is not always H, but an extension of H by G. A powerful and physically transparent construction of EG and HG, where G = U(1) and H1(Q,Z) = 0, has been developed using the path space P. (P consists of paths on Q from a fixed point). In this paper the authors show how to construct EG and HG when G is U(1) or U(1) x π1(Q) and there is no restriction on H1(Q,Z). The method is illustrated with concrete examples, such as a system of charges and monopoles. The method is illustrated with concrete examples, such as a system of charges and monopoles. The authors argue also that P is a sort of superbundle from which a large variety of bundles can be obtained by imposing suitable equivalence relations

  10. Photonic bandgap fiber bundle spectrometer

    CERN Document Server

    Hang, Qu; Syed, Imran; Guo, Ning; Skorobogatiy, Maksim

    2010-01-01

    We experimentally demonstrate an all-fiber spectrometer consisting of a photonic bandgap fiber bundle and a black and white CCD camera. Photonic crystal fibers used in this work are the large solid core all-plastic Bragg fibers designed for operation in the visible spectral range and featuring bandgaps of 60nm - 180nm-wide. 100 Bragg fibers were chosen to have complimentary and partially overlapping bandgaps covering a 400nm-840nm spectral range. The fiber bundle used in our work is equivalent in its function to a set of 100 optical filters densely packed in the area of ~1cm2. Black and white CCD camera is then used to capture spectrally "binned" image of the incoming light at the output facet of a fiber bundle. To reconstruct the test spectrum from a single CCD image we developed an algorithm based on pseudo-inversion of the spectrometer transmission matrix. We then study resolution limit of this spectroscopic system by testing its performance using spectrally narrow test peaks (FWHM 5nm-25nm) centered at va...

  11. Initial performance results of the DIII-D Divertor 2000

    International Nuclear Information System (INIS)

    A major upgrade of the DIII-D divertor, with the goal of enhancing impurity and density control and increasing the thermal pulse length limit of advanced tokamak (AT) plasmas has been successfully completed and commissioned. The integrated system that includes independent cryopumps at both the inner and the outer legs of the divertor, private flux region and outboard baffles, and improved graphite divertor armor, has been successfully applied to a variety of plasma conditions. Comparison of similar discharges before and after the upgrades show that with the new divertor the core plasma neutral source and carbon content are lower by as much as 50%. Calculations supported by preliminary infra-red (IR) camera measurements show that the new graphite armor design increases the limit on the discharge duration, due to temperature of the tile edges reaching sublimation point, by an order of magnitude. With the new system we have been able to control the density of high confinement H-mode plasmas to less than 1/3 of the Greenwald limit. It is observed that with divertor pumping during the current ramp phase the wall particle inventory and consequently the density rise after the H-mode transition can be significantly reduced

  12. Coherence imaging of flows in the DIII-D divertor

    Energy Technology Data Exchange (ETDEWEB)

    Howard, J.; Diallo, A.; Creese, M. [Plasma Research Laboratory, The Australian National University, Canberra (Australia); Allen, S.L.; Ellis, R.M.; Meyer, W.; Fenstermacher, M.E.; Porter, G.D. [Lawrence Livermore National Laboratory at General Atomics, San Diego (United States); Brooks, N.H.; Van Zeeland, M.E.; Boivin, R.L. [General Atomics, San Diego (United States)

    2011-03-15

    Various spatial heterodyne polarization interferometers for spectrally-resolved optical imaging of edge and core parameters in high temperature magnetized plasmas are described. Applications for such ''coherence imaging'' (CI) systems include imaging motional Stark effect and Zeeman effect polarimetry for determination of the magnetic field pitch angle, and passive and active (charge exchange recombination spectroscopy - CXRS) Doppler imaging of plasma temperature and flow. In this paper we describe spatial heterodyne coherence imaging systems and present first results of Doppler flow imaging in the DIII-D divertor. Instruments have been installed for imaging flows in the divertor and scrape-off-layer in the DIII-D tokamak and also for Doppler imaging on the H-1 heliac [1]. In the former case, single snapshot interferometric images of the plasma in CII 514nm, and CIII 465nm emission have been demodulated to obtain flow and ion temperature projections in both the scrape-off-layer and divertor. Flow field amplitudes in the divertor are found to be broad agreement with UEDGE modeling [2], and point the way towards experiments that address important divertor transport issues in future (copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  13. Analysis of sweeping heat loads on divertor plate materials

    International Nuclear Information System (INIS)

    The heat flux on the divertor plate of a fusion reactor is probably one of the most limiting constraints on its lifetime. The current heat flux profile on the outer divertor plate of a device like ITER is highly peaked with narrow profile. The peak heat flux can be as high as 30--40 MW/m2 with full width at half maximum (FWHM) is in the order of a few centimeters. Sweeping the separatrix along the divertor plate is one of the options proposed to reduce the thermomechanical effects of this highly peaked narrow profile distribution. The effectiveness of the sweeping process is investigated parametrically for various design values. The optimum sweeping parameters of a particular heat load will depend on the design of the divertor plate as well as on the profile of such a heat load. In general, moving a highly peaked heat load results in substantial reduction of the thermomechanical effects on the divertor plate. 3 refs., 8 figs

  14. Initial Development of the NSTX-U Snowflake Divertor Control

    Science.gov (United States)

    Vail, Patrick; Kolemen, Egemen; Welander, Anders; Lanctot, Matthew

    2015-11-01

    A feedback control system has been implemented at NSTX-U for real-time detection and manipulation of snowflake divertor (SFD) magnetic configurations. The SFD is an alternative magnetic divertor concept that is characterized by a second-order null formed by two x-points in close proximity. The SFD is an attractive option for heat flux mitigation for NSTX-U in which unmitigated peak heat fluxes in standard divertor operation near 20 MW/m2 may compromise plasma-facing components. The real-time control system at NSTX-U is capable of simultaneous control of multiple SFD parameters, such as the separation between the two x-points in the divertor region and their orientation. Control of SFD configurations in NSTX-U has been simulated in TOKSYS using the upgraded sets of poloidal field coils in both the upper and lower divertor regions. Performance of the real-time control system and its effect on plasma performance will be assessed experimentally as an initial step toward the development of the SFD concept at NSTX-U. Supported by the US DOE under DE-AC02-09CH11466.

  15. A novel approach to magnetic divertor configuration design

    International Nuclear Information System (INIS)

    Divertor exhaust system design and analysis tools are crucial to evolve from experimental fusion reactors towards commercial power plants. In addition to material research and dedicated vessel geometry design, improved magnetic configurations can contribute to sustaining the diverted heat loads. Yet, computational design of the magnetic divertor is a challenging process involving a magnetic equilibrium solver, a plasma edge grid generator and a computationally demanding plasma edge simulation. In this paper, an integrated approach to efficient sensitivity calculations is discussed and applied to a set of slightly reduced divertor models. Sensitivities of target heat load performance to the shaping coil currents are directly evaluated. Using adjoint methods, the cost for a sensitivity evaluation is reduced to about two times the simulation cost of one specific configuration. Further, the use of these sensitivities in an optimal design framework is illustrated by a case with realistic Joint European Torus (JET) configurational parameters

  16. Modeling of Alcator C-Mod Divertor Baffling Experiments

    Energy Technology Data Exchange (ETDEWEB)

    D. P. Stotler; C. S. Pitcher; C. J. Boswell; T. K. Chung; B. LaBombard; B. Lipschultz; J. L. Terry; R. J. Kanzleiter

    2000-11-29

    A specific Alcator C-Mod discharge from the series of divertor baffling experiments is simulated with the DEGAS 2 Monte Carlo neutral transport code. A simple two-point plasma model is used to describe the plasma variation between Langmuir probe locations. A range of conductances for the bypass between the divertor plenum and the main chamber are considered. The experimentally observed insensitivity of the neutral current flowing through the bypass and of the D alpha emissions to the magnitude of the conductance is reproduced. The current of atoms in this regime is being limited by atomic physics processes and not the bypass conductance. The simulated trends in the divertor pressure, bypass current, and D alpha emission agree only qualitatively with the experimental measurements, however. Possible explanations for the quantitative differences are discussed.

  17. Understanding atomic hydrogen behaviour in pumped divertor plasmas

    International Nuclear Information System (INIS)

    In order to set up a data base and diagnostic capability for understanding atomic hydrogen behaviour in pumped divertor plasmas, an experiment and a feasibility study using a novel laser-induced fluorescence (LIF) technique were performed. For the former, combined measurements of LIF tuned to Hα and emission intensities at Hα/Hβ were carried out on the compact helical system (CHS). The comparison of the measured data and a particle simulation code revealed atomic hydrogen behaviour quantitatively, providing a full estimate of toroidally and poloidally asymmetric distributions of hydrogen atoms. In order to supplement data base around the pumped divertor region, the applicability of an LIF technique which uses two-photon excitation from the ground state examined, based on the real optical constraints of the envisaged JET pumped divertor. It was concluded that ii is feasible and the only remaining problem is not a serious one. (orig.)

  18. Hydrogen recycling and transport in the helical divertor of TEXTOR

    International Nuclear Information System (INIS)

    The aim of this thesis was to investigate the hydrogen recycling at the target plates of the helical divertor in TEXTOR and by this the capability of this divertor configuration to access such favourable operational regimes. In order to study the different divertor density regimes in TEXTOR, discharges were performed in which the total plasma density was increased continuously up to the density limit. The recycling was investigated in a fixed helical divertor structure where four helical strike points with a poloidal width of about 8-10 cm are created at the divertor target plates. The experimental investigation of the hydrogen recycling was carried out using mainly spectroscopic methods supplemented by Langmuir probe, interferometric and atomic beam measurements. In the framework of this thesis a spectroscopic multi camera system has been built that facilitates the simultaneous observation of four different spectral lines, recording images of the divertor target plates and the plasma volume close to the target. The system facilitates the simultaneous measurement of the poloidal and toroidal pattern of the recycling flux at the divertor target without the need for sweeping the plasma structure. The simultaneous observation of different spectral lines reduces the uncertainty in the analysis based on several lines, as the contribution from uncertainties in the reproducibility of plasma parameters in different discharges are eliminated and only the uncertainty of the measurement method limits the accuracy. The spatial resolution of the system in poloidal and toroidal direction (0.8 mm±0.01 mm) is small compared to the separation of the helical strike points, the capability of the measurement method to resolve these structures is therefore limited by the line-of-sight integration and the penetration depth of the light emitting species. The measurements showed that the recycling flux increases linearly with increasing plasma density, a high recycling regime is not

  19. A survey of problems in divertor and edge plasma theory

    International Nuclear Information System (INIS)

    Theoretical physics problems related to divertor design are presented, organized by the region in which they occur. Some of the open questions in edge physics are presented from a theoretician's point of view. After a cursory sketch of the fluid models of the edge plasma and their numerical realization, the following topics are taken up: time-dependent problems, non-axisymmetric effects, anomalous transport in the scrape-off layer, edge kinetic theory, sheath effects and boundary conditions in divertors, electric field effects, atomic and molecular data issues, impurity transport in the divertor region, poloidally localized power dissipation (MARFEs and dense gas targets), helium ash removal, and neutral transport. The report ends with a summary of selected problems of particular significance and a brief bibliography of survey articles and related conference proceedings

  20. A novel approach to magnetic divertor configuration design

    Energy Technology Data Exchange (ETDEWEB)

    Blommaert, M., E-mail: m.blommaert@fz-juelich.de [Institute of Energy and Climate Research, Plasma Physics (IEK-4), FZ Jülich GmbH, D-52425 Jülich (Germany); Baelmans, M., E-mail: martine.baelmans@kuleuven.be [KU Leuven, Department of Mechanical Engineering, 3001 Leuven (Belgium); Dekeyser, W. [Institute of Energy and Climate Research, Plasma Physics (IEK-4), FZ Jülich GmbH, D-52425 Jülich (Germany); KU Leuven, Department of Mechanical Engineering, 3001 Leuven (Belgium); Gauger, N.R. [RWTH Aachen, Department of Mathematics and Center for Computational Engineering Science, D-52062 Aachen (Germany); Reiter, D. [Institute of Energy and Climate Research, Plasma Physics (IEK-4), FZ Jülich GmbH, D-52425 Jülich (Germany)

    2015-08-15

    Divertor exhaust system design and analysis tools are crucial to evolve from experimental fusion reactors towards commercial power plants. In addition to material research and dedicated vessel geometry design, improved magnetic configurations can contribute to sustaining the diverted heat loads. Yet, computational design of the magnetic divertor is a challenging process involving a magnetic equilibrium solver, a plasma edge grid generator and a computationally demanding plasma edge simulation. In this paper, an integrated approach to efficient sensitivity calculations is discussed and applied to a set of slightly reduced divertor models. Sensitivities of target heat load performance to the shaping coil currents are directly evaluated. Using adjoint methods, the cost for a sensitivity evaluation is reduced to about two times the simulation cost of one specific configuration. Further, the use of these sensitivities in an optimal design framework is illustrated by a case with realistic Joint European Torus (JET) configurational parameters.

  1. Simulations of NSTX with a liquid lithium divertor module

    International Nuclear Information System (INIS)

    A strategy to develop self-consistent simulations of the behavior of lithium in the Liquid Lithium Divertor (LLD) module to be installed in NSTX is described. In this initial stage of the plan, the UEDGE edge plasma transport code is used to simulate an existing NSTX shot, with UEDGE's transport coefficients set using midplane and divertor diagnostic data. The LLD is incorporated into the simulations as a reduction in the recycling coefficient over the outer divertor. Heat transfer calculations performed using the resulting heat flux profiles provide preliminary estimates on operating limits for the LLD as well as input data for subsequent steps in the LLD modeling effort (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  2. Hydrogen recycling and transport in the helical divertor of TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Clever, Meike

    2010-07-01

    The aim of this thesis was to investigate the hydrogen recycling at the target plates of the helical divertor in TEXTOR and by this the capability of this divertor configuration to access such favourable operational regimes. In order to study the different divertor density regimes in TEXTOR, discharges were performed in which the total plasma density was increased continuously up to the density limit. The recycling was investigated in a fixed helical divertor structure where four helical strike points with a poloidal width of about 8-10 cm are created at the divertor target plates. The experimental investigation of the hydrogen recycling was carried out using mainly spectroscopic methods supplemented by Langmuir probe, interferometric and atomic beam measurements. In the framework of this thesis a spectroscopic multi camera system has been built that facilitates the simultaneous observation of four different spectral lines, recording images of the divertor target plates and the plasma volume close to the target. The system facilitates the simultaneous measurement of the poloidal and toroidal pattern of the recycling flux at the divertor target without the need for sweeping the plasma structure. The simultaneous observation of different spectral lines reduces the uncertainty in the analysis based on several lines, as the contribution from uncertainties in the reproducibility of plasma parameters in different discharges are eliminated and only the uncertainty of the measurement method limits the accuracy. The spatial resolution of the system in poloidal and toroidal direction (0.8 mm{+-}0.01 mm) is small compared to the separation of the helical strike points, the capability of the measurement method to resolve these structures is therefore limited by the line-of-sight integration and the penetration depth of the light emitting species. The measurements showed that the recycling flux increases linearly with increasing plasma density, a high recycling regime is not

  3. Plasma transport in a simulated magnetic-divertor configuration

    Energy Technology Data Exchange (ETDEWEB)

    Strawitch, C. M.

    1981-03-01

    The transport properties of plasma on magnetic field lines that intersect a conducting plate are studied experimentally in the Wisconsin internal ring D.C. machine. The magnetic geometry is intended to simulate certain aspects of plasma phenomena that may take place in a tokamak divertor. It is found by a variety of measurements that the cross field transport is non-ambipolar; this may have important implications in heat loading considerations in tokamak divertors. The undesirable effects of nonambipolar flow make it preferable to be able to eliminate it. However, we find that though the non-ambipolarity may be reduced, it is difficult to eliminate entirely. The plasma flow velocity parallel to the magnetic field is found to be near the ion acoustic velocity in all cases. The experimental density and electron temperature profiles are compared to the solutions to a one dimensional transport model that is commonly used in divertor theory.

  4. Tungsten Divertor Target Technology and Test Facilities Development

    International Nuclear Information System (INIS)

    Full text: Tungsten divertor target technology development is in progress at IPR for water-cooled divertors of ITER-like tokamak. Test mock-ups are fabricated using tungsten materials in macro-brush as well as mono-block fashion. Vacuum brazing technique is used for macro-brush fabrication whereas high pressure high temperature diffusion bonding technique is used for mono-block fabrication. Experimental facilities are also being set-up at IPR for Non-destructive testing and high heat flux testing of divertor targets. Present paper describes recent results on high heat flux testing of the test mock-ups and briefly mention about some of the experimental test facilities being set-up at IPR. (author)

  5. Plasma transport in a simulated magnetic-divertor configuration

    International Nuclear Information System (INIS)

    The transport properties of plasma on magnetic field lines that intersect a conducting plate are studied experimentally in the Wisconsin internal ring D.C. machine. The magnetic geometry is intended to simulate certain aspects of plasma phenomena that may take place in a tokamak divertor. It is found by a variety of measurements that the cross field transport is non-ambipolar; this may have important implications in heat loading considerations in tokamak divertors. The undesirable effects of nonambipolar flow make it preferable to be able to eliminate it. However, we find that though the non-ambipolarity may be reduced, it is difficult to eliminate entirely. The plasma flow velocity parallel to the magnetic field is found to be near the ion acoustic velocity in all cases. The experimental density and electron temperature profiles are compared to the solutions to a one dimensional transport model that is commonly used in divertor theory

  6. ITER full tungsten divertor qualification program and progress

    International Nuclear Information System (INIS)

    The full tungsten divertor qualification program was defined for the R and D activity in domestic agencies. The qualification program consists of two steps: (i) technology development and validation and (ii) a full-scale demonstration. Small-scale mock-ups were manufactured in Japanese and European industries and delivered to the ITER divertor test facility in Russia for high heat flux testing. In parallel activity to the qualification program, both domestic agencies demonstrated that W monoblock technologies withstanding up to 20 MW m−2 were available. (paper)

  7. Divertor development for a future fusion power plant

    International Nuclear Information System (INIS)

    Nuclear fusion is considered as a future source of sustainable energy supply. In the first chapter, the physical principle of magnetic plasma confinement, and the function of a tokamak are described. Since the discovery of the H-mode in ASDEX experiment ''Divertor I'' in 1982, the divertor has been an integral part of all modern tokamaks and stellarators, not least the ITER machine. The goal of this work is to develop a feasible divertor design for a fusion power plant to be built after ITER. This task is particularly challenging because a fusion power plant formulates much greater demands on the structural material and the design than ITER in terms of neutron wall load and radiation. First several divertor concepts proposed in the literature e.g. the Power Plant Conceptual Study (PPCS) using different coolants are reviewed and analyzed with respect to their performance. As a result helium cooled divertor concept exhibited the best potential to come up to the highest safety requirements and therefore has been chosen for the design process. From the third chapter the necessary steps towards this goal are described. First, the boundary conditions for the arrangement of a divertor with respect to the fusion plasma are discussed, as this determines the main thermal and neutronic load parameters. Based on the loads material selection criteria are inherently formulated. In the next step, the reference design is defined in accordance with the established functional design specifications. The developed concept is of modular nature and consists of cooling fingers of tungsten using an impingement cooling in order to achieve a heat dissipation of 10 MW/m2. In the next step, the design was subjected to the thermal-hydraulic and thermo-mechanical calculations in order to analyze and improve the performance and the manufacturing technologies. Based on these results, a prototype was produced and experimentally tested on their cooling capacity, their thermo-cyclic loading behavior

  8. An analytic model for flow reversal in divertor plasmas

    International Nuclear Information System (INIS)

    An analytic model is developed and used to study the phenomenon of flow reversal which is observed in two-dimensional simulations of divertor plasmas. The effect is shown to be caused by the radial spread of neutral particles emitted from the divertor target which can lead to a strong peaking of the ionization source at certain radial locations. The results indicate that flow reversal over a portion of the width of the scrape-off layer is inevitable in high recycling conditions. Implications for impurity transport and particle removal in reactors are discussed

  9. PARAMETRIC SENSITIVITY STUDY OF A FUSION REACTOR DIVERTOR COOLING FINGER

    OpenAIRE

    Martin, Oliver; SIMONOVSKI IGOR

    2012-01-01

    In this paper the results of coupled thermal-mechanical Finite Element (FE) analysis on the design of a fusion reactor divertor cooling finger are presented. Beside its main purpose, to remove alpha particles, helium and other impurities from the plasma stream, a divertor has to remove approximately 15% of the total thermal power of the fusion reactor. The aim of the analysis is to assess the influence of a number of physical properties of the brazing layer (BL) of the cooling finger on the o...

  10. Quadratic bundle and nonlinear equations

    International Nuclear Information System (INIS)

    The paper is aimed at giving an exhaustive description of the nonlinear evolution equations (NLEE), connected with the quadratic bundle (the spectral parameter lambda, which enters quadratically into the equations) and at describing Hamiltonian structure of these equations. The equations are solved through the inverse scattering method (ISM). The basic formulae for the scattering problem are given. The spectral expansion of the integrodifferential operator is used so that its eigenfunctions are the squared solutions of the equation. By using the notions of Hamiltonian structure hierarchy and gauge transformations it is shown how to single out physically interesting NLEE

  11. Static stress analysis of CANFLEX fuel bundles

    International Nuclear Information System (INIS)

    The static stress analysis of CANFLEX bundles is performed to evaluate the fuel structural integrity during the refuelling service. The structure analysis is carried out by predicting the drag force, stress and displacements of the fuel bundle. By the comparison of strength tests and analysis results, the displacement values are well agreed within 15%. The analysis shows that the CANFLEX fuel bundle keep its structural integrity. 24 figs., 6 tabs., 12 refs. (Author) .new

  12. Damping Properties of the Hair Bundle

    OpenAIRE

    Baumgart, Johannes; Kozlov, Andrei S.; Risler, Thomas; Hudspeth, A. James

    2015-01-01

    The viscous liquid surrounding a hair bundle dissipates energy and dampens oscillations, which poses a fundamental physical challenge to the high sensitivity and sharp frequency selectivity of hearing. To identify the mechanical forces at play, we constructed a detailed finite-element model of the hair bundle. Based on data from the hair bundle of the bullfrog's sacculus, this model treats the interaction of stereocilia both with the surrounding liquid and with the liquid in the narrow gaps b...

  13. Tying, Bundling, and Loyalty/Requirement Rebates

    OpenAIRE

    Nicholas Economides

    2011-01-01

    I discuss the impact of tying, bundling, and loyalty/requirement rebates on consumer surplus in the affected markets. I show that the Chicago School Theory of a single monopoly surplus that justifies tying, bundling, and loyalty/requirement rebates on the basis of efficiency typically fails. Thus, tying, bundling, and loyalty/requirement rebates can be used to extract consumer surplus and enhance profit of firms with market power. I discuss the various setups when this occurs.

  14. Bundling and Competition on the Internet

    OpenAIRE

    Yannis Bakos; Erik Brynjolfsson

    2000-01-01

    The Internet has signi.cantly reduced the marginal cost of producing and distributing digital information goods. It also coincides with the emergence of new competitive strategies such as large-scale bundling. In this paper, we show that bundling can create “economies of aggregation” for information goods if their marginal costs are very low, even in the absence of network externalities or economies of scale or scope. We extend the Bakos-Brynjolfsson bundling model (1999) to settings with sev...

  15. Bundling and joint marketing by rival firms

    OpenAIRE

    Jeitschko, Thomas D.; Jung, Yeonjei; Kim, Jaesoo

    2014-01-01

    We study joint marketing arrangements by competing firms who engage in price discrimination between consumers who patronize only one firm (single purchasing) and those who purchase from both competitors (bundle purchasers). Two types of joint marketing are considered. Firms either commit to a component-price that applies to bundle-purchasers and then firms set stand-alone prices for single purchasers; or firms commit to a rebate off their stand alone price that will be applied to bundle-purch...

  16. Statistical Constitutive Equation of Aramid Fiber Bundles

    Institute of Scientific and Technical Information of China (English)

    熊杰; 顾伯洪; 王善元

    2003-01-01

    Tensile impact tests of aramid (Twaron) fiber bundles were carried om under high strain rates with a wide range of 0. 01/s~1000/s by using MTS and bar-bar tensile impact apparatus. Based on the statistical constitutive model of fiber bundles, statistical constitutive equations of aramid fiber bundles are derived from statistical analysis of test data at different strain rates. Comparison between the theoretical predictions and experimental data indicates statistical constitutive equations fit well with the experimental data, and statistical constitutive equations of fiber bundles at different strain rates are valid.

  17. Hydraulic characteristics of HANARO fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Cho, S.; Chung, H. J.; Chun, S. Y.; Yang, S. K.; Chung, M. K. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper presents the hydraulic characteristics measured by using LDV (Laser Doppler Velocimetry) in subchannels of HANARO, KAERI research reactor, fuel bundle. The fuel bundle consists of 18 axially finned rods with 3 spacer grids, which are arranged in cylindrical configuration. The effects of the spacer grids on the turbulent flow were investigated by the experimental results. Pressure drops for each component of the fuel bundle were measured, and the friction factors of fuel bundle and loss coefficients for the spacer grids were estimated from the measured pressure drops. Implications regarding the turbulent thermal mixing were discussed. Vibration test results measured by using laser vibrometer were presented. 9 refs., 12 figs. (Author)

  18. Visible wide angle view imaging system of KTM tokamak based on multielement image fiber bundle

    Energy Technology Data Exchange (ETDEWEB)

    Chektybayev, B., E-mail: Chektybaev@nnc.kz; Shapovalov, G.; Kolodeshnikov, A. [Institute of Atomic Energy Branch of National Nuclear Center, Kurchatov (Kazakhstan)

    2015-05-15

    In the paper, new visible wide angle view imaging system of KTM tokamak is described. The system has been designed to observe processes inside of plasma and the processes occurring due to plasma-wall interactions through the long equatorial port. Imaging system is designed based on special image fiber bundle and entrance wide angle lens, which provide image of large section of the vacuum chamber, both poloidal half-section and divertor through the sufficiently long equatorial port. The system also consists of two video cameras: slow and fast with image intensifier. Commercial equipment had been used in design of the system that allowed reducing the cost and time for research and development. The paper also discusses advantages and disadvantages of the system in comparison with conventional endoscopes based on a lens system and considers its promising utilization in future tokamaks and future steady state fusion reactors.

  19. Visible wide angle view imaging system of KTM tokamak based on multielement image fiber bundle.

    Science.gov (United States)

    Chektybayev, B; Shapovalov, G; Kolodeshnikov, A

    2015-05-01

    In the paper, new visible wide angle view imaging system of KTM tokamak is described. The system has been designed to observe processes inside of plasma and the processes occurring due to plasma-wall interactions through the long equatorial port. Imaging system is designed based on special image fiber bundle and entrance wide angle lens, which provide image of large section of the vacuum chamber, both poloidal half-section and divertor through the sufficiently long equatorial port. The system also consists of two video cameras: slow and fast with image intensifier. Commercial equipment had been used in design of the system that allowed reducing the cost and time for research and development. The paper also discusses advantages and disadvantages of the system in comparison with conventional endoscopes based on a lens system and considers its promising utilization in future tokamaks and future steady state fusion reactors. PMID:26026523

  20. Thermal Analysis of the Divertor Primary Heat Transfer System Piping During the Gas Baking Process

    Energy Technology Data Exchange (ETDEWEB)

    Yoder Jr, Graydon L [ORNL; Harvey, Karen [ORNL; Ferrada, Juan J [ORNL

    2011-02-01

    A preliminary analysis has been performed examining the temperature distribution in the Divertor Primary Heat Transfer System (PHTS) piping and the divertor itself during the gas baking process. During gas baking, it is required that the divertor reach a temperature of 350 C. Thermal losses in the piping and from the divertor itself require that the gas supply temperature be maintained above that temperature in order to ensure that all of the divertor components reach the required temperature. The analysis described in this report was conducted in order to estimate the required supply temperature from the gas heater.

  1. Three-dimensional neutronics and shielding analyses for the ITER divertor

    International Nuclear Information System (INIS)

    3-D neutronics and shielding analyses have been performed for the divertor region of the ITER interim design. The peak neutron wall loading in the divertor region is 0.6 MW/m2 at the divertor cassette dome. The total nuclear heating in the 60 divertor cassettes is 102.4 MW. The peak helium production in the VV behind the pumping ducts is 0.5 He appm/FPY implying that rewelding might be feasible. The total nuclear heating in the parts of the TF coils in the divertor region is only 2.1 kW. 5 refs., 4 figs., 5 tabs

  2. Extension of holomorphic bundles to the disc (and Serre's Problem on Stein bundles)

    OpenAIRE

    Rosay, Jean-Pierre

    2006-01-01

    We show how to extend some holomorphic bundles with fifer C^2 and base an open set in C, to bundles on the Riemann Sphere, by an extremely simple technique. In particular, it applies to examples of non-Stein bundles constructed by Skoda and Demailly. It gives an example on C, with polynomial transition automorphisms.

  3. Effects of divertor plate biasing on radial and poloidal edge fluxes in the TdeV

    International Nuclear Information System (INIS)

    The divertor plates of TdeV, a tokamak with a double-null divertor and closed divertor chambers, have been electrically biased with respect to the walls. The paper discusses the resulting effects on the edge electron density profile, on the neutral pressures and impurity fluxes in the main vacuum chamber and the divertor chambers, and on the plasma flow to the divertors. As a function of the bias voltage, which was varied between - 180 V and + 160 V, the electron density scrape-off width and the wall impurity influxes increase monotonically; the flows to the top and bottom divertors vary strongly, in qualitative agreement with an E-vector x B-vector/B2 rotation, but not symmetrically. With negative biasing, the electrostatic barrier and the rotation combine to give a strong improvement of the divertor efficiency. (author). 30 refs, 10 figs

  4. Controlled detachment and particle transport in the divertor plasma in TdeV

    International Nuclear Information System (INIS)

    At high densities, the plasma detaches from the outboard divertor plates in TdeV. The signatures are a reduction of the ion flux to the divertor plate, movement of the radiating zone from the plate toward the X-point, a pressure gradient between an ionization front and the target plate, and strong cross-field transport in the divertor. A toroidally-viewing TV imaging system allows us to observe local interactions between the divertor plasma and the different divertor plates. As the plasma detaches, the gas pressure in the divertor continues to rise, and there is evidence for molecular processes in the cold plasma near the divertor plates. Auxiliary heating increases the power and particle flow across the separatrix; our results suggest that detachment depends on the energy transported per particle. Simulations using the B2/EIRENE and DIVIMP codes give reasonable agreement with the measurements for the attached phase. (orig.)

  5. Controlled detachment and particle transport in the divertor plasma in TdeV

    Energy Technology Data Exchange (ETDEWEB)

    Stansfield, B.L. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Meo, F. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Abel, G. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Boucher, C. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Gauvreau, J.-L. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Gunn, J.P. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Haddad, E. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Lachambre, J.-L. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Mailloux, J. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Marchand, R. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Ratel, G. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Richard, N. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Shoucri, M.M. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Terreault, B. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Beaudry, S. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Decoste, R. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Pacher, G.W. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Zuzak, W. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Elder, J.D. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada); Stangeby, P.C. [Centre Canadien de Fusion Magnetique, Varennes, PQ (Canada)

    1997-02-01

    At high densities, the plasma detaches from the outboard divertor plates in TdeV. The signatures are a reduction of the ion flux to the divertor plate, movement of the radiating zone from the plate toward the X-point, a pressure gradient between an ionization front and the target plate, and strong cross-field transport in the divertor. A toroidally-viewing TV imaging system allows us to observe local interactions between the divertor plasma and the different divertor plates. As the plasma detaches, the gas pressure in the divertor continues to rise, and there is evidence for molecular processes in the cold plasma near the divertor plates. Auxiliary heating increases the power and particle flow across the separatrix; our results suggest that detachment depends on the energy transported per particle. Simulations using the B2/EIRENE and DIVIMP codes give reasonable agreement with the measurements for the attached phase. (orig.).

  6. Energy and particle control characteristics of the ASDEX Upgrade 'LYRA' divertor

    International Nuclear Information System (INIS)

    In 1997 the new 'LYRA' divertor went into operation at ASDEX Upgrade and the neutral beam heating power was increased to 20 MW by installation of a second injector. This leads to the relatively high value of P/R of 12 MW/m. It has been shown that the ASDEX Upgrade LYRA divertor is capable of handling such high heating powers. Measurements presented in this paper reveal a reduction of the maximum heat flux in the LYRA divertor by more than a factor of two compared to the open Divertor I. This reduction is caused by radiative losses inside the divertor region. Carbon radiation cools the divertor plasma down to a few eV where hydrogen radiation losses become significant. They are increased due to an effective reflection of neutrals into the hot separatrix region. B2-Eirene modelling of the performed experiments supports the experimental findings and refines the understanding of loss processes in the divertor region. (and others)

  7. Design study on divertor plates of Large Helical Device (LHD)

    International Nuclear Information System (INIS)

    A conceptual design has been completed for the divertor plates of the Large Helical Device (LHD, R = 3.9 m, ap = 50 ∼ 60 cm, Bh = 3 ∼ 4T/ superconducting coils of NbTi) and the detailed technical design is now in progress. The design concept and the status of research and development (R and D) programs are described. (author)

  8. Edge plasma control by local island divertor in LHD

    International Nuclear Information System (INIS)

    In the Large Helical Device (LHD) program, one of the key research issues is to enhance helical plasma performance through the edge plasma control. For the first time in the LHD program, the edge plasma control was performed with a local island divertor (LID) that is a closed divertor, utilizing an m/n 1/1 island generated externally by 20 small perturbation coils, and fundamental LID functions were demonstrated experimentally. It was found that the outward heat and particle fluxes crossing the island separatrix flow along the field lines to the backside of the divertor head, where carbon plates are placed to receive the heat and particle loads. Accordingly high efficient pumping was demonstrated, which is considered to be the key in realizing high temperature divertor operation, resulting in an improvement of energy confinement. In the present experiment, a factor of ∼1.2 improvement of the energy confinement time, τE, was observed at a magnetic axis position, Rax, of 3.75 m over the International Stellarator Scaling 95. Results of edge modeling are also presented by using the EMC3-EIRENE code. (author)

  9. Material and design considerations for the carbon armored ITER divertor

    International Nuclear Information System (INIS)

    The properties of materials for the carbon armored ITER divertor were evaluated from literature and manufacturers' documentation. Most of these data, however, have been not known or not published yet. We have evaluated an optimum data set of the candidate materials of the ITER divertor, which were needed for finite element analyses (FEM). The materials evaluated are as follows; MFC-1, CX2002U, SEP-N112, P-130, IG-430U for the carbon based materials, and Oxygen Free Copper (OFCu), Dispersion Strengthened Copper (DSCu), TZM, W5Re and W-Cu as a heat sink material. It should be noted that W-Cu is first proposed for a heat sink application of the ITER divertor plate. The finite element analyses were performed for the residual stress induced by brazing, thermal response and thermal stresses under a uniform heat flux of 15 MW/m2 to the plasma facing surface. The stress free temperature of 750degC is assumed for the residual stress by brazing. Ten different geometries of the divertor were considered in the analyses including possible material combinations. The FEM results show that the material combinations of MFC-1 and W-30Cu or DSUc in the flat-plate geometry satisfy the presently accepted ITER requirements. The combinations of CX2002U and TZM or W5Re is considered a good choice in terms of residual and thermal stresses, whereas the surface temperature exceeds the ITER requirements. (author) 106 refs

  10. Electron and molecular ion collisions relevant to divertor plasma

    International Nuclear Information System (INIS)

    We introduce the concept of the multi-channel quantum defect theory (MQDT) and show the outline of the MQDT newly extended to include the dissociative states. We investigate some molecular processes relevant to the divertor plasma by using the MQDT: the dissociative recombination, dissociative excitation, and rotation-vibrational transition in the hydrogen molecular ion and electron collisions. (author)

  11. Modeling results for a linear simulator of a divertor

    Energy Technology Data Exchange (ETDEWEB)

    Hooper, E.B.; Brown, M.D.; Byers, J.A.; Casper, T.A.; Cohen, B.I.; Cohen, R.H.; Jackson, M.C.; Kaiser, T.B.; Molvik, A.W.; Nevins, W.M.; Nilson, D.G.; Pearlstein, L.D.; Rognlien, T.D.

    1993-06-23

    A divertor simulator, IDEAL, has been proposed by S. Cohen to study the difficult power-handling requirements of the tokamak program in general and the ITER program in particular. Projections of the power density in the ITER divertor reach {approximately} 1 Gw/m{sup 2} along the magnetic fieldlines and > 10 MW/m{sup 2} on a surface inclined at a shallow angle to the fieldlines. These power densities are substantially greater than can be handled reliably on the surface, so new techniques are required to reduce the power density to a reasonable level. Although the divertor physics must be demonstrated in tokamaks, a linear device could contribute to the development because of its flexibility, the easy access to the plasma and to tested components, and long pulse operation (essentially cw). However, a decision to build a simulator requires not just the recognition of its programmatic value, but also confidence that it can meet the required parameters at an affordable cost. Accordingly, as reported here, it was decided to examine the physics of the proposed device, including kinetic effects resulting from the intense heating required to reach the plasma parameters, and to conduct an independent cost estimate. The detailed role of the simulator in a divertor program is not explored in this report.

  12. Modeling results for a linear simulator of a divertor

    International Nuclear Information System (INIS)

    A divertor simulator, IDEAL, has been proposed by S. Cohen to study the difficult power-handling requirements of the tokamak program in general and the ITER program in particular. Projections of the power density in the ITER divertor reach ∼ 1 Gw/m2 along the magnetic fieldlines and > 10 MW/m2 on a surface inclined at a shallow angle to the fieldlines. These power densities are substantially greater than can be handled reliably on the surface, so new techniques are required to reduce the power density to a reasonable level. Although the divertor physics must be demonstrated in tokamaks, a linear device could contribute to the development because of its flexibility, the easy access to the plasma and to tested components, and long pulse operation (essentially cw). However, a decision to build a simulator requires not just the recognition of its programmatic value, but also confidence that it can meet the required parameters at an affordable cost. Accordingly, as reported here, it was decided to examine the physics of the proposed device, including kinetic effects resulting from the intense heating required to reach the plasma parameters, and to conduct an independent cost estimate. The detailed role of the simulator in a divertor program is not explored in this report

  13. Plasma/neutral gas transport in divertors and limiters

    International Nuclear Information System (INIS)

    The engineering design of the divertor and first wall region of fusion reactors requires accurate knowledge of the energies and particle fluxes striking these surfaces. Simple calculations indicate that approx. 10 MW/m2 heat fluxes and approx. 1 cm/yr erosion rates are possible, but there remain fundamental physics questions that bear directly on the engineering design. The purpose of this study was to treat hydrogen plasma and neutral gas transport in divertors and pumped limiters in sufficient detail to answer some of the questions as to the actual conditions that will be expected in fusion reactors. This was accomplished in four parts: (1) a review of relevant atomic processes to establish the dominant interactions and their data base; (2) a steady-state coupled O-D model of the plasma core, scrape-off layer and divertor exhaust to determine gross modes of operation and edge conditions; (3) a 1-D kinetic transport model to investigate the case of collisionless divertor exhaust, including non-Maxwellian ions and neutral atoms, highly collisional electrons, and a self-consistent electric field; and (4) a 3-D Monte Carlo treatment of neutral transport to correctly account for geometric effects

  14. Edge plasma control by local island divertor in LHD

    International Nuclear Information System (INIS)

    In the Large Helical Device (LHD) program, one of the key research issues is to enhance helical plasma performance through the edge plasma control. For the first time in the LHD program, the edge plasma control was performed with a local island divertor (LID) that is a closed divertor, utilizing an m/n=1/1 island generated externally by 20 small perturbation coils, and fundamental LID functions were demonstrated experimentally. It was found that the outward heat and particle fluxes crossing the island separatrix flow along the field lines to the backside of the divertor head, where carbon plates are placed to receive the heat and particle loads. Accordingly high efficient pumping was demonstrated, which is considered to be the key in realizing high temperature divertor operation, resulting in an improvement of energy confinement. In the present experiment, relatively good energy confinement is achieved in the high density regime at a magnetic axis position, Rax, of 3.75 m. Results of edge modelling are also presented by using the EMC3-EIRENE code. (author)

  15. Taming the plasma-material interface with the snowflake divertor.

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V A

    2015-04-24

    Experiments in several tokamaks have provided increasing support for the snowflake configuration as a viable tokamak heat exhaust concept. This white paper summarizes the snowflake properties predicted theoretically and studied experimentally, and identifies outstanding issues to be resolved in existing and future facilities before the snowflake divertor can qualify for the reactor interface.

  16. Spectroscopic study of JT-60U divertor plasma

    International Nuclear Information System (INIS)

    Particle behavior in the JT-60U divertor plasmas has been studied spectroscopically. Doppler profiles of the Dα line have been investigated for understanding of atomic and molecular processes in deuterium particle recycling and Dα line emission. Near the divertor plates, dissociative excitation from deuterium molecules and molecular ions plays an important role for the line emission. By investigation of spectral profiles of the He I line (667.8 nm), Doppler broadening due to elastic scattering by protons has been found. It is estimated that the penetration probability of the helium atoms from the divertor plates to the main plasma and the helium atom flux to the gap for pumping increase by 30% due to the elastic scattering. Intensity distribution of the CD band (around 430.5 nm) has been compared between the W-shaped divertor with a dome in the private flux region and the previous open one. The dome prevents the upstream transport of hydrocarbon impurity produced by chemical sputtering. (author)

  17. Progress in ergodic divertor operation on Tore Supra

    International Nuclear Information System (INIS)

    Upgrade of the Tore ergodic divertor has led to significant progress in ergodic divertor physics. The disruptive limit governed by the stochastization of the outer magnetic surfaces is found to occur for a value of the Chirikov parameter reaching 2 on the magnetic surface q = 2 + 3 / 12. This experimentally observed robustness allows one to operate at very low safety factor on the separatrix (q ∼ 2). Numerical analysis of ballooning turbulence in a stochastic layer indicates that the decay of the density fluctuations is in associated with an increase of the fluctuating electric drift velocity. The bottom line is then an enhanced cross-field transport in the vicinity of the target plates. This lowering of confinement appears to be compensated by an intrinsic transport barrier on the electron temperature. The 3-D response of the temperature field is computed with a fluid code. The intrinsic transport barrier at the separatrix, reported experimentally, can be recovered together with small amplitude temperature modulations in the divertor volume. Experimental evidence of the 3 density regimes (linear, high recycling and detachment) is reported. The low critical density values for these transitions indicate that similar parallel physics govern the axisymmetric and ergodic divertor, despite the open configuration of the latter. Measurement and understanding of these density regimes provide a means for feedback control of plasma density and an improvement in ICRH coupling scenarios. Experimental data also indicated that particle control with the vented target plates is effective. Increase of impurity control and radiation efficiency are recalled. Global power balance has been analysed. These results confirm the enhanced radiation capacity of the ergodic divertor. (author)

  18. PFC integration on Tore-Supra WEST divertor

    International Nuclear Information System (INIS)

    Full text of publication follows. In the context of the Tokamak Tore Supra evolution, the CEA Cadarache aims at transforming it into a test bench for ITER plasma facing components. This project named WEST (Tungsten Environment in Steady state Tokamak) is especially focused on the divertor target. The modification of the machine, by adding two axisymmetric divertors will make feasible an H-mode, and an X-point close to the lower divertor. This environment will allow exposing the divertor components up to 20 MW/m2 heat flux during long pulse. These specifications are well suited to test the actively cooled tungsten target elements, respecting the ITER design. One challenge in such machine evolution is to integrate components in an existing vacuum vessel in order to obtain the best achievable performance. The divertors coils are designed regarding the magnetic specifications, the plasma facing components are placed according to the plasma shape, and then the interfaces have to be managed regarding the remaining space. Moreover in this layer, many important smaller components have to be integrated as cooling pipes, magnetic diagnostics, gas injection, Langmuir probe, etc. This paper deals with the integrated design of ITER tungsten target elements into the WEST environment considering magnetic, electric, thermal and mechanical loads. The feasibility of installation and maintenance has to be strongly considered as PFC will be replaced several times. The ports size allows entering a 30 degrees sector of pre-installed tungsten targets which will be plugged as quickly and easily has possible. The main feature of steady state operations is the active cooling, which lead to have many embedded cooling channels and bulky pipes on the PFC module. It means to take care of the many connections and sealing between vacuum and water. (authors)

  19. Principal Bundles on the Projective Line

    Indian Academy of Sciences (India)

    V B Mehta; S Subramanian

    2002-08-01

    We classify principal -bundles on the projective line over an arbitrary field of characteristic ≠ 2 or 3, where is a reductive group. If such a bundle is trivial at a -rational point, then the structure group can be reduced to a maximal torus.

  20. The Verlinde formula for Higgs bundles

    CERN Document Server

    Andersen, Jørgen Ellegaard; Pei, Du

    2016-01-01

    We propose and prove the Verlinde formula for the quantization of the Higgs bundle moduli spaces and stacks for any simple and simply-connected group. This generalizes the equivariant Verlinde formula for the case of $SU(n)$ proposed previously by the second and third author. We further establish a Verlinde formula for the quantization of parabolic Higgs bundle moduli spaces and stacks.

  1. CANFLEX fuel bundle strength tests (test report)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Seok Kyu; Chung, C. H.; Kim, B. D.

    1997-08-01

    This document outlines the test results for the strength tests of the CANFLEX fuel bundle. Strength tests are performed to determine and verify the amount of the bundle shape distortion which is against the side-stops when the bundles are refuelling. There are two cases of strength test; one is the double side-stop test which simulates the normal bundle refuelling and the other is the single side-stop test which simulates the abnormal refuelling. the strength test specification requires that the fuel bundle against the side-stop(s) simulators for this test were fabricated and the flow rates were controlled to provide the required conservative hydraulic forces. The test rig conditions of 120 deg C, 11.2 MPa were retained for 15 minutes after the flow rate was controlled during the test in two cases, respectively. The bundle loading angles of number 13- number 15 among the 15 bundles were 67.5 deg CCW and others were loaded randomly. After the tests, the bundle shapes against the side-stops were measured and inspected carefully. The important test procedures and measurements were discussed as follows. (author). 5 refs., 22 tabs., 5 figs.

  2. CANFLEX fuel bundle strength tests (test report)

    International Nuclear Information System (INIS)

    This document outlines the test results for the strength tests of the CANFLEX fuel bundle. Strength tests are performed to determine and verify the amount of the bundle shape distortion which is against the side-stops when the bundles are refuelling. There are two cases of strength test; one is the double side-stop test which simulates the normal bundle refuelling and the other is the single side-stop test which simulates the abnormal refuelling. the strength test specification requires that the fuel bundle against the side-stop(s) simulators for this test were fabricated and the flow rates were controlled to provide the required conservative hydraulic forces. The test rig conditions of 120 deg C, 11.2 MPa were retained for 15 minutes after the flow rate was controlled during the test in two cases, respectively. The bundle loading angles of number 13- number 15 among the 15 bundles were 67.5 deg CCW and others were loaded randomly. After the tests, the bundle shapes against the side-stops were measured and inspected carefully. The important test procedures and measurements were discussed as follows. (author). 5 refs., 22 tabs., 5 figs

  3. k-Gerbes, Line Bundles and Anomalies

    CERN Document Server

    Ekstrand, C

    2000-01-01

    We use sets of trivial line bundles for the realization of gerbes. For1-gerbes the structure arises naturally for the Weyl fermion vacuum bundle at afixed time. The Schwinger term is an obstruction in the triviality of a1-gerbe.

  4. k-Gerbes, Line Bundles and Anomalies

    International Nuclear Information System (INIS)

    We use sets of trivial line bundles for the realization of gerbes. For 1-gerbes the structure arises naturally for the Weyl fermion vacuum bundle at a fixed time. The Schwinger term is an obstruction in the triviality of a 1-gerbe. (author)

  5. k-Gerbes, Line Bundles and Anomalies

    OpenAIRE

    Ekstrand, C.

    2000-01-01

    We use sets of trivial line bundles for the realization of gerbes. For 1-gerbes the structure arises naturally for the Weyl fermion vacuum bundle at a fixed time. The Schwinger term is an obstruction in the triviality of a 1-gerbe.

  6. Heights for line bundles on arithmetic surfaces

    OpenAIRE

    Jahnel, Joerg

    1995-01-01

    For line bundles on arithmetic varieties we construct height functions using arithmetic intersection theory. In the case of an arithmetic surface, generically of genus g, for line bundles of degree g equivalence is shown to the height on the Jacobian defined by the Theta divisor.

  7. Damping Properties of the Hair Bundle

    Science.gov (United States)

    Baumgart, Johannes; Kozlov, Andrei S.; Risler, Thomas; Hudspeth, A. J.

    2011-11-01

    The viscous liquid surrounding a hair bundle dissipates energy and dampens oscillations, which poses a fundamental physical challenge to the high sensitivity and sharp frequency selectivity of hearing. To identify the mechanical forces at play, we constructed a detailed finite-element model of the hair bundle. Based on data from the hair bundle of the bullfrog's sacculus, this model treats the interaction of stereocilia both with the surrounding liquid and with the liquid in the narrow gaps between the individual stereocilia. The investigation revealed that grouping stereocilia in a bundle dramatically reduces the total drag. During hair-bundle deflections, the tip links potentially induce drag by causing small but very dissipative relative motions between stereocilia; this effect is offset by the horizontal top connectors that restrain such relative movements at low frequencies. For higher frequencies the coupling liquid is sufficient to assure that the hair bundle moves as a unit with a low total drag. This work reveals the mechanical characteristics originating from hair-bundle morphology and shows quantitatively how a hair bundle is adapted for sensitive mechanotransduction.

  8. Fock modules and noncommutative line bundles

    Science.gov (United States)

    Landi, Giovanni

    2016-09-01

    To a line bundle over a noncommutative space there is naturally associated a Fock module. The algebra of corresponding creation and annihilation operators is the total space algebra of a principal U(1) -bundle over the noncommutative space. We describe the general construction and illustrate it with examples.

  9. New achievements of the Divertor Test Platform programme for the ITER divertor remote maintenance R and D

    International Nuclear Information System (INIS)

    The divertor assembly for the ITER fusion reactor consists of a number of rail mounted cassettes (54 now in ITER FEAT) located in the bottom region of the vacuum vessel. These cassettes shall be removed/installed remotely during the life of the reactor by means of specific devices. To demonstrate and optimise the feasibility of the in-vessel maintenance process the Divertor Test Platform (DTP) has been established at the ENEA Research Centre in Brasimone, Italy, as a major part of the large ITER R and D project L7. A first set of tests has been already carried out and reported during 1998, when the basic feasibility of the divertor replacement was demonstrated. In the present period (January 1999-July 2000), new activities, including both site tests and other 'external' R and D works, have been carried out in order to refine and improve the ITER divertor maintenance scenario. These include the study of abnormal maintenance operations and of possible handling equipment failure and its consequences; the procurement and testing of new sub-systems (e.g. a force reflection manipulator arm), and the development of remote handling techniques including a virtual reality system. Following a short description of the DTP, this paper reports on the new results and achievements, draws the relevant conclusions, and finally discusses future activities

  10. Dirac structures and Dixmier-Douady bundles

    CERN Document Server

    Alekseev, A

    2009-01-01

    A Dirac structure on a vector bundle V is a maximal isotropic subbundle E of the direct sum of V with its dual. We show how to associate to any Dirac structure a Dixmier-Douady bundle A, that is, a Z/2Z-graded bundle of C*-algebras with typical fiber the compact operators on a Hilbert space. The construction has good functorial properties, relative to Morita morphisms of Dixmier-Douady bundles. As applications, we show that the `spin' Dixmier-Douady bundle over a compact, connected Lie group (as constructed by Atiyah-Segal) is multiplicative, and we obtain a canonical `twisted Spin-c-structure' on spaces with group valued moment maps.

  11. Bringing the CANFLEX fuel bundle to market

    International Nuclear Information System (INIS)

    CANFLEX is a 43-element CANDU fuel bundle, under joint development by AECL and KAERI, to facilitate the use of various advanced fuel cycles in CANDU reactors through the provision of enhanced operating margins. The bundle uses two element diameters (13.5 and 11.5 mm ) to reduce element ratings by 20%, and includes the use of critical-heat-flux (CHF) enhancing appendages to increase the minimum CHF ratio or dryout margin of the bundle. Test programs are underway to demonstrate: the irradiation behaviour, hydraulic characteristics and reactor physics properties of the bundle, along with a test program to demonstrate the ability of the bundle to be handled by CANDU-6 fuelling machines. A fuel design manual and safety analysis reports have been drafted, and both analyses, plus discussions with utilities are underway for a demonstration irradiation in a CANDU-6 reactor. (author)

  12. Line bundle embeddings for heterotic theories

    Science.gov (United States)

    Nibbelin, Stefan Groot; Ruehle, Fabian

    2016-04-01

    In heterotic string theories consistency requires the introduction of a non-trivial vector bundle. This bundle breaks the original ten-dimensional gauge groups E8 × E8 or SO(32) for the supersymmetric heterotic string theories and SO(16) × SO(16) for the non-supersymmetric tachyon-free theory to smaller subgroups. A vast number of MSSM-like models have been constructed up to now, most of which describe the vector bundle as a sum of line bundles. However, there are several different ways of describing these line bundles and their embedding in the ten-dimensional gauge group. We recall and extend these different descriptions and explain how they can be translated into each other.

  13. CANDU fuel bundle skin friction factor

    International Nuclear Information System (INIS)

    Single-phase, incompressible fluid flow skin friction factor correlations, primarily for CANDU 37-rod fuel bundles, were reviewed. The correlations originated from curve-fits to flow test data, mostly with new fuel bundles in new pressure tubes (flow tubes), without internal heating. Skin friction in tubes containing fuel bundles (noncircular flow geometry) was compared to that in equivalent diameter smooth circular tubes. At Reynolds numbers typical of normal flows in CANDU fuel channels, the skin friction in tubes containing bundles is 8 to 15% higher than in equivalent diameter smooth circular tubes. Since the correlations are based on scattered results from measurements, the skin friction with bundles may be even higher than indicated above. The information permits over- or under-prediction of the skin friction, or choosing an intermediate value of friction, with allowance for surface roughnesses, in thermal-hydraulic analyses of CANDU heat transport systems. (author) 9 refs., 2 figs

  14. Line bundle embeddings for heterotic theories

    CERN Document Server

    Nibbelink, Stefan Groot

    2016-01-01

    In heterotic theories consistency requires the introduction of a non-trivial vector bundle. This bundle breaks the original ten-dimensional gauge groups E_8 x E_8 or SO(32) for the supersymmetric heterotic theories and SO(16) x SO(16) for the non-supersymmetric tachyon-free theory to smaller subgroups. A vast number of MSSM-like models have been constructed up to now, most of which describe the vector bundle as a sum of line bundles. However, there are several different ways of describing these line bundles and their embedding in the ten-dimensional gauge group. We recall and extend these different descriptions and explain how they can be translated into each other.

  15. Canonical singular hermitian metrics on relative logcanonical bundles

    OpenAIRE

    Tsuji, Hajime

    2010-01-01

    This supersedes 0704.0566. We prove the invariance of logarithmic plurigenera for a projective family of KLT pairs and the adjoint line bundle of KLT line bundles. The proof uses the canonical singular hermitian metrics on relative logcanonical bundles.

  16. On Harder–Narasimhan Reductions for Higgs Principal Bundles

    Indian Academy of Sciences (India)

    Arijit Dey; R Parthasarathi

    2005-05-01

    The existence and uniqueness of – reduction for the Higgs principal bundles over nonsingular projective variety is shown. We also extend the notion of – reduction for (, )-bundles and ramified -bundles over a smooth curve.

  17. Effect of bundle size on BWR fuel bundle critical power performance

    International Nuclear Information System (INIS)

    Effect of the bundle size on the BWR fuel bundle critical power performance was studied. For this purpose, critical power tests were conducted with both 6 x 6 (36 heater rods) and 12 x 12 (144 heater rods) size bundles in the GE ATLAS heat transfer test facility located in San Jose, California. All the bundle geometries such as rod diameter, rod pitch and rod space design are the same except size of flow channel. Two types of critical power tests were performed. One is the critical power test with uniform local peaking pattern for direct comparison of the small and large bundle critical power. Other is the critical power test for lattice positions in the bundle. In this test, power of a group of four rods (2 x 2 array) in a lattice region was peaked higher to probe the critical power of that lattice position in the bundle. In addition, the test data were compared to the COBRAG calculations. COBRAG is a detailed subchannel analysis code for BWR fuel bundle developed by GE Nuclear Energy. Based on these comparisons the subchannel model was refined to accurately predict the data obtained in this test program, thus validating the code capability of handling the effects of bundle size on bundle critical power for use in the study of the thermal hydraulic performance of the future advance BWR fuel bundle design. The author describes the experimental portion of the study program

  18. Gauge symmetries and fibre bundles

    International Nuclear Information System (INIS)

    The matter is organized as follows. After a brief introduction to the concept of gauge invariance and its relationship to determinism, we introduce in chapters 3 and 4 the notion of fibre bundles in the context of a discussion on spinning point particles and Dirac monopoles. Chapter 3 deals with a non relativistic treatment of the spinning particle. The non trivial extension to relativistic spinning particles is dealt with in Chapter 5. The free particle system as well as interactions with external electromagnetic and gravitational fields are discussed in detail. In chapter 5 we also elaborate on a remarkable relationship between the charge-monopole system and the system of a massless particle with spin. The classical description of Yang-Mills particles with internal degrees of freedom, such as isospin or colour, is given in chapter 6. We apply the above in a discussion of the classical scattering of particles off a 't Hooft-Polyakov monopole. In chapter 7 we elaborate on a Kaluza-Klein description of particles with internal degrees of freedom. The canonical formalism and the quantization of most of the preceeding systems are discussed in chapter 8. The dynamical systems given in chapters 3-7 are formulated on group manifolds. The procedure for obtaining the extension to super-group manifolds is briefly discussed in chapter 9. In chapter 10, we show that if a system admits only local Lagrangians for a configuration space Q, then under certain conditions, it admits a global Lagrangian when Q is enlarged to a suitable U(1) bundle over Q. Conditions under which a symplectic form is derivable from a Lagrangian are also found. (orig./HSI)

  19. The control of divertor carbon erosion/redeposition in the DIII-D tokamak

    International Nuclear Information System (INIS)

    The DIII-D tokamak has demonstrated an operational scenario where the graphite-covered divertor is free of net erosion. Reduction of divertor carbon erosion is accomplished using a low temperature (detached) divertor plasma that eliminates physical sputtering. Likewise, the carbon source rate arising from chemical erosion is found to be very low in the detached divertor. Near strikepoint regions, the rate of carbon deposition is ∼3 cm/burn-year, with a corresponding hydrogenic codeposition rate >1kg/m2/burn-year; rates both problematic for steady-state fusion reactors. The carbon net deposition rate in the divertor is consistent with carbon arriving from the core plasma region. Carbon influx from the main wall is measured to be relatively large in the high-density detached regime and is of sufficient magnitude to account for the deposition rate in the divertor. Divertor redeposition is therefore determined by non-divertor erosion and transport. Despite the success in reducing divertor erosion on DIII-D with detachment, no significant reduction is found in the core plasma carbon density, illustrating the importance of non-divertor erosion and the complex coupling between erosion/redeposition and impurity plasma transport. (author)

  20. Preliminary concept design of the divertor remote handling system for DEMO power plant

    Energy Technology Data Exchange (ETDEWEB)

    Carfora, D., E-mail: dario.carfora@gmail.com [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); ENEA/CREATE/University of Naples Federico II, 80125 Naples (Italy); Di Gironimo, G. [ENEA/CREATE/University of Naples Federico II, 80125 Naples (Italy); Järvenpää, J. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Huhtala, K. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T.; Siuko, M. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland)

    2014-11-15

    Highlights: • Concept design of the RH system for the DEMO fusion power plant. • Divertor Mover: Hydraulic telescopic boom concept design. An alternative solution to ITER rack and pinion divertor mover (CMM). • Divertor cassettes end effector studies. • Transportation cask conceptual studies and logistic. - Abstract: This paper is based on the remote maintenance system project (WPRM) for the demonstration fusion power reactor (DEMO). Following ITER, DEMO aims to confirm the capability of generating several hundred of MW of net electricity by 2050. The main objective of these activities is to develop an efficient and reliable remote handling (RH) system for replacing the divertor cassettes. This paper presents the preliminary results of the concept design of the divertor RH system. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections of 4 m each, and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel. Two alternative design of the end effector to grip and manipulate the divertor cassette are also presented in this work. Both the concepts are hydraulically actuated, basing on the ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate. The main objective of this paper is to illustrate the feasibility of DEMO divertor remote maintenance operations.

  1. Preliminary concept design of the divertor remote handling system for DEMO power plant

    International Nuclear Information System (INIS)

    Highlights: • Concept design of the RH system for the DEMO fusion power plant. • Divertor Mover: Hydraulic telescopic boom concept design. An alternative solution to ITER rack and pinion divertor mover (CMM). • Divertor cassettes end effector studies. • Transportation cask conceptual studies and logistic. - Abstract: This paper is based on the remote maintenance system project (WPRM) for the demonstration fusion power reactor (DEMO). Following ITER, DEMO aims to confirm the capability of generating several hundred of MW of net electricity by 2050. The main objective of these activities is to develop an efficient and reliable remote handling (RH) system for replacing the divertor cassettes. This paper presents the preliminary results of the concept design of the divertor RH system. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections of 4 m each, and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel. Two alternative design of the end effector to grip and manipulate the divertor cassette are also presented in this work. Both the concepts are hydraulically actuated, basing on the ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate. The main objective of this paper is to illustrate the feasibility of DEMO divertor remote maintenance operations

  2. Diagnostic options for radiative divertor feedback control on NSTX-Ua)

    Science.gov (United States)

    Soukhanovskii, V. A.; Gerhardt, S. P.; Kaita, R.; McLean, A. G.; Raman, R.

    2012-10-01

    A radiative divertor technique is used in present tokamak experiments and planned for ITER to mitigate high heat loads on divertor plasma-facing components (PFCs) to prevent excessive material erosion and thermal damage. In NSTX, a large spherical tokamak with lithium-coated graphite PFCs and high divertor heat flux (qpeak ⩽ 15 MW/m2), radiative divertor experiments have demonstrated a significant reduction of divertor peak heat flux simultaneously with good core H-mode confinement using pre-programmed D2 or CD4 gas injections. In this work diagnostic options for a new real-time feedback control system for active radiative divertor detachment control in NSTX-U, where steady-state peak divertor heat fluxes are projected to reach 20-30 MW/m2, are discussed. Based on the NSTX divertor detachment measurements and analysis, the control diagnostic signals available for NSTX-U include divertor radiated power, neutral pressure, spectroscopic deuterium recombination signatures, infrared thermography of PFC surfaces, and thermoelectric scrape-off layer current. In addition, spectroscopic "security" monitoring of possible confinement or pedestal degradation is recommended. These signals would be implemented in a digital plasma control system to manage the divertor detachment process via an actuator (impurity gas seeding rate).

  3. Diagnostic options for radiative divertor feedback control on NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V. A.; McLean, A. G. [Lawrence Livermore National Laboratory, Livermore, California, 94550 (United States); Gerhardt, S. P.; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Raman, R. [University of Washington, Seattle, Washington 98195 (United States)

    2012-10-15

    A radiative divertor technique is used in present tokamak experiments and planned for ITER to mitigate high heat loads on divertor plasma-facing components (PFCs) to prevent excessive material erosion and thermal damage. In NSTX, a large spherical tokamak with lithium-coated graphite PFCs and high divertor heat flux (q{sub peak} Less-Than-Or-Slanted-Equal-To 15 MW/m{sup 2}), radiative divertor experiments have demonstrated a significant reduction of divertor peak heat flux simultaneously with good core H-mode confinement using pre-programmed D{sub 2} or CD{sub 4} gas injections. In this work diagnostic options for a new real-time feedback control system for active radiative divertor detachment control in NSTX-U, where steady-state peak divertor heat fluxes are projected to reach 20-30 MW/m{sup 2}, are discussed. Based on the NSTX divertor detachment measurements and analysis, the control diagnostic signals available for NSTX-U include divertor radiated power, neutral pressure, spectroscopic deuterium recombination signatures, infrared thermography of PFC surfaces, and thermoelectric scrape-off layer current. In addition, spectroscopic 'security' monitoring of possible confinement or pedestal degradation is recommended. These signals would be implemented in a digital plasma control system to manage the divertor detachment process via an actuator (impurity gas seeding rate).

  4. Diagnostic options for radiative divertor feedback control on NSTX-U

    International Nuclear Information System (INIS)

    A radiative divertor technique is used in present tokamak experiments and planned for ITER to mitigate high heat loads on divertor plasma-facing components (PFCs) to prevent excessive material erosion and thermal damage. In NSTX, a large spherical tokamak with lithium-coated graphite PFCs and high divertor heat flux (qpeak⩽ 15 MW/m2), radiative divertor experiments have demonstrated a significant reduction of divertor peak heat flux simultaneously with good core H-mode confinement using pre-programmed D2 or CD4 gas injections. In this work diagnostic options for a new real-time feedback control system for active radiative divertor detachment control in NSTX-U, where steady-state peak divertor heat fluxes are projected to reach 20–30 MW/m2, are discussed. Based on the NSTX divertor detachment measurements and analysis, the control diagnostic signals available for NSTX-U include divertor radiated power, neutral pressure, spectroscopic deuterium recombination signatures, infrared thermography of PFC surfaces, and thermoelectric scrape-off layer current. In addition, spectroscopic “security” monitoring of possible confinement or pedestal degradation is recommended. These signals would be implemented in a digital plasma control system to manage the divertor detachment process via an actuator (impurity gas seeding rate).

  5. Latest results and future plans from JT-60U and JFT-2M divertor research

    International Nuclear Information System (INIS)

    The JT-60U divertor research program is designed to address key issues for ITER physics R and D. Radiative cooling divertor is investigated to reduce the heat load on the divertor plates in L- and ELMy H-mode discharges. The mechanism of radiation losses is spectroscopically studied with visible spectroscopy. The peak heat flux density reaches 300-400 MW/m2 and most of the power is deposited within a few milliseconds due to ELMs. In high density plasmas, the physically sputtered impurities from target plates are reduced and chemical sputtering by neutral particles which strike the divertor plates in the private region becomes dominant. Helium exhaust from the core plasma is observed due to wall pumping caused by Solid Target Boronization. A new W-shaped pumped divertor in JT-60U providing a dense and cold divertor plasma is the basis to develop the integrated performance in quasi-steady state. Divertor configuration was modified an open divertor to a closed divertor and initial results on the closed divertor has been obtained in JFT-2M. (author)

  6. Preliminary report: NIF laser bundle review

    International Nuclear Information System (INIS)

    As requested in the guidance memo 1, this committe determined whether there are compelling reasons to recommend a change from the NIF CDR baseline laser. The baseline bundle design based on a tradeoff between cost and technical risk, which is replicated four times to create the required 192 beams. The baseline amplifier design uses bottom loading 1x4 slab and flashlamp cassettes for amplifier maintenance and large vacuum enclosures (2.5m high x 7m wide in cross-section for each of the two spatial filters in each of the four bundles. The laser beams are arranged in two laser bays configured in a u-shape around the target area. The entire bundle review effort was performed in a very short time (six weeks) and with limited resources (15 personnel part-time). This should be compared to the effort that produced the CDR design (12 months, 50 to 100 personnel). This committee considered three alternate bundle configurations (2x2, 4x2, and 4x4 bundles), and evaluated each bundle against the baseline design using the seven requested issues in the guidance memo: Cost; schedule; performance risk; maintainability/operability; hardware failure cost exposure; activation; and design flexibility. The issues were reviewed to identify differences between each alternate bundle configuration and the baseline

  7. Prioritary omalous bundles on Hirzebruch surfaces

    Science.gov (United States)

    Aprodu, Marian; Marchitan, Marius

    2016-01-01

    An irreducible algebraic stack is called unirational if there exists a surjective morphism, representable by algebraic spaces, from a rational variety to an open substack. We prove unirationality of the stack of prioritary omalous bundles on Hirzebruch surfaces, which implies also the unirationality of the moduli space of omalous H-stable bundles for any ample line bundle H on a Hirzebruch surface (compare with Costa and Miro-Ŕoig, 2002). To this end, we find an explicit description of the duals of omalous rank-two bundles with a vanishing condition in terms of monads. Since these bundles are prioritary, we conclude that the stack of prioritary omalous bundles on a Hirzebruch surface different from P1 ×P1 is dominated by an irreducible section of a Segre variety, and this linear section is rational (Ionescu, 2015). In the case of the space quadric, the stack has been explicitly described by N. Buchdahl. As a main tool we use Buchdahl's Beilinson-type spectral sequence. Monad descriptions of omalous bundles on hypersurfaces in P4, Calabi-Yau complete intersection, blowups of the projective plane and Segre varieties have been recently obtained by A.A. Henni and M. Jardim (Henni and Jardim, 2013), and monads on Hirzebruch surfaces have been applied in a different context in Bartocci et al. (2015).

  8. Singular hermitian metrics on vector bundles

    CERN Document Server

    De Cataldo, M A A

    1997-01-01

    We introduce a notion of singular hermitian metrics (s.h.m.) for holomorphic vector bundles and define positivity in view of $L^2$-estimates. Associated with a suitably positive s.h.m. there is a (coherent) sheaf 0-th kernel of a certain $d''$-complex. We prove a vanishing theorem for the cohomology of this sheaf. All this generalizes to the case of higher rank known results of Nadel for the case of line bundles. We introduce a new semi-positivity notion, $t$-nefness, for vector bundles, establish some of its basic properties and prove that on curves it coincides with ordinary nefness. We particularize the results on s.h.m. to the case of vector bundles of the form $E=F \\otimes L$, where $F$ is a $t$-nef vector bundle and $L$ is a positive (in the sense of currents) line bundle. As applications we generalize to the higher rank case 1) Kawamata-Viehweg Vanishing Theorem, 2) the effective results concerning the global generation of jets for the adjoint to powers of ample line bundles, and 3) Matsusaka Big Theor...

  9. Deformations of the generalised Picard bundle

    International Nuclear Information System (INIS)

    Let X be a nonsingular algebraic curve of genus g ≥ 3, and let Mξ denote the moduli space of stable vector bundles of rank n ≥ 2 and degree d with fixed determinant ξ over X such that n and d are coprime. We assume that if g = 3 then n ≥ 4 and if g = 4 then n ≥ 3, and suppose further that n0, d0 are integers such that n0 ≥ 1 and nd0 + n0d > nn0(2g - 2). Let E be a semistable vector bundle over X of rank n0 and degree d0. The generalised Picard bundle Wξ(E) is by definition the vector bundle over Mξ defined by the direct image pMξ *(Uξ x pX*E) where Uξ is a universal vector bundle over X x Mξ. We obtain an inversion formula allowing us to recover E from Wξ(E) and show that the space of infinitesimal deformations of Wξ(E) is isomorphic to H1(X, End(E)). This construction gives a locally complete family of vector bundles over Mξ parametrised by the moduli space M(n0,d0) of stable bundles of rank n0 and degree d0 over X. If (n0,d0) = 1 and Wξ(E) is stable for all E is an element of M(n0,d0), the construction determines an isomorphism from M(n0,d0) to a connected component M0 of a moduli space of stable sheaves over Mξ. This applies in particular when n0 = 1, in which case M0 is isomorphic to the Jacobian J of X as a polarised variety. The paper as a whole is a generalisation of results of Kempf and Mukai on Picard bundles over J, and is also related to a paper of Tyurin on the geometry of moduli of vector bundles. (author)

  10. Modeling of divertor geometry effects in China fusion engineering testing reactor by SOLPS/B2-Eirene

    International Nuclear Information System (INIS)

    The China Fusion Engineering Testing Reactor (CFETR) is currently under design. The SOLPS/B2-Eirene code package is utilized for the design and optimization of the divertor geometry for CFETR. Detailed modeling is carried out for an ITER-like divertor configuration and one with relatively open inner divertor structure, to assess, in particular, peak power loading on the divertor target, which is a key issue for the operation of a next-step fusion machine, such as ITER and CFETR. As expected, the divertor peak heat flux greatly exceeds the maximum steady-state heat load of 10 MW/m2, which is a limit dictated by engineering, for both divertor configurations with a wide range of edge plasma conditions. Ar puffing is effective at reducing divertor peak heat fluxes below 10 MW/m2 even at relatively low densities for both cases, favoring the divertor configuration with more open inner divertor structure

  11. Geometry of quantum principal bundles, 1

    CERN Document Server

    Durdevic, M

    1995-01-01

    A theory of principal bundles possessing quantum structure groups and classical base manifolds is presented. Structural analysis of such quantum principal bundles is performed. A differential calculus is constructed, combining differential forms on the base manifold with an appropriate differential calculus on the structure quantum group. Relations between the calculus on the group and the calculus on the bundle are investigated. A concept of (pseudo)tensoriality is formulated. The formalism of connections is developed. In particular, operators of horizontal projection, covariant derivative and curvature are constructed and analyzed. Generalizations of the first structure equation and of the Bianchi identity are found. Illustrative examples are presented.

  12. Weak equivalence classes of complex vector bundles

    OpenAIRE

    Hông-Vân Lê

    2006-01-01

    For any complex vector bundle Ek of rank k over a manifold Mm with Chern classes ci Î H2i(Mm, Z) and any non-negative integers l1, . . ., lk we show the existence of a positive number p(m, k) and the existence of a complex vector bundle Êk over Mm whose Chern classes are p(m, k) × li × ci Î H2i(Mm, Z). We also discuss a version of this statement for holomorphic vector bundles over projective algebraic manifolds.

  13. Assembly mechanism for nuclear fuel bundles

    International Nuclear Information System (INIS)

    A description is given of a nuclear power reactor fuel bundle having tie rods fastened to a lower tie plate and passing through openings in the upper tie plate with the assembled bundle secured by rotatable locking sleeves which engage slots provided in the upper tie plate. Pressure exerted by helical springs mounted around each of the fuel rods urge the upper tie plate against the locking sleeves. The bundle may be disassembled after depressing the upper tie plate and rotating the locking sleeves to the unlocked position

  14. Vector supersymmetry in the universal bundle

    International Nuclear Information System (INIS)

    We present a vector supersymmetry for Witten-type topological gauge theories, and examine its algebra in terms of a superconnection formalism. When covariant constraints on the supercurvature are chosen, a correspondence is established with the universal bundle construction of Atiyah and Singer. The vector supersymmetry represents a certain shift operator in the curvature of the universal bundle, and can be used to generate the hierarchy of observables in these theories. This formalism should lead to the construction of vector supergravity theories, and perhaps to the gravitational analogue of the universal bundle. (orig.)

  15. Bundle duct interaction studies for fuel assemblies

    International Nuclear Information System (INIS)

    It is known that the wire-wrapped rods and duct in an LMFBR are undergoing a gradual structural distortion from the initially uniform geometry under the combined effects of thermal expansion and irradiation induced swelling and creep. These deformations have a significant effect on flow characteristics, thus causing changes in thermal behavior such as cladding temperature and temperature distribution within a bundle. The temperature distribution may further enhance or retard irradiation induced deformation of the bundle. This report summarizes the results of the continuing effort in investigating the bundle-duct interaction, focusing on the need for the large development plant

  16. An automated approach to magnetic divertor configuration design

    Science.gov (United States)

    Blommaert, M.; Dekeyser, W.; Baelmans, M.; Gauger, N. R.; Reiter, D.

    2015-01-01

    Automated methods based on optimization can greatly assist computational engineering design in many areas. In this paper an optimization approach to the magnetic design of a nuclear fusion reactor divertor is proposed and applied to a tokamak edge magnetic configuration in a first feasibility study. The approach is based on reduced models for magnetic field and plasma edge, which are integrated with a grid generator into one sensitivity code. The design objective chosen here for demonstrative purposes is to spread the divertor target heat load as much as possible over the entire target area. Constraints on the separatrix position are introduced to eliminate physically irrelevant magnetic field configurations during the optimization cycle. A gradient projection method is used to ensure stable cost function evaluations during optimization. The concept is applied to a configuration with typical Joint European Torus (JET) parameters and it automatically provides plausible configurations with reduced heat load.

  17. Alignment systems for pumped divertor installation at JET

    International Nuclear Information System (INIS)

    The installation of the JET Pumped Divertor, designed to study impurity control, has recently been completed. The main components are four magnetic coils, forty eight divertor plate assemblies, one toroidal cryopump, eight ICRH antennae, sixteen inner wall guard limiters and twelve poloidal limiters. Due to the high thermal loads, accurate positioning of plasma facing components to the magnetic centre of the machine was a major requirement. Typically alignment within ± 2 mm was required, with steps between tiles on a component being controlled to ± 0.25 mm. In some cases a set of components was required to be concentric, while also lying within a narrow band defined by the position of some other components. A typical example of this was the positioning of the poloidal limiters, which perform the dual function of limiting the plasma and also protecting the antennae. Clearly, a measuring system accurate to better than ± 0.5 mm was required. (author) 4 refs.; 3 figs

  18. Alignment systems for pumped divertor installation at JET

    International Nuclear Information System (INIS)

    The installation of the JET Pumped Divertor, designed to study impurity control, has recently been completed. The main components are four magnetic coils, forty eight divertor plate assemblies, one toroidal cryopump, eight ICRH antennae, sixteen inner wall guard limiters and twelve poloidal limiters. Due to the high thermal loads, accurate positioning of plasma facing components to the magnetic centre of the machine was a major requirement. Typically alignment within ± 2 mm was required, with steps between tiles on a component being controlled to ± 0.25 mm. In some cases a set of components was required to be concentric, while also lying within a narrow band defined by the position of some other components. A typical example of this was the positioning of the poloidal limiters, which perform the dual function of limiting the plasma and also protecting the antennae. Clearly, a measuring system accurate to better than ± 0.5 mm was required. (orig.)

  19. Measurements of divertor impurity concentrations on DIII-D

    International Nuclear Information System (INIS)

    Carbon emissions in the DIII-D divertor during partial detachment have been measured, and the deduced radiated power and the temporal behavior of the impurity emissions from spectroscopy are in good agreement with bolometer measurements. Effective electron temperatures from line ratios for CIV (9-11 eV) and CIII (6-8 eV) are correlated with DTS measured electron temperatures to determine the spatial location of the carbon radiation zone. During PDD operation, the bulk of the divertor radiation is emitted from CIV near the X- point while deuterium radiation is strongest near the outer strikepoint. The carbon ion concentrations are in the range of 1% - 4% of the electron density

  20. Divertor detachment, He exhaust and compact toroid injection on TdeV

    International Nuclear Information System (INIS)

    Progressive detachment with increasing density is shown to proceed with a marked reduction of the ion flux to the divertor plates, a pressure gradient between a ionization front and the plate, and strong cross-field transport in the divertor. The divertor He exhaust is not affected by detachment although the He enrichment remains low but constant. A moderate density of n-bare ∼ 5 x 1019 m-3 seems to be sufficient both for efficient peak power load reduction at the plate and good He exhaust through the divertor. Simulations indicate possible divertor geometry improvements which will soon be verified experimentally in the new TdeV-96 divertor upgrade. Finally, central fuelling with compact toroid injection is reported with no detrimental effects on the plasma. (author). 16 refs, 8 figs

  1. Results of the H-mode experiments with JT-60 outer and lower divertors

    International Nuclear Information System (INIS)

    In JT-60, hydrogen H-mode experiments with outer and lower divertors were performed. In the outer divertor, H-mode were obtained, similar to the ones observed in the other lower/upper divertors. Its threshold absorbed power and electron density were 16 MW and 1.8 x 1019m-3. In the two combined heatings with NB+ICRF and NB+LHRF, H-mode discharges are also obtained. Moreover, in new configuration of lower divertor, H-mode phases without and with ELM are obtained. Typical results of the lower divertor are shown to compare the H-mode characteristics between the two configurations. Improvement of the energy confinement time in the two divertors was limited to 10 %. Analyses on ballooning/interchange instabilities were carried out with precise equlibria of JT-60. These results showed that the both modes were enough stable. (author)

  2. The control of convection by fuelling and pumping in the JET pumped divertor

    Energy Technology Data Exchange (ETDEWEB)

    Harbour, P.J.; Andrew, P.; Campbell, D.; Clement, S.; Davies, S.; Ehrenberg, J.; Erents, S.K.; Gondhalekar, A.; Gadeberg, M.; Gottardi, N.; Von Hellermann, M.; Horton, L.; Loarte, A.; Lowry, C.; Maggi, C.; McCormick, K.; O`Brien, D.; Reichle, R.; Saibene, G.; Simonini, R.; Spence, J.; Stamp, M.; Stork, D.; Taroni, A.; Vlases, G. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

    1994-07-01

    Convection from the scrape-off layer (SOL) to the divertor will control core impurities, if it retains them in a cold, dense, divertor plasma. This implies a high impurity concentration in the divertor, low at its entrance. Particle flux into the divertor entrance can be varied systematically in JET, using the new fuelling and pumping systems. The convection ratio has been estimated for various conditions of operation. Particle convection into the divertor should increase thermal convection, decreasing thermal conduction, and temperature and density gradients along the magnetic field, hence increasing the frictional force and decreasing the thermal force on impurities. Changes in convection in the SOL, caused by gaseous fuelling, have been studied, both experimentally in the JET Mk I divertor and with EDGE2/NIMBUS. 1 ref., 4 figs., 1 tab.

  3. Active control of divertor heat and particle fluxes in EAST towards advanced steady state operations

    International Nuclear Information System (INIS)

    Significant progress has been made in EAST towards advanced steady state operations by active control of divertor heat and particle fluxes. Many innovative techniques have been developed to mitigate transient ELM and stationary heat fluxes on the divertor target plates. It has been found that lower hybrid current drive (LHCD) can lead to edge plasma ergodization, striation of the stationary heat flux and lower ELM transient heat and particle fluxes. With multi-pulse supersonic molecular beam injection (SMBI) to quantitatively regulate the divertor particle flux, the divertor power footprint pattern can be actively modified. H-modes have been extended over 30 s in EAST with the divertor peak heat flux and the target temperature being controlled well below 2 MW/m2 and 250 °C, respectively, by integrating these new methods, coupled with advanced lithium wall conditioning and internal divertor pumping, along with an edge coherent mode to provide continuous particle and power exhaust

  4. Divertor heat and particle control experiments on the DIII-D tokamak

    International Nuclear Information System (INIS)

    In this paper we present a summary of recent DIII-D divertor physics activity and plans for future divertor upgrades. During the past year, DIII-D experimental effort was focused on areas of active heat and particle control and divertor target erosion studies. Using the DIII-D Advanced Divertor system we have succeeded for the first time to control the plasma density and demonstrate helium exhaust in H-mode plasmas. Divertor heat flux control by means of D2 gas puffing and impurity injection were studied separately and in, both cases up to a factor of five reduction of the divertor peak heat flux was observed. Using the DiMES sample transfer system we have obtained erosion data on various material samples in well diagnosed plasmas and compared the results with predictions of numerical models

  5. Active control of divertor heat and particle fluxes in EAST towards advanced steady state operations

    Energy Technology Data Exchange (ETDEWEB)

    Wang, L., E-mail: lwang@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Dalian University of Technology, Dalian 116024 (China); Guo, H.Y. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); General Atomics, P. O. Box 85608, San Diego, CA 92186 (United States); Li, J.; Wan, B.N.; Gong, X.Z.; Zhang, X.D.; Hu, J.S. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Liang, Y. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Association EURATOM-FZJ, D-52425 Jülich (Germany); Xu, G.S. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Zou, X.L. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Loarte, A. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Maingi, R.; Menard, J.E. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Luo, G.N.; Gao, X.; Hu, L.Q.; Gan, K.F.; Liu, S.C.; Wang, H.Q.; Chen, R. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); and others

    2015-08-15

    Significant progress has been made in EAST towards advanced steady state operations by active control of divertor heat and particle fluxes. Many innovative techniques have been developed to mitigate transient ELM and stationary heat fluxes on the divertor target plates. It has been found that lower hybrid current drive (LHCD) can lead to edge plasma ergodization, striation of the stationary heat flux and lower ELM transient heat and particle fluxes. With multi-pulse supersonic molecular beam injection (SMBI) to quantitatively regulate the divertor particle flux, the divertor power footprint pattern can be actively modified. H-modes have been extended over 30 s in EAST with the divertor peak heat flux and the target temperature being controlled well below 2 MW/m{sup 2} and 250 °C, respectively, by integrating these new methods, coupled with advanced lithium wall conditioning and internal divertor pumping, along with an edge coherent mode to provide continuous particle and power exhaust.

  6. Effect of localized gas puffing on divertor plasma behavior in EAST

    International Nuclear Information System (INIS)

    Localized D2 puffing from various divertor locations has been carried out under double null (DN) and lower single null (LSN) divertor configurations to investigate the effect of gas puff locations on the divertor behavior in ohmic L-mode discharges in EAST. Localized gas puffing from the dome has a higher fueling efficiency than that from the inner and outer targets for both DN and LSN configurations. Under the DN configuration, gas puffing from the inner target exhibits a much better fueling efficiency than that from the outer target. In contrast, the gas fueling efficiency shows little difference between the inner and outer divertor gas puff locations in the LSN configuration. In LSN, localized gas puffing from the outer divertor target tends to promote detachment at the outer target. This will be employed as a means to control heat fluxes to the outer divertor target plates for high power long pulse operations.

  7. Multiscale study on hydrogen mobility in metallic fusion divertor material

    OpenAIRE

    Heinola, Kalle

    2010-01-01

    For achieving efficient fusion energy production, the plasma-facing wall materials of the fusion reactor should ensure long time operation. In the next step fusion device, ITER, the first wall region facing the highest heat and particle load, i.e. the divertor area, will mainly consist of tiles based on tungsten. During the reactor operation, the tungsten material is slowly but inevitably saturated with tritium. Tritium is the relatively short-lived hydrogen isotope used in the fusion reactio...

  8. An automated approach to magnetic divertor configuration design

    OpenAIRE

    Blommaert, Maarten; Dekeyser, Wouter; Baelmans, Martine; Gauger, Nicolas Ralph; Reiter, Detlev

    2015-01-01

    Automated methods based on optimization can greatly assist computational engineering design in many areas. In this paper an optimization approach to the magnetic design of a nuclear fusion reactor divertor is proposed and applied to a tokamak edge magnetic configuration in a first feasibility study. The approach is based on reduced models for magnetic field and plasma edge, which are integrated with a grid generator into one sensitivity code. The design objective chosen here for demonstrat...

  9. Method of plasma impurity control without magnetic divertor

    International Nuclear Information System (INIS)

    A method is proposed for controlling impurity generation in a tokomak by skimming and pumping the scrape-off. This method avoids many of the complications of a magnetic divertor, such as specially configured magnetic fields, toroidal symmetry, and inefficient use of toroidal field volume. Estimates are given for operating parameters. Impurity reductions of as much as a factor of 10 should be achievable. The necessary high-capacity pump would employ either titanium gettering or cryocondensation

  10. Modeling of a poloidally symmetric toroidal field divertor in a reversed--field-pinch plasma machine

    International Nuclear Information System (INIS)

    Magnetic divertors have been shown to be successful in minimizing plasma-wall interactions and in leading to high confinement regimes in Tokamaks. This leads to the hope that similar benefits may occur in an Reversed-Field-Pinch (RPF) fitted with a divertor. Previous experiments using divertors in a RFP have used a poloidal field divertor configuration such as is used in Tokamaks. This study investigates another approach; namely a toroidal field divertor. In this study a simple model of a poloidally symmetric toroidal field divertor is developed and used in a study of stochastic effects due to the divertor and in a 3-D magnetohydrodynamic (MHD) code to study the response of the plasma to the large poloidal m = 0 perturbations caused by the divertor coils. It is found that the topology of the RFP-divertor system is much more complex than had been expected. Stochasticity is enhanced in the outer edge region of the plasma because of this geometrical complexity. The way of the RFP reaches an equilibrium in this complex system is investigated with the 3-D relaxation code, DEBS (authored by Dalton Schnack). This code showed that the divertor will not hinder the formation of a reversed toroidal field in the plasma, and that the dynamics of its formation is altered when toroidal effects are considered. The plasma develops flows and currents in the throat of the divertor in response to the vacuum-like divertor fields. These flows and currents help to restore the force free character of the plasma

  11. Effect of nozzle sizes on jet impingement heat transfer in He-cooled divertor

    OpenAIRE

    Končar, Boštjan; Norajitra, Prachai; Oblak, Klemen

    2009-01-01

    Abstract The use of impinging jets for divertor cooling in the conceptual fusion power plant is attracting much attention due to its very high heat removal capability and moderate pumping power requirement. The latest and the most advanced divertor concept is based on modular design cooled by helium impinging jets. To reduce the thermal stresses, the plasma-facing side of the divertor is build up of numerous small cooling fingers cooled by an array of helium jets. In this study the...

  12. Particle recirculation in the ergodic divertor of Tore Supra

    International Nuclear Information System (INIS)

    The present paper addresses the issue of particle recirculation in discharges where low energy flux to ergodic divertor target plates is achieved, in highly radiating detached ohmic plasmas. Plasma temperature and particle flux are measured by flush-mounted probes in the divertor plates, and by an upstream fast scanning Mach probe. The scalings with core density of the ion flux and electron temperature are well described by the simple two-point model used in axisymmetric poloidal divertors. The detachment signature is a pressure drop that occurs when the edge temperature falls below 10 eV. The parallel ion flux gradient is always positive, indicating that recombination is unlikely to play an important role in detachment. Visible spectroscopy of a neutralizer plate shows that attainment of cold detached plasmas near the density limit coincides with an abrupt increase of fueling for both deuterium and impurities. A feedback algorithm based on real time Langmuir probe measurements has been developed to monitor detachment and avoid disruptions. (authors)

  13. Particle recirculation in the ergodic divertor of Tore Supra

    International Nuclear Information System (INIS)

    The present paper addresses the issue of particle recirculation in discharges where low-energy flux to ergodic divertor target plates is achieved in highly-radiating detached ohmic plasmas. Plasma temperature and particle flux are measured by flush-mounted probes in the divertor plates and by an upstream fast scanning Mach probe. The scalings with core density of the ion flux and electron temperature are well described by the simple two-point model used in axisymmetric poloidal divertors. The detachment signature is a pressure drop that occurs when the edge temperature falls below 10 eV. The parallel ion flux gradient is always positive, indicating that recombination is unlikely to play an important role in detachment. Visible spectroscopy of a neutralizer plate shows that attainment of cold detached plasmas near the density limit coincides with an abrupt increase of fuelling efficiency for both deuterium and impurities. A feedback algorithm based on real-time Langmuir probe measurements has been developed to monitor detachment and avoid disruptions. (author)

  14. Radiative divertor plasmas with convection in DIII-D

    International Nuclear Information System (INIS)

    The radiation of divertor heat flux on DIII-D [J. Luxon et al., in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion (International Atomic Energy Agency, Vienna, 1987), p. 159] is shown to greatly exceed the limits imposed by assumptions of energy transport dominated by electron thermal conduction parallel to the magnetic field. Approximately 90% of the power flowing into the divertor is dissipated through low-Z radiation and plasma recombination. The dissipation is made possible by an extended region of low electron temperature in the divertor. A one-dimensional analysis of the parallel heat flux finds that the electron temperature profile is incompatible with conduction-dominated parallel transport. Plasma flow at up to the ion acoustic speed, produced by upstream ionization, can account for the parallel heat flux. Modeling with the two-dimensional fluid code UEDGE [T. Rognlien, J. L. Milovich, M. E. Rensink, and G. D. Porter, J. Nucl. Mater. 196 endash 198, 347 (1992)] has reproduced many of the observed experimental features. copyright 1998 American Institute of Physics

  15. Analytical calculations for impurity seeded partially detached divertor conditions

    Science.gov (United States)

    Kallenbach, A.; Bernert, M.; Dux, R.; Reimold, F.; Wischmeier, M.; ASDEX Upgrade Team

    2016-04-01

    A simple analytical formula for the impurity seeded partially detached divertor operational point has been developed using 1D modelling. The inclusion of charge exchange momentum loss terms improves the 1D modelling for ASDEX Upgrade conditions and its extrapolation to larger devices. The investigations are concentrated around a partially detached divertor working point of low heat flux and an electron temperature around 2.5 eV at the target which are required to maintain low sputtering rates at a tungsten target plate. An experimental formula for the onset of detachment by nitrogen seeding in ASDEX Upgrade is well reproduced, and predictions are given for N, Ne and Ar seeding for variable device size. Moderate deviations from a linear {{P}\\text{sep}}/R size dependence of the detachment threshold are seen in the modelling caused by upstream radiation at longer field line lengths. The presented formula allows the prediction of the neutral gas or seed impurity pressure which is required to achieve partial detachment for a given {{P}\\text{sep}} in devices with a closed divertor similar to the geometry in ASDEX Upgrade.

  16. Parametric study of FER first wall and divertor plate performance

    International Nuclear Information System (INIS)

    Thermal, mechanical, and lifetime performance of various first wall and divertor plate materials were examined over a broad range of conditions, representative of those considered for next-generation tokamaks such as FER. Candidate plasma side materials include beryllium, graphite, silicon carbide, molybdenum, tantalum, and tungsten. Copper, copper alloy C17510, austenitic stainless steel (316SS), ferritic stainless steel (HT-9), vanadium alloy V-15Cr-5Ti, and molybdenum alloy TZM were considered as candidate heat sink/structural materials. Performance was examined at heat fluxes ranging from 0.05 MW/m2 for the first wall up to 5.0 MW/m2 for the divertor plate. Ion flux, plasma edge temperature, burn time per pulse, and number of operating cycles were the other major parameters varied in this study. The analysis model used for these studies includes: (1) a thermal model; (2) a thermal stress model; (3) a disruption erosion model; (4) a sputtering erosion model; and (5) a fatique lifetime model. Results show that recommended first wall and divertor plate designs perform adequately over most of the range of conditions considered for FER design options. Thermal shock of the plasma facing material during intense disruption heating and radiation damage and temperature limitations for graphite are identified as major concerns reguiring experimental investigation. (author)

  17. Plasma parameters in the COMPASS divertor during Ohmic plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Dimitrova, M. [Institute of Plasma Physics, Academy of Sciences of the Czech Republic v.v.i., Prague (Czech Republic); Emil Djakov Institute of Electronics, Bulgarian Academy of Sciences, Sofia (Bulgaria); Dejarnac, R.; Stoeckel, J.; Havlicek, J.; Janky, F.; Panek, R. [Institute of Plasma Physics, Academy of Sciences of the Czech Republic v.v.i., Prague (Czech Republic); Popov, Ts.K. [Faculty of Physics, St. Kl. Ohridski University of Sofia (Bulgaria); Ivanova, P.; Vasileva, E. [Emil Djakov Institute of Electronics, Bulgarian Academy of Sciences, Sofia (Bulgaria); Kovacic, J. [Jozef Stefan Institute, Ljubljana (Slovenia)

    2014-04-15

    This paper reports on probe measurements of the electron energy distribution function and plasma potential in the divertor region of the COMPASS tokamak during D-shaped plasmas. The probe data have been processed using the novel first-derivative technique. A comparison with the results obtained by processing the same data with the classical probe technique, which assumes Maxwellian electron energy distribution functions is presented and discussed. In the vicinity of the inner and outer strike points of the divertor the electron energy distribution function can be approximated by a bi-Maxwellian, with a dominating low-energy electron population (4-7 eV) and a minority of higher energy electrons (12-25 eV). In the private flux region between the two strike points the electron energy distribution function is found to be Maxwellian with temperatures in the range of 7-10 eV. The comparative analysis using both techniques has allowed a better insight into the underlying physical processes at the divertor region of the COMPASS tokamak. (copyright 2014 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  18. Non-linear effects on neutral gas transport in divertors

    International Nuclear Information System (INIS)

    The effects of neutral particles on the condition of the plasma edge play a key role in divertor and limiter physics. In computational models they are usually treated in the linear test particle approximation. However, in some divertor concepts a large neutral gas pressure is required in the divertor chamber to provide sufficient neutral-plasma interaction in the plasma fan (momentum removal and energy dissipation) and to permit adequate pumping performance. In such regimes viscous effects in the neutral gas may become relevant. We have extended the EIRENE code to solve the Boltzmann equation with a non-linear BGK-model collision term added to its standard linear collision integrals. The linear in-elastic collision integrals are reconsidered with respect to volume recombination and momentum removal efficiency from the plasma. The numerical procedure in the EIRENE Monte Carlo code is outlined. A simple test application (Couette flow) shows that the procedure works properly. First numerical studies have been carried out and the results are discussed. (orig.)

  19. Activation analysis of coolant water in ITER blanket and divertor

    International Nuclear Information System (INIS)

    Coolant water in blankets and divertor cassettes will be activated by neutrons during ITER operation. 16N and 17N are determined to be the most important activation products in the coolant water in terms of their impact on ITER design and performance. In this study, the geometry of cooling channels in blanket module 4 was described precisely in the ITER neutronics model ‘Alite-4’ based on the latest CAD model converted using MCAM developed by FDS Team. The 16N and 17N concentration distribution in the blanket, divertor cassette and their primary heat transport systems were calculated by MCNP with data library FENDL2.1. The activation of cooling pipes induced 17N decay neutrons was analyzed and compared with that induced by fusion neutrons, using FISPACT-2007 with data library EAF-2007. The outlet concentration of blanket and divertor cooling systems was 1.37 × 1010 nuclide/cm3 and 1.05 × 1010 nuclide/cm3 of 16N, 8.93 × 106 nuclide/cm3 and 0.33 × 105 nuclide/cm3 of 17N. The decay gamma-rays from 16N in activated water could be a problem for cryogenic equipments inside the cryostat. Near the cryostat, the activation of pipes from 17N decay neutrons was much lower than that from fusion neutrons.

  20. Effect of Testing Conditions on Fibre-Bundle Tensile Properties Part Ⅰ: Sample Preparation, Bundle Mass and Fibre Alignment of Wool Bundles

    Institute of Scientific and Technical Information of China (English)

    YU Wei-dong; YAN Hao-jing; Ron Postle; Yang Shouren

    2002-01-01

    Due to the effects of samples and testing conditions on fibre-bundle tensile behaviour, it is necessary to investigate the relationships between experimental factors and tensile properties for the fibre-bumdle tensile tester (TENSOR). The effects of bundle sample preparation, fibre bundle mass and fibre alignment have been tested. The experimental results indicated that (1) the low damage in combing and no free-end fibres in the cut bundle are most important for the sample preparation; (2) the reasonable bundle mass is 400- 700tex, but the tensile properties measured should bemodified with the bundle mass because a small amount of bundle mass causes the scatter results, while the larger is the bundle mass, the more difficult to comb fibres parallel and to clamp fibre evenly; and (3) the fibre irregular arrangement forms a slack bundle resulting in interaction between fibres, which will affect the reproducibility and accuracy of the tensile testing.

  1. Self-mapping degrees of torus bundles and torus semi-bundles

    OpenAIRE

    Sun, Hongbin; Wang, Shicheng; Wu, Jianchun

    2010-01-01

    Each closed oriented 3-manifold $M$ is naturally associated with a set of integers $D(M)$, the degrees of all self-maps on $M$. $D(M)$ is determined for each torus bundle and torus semi-bundle $M$. The structure of torus semi-bundle is studied in detail. The paper is a part of a project to determine $D(M)$ for all 3-manifolds in Thurston's picture.

  2. In-pool damaged fuel bundle recovery

    International Nuclear Information System (INIS)

    While preparing to rerack the Oyster Creek Nuclear Generating Station, GPU Nuclear had need to move a damaged fuel bundle. This bundle had no upper tie plate and could not be moved in the normal manner. GPU Nuclear formed a small, dedicated project team to disassemble, package, and move this damaged bundle. The team was composed of key personnel from GPU Nuclear Fuels Projects, OCNGS Operations and Proto-Power/Bisco, a specialty contractor who has fuel bundle reconstitution and rod consolidation experience, remote tooling, underwater video systems and experienced technicians. Proven tooling, clear procedures and a simple approach were important, but the key element was the spirit of teamwork and leadership exhibited by the people involved. In spite of several emergent problems which a task of this nature presents, this small, close knit utility/vendor team completed the work on schedule and within the exposure and cost budgets

  3. In-pool damaged fuel bundle recovery

    International Nuclear Information System (INIS)

    While preparing to rerack the Oyster Creek Nuclear Generating Station, GPU Nuclear had need to move a damaged fuel bundle. This bundle had no upper tie plate and could not be moved in the normal manner. GPU Nuclear formed a small, dedicated project team to disassemble, package and move this damaged bundle. The team was composed of key personnel from GPU Nuclear Fuels Projects, OCNGS Operations and Proto-Power / Bisco, a specialty contractor who has fuel bundle reconstitution and rod consolidation experience, remote tooling, underwater video systems and experienced technicians. Proven tooling, clear procedures and a simple approach were important, but the key element was the spirit of teamwork and leadership exhibited by the people involved

  4. Nuclear fuel bundle disassembly and assembly tool

    International Nuclear Information System (INIS)

    A nuclear power reactor fuel bundle is described which has a plurality of tubular fuel rods disposed in parallel array between two transverse tie plates. It is secured against disassembly by one or more locking forks which engage slots in tie rods which position the transverse plates. Springs mounted on the fuel and tie rods are compressed when the bundle is assembled thereby maintaining a continual pressure against the locking forks. Force applied in opposition to the springs permits withdrawal of the locking forks so that one tie plate may be removed, giving access to the fuel rods. An assembly and disassembly tool facilitates removal of the locking forks when the bundle is to be disassembled and the placing of the forks during assembly of the bundle. (U.S.)

  5. Quantum Bundle Description of Quantum Projective Spaces

    Science.gov (United States)

    Ó Buachalla, Réamonn

    2012-12-01

    We realise Heckenberger and Kolb's canonical calculus on quantum projective ( N - 1)-space C q [ C p N-1] as the restriction of a distinguished quotient of the standard bicovariant calculus for the quantum special unitary group C q [ SU N ]. We introduce a calculus on the quantum sphere C q [ S 2 N-1] in the same way. With respect to these choices of calculi, we present C q [ C p N-1] as the base space of two different quantum principal bundles, one with total space C q [ SU N ], and the other with total space C q [ S 2 N-1]. We go on to give C q [ C p N-1] the structure of a quantum framed manifold. More specifically, we describe the module of one-forms of Heckenberger and Kolb's calculus as an associated vector bundle to the principal bundle with total space C q [ SU N ]. Finally, we construct strong connections for both bundles.

  6. Noncommutative principal bundles through twist deformation

    CERN Document Server

    Aschieri, Paolo; Pagani, Chiara; Schenkel, Alexander

    2016-01-01

    We construct noncommutative principal bundles deforming principal bundles with a Drinfeld twist (2-cocycle). If the twist is associated with the structure group then we have a deformation of the fibers. If the twist is associated with the automorphism group of the principal bundle, then we obtain noncommutative deformations of the base space as well. Combining the two twist deformations we obtain noncommutative principal bundles with both noncommutative fibers and base space. More in general, the natural isomorphisms proving the equivalence of a closed monoidal category of modules and its twist related one are used to obtain new Hopf-Galois extensions as twists of Hopf-Galois extensions. A sheaf approach is also considered, and examples presented.

  7. Observation of the heteroclinic tangles in the heat flux pattern of the ergodic divertor at TEXTOR

    International Nuclear Information System (INIS)

    A fine structure of open chaotic field lines, namely, a heteroclinic tangle, in the ergodic divertor has been observed by measurements of heat deposition pattern on the divertor plates at TEXTOR. Calculations show that magnetic footprints on the divertor plates are formed by open field lines coming from the plasma along narrow stripe regions called fingers. The latter are determined by the structure of stable and unstable manifolds of the outermost resonant magnetic island. This fact is confirmed by observations of the bifurcations of the heat flux pattern on the divertor plates with changing edge safety factor

  8. Copper matrix composites as heat sink materials for water-cooled divertor target

    OpenAIRE

    Jeong-Ha You

    2015-01-01

    According to the recent high heat flux (HHF) qualification tests of ITER divertor target mock-ups and the preliminary design studies of DEMO divertor target, the performance of CuCrZr alloy, the baseline heat sink material for DEMO divertor, seems to only marginally cover the envisaged operation regime. The structural integrity of the CuCrZr heat sink was shown to be affected by plastic fatigue at 20 MW/m². The relatively high neutron irradiation dose expected for the DEMO divertor target is ...

  9. Experimental studies on an axisymmetric divertor in DIVA(JFT-2a)

    International Nuclear Information System (INIS)

    DIVA(JFT-2a) is the first tokamak with an axisymmetric divertor in the world. Objectives of the experiments were i) Plasma production and confinement in a tokamak with a separatrix magnetic surface, and ii) divertor effects on radiation loss and plasma confinement. The results so far are as follows: i) The equilibrium with a separatrix magnetic surface is stable during the discharge. ii) There is an ergodic region near the separatrix magnetic surface due to non-axisymmetric magnetic perturbations. iii) The divertor reduces radiation loss and increases energy confinement time. iv) The divertor does not affect the transport process in the main plasma. (author)

  10. Limiter and divertor systems - conceptual and mechanical design for Aditya Tokamak upgrade

    International Nuclear Information System (INIS)

    Existing Aditya tokamak with limiter configuration is being upgraded into a machine to have both the limiter and divertor configurations. Necessary modifications have been carried out to accommodate divertor coils by replacing the old vacuum vessel with a new circular section vacuum vessel. The upgraded Aditya tokamak will have different set of limiters and divertors, such as Safety limiter, Toroidal Inner limiter, outer limiter of smaller toroidal extent, Upper and lower divertor plates. The limiter and divertor locations inside the Aditya tokamak upgrade are decided based on the numerical simulation of the plasma equilibrium profiles. Initially graphite will be used as plasma facing material (PFM) in all the limiter and divertor plates. The dimensions of the limiter and divertor tiles are decided based on their installation inside the vacuum vessel as well as on the total plasma heat loads (∼ 1 MW) falling on them. Depending upon the heat loads; the thickness of graphite tiles for limiter and divertor plates is estimated. Shaped graphite tiles will be fixed on specially designed support structures made out of SS-304L inside the torus shaped vacuum vessel. In this paper mechanical structural design of limiter and divertor of Aditya Upgrade Tokamak is presented. (author)

  11. Comparison of Ne and Ar seeded radiative divertor plasmas in JT-60U

    Energy Technology Data Exchange (ETDEWEB)

    Nakano, T., E-mail: nakano.tomohide@jaea.go.jp

    2015-08-15

    In H-mode plasmas with Ne, Ar and a mixture of Ne and Ar injection, the divertor radiation power fractions amongst these impurities in addition to an intrinsic impurity, C, are investigated. In plasmas with the inner divertor plasma attached, carbon is the biggest radiator, whichever impurity, Ne, Ar or a mixture of Ar and Ne is injected. In contrast, in plasmas with the inner divertor plasma detached, Ne is the biggest radiator due to a significantly high recombination radiation from Ne VIII. Ar is always a minor contributor in plasmas with the inner divertor both attached and detached.

  12. Divertor plasma conditions and neutral dynamics in horizontal and vertical divertor configurations in JET-ILW low confinement mode plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Groth, M., E-mail: mathias.groth@aalto.fi [Aalto University, Association EURATOM-Tekes, Otakaari 4, Espoo (Finland); Brezinsek, S. [Forschungszentrum Jülich, IEK4 – Plasma Physik, Jülich (Germany); Belo, P. [Institute of Plasmas and Nuclear Fusion, Association EURATOM/IST, Lisbon (Portugal); Culham Centre for Fusion Energy, Association EURATOM-CCFE, Abingdon (United Kingdom); Brix, M. [Culham Centre for Fusion Energy, Association EURATOM-CCFE, Abingdon (United Kingdom); Calabro, G. [Association EURATOM-ENEA, Frascati (Italy); Chankin, A. [Max-Planck Institute for Plasma Physics, EURATOM Association, Garching (Germany); Clever, M.; Coenen, J.W. [Forschungszentrum Jülich, IEK4 – Plasma Physik, Jülich (Germany); Corrigan, G. [Culham Centre for Fusion Energy, Association EURATOM-CCFE, Abingdon (United Kingdom); Drewelow, P. [Max-Planck-Institute for Plasma Physics, EURATOM Association, Greifswald (Germany); Guillemaut, C. [Association EURATOM CEA, CEA/DSM/IRFM, Cadarache (France); Harting, D. [Culham Centre for Fusion Energy, Association EURATOM-CCFE, Abingdon (United Kingdom); Huber, A. [Forschungszentrum Jülich, IEK4 – Plasma Physik, Jülich (Germany); Jachmich, S. [Association ‘Euratom-Belgian state’, Ecole Royale Militaire, Brussels (Belgium); Järvinen, A. [Aalto University, Association EURATOM-Tekes, Otakaari 4, Espoo (Finland); Kruezi, U.; Lawson, K.D. [Culham Centre for Fusion Energy, Association EURATOM-CCFE, Abingdon (United Kingdom); Lehnen, M. [Forschungszentrum Jülich, IEK4 – Plasma Physik, Jülich (Germany); ITER Organisation, 13115 Saint-Paul-Lez-Durance (France); and others

    2015-08-15

    Measurements of the plasma conditions at the low field side target plate in JET ITER-like wall ohmic and low confinement mode plasmas show minor differences in divertor plasma configurations with horizontally and vertically inclined targets. Both the reduction of the electron temperature in the vicinity of the strike points and the rollover of the ion current to the plates follow the same functional dependence on the density at the low field side midplane. Configurations with vertically inclined target plates, however, produce twice as high sub-divertor pressures for the same upstream density. Simulations with the EDGE2D-EIRENE code package predict significantly lower plasma temperatures at the low field side target in vertical than in horizontal target configurations. Including cross-field drifts and imposing a pumping by-pass leak at the low-field side plate can still not recover the experimental observations.

  13. Crosstalk analysis of carbon nanotube bundle interconnects

    OpenAIRE

    Zhang, Kailiang; Tian, Bo; Zhu, Xiaosong; WANG, FANG; Wei, Jun

    2012-01-01

    Carbon nanotube (CNT) has been considered as an ideal interconnect material for replacing copper for future nanoscale IC technology due to its outstanding current carrying capability, thermal conductivity, and mechanical robustness. In this paper, crosstalk problems for single-walled carbon nanotube (SWCNT) bundle interconnects are investigated; the interconnect parameters for SWCNT bundle are calculated first, and then the equivalent circuit has been developed to perform the crosstalk analys...

  14. A Geometric Approach to Noncommutative Principal Bundles

    CERN Document Server

    Wagner, Stefan

    2011-01-01

    From a geometrical point of view it is, so far, not sufficiently well understood what should be a "noncommutative principal bundle". Still, there is a well-developed abstract algebraic approach using the theory of Hopf algebras. An important handicap of this approach is the ignorance of topological and geometrical aspects. The aim of this thesis is to develop a geometrically oriented approach to the noncommutative geometry of principal bundles based on dynamical systems and the representation theory of the corresponding transformation group.

  15. Parahoric bundles on a compact Riemann surface

    CERN Document Server

    Balaji, V

    2010-01-01

    Let $X$ be a compact Riemann surface of genus $g \\geq 2$. The aim of this paper is to study homomorphisms of certain discrete subgroups of $PSL(2, {\\mathbb R})$ into maximal compact subgroups of semisimple simply connected algebraic groups and relate them to torsors under a Bruhat-Tits group scheme. We also construct the moduli spaces of semistable parahoric bundles. These results generalize the theorem of Mehta and Seshadri on parabolic vector bundles.

  16. Evaluation on BDI of large diameter pin bundles by out-of-pile bundle compression test

    International Nuclear Information System (INIS)

    Bundle-duct interaction (BDI) in core fuel subassemblies in fast reactors (FRs) is a limiting factor for fuel burnup. Since the large diameter fuel pin is generally believed to be a measure to improve FR fuel performance, the out-of-pile bundle compression test with large diameter pins (φ8.5mm and (φ 10.4mm) was performed to evaluate BDI in these bundles. In the compression test, bundle cross-sectional images (CT images) under BDI condition were obtained by using the X-ray computer tomography. In the main study, the CT images were numerically analyzed to evaluate deformation of the large diameter pin bundle due to BDI. The CT image analysis results revealed that pin-to-duct contact did not occur when the flat-to-flat bundle compression level reached one wire diameter (BDI level of 1dw), which indicates that BDI in large diameter pin bundles was mitigated similarly to the currently used small diameter pin bundles. In addition, the mitigation mechanism for BDI, which delays initiation of pin-to-duct contact, was investigated by using the computer code analysis. The code analysis results showed that cladding oval-distortion acted as a major mitigation mechanism for BDI as in the case of small pin diameter bundles. (author)

  17. Annular burnout data from rod bundle experiments

    International Nuclear Information System (INIS)

    Burnout data for annular flow in a rod bundle are presented for both transient and steady-state conditions. Tests were performed at the Oak Ridge National Laboratory in the Thermal Hydraulic Test Facility (THTF), a pressurized-water loop containing an electrically heated 64-rod bundle. The bundle configuration is typical of later generation pressurized-water reactors with 17 x 17 fuel arrays. Both axial and radial power profiles are flat. All experiments were carried out in upflow with subcooled inlet conditions, insuring accurate flow measurement. Conditions within the bundle were typical of those which could be encountered during a nuclear reactor loss-of-coolant accident. Level average fluid conditions within the test section were calculated using steady-state mass and energy conservation considerations for the steady-state tests and a transient, homogeneous, equilibrium computer code for the transient tests. Unlike tube dryout, burnout within a rod bundle does not necessarily occur at one distinct axial level. The location of individual rod dryout was determined by scanning rods axially and locating the position where rod superheat increased from approx. =0 to 30 K or greater. Thermocouple instrumentation within the bundle allows the location of dryout to be determined to within approximately +.5 cm for many of the tests

  18. K-Theories for Certain Infinite Rank Bundles

    OpenAIRE

    Larrain-Hubach, Andres

    2011-01-01

    Several authors have recently constructed characteristic classes for classes of infinite rank vector bundles appearing in topology and physics. These include the tangent bundle to the space of maps between closed manifolds, the infinite rank bundles in the families index theorem, and bundles with pseudodifferential operators as structure group. In this paper, we construct the corresponding K-theories for these types of bundles. We develop the formalism of these theories and use their Chern ch...

  19. Effect of left bundle branch block on TIMI frame count

    OpenAIRE

    Hatice Tolunay; Ahmet Kasapkara; İsa Öner Yüksel; Nurcan Başar; Ayşe Saatcı Yaşar; Mehmet Bilge

    2010-01-01

    Aim: Left bundle branch block is an independent risk factorfor cardiac mortality. In this study we aimed to evaluatecoronary blood flow with TIMI frame count in patients with left bundle branch block and angiographically proven normal coronary arteries.Materials and methods: We retrospectively studied 17 patients with left bundle branch block and as a control group 16 patients without left bundle branch block. All patientshad angiographically proven normal coronary arteries.Left bundle branch...

  20. Product-bundling and Incentives for Merger and Strategic Alliance

    OpenAIRE

    Sue Mialon

    2009-01-01

    This paper analyzes firms' choice between a merger and a strategic alliance in bundling their product with other complementary products. We consider a framework in which firms can improve profits only from product-bundling. While mixed bundling is not profitable, pure bundling is because pure bundling reduces consumers' choices, and thus softens competition among firms. Firms benefit the most from this reduced competition if they form an alliance. Firms do not gain as much from a merger becau...

  1. Mechanism of Actin Filament Bundling by Fascin

    Energy Technology Data Exchange (ETDEWEB)

    Jansen, Silvia; Collins, Agnieszka; Yang, Changsong; Rebowski, Grzegorz; Svitkina, Tatyana; Dominguez, Roberto (UPENN); (UPENN-MED)

    2013-03-07

    Fascin is the main actin filament bundling protein in filopodia. Because of the important role filopodia play in cell migration, fascin is emerging as a major target for cancer drug discovery. However, an understanding of the mechanism of bundle formation by fascin is critically lacking. Fascin consists of four {beta}-trefoil domains. Here, we show that fascin contains two major actin-binding sites, coinciding with regions of high sequence conservation in {beta}-trefoil domains 1 and 3. The site in {beta}-trefoil-1 is located near the binding site of the fascin inhibitor macroketone and comprises residue Ser-39, whose phosphorylation by protein kinase C down-regulates actin bundling and formation of filopodia. The site in {beta}-trefoil-3 is related by pseudo-2-fold symmetry to that in {beta}-trefoil-1. The two sites are {approx}5 nm apart, resulting in a distance between actin filaments in the bundle of {approx}8.1 nm. Residue mutations in both sites disrupt bundle formation in vitro as assessed by co-sedimentation with actin and electron microscopy and severely impair formation of filopodia in cells as determined by rescue experiments in fascin-depleted cells. Mutations of other areas of the fascin surface also affect actin bundling and formation of filopodia albeit to a lesser extent, suggesting that, in addition to the two major actin-binding sites, fascin makes secondary contacts with other filaments in the bundle. In a high resolution crystal structure of fascin, molecules of glycerol and polyethylene glycol are bound in pockets located within the two major actin-binding sites. These molecules could guide the rational design of new anticancer fascin inhibitors.

  2. Analysis of the Bundle Duct Interaction using the FBR fuel pin bundle deformation analysis code 'BAMBOO'

    International Nuclear Information System (INIS)

    PNC has been developing a computer code 'BAMBOO' to analyze the wire spaced FBR fuel pin bundle deformation under the BDI (Bundle Duct Interaction) condition by means of the three dimensional F.E.M. This code analyzes fuel pins' bowing and oval deformations which are dominant deformation behaviors of the fuel pin bundle under the BDI condition. In this study the 'BAMBOO' code is validated on the out-of-pile compression test of the FBR bundle (compression test) by comparing the results of the code analysis with the compression test results, and the highly irradiated (≥2.1x1027 n/m2, E > 0.1 MeV) bundle deformation behaviors are investigated from the viewpoint of the similarity to those in the compression test based on the analytical results of the code. (1) The calculated pin-to-duct minimum clearances as a function of the BDI levels in the compression test analysis agree with the experimental values evaluated from the CT image analysis of the bundle cross-section in the compression test within ±0.2 mm. And the calculated values of the fuel pins' oval deformations agree with the experimental values based on the pin diameter measurements done after the compression test within ±0.05 mm. (2) By comparing the irradiation induced bundle deformation with the bundle deformation in the compression test based on the code analysis, it is confirmed that the changes of the pin-to-duct minimum clearances with the BDI levels show equivalent trends between the both bundle deformations. And in this code analysis of the irradiation induced bundle deformation, contact loads between the fuel pins and the pacer wires are extremely small (below 10 kgf) even at about 3 dw of the BDI level compared to those in the compression test analysis. (J.P.N.)

  3. NIF laser bundle review. Final report

    International Nuclear Information System (INIS)

    We performed additional bundle review effort subsequent to the completion of the preliminary report and are revising our original recommendations. We now recommend that the NIF baseline laser bundle size be changed to the 4x2 bundle configuration. There are several 4x2 bundle configurations that could be constructed at a cost similar to that of the baseline 4x12 (from $11M more to about $11M less than the baseline; unescalated, no contingency) and provide significant system improvements. We recommend that the building cost estimates (particularly for the in-line building options) be verified by an architect/engineer (A/E) firm knowledgeable about building design. If our cost estimates of the in-line building are accurate and therefore result in a change from the baseline U-shaped building layout, the acceptability of the in-line configuration must be reviewed from an operations viewpoint. We recommend that installation, operation, and maintenance of all laser components be reviewed to better determine the necessity of aisles, which add to the building cost significantly. The need for beam expansion must also be determined since it affects the type of bundle packing that can be used and increases the minimum laser bay width. The U-turn laser architecture (if proven viable) offers a reduction in building costs since this laser design is shorter than the baseline switched design and requires a shorter laser bay

  4. Investigation of scrape-off layer and divertor heat transport in ASDEX Upgrade L-mode

    Science.gov (United States)

    Sieglin, B.; Eich, T.; Faitsch, M.; Herrmann, A.; Scarabosio, A.; the ASDEX Upgrade Team

    2016-05-01

    Power exhaust is one of the major challenges for the development of a fusion power plant. Predictions based upon a multimachine database give a scrape-off layer power fall-off length {λq}≤slant 1 mm for large fusion devices such as ITER. The power deposition profile on the target is broadened in the divertor by heat transport perpendicular to the magnetic field lines. This profile broadening is described by the power spreading S. Hence both {λq} and S need to be understood in order to estimate the expected divertor heat load for future fusion devices. For the investigation of S and {λq} L-Mode discharges with stable divertor conditions in hydrogen and deuterium were conducted in ASDEX Upgrade. A strong dependence of S on the divertor electron temperature and density is found which is the result of the competition between parallel electron heat conductivity and perpendicular diffusion in the divertor region. For high divertor temperatures it is found that the ion gyro radius at the divertor target needs to be considered. The dependence of the in/out asymmetry of the divertor power load on the electron density is investigated. The influence of the main ion species on the asymmetric behaviour is shown for hydrogen, deuterium and helium. A possible explanation for the observed asymmetry behaviour based on vertical drifts is proposed.

  5. Installation, features, and capabilities of the DIII-D advanced tokamak radiative divertors

    Energy Technology Data Exchange (ETDEWEB)

    Friend, M.E. E-mail: friend@fusion.gat.com; Bozek, A.S.; Baxi, C.B.; O' Neill, R.C.; Reis, E.E.; Mahdavi, M.A

    2001-10-01

    The DIII-D program has completed a series of density control and plasma core confinement experiments this past year. These experiments were designed to investigate the performance of baffled and open divertors with single-null plasmas and particle control in double-null plasmas. The experiments utilized all three of the DIII-D divertor assemblies located in the lower outer corner, the upper outer corner, and the upper inner corner of the vessel, which were installed last year. Each divertor consists of a liquid helium cryopump, a shielded protective ring, and a gas puff system. The divertors were designed to optimize pumping performance and to withstand the electromagnetic loads from both halo and toroidal, induced currents. With theoretical pumping speeds varying from 15,000 to 32,000 l/s, the cryopumps, combined with the baffle structures, collect particles and prevent them from recirculating back into the plasma core. The intent of the gas puff systems is to inject neutral gases in and around the divertors to minimize the heat flux on the divertors, minimizing the impurities generated by the excessive heating of the divertor graphite tiles. This hardware permits either single- or double-null plasma experiments and enables continued research of well confined high beta divertor plasmas with noninductive current drive, which is one of the primary research goals of DIII-D.

  6. Near-infrared spectroscopy for divertor plasma diagnosis and control in DIII-D tokamak

    International Nuclear Information System (INIS)

    New near infrared (NIR) spectroscopic measurements performed in the DIII-D tokamak divertor plasma suggest new viable diagnostic applications: divertor recycling and low-Z impurity flux measurements, a spectral survey for divertor Thomson scattering (DTS) diagnostic, and Te monitoring for divertor detachment control. A commercial 0.3 m spectrometer coupled to an imaging lens via optical fiber and a InGaAs 1024 pixel array detector enabled deuterium and impurity emission measurements in the range 800–2300 nm. The first full NIR survey identified D, He, B, Li, C, N, O, Ne lines and provided plasma Te, ne estimates from deuterium Paschen and Brackett series intensity and Stark line broadening analysis. The range 1.000–1.060 mm was surveyed in high-density and neon seeded divertor plasmas for spectral background emission studies for λ = 1.064 μm laser-based DTS development. The ratio of adjacent deuterium Paschen-α and Brackett Br9 lines in recombining divertor plasmas is studied for divertor Te monitoring aimed at divertor detachment real-time feedback control

  7. Near-infrared spectroscopy for divertor plasma diagnosis and control in DIII-D tokamak.

    Science.gov (United States)

    Soukhanovskii, V A; McLean, A G; Allen, S L

    2014-11-01

    New near infrared (NIR) spectroscopic measurements performed in the DIII-D tokamak divertor plasma suggest new viable diagnostic applications: divertor recycling and low-Z impurity flux measurements, a spectral survey for divertor Thomson scattering (DTS) diagnostic, and Te monitoring for divertor detachment control. A commercial 0.3 m spectrometer coupled to an imaging lens via optical fiber and a InGaAs 1024 pixel array detector enabled deuterium and impurity emission measurements in the range 800-2300 nm. The first full NIR survey identified D, He, B, Li, C, N, O, Ne lines and provided plasma Te, ne estimates from deuterium Paschen and Brackett series intensity and Stark line broadening analysis. The range 1.000-1.060 mm was surveyed in high-density and neon seeded divertor plasmas for spectral background emission studies for λ = 1.064 μm laser-based DTS development. The ratio of adjacent deuterium Paschen-α and Brackett Br9 lines in recombining divertor plasmas is studied for divertor Te monitoring aimed at divertor detachment real-time feedback control. PMID:25430325

  8. Divertor coil power supply in Aditya Tokamak for improved plasma operation

    International Nuclear Information System (INIS)

    The existing Aditya tokamak, a medium sized tokamak with limiter configuration is being upgraded to a Tokamak with divertor configuration. This moderate field Tokamak is capable of producing 250 kA of plasma current with 300 ms duration. Two new sets of diverter coils will be added to the system with an objective of producing double null plasmas in Aditya Upgrade Tokamak. Diverter coils, made up of continuously transposed conductor, are low voltage high current carrying poloidal field coils. One set of inner divertor coil has radius of 460 mm containing 6 turns and the other set of 1075 mm radius coil with 1 turn makes the outer divertor coils. The simulated plasma double null equilibrium demands 150 kAT of NI for the inner divertor coils and 10 - 20 kAT of NI for outer divertor coils. To energize the divertor coils with required power, a pulsed DC power supply of 3 MW (100V, 30 kA) has been designed. The designed pulsed DC power supply will be a 3-phase, 12-pulse rectifier based convertor power supply having a duty cycle of 300 ms on-time and 15 minutes off-time. The current rise time in the divertor coils will be ∼ 0.6 MA/sec. Detailed design of the divertor power supply with active controls for real time control of the plasma shape will be discussed in this paper. (author)

  9. Particle and power deposition on divertor targets in EAST H-mode plasmas

    DEFF Research Database (Denmark)

    Wang, L.; Xu, G.S.; Guo, H.Y.;

    2012-01-01

    The effects of edge-localized modes (ELMs) on divertor particle and heat fluxes were investigated for the first time in the Experimental Advanced Superconducting Tokamak (EAST). The experiments were carried out with both double null and lower single null divertor configurations, and comparisons w...

  10. Alcator C-Mod: A high-field divertor tokamak

    Science.gov (United States)

    Lipschultz, B.; Becker, H.; Bonoli, P.; Coleman, J.; Fiore, C.; Golovato, S.; Granetz, R.; Greenwald, M.; Gwinn, D.; Humphries, D.; Hutchinson, I.; Irby, J.; Marmar, E.; Montgomery, D. B.; Najmabadi, F.; Parker, R.; Porkolab, M.; Rice, J.; Sevillano, E.; Takase, Y.; Terry, J.; Watterson, R.; Wolfe, S.

    1989-04-01

    The Alcator C-Mod tokamak is a new device presently under construction at Massachusetts Institute of Technology (M.I.T.) which is scheduled to begin operation in mid-1990. The projected operating parameters are as follows: Toroidal field of 9 T; Ip ≤ 3 MA, R = 66.5 cm, a = 21 cm, κ ≤ 2.0, δ ≤ 0.5, ne ≤ 10 21m-3, PICRF ≤ 6 MW. The divertor configuration includes mechanical baffling as opposed to an 'open' geometry. Under strictly ohmic heating conditions, central Ti and Te are predicted to be in the range 2.5-3.5 keV over the density range (4-8) × 10 20m-3. With the addition of 6 MW of ICRF heating, Ti should vary from 4-8 keV over the same density range (assuming either Kaye-Goldston or Neo-Alcator scalings for electron confinement). Based on edge plasma characterizations from Alcator-C and divertor tokamaks, the scrape-off layer (SOL) properties are predicted to be: λn ≈ 10mm, density at the divertor plate < 2 × 10 21m-3, H 0 ionization mean free path between 1 and 10 mm. Maximum heat loads on various internal components are predicted to be in the range 5-10 MW/m 2. The flexibility of the poloidal field system in forming a number of flux surface geometries will provide further comparisons of the relative impurity control capabilities of double-null, single-null and limiter plasmas.

  11. The installation of the JET Mki divertor - features and achievements

    International Nuclear Information System (INIS)

    The installation of the MkI divertor involved the removal of 25 tonnes and installing nearly 80 tonnes of equipment into the JET vacuum vessel. The work was carried out over a twenty-two and a half month shutdown period requiring over 50,000 man hours of in-vessel effort. JET's requirements were successfully achieved in terms of component installation, dimensional accuracy and timescale. The work was organised into three stages: Stage One dealt with the stripout of all components and the removal from the vessel wall of many welded bosses. Two external toroidal coils were also replaced. This work was carried out in a Beryllium and Tritium contaminated environment with a high radiation level. An extensive cleaning and decontamination exercise completed the stage. Stage Two involved the fabrication of the four divertor coils at the end of which another full vessel cleaning was carried out. Stage Three was the installation of the pumped divertor and the associated wall and roof components. This paper reviews the overall concept of the shutdown and looks at the organisations, training and logistical support required. Certain key features will be described. These include the cleaning and decontamination of the vessel, without this full pressurised suits would have been needed for phases two and three, thus extending the programme to an unacceptable length. Material handling in the vessel which ranged from temporarily supporting the 24 tonne weight of the four coils from the roof of the vessel to an in-vessel crane and mobile personnel carriers. Finally, the installation and correct alignment of the target plates on which the success of the shutdown largely depended. (orig.)

  12. Compatibility of the radiating divertor with high performance plasmas in DIII-D

    International Nuclear Information System (INIS)

    Full text: We report on recent DIII-D experiments that successfully applied a radiating divertor scenario to high performance 'hybrid' plasmas [T.C. Luce, et al., Nucl. Fusion 43 (2003) 321]. In the puff-and-pump approach [M.J. Schaffer, et al., Nucl. Mater. 241-243 (1997) 585] used here, argon was injected near the outer divertor target, plasma flows into both the inner and outer divertors were enhanced by a combination of particle pumping near both divertor targets and deuterium gas puffing upstream of the divertor targets, and a 'dome' structure in the private flux region isolated the inner divertor from the outer divertor. Good hybrid conditions were maintained (e.g. energy confinement time normalized to ITER89p ≥ 2 and normalized plasma β ≅ 2.4), and the argon accumulation in the main plasma was modest. The peak heat flux at the outer divertor target was reduced by a factor of ≅ 2.5, while the peak heat flux at the inner target fell by only ∼20%. This was largely due to a much higher argon concentration near the outer divertor target than near the inner target (∼7 times). Exhaust enrichment (ER) as high as 64 were obtained, and ER was insensitive to the argon injection rate. (ER is defined as the ratio of the neutral argon pressure in the baffle plenum to the atomic-equivalent pressure of deuterium in the baffle plenum, divided by the ratio of argon density to electron density in the main plasma.) The asymmetry in the argon distribution and the favorable enrichment values arose largely from the closed and partitioned divertor geometry and from the frictional forces due to the enhanced divertor flow, which impeded the escape of argon from the outer divertor. Although the argon density profiles were more peaked than the electron profiles at high argon injection rates, the emissivity profiles in the main plasma remained 'hollow'. Our results suggest that independent control of both the radiating properties at the inner and outer divertor targets can be

  13. Conceptual Design for a Bulk Tungsten Divertor Tile in JET

    International Nuclear Information System (INIS)

    With ITER on the verge of being build, the ITER-like Wall project (ILW) for JET aims at providing the plasma chamber of the tokamak with an environment of mixed materials which will be relevant to the support of decisions to the first wall construction and, from the point of view of plasma physics, to the corresponding investigations of possible plasma configuration and plasma-wall interaction. In both respects, tungsten plays a key role in the divertor cladding whereas beryllium will be used for the vessel's first wall. For the central tile, also called LB-SRP for '' Load-Bearing Septum Replacement Plate '', resort to bulk tungsten is envisaged in order to cope with the high loads expected (up to 10 MW/m2 for about 10 s). This is indeed the preferred plasma-facing component for positioning the outer strike-point in the divertor. Forschungszentrum Juelich has developed a conceptual design for this tile, based on an assembly of tungsten blades or lamellae. It was selected in the frame of an extensive R-and-D study in search of a suitable, inertially cooled component(T. Hirai et al., R-and-D on full tungsten divertor and beryllium wall for JET ITER-like Wall Project: this conference). As reported elsewhere, the design is actually driven by electromagnetic considerations in the first place(S. Sadakov et al., Detailed electromagnetic analysis for optimisation of a tungsten divertor plate for JET: this conference). The lamellae are grouped in four stacks per tile which are independently attached to an equally re-designed supporting structure. A so-called adapter plate, also a new design, takes care of an appropriate interface to the base carrier of JET, onto which modules of two tiles are positioned and screwed by remote handling (RH) procedures. The compatibility of the design on the whole with RH requirements is another essential ingredient which was duly taken into account throughout. The concept and the underlying philosophy will be presented along with important

  14. Equilibrium configuration for a high current pumped divertor

    International Nuclear Information System (INIS)

    A realistic design of a pumped divertor plasma configuration to be fitted to the JET vessel can be obtained as a compromise among various geometrical, physical and technical constraints. The possibility of reaching a satisfactory solution has been analysed for plasmas up to 6 MA. Optimisation of the plasma coupling to the RF antennae requires a largely asymmetric distribution of ampere turns in the PF coils and some mechanical flexibility. The calculations presented were carried out using the specially developed JET equilibrium and configuration analysis codes. (U.K.)

  15. Manufacture and installation of JET MKII divertor support structure

    International Nuclear Information System (INIS)

    The water cooled support structure, comprising twenty-four modules is the main component of the JET MKII divertor system. It is to be installed in the vacuum vessel with high accuracy with respect to the magnetic center and the other in-vessel components. The paper describes the design and manufacturing cycle including the required tolerances, the assembly and installation method and the material production process required to ensure the accuracy and reliability of the MKII support structure system. The water cooling holes, machined into the support structure require the procurement of special material to prevent risks of leaks inside the vacuum vessel

  16. Divertor retention of metallic impurities during neutralization plate biasing on TdeV

    International Nuclear Information System (INIS)

    Laser ablation injection of aluminium is used to measure the retention of metallic impurities in the lower poloidal divertor of TdeV. A detailed calibration of the ablation process allows the determination of the quantity and velocity distribution of the injected particles. The experiment measures the flow of the injected particles from the divertor to the main plasma. Negative biasing of the divertor neutralization plates is shown to improve the retention in the active divertor by a factor of at least four at -200 V. A simple model is developed to show that the improved confinement is due to the increased poloidal flux to the divertor during biasing. (author). 32 refs, 9 figs

  17. Analysis on EAST LHCD operation space by using simple Core-SOL-Divertor model

    International Nuclear Information System (INIS)

    A simple Core-SOL-Divertor model (CSD model) has been developed to investigate qualitatively the overall features of the operational space for the integrated core and edge plasma. In the CSD model, the core plasma model of ITER physics guidelines and the two-point SOL-divertor model are applied. This CSD model is validated by the two dimensional divertor transport code (B2-EIRINE) and by the JT-60U divertor recycling database, and this model is applicable to the low- and high-recycling state of the divertor plasma. The CSD model is applied to the study of the EAST operational space with lower hybrid current drive experiments under various kinds of trade-off for the basic plasma parameters, and the relationship between the operational space and the plasma discharge duration is also discussed. (author)

  18. Analysis on EAST LHCD Operation Space by Using Simple Core-SOL-Divertor Model

    International Nuclear Information System (INIS)

    A simple core-SOL-divertor model (CSD model) was developed to investigate qualitatively the overall features of the operational space for the integrated core and edge plasma. In the CSD model, the core plasma model of ITER physics guidelines and the two-point SOL-divertor model are applied. This CSD model is validated by the two dimensional divertor transport code (B2-EIRINE) and by the JT-60U divertor recycling database, and this model is applicable to the low- and high-recycling state of the divertor plasma. The CSD model is applied to the study of the EAST operational space with lower hybrid current drive under various kinds of trade-off for the basic plasma parameters, and the relationship between the operational space and the plasma discharge duration is also discussed. (magnetically confined plasma)

  19. Model-based radiation scalings for the ITER-like divertors of JET and ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Aho-Mantila, L., E-mail: leena.aho-mantila@vtt.fi [VTT Technical Research Centre of Finland, FI-02044 VTT (Finland); Bonnin, X. [LSPM – CNRS, Université Paris 13, Sorbonne Paris Cité, F-93430 Villetaneuse (France); Coster, D.P. [Max-Planck Institut für Plasmaphysik, D-85748 Garching (Germany); Lowry, C. [EFDA JET CSU, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Wischmeier, M. [Max-Planck Institut für Plasmaphysik, D-85748 Garching (Germany); Brezinsek, S. [Forschungszentrum Jülich, Institut für Energie- und Klimaforschung Plasmaphysik, 52425 Jülich (Germany); Federici, G. [EFDA PPP& T Department, D-85748 Garching (Germany)

    2015-08-15

    Effects of N-seeding in L-mode experiments in ASDEX Upgrade and JET are analysed numerically with the SOLPS5.0 code package. The modelling yields 3 qualitatively different radiative regimes with increasing N concentration, when initially attached outer divertor conditions are studied. The radiation pattern is observed to evolve asymmetrically, with radiation increasing first in the inner divertor, then in the outer divertor, and finally on closed field lines above the X-point. The properties of these radiative regimes are observed to be sensitive to cross-field drifts and they differ between the two devices. The modelled scaling of the divertor radiated power with the divertor neutral pressure is similar to an experimental scaling law for H-mode radiation. The same parametric dependencies are not observed in simulations without drifts.

  20. Facilities for technology testing of ITER divertor concepts, models, and prototypes in a plasma environment

    International Nuclear Information System (INIS)

    The exhaust of power and fusion-reaction products from ITER plasma are critical physics and technology issues from performance, safety, and reliability perspectives. Because of inadequate pulse length, fluence, flux, scrape-off layer plasma temperature and density, and other parameters, the present generation of tokamaks, linear plasma devices, or energetic beam facilities are unable to perform adequate technology testing of divertor components, though they are essential contributors to many physics issues such as edge-plasma transport and disruption effects and control. This Technical Requirements Documents presents a description of the capabilities and parameters divertor test facilities should have to perform accelerated life testing on predominantly technological divertor issues such as basic divertor concepts, heat load limits, thermal fatigue, tritium inventory and erosion/redeposition. The cost effectiveness of such divertor technology testing is also discussed

  1. Comparison of observed divertor heat flux and modeling results at LHD

    International Nuclear Information System (INIS)

    The divertor strike line pattern on the helical divertor of LHD was observed with an infra red camera. The derived heat flux pattern show multiple distinct strike lines depending on the equilibrium magnetic configuration. Predictions of such divertor heat loads thus require a modeling of the magnetic configuration and the heat transport in the magnetic edge. Equilibrium magnetic topologies were analyzed with HINT2, while the plasma fluid model code EMC3 was used to simulate the energy transport in the edge. The measured multi peak structure of the divertor heat flux is correlated to the intersection points of elongated loop shaped flux tubes of long LC field lines. But the fluid model could not recreate the total energy load and the multiple heat flux peaks on the divertor. A Variation in the plasma density ne as a transport parameter in order to fit the simulated heat flux to the measured one shows a contradicting tendency. (author)

  2. Tungsten erosion and redeposition in the all-tungsten divertor of ASDEX Upgrade

    International Nuclear Information System (INIS)

    Net erosion and deposition of tungsten (W) in the ASDEX Upgrade divertor were determined after the 2007 campaign by using thin W marker stripes. ASDEX Upgrade had full-W plasma-facing components during this campaign. The inner divertor and the roof baffle were net W deposition areas with a maximum deposition of about 1x1018 W-atoms cm-2 in the private flux region below the inner strike point. Net erosion of W was observed in the whole outer divertor, with the largest erosion close to the outer strike point. Only a small fraction of the W eroded in the main chamber and in the outer divertor was found in redeposits in the inner divertor, while a large fraction was either redeposited at unidentified places in the main chamber or has formed dust.

  3. Tungsten erosion and redeposition in the all-tungsten divertor of ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Mayer, M; Krieger, K; Matern, G; Neu, R; Rasinski, M; Rohde, V; Sugiyama, K; Wiltner, A [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Boltzmannstrasse 2, 85748 Garching (Germany); Andrzejczuk, M; Fortuna-Zalesna, E; Kurzydlowski, K J; Zielinski, W [Faculty of Materials Science and Engineering, Warsaw University of Technology, Association EURATOM-IPPLM, 02-507 Warsaw (Poland); Hakola, A; Koivuranta, S; Likonen, J [VTT Materials for Power Engineering, EURATOM Association, PO Box 1000, FI-02044 VTT (Finland); Ramos, G [CICATA-Qro, Instituto Politecnico Nacional, Queretaro (Mexico); Dux, R, E-mail: matej.mayer@ipp.mpg.de

    2009-12-15

    Net erosion and deposition of tungsten (W) in the ASDEX Upgrade divertor were determined after the 2007 campaign by using thin W marker stripes. ASDEX Upgrade had full-W plasma-facing components during this campaign. The inner divertor and the roof baffle were net W deposition areas with a maximum deposition of about 1x10{sup 18} W-atoms cm{sup -2} in the private flux region below the inner strike point. Net erosion of W was observed in the whole outer divertor, with the largest erosion close to the outer strike point. Only a small fraction of the W eroded in the main chamber and in the outer divertor was found in redeposits in the inner divertor, while a large fraction was either redeposited at unidentified places in the main chamber or has formed dust.

  4. Turbulent flow through two asymmetric rod bundles

    International Nuclear Information System (INIS)

    Measurements of the mean velocity, of the wall shear stresses, and of the turbulence have been performed in four wall subchannels of rod bundles of four parallel rods enclosed in a rectangular channel. The pitch-to-diameter ratio was P/D=1.148 and the wall-to-diameter ratios ranged from 1.045 to 1.252. The full Reynolds stress tensor has been determined by hot-wire technique. The results of the turbulences intensities show that the flow through rod bundles differs widely from flow through circular tubes. More sophisticated analytical tools than presently available are required to predict turbulent flow through rod bundles with sufficient accuracy

  5. Assembly mechanism for nuclear fuel bundles

    International Nuclear Information System (INIS)

    In a nuclear power reactor fuel bundle having tie rods fastened to a lower tie plate and passing through openings in the upper tie plate, the assembled bundle is secured by locking lugs fixed to rotatable locking sleeves which engage the upper tie plate. Pressure exerted by helical springs mounted around each of the tie rods urge retaining lugs fixed to a retaining sleeve associated with respective tie rods into a position with respect to the locking sleeve to prevent accidental disengagement of the upper plate from the locking lugs. The bundle may be disassembled by depressing the retaining sleeves and rotating the locking lugs to the disengaged position, and then removing the upper tie plate

  6. Porous Silicon and Denim Fiber Bundle Characterization

    Science.gov (United States)

    Deuro, Randi Ellen

    My thesis research aims to characterize and exploit materials in an efficient, rapid, non-destructive manner. Part I of this document summarizes my research on porous silicon (pSi) design, fabrication, and surface modification for use as a novel chemical sensor. The optimization of fabrication process parameters (etching time, etching solution, electrode shape, and the fixing process) on pSi photoluminescence (PL) is presented. I have also investigated the effects of analyte vapors (acetonitrile, toluene, methanol, acetone) on the pSi PL and surface chemistry using luminescence and Fourier-transform infrared (FT-IR) spectroscopy and microscopy methods. The mechanism and benefits of one method of pSi surface modification and protection (ultraviolet (UV) hydrosilylation) will also be presented. Finally, high thorough-put methods of pSi sensor production are described. In Part II of this document, I introduce a novel technique for analyzing and discriminating among denim fiber bundles. An investigation into the benefits of luminescence-based multispectral imaging (LMSI) for denim fiber bundle identification has been conducted. I explore the power of nitromethane (CH 3NO2) based quenching in fiber bundle classification and identify the quenching mechanism. The luminescence spectra (450 - 850 nm) and images from the denim fiber bundles were obtained while exciting at 325 nm or 405 nm. Here, LMSI data were recorded in < 10 s and subsequently assessed by principal component analysis (PCA) and rendered red, green, blue (RGB) component histograms. The results show that LMSI data can be used to rapidly and uniquely classify all the fiber bundle types studied in this research. These non-destructive techniques eliminate extensive sample preparation and allow for rapid multispectral image collection, analysis, and assessment. The quenching data also revealed that the dye molecules within the individual fiber bundles exhibited dramatically different accessibilities to CH 3NO2.

  7. Bundling in semiflexible polymers: A theoretical overview.

    Science.gov (United States)

    Benetatos, Panayotis; Jho, YongSeok

    2016-06-01

    Supramolecular assemblies of polymers are key modules to sustain the structure of cells and their function. The main elements of these assemblies are charged semiflexible polymers (polyelectrolytes) generally interacting via a long(er)-range repulsion and a short(er)-range attraction. The most common supramolecular structure formed by these polymers is the bundle. In the present paper, we critically review some recent theoretical and computational advances on the problem of bundle formation, and point a few promising directions for future work. PMID:26813628

  8. A bundle of sticks in my garden

    OpenAIRE

    Farran, Sue

    2012-01-01

    The English law of property is often described as a ‘bundle of sticks’ in which each ‘stick’ represents a particular right. Gardens challenge these rights and wreak havoc on the ‘bundle of sticks’. This paper looks at the twenty-first century manifestations of community engagement with ground and explores how ‘gardening’ is undermining concepts of ownership, possession and management of land and how the fence between what is private and what is public is being encroached and challenged by com...

  9. Characteristic classes of quantum principal bundles

    CERN Document Server

    Durdevic, M

    1995-01-01

    A noncommutative-geometric generalization of classical Weil theory of characteristic classes is presented, in the conceptual framework of quantum principal bundles. A particular care is given to the case when the bundle does not admit regular connections. A cohomological description of the domain of the Weil homomorphism is given. Relations between universal characteristic classes for the regular and the general case are analyzed. In analogy with classical geometry, a natural spectral sequence is introduced and investigated. The appropriate counterpart of the Chern character is constructed, for structures admitting regular connections. Illustrative examples and constructions are presented.

  10. TRIGA spent fuel bundles safe storage

    Energy Technology Data Exchange (ETDEWEB)

    Negut, G.; Covaci, St. [Institute for Nuclear Research, Research Reactor Dept., Pitesti (Romania); Prisecaru, I.; Dupleac, D. [Bucharest Univ. Politehnica, Power and Nuclear Engineering Dept., Bucharest (Romania)

    2007-07-01

    TRIGA-SSR is a steady state research and material test reactor that has been in operation since 1980. The original TRIGA fuel was HEU (highly enriched uranium) with a U{sup 235} enrichment of 93 per cent. Almost all TRIGA HEU fuel bundles are now burned-up. Part of the spent fuel was loaded and transferred to US, in a Romania - DOE arrangement. The rest of the TRIGA fuel bundles have to be temporarily stored in the TRIGA facility. As the storage conditions had to be established with caution, neutron and thermal hydraulic evaluations of the storage conditions were required. Some criticality evaluations were made based on the SAR (Safety Analysis Report) data. Fuel constant axial temperature approximation effect is usual for criticality computations. TRIGA-SSR fuel bundle geometry and materials model for SCALE5-CSAS module allows the introduction of a fuel temperature dependency for the entire fuel active height, using different materials for each fuel bundle region. Previous RELAP5 thermal hydraulic computations for an axial and radial power distribution in the TRIGA fuel pin were done. Fuel constant temperature approximation overestimates pin factors for every core operating at high temperatures. From the thermal hydraulic point of view the worst condition of the storage grid occurs when the transfer channel is accidentally emptied of water from the pool, or the bundle is handled accidentally to remain in air. All the residual heat from the bundles has to be removed without fuel overheating and clad failure. RELAP5 computer code for residual heat removal was used in the assessment of residual heat removal. We made a couple of evaluations of TRIGA bundle clad temperatures in air cooling conditions, with different residual heat levels. The criticality computations have shown that the spent TRIGA fuel bundles storage grid is strongly sub-critical with k(eff) = 0.5951. So, there is no danger for a criticality accident for this storage grid type. The assessment is done

  11. Scaling Shift in Multicracked Fiber Bundles

    Science.gov (United States)

    Manca, Fabio; Giordano, Stefano; Palla, Pier Luca; Cleri, Fabrizio

    2014-12-01

    Bundles of fibers, wires, or filaments are ubiquitous structures in both natural and artificial materials. We investigate the bundle degradation induced by an external damaging action through a theoretical model describing an assembly of parallel fibers, progressively damaged by a random population of cracks. Fibers in our model interact by means of a lateral linear coupling, thus retaining structural integrity even after substantial damage. Monte Carlo simulations of the Young's modulus degradation for increasing crack density demonstrate a remarkable scaling shift between an exponential and a power-law regime. Analytical solutions of the model confirm this behavior, and provide a thorough understanding of the underlying physics.

  12. Safe Harbors for Quantity Discounts and Bundling

    OpenAIRE

    Dennis W. Carlton; Michael Waldman

    2008-01-01

    The courts and analysts continue to struggle to articulate safe harbors for a wide variety of common business pricing practices in which either a single product is sold at a discount if purchased in bulk or in which multiple products are bundled together at prices different from the ones that would emerge if the products were purchased separately. The phenomenon of tying in which the sale of one product is conditioned on the purchase of another is closely related to bundling. Its analysis rel...

  13. TRIGA spent fuel bundles safe storage

    International Nuclear Information System (INIS)

    TRIGA-SSR is a steady state research and material test reactor that has been in operation since 1980. The original TRIGA fuel was HEU (highly enriched uranium) with a U235 enrichment of 93 per cent. Almost all TRIGA HEU fuel bundles are now burned-up. Part of the spent fuel was loaded and transferred to US, in a Romania - DOE arrangement. The rest of the TRIGA fuel bundles have to be temporarily stored in the TRIGA facility. As the storage conditions had to be established with caution, neutron and thermal hydraulic evaluations of the storage conditions were required. Some criticality evaluations were made based on the SAR (Safety Analysis Report) data. Fuel constant axial temperature approximation effect is usual for criticality computations. TRIGA-SSR fuel bundle geometry and materials model for SCALE5-CSAS module allows the introduction of a fuel temperature dependency for the entire fuel active height, using different materials for each fuel bundle region. Previous RELAP5 thermal hydraulic computations for an axial and radial power distribution in the TRIGA fuel pin were done. Fuel constant temperature approximation overestimates pin factors for every core operating at high temperatures. From the thermal hydraulic point of view the worst condition of the storage grid occurs when the transfer channel is accidentally emptied of water from the pool, or the bundle is handled accidentally to remain in air. All the residual heat from the bundles has to be removed without fuel overheating and clad failure. RELAP5 computer code for residual heat removal was used in the assessment of residual heat removal. We made a couple of evaluations of TRIGA bundle clad temperatures in air cooling conditions, with different residual heat levels. The criticality computations have shown that the spent TRIGA fuel bundles storage grid is strongly sub-critical with k(eff) = 0.5951. So, there is no danger for a criticality accident for this storage grid type. The assessment is done for

  14. Research proposal on : amplitude modulated reflectometry system for JET divertor

    International Nuclear Information System (INIS)

    Amplitude Modulated reflectometry is presented here as a tool for density profile measurements in the JET divertor plasmas. One of the main problems which has been presented in most reflectometers during the last years is the need for a coherent tracking of the phase delay: fast density fluctuations and strong modulation on the amplitude of the reflected signal usually bring to fringe jumps' in the phase signal, which are a big problem when the phase values are much larger than 2 pi. The conditions in the JET divertor plasmas: plasma geometry, access and long oversized broad-band waveguide paths makes very difficult the phase measurements at the millimeter wave range. AM reflectometry is to some extension an intermediate solution between the classical phase delay reflectometry, so far applied to small distances, and the time domain reflectometry, used for ionospheric studies and recently also proposed for fusion plasma. the main advantage is to allow the use of millimeter wave reflectometry with moderate phase shifts (approx 2 pi). (author)

  15. Tore Supra divertor screening efficiency during density regime experiments

    International Nuclear Information System (INIS)

    The Tore Supra ergodic divertor (ED) screening efficiency has been investigated in density regime experiments. The ED screening efficiency is analysed by using the 'tightness' concept, which is the ratio of the density on the ED neutraliser plates to the volume averaged plasma density. Tightness is studied as a function of different plasma edge parameters, such as Tdiv, ED magnetic perturbation (Δ), plasma composition, location of recycling source, and additional power. Tightness is shown to increase with Δ, Pdiv0.55/(1-Fr)1.22, and 1/Tdiv0.5. These trends are well explained by a simple 0-D model, where the particle confinement time in the ergodized peripheral region is very small. Finally, tightness increases with the power conducted onto the ED plates. Since ED plasmas have low Pdiv, their tightness value remains low compared to that obtained with axisymmetric divertors for which Pdiv is considerably larger. Increasing Pdiv will result in an improved tightness and a better particle control

  16. Tore Supra divertor screening efficiency during density regime experiments

    Energy Technology Data Exchange (ETDEWEB)

    Grisolia, C. E-mail: grisolia@drfc.cad.cea.fr; Ghendrih, Ph.; Gunn, J.; Loarer, T.; Monier-Garbet, P.; De Michelis, C.; Costanzo, L.; Pascal, J.Y

    2001-03-01

    The Tore Supra ergodic divertor (ED) screening efficiency has been investigated in density regime experiments. The ED screening efficiency is analysed by using the 'tightness' concept, which is the ratio of the density on the ED neutraliser plates to the volume averaged plasma density. Tightness is studied as a function of different plasma edge parameters, such as T{sub div}, ED magnetic perturbation ({delta}), plasma composition, location of recycling source, and additional power. Tightness is shown to increase with {delta}, P{sub div}{sup 0.55}/(1-Fr){sup 1.22}, and 1/T{sub div}{sup 0.5}. These trends are well explained by a simple 0-D model, where the particle confinement time in the ergodized peripheral region is very small. Finally, tightness increases with the power conducted onto the ED plates. Since ED plasmas have low P{sub div}, their tightness value remains low compared to that obtained with axisymmetric divertors for which P{sub div} is considerably larger. Increasing P{sub div} will result in an improved tightness and a better particle control.

  17. First EMC3-Eirene simulations of the TCV snowflake divertor

    International Nuclear Information System (INIS)

    One of the approaches to solve the heat load problem in a divertor tokamak is the so called ‘snowflake’ (SF) configuration, a magnetic equilibrium with two nearby x-points and two additional divertor legs. Here we report on the first EMC3-Eirene simulations of plasma- and neutral particle transport in the scrape-off layer of a series of TCV SF equilibria with different values of σ, the distance between the x-points normalized to the minor radius of the plasma. The constant cross-field transport coefficients were chosen such that the power- and particle deposition profiles at the primary inner strike point (SP) match the Langmuir probe measurements for the σ = 0.1 case. At the secondary SP on the floor, however, a significantly larger power flux than that predicted by the simulation was measured by the probes, indicating an enhanced transport across the primary separatrix. As the ideal SF configuration (σ = 0) is approached, the density as well as the radiation maximum are predicted to move from the target plates upward to the x-point by the simulation. (paper)

  18. Axisymmetric curvature-driven instability in a model divertor geometry

    International Nuclear Information System (INIS)

    A model problem is presented which qualitatively describes a pressure-driven instability which can occur near the null-point in the divertor region of a tokamak where the poloidal field becomes small. The model problem is described by a horizontal slot with a vertical magnetic field which plays the role of the poloidal field. Line-tying boundary conditions are applied at the planes defining the slot. A toroidal field lying parallel to the planes is assumed to be very strong, thereby constraining the possible structure of the perturbations. Axisymmetric perturbations which leave the toroidal field unperturbed are analyzed. Ideal magnetohydrodynamics is used, and the instability threshold is determined by the energy principle. Because of the boundary conditions, the Euler equation is, in general, non-separable except at marginal stability. This problem may be useful in understanding the source of heat transport into the private flux region in a snowflake divertor which possesses a large region of small poloidal field, and for code benchmarking as it yields simple analytic results in an interesting geometry

  19. Power balance in the divertor-tokamak DIVA

    International Nuclear Information System (INIS)

    Power balances of Ohmically and radio-frequency (RF) heated plasmas including a boundary (scrape-off layer) plasma are investigated in the divertor-tokamak DIVA. First, methods of measurement of the boundary plasma are described. These are applied to the divertor plasma in the case of Ohmic heating. The results clarify characteristics of the boundary plasma of a conventional tokamak. The scaling law for the boundary plasma is derived in consideration of the power balance including the boundary plasma. Heat flux to material surfaces is investigated in detail; the relationship between heat flux, particle flux and electron velocity distribution is clarified. Gross power balance is investigated by measurements of total heat flux to the wall and total radiation loss including charge-exchange loss. These results provide experimental evidence for the above scaling law. Finally, power balance during the Ion-Cyclotron Range of Frequency (ICRF) heating is described. Optimum heating conditions of the ICRF heating in the two-ion hybrid regime are surveyed. For the optimum heating conditions, gross power balance including the boundary plasma is considered, in which the heating efficiency is derived. Radial profile of the RF-heating power, the ratio of the heating power to each species and the transport of RF-heated ions are clarified in the power balance. (author)

  20. Impact of bundle deformation on CHF: ASSERT-PV assessment of extended burnup Bruce B bundle G85159W

    International Nuclear Information System (INIS)

    This paper presents a subchannel thermalhydraulic analysis of the effect on critical heat flux (CHF) of bundle deformation such as element bow and diametral creep. The bundle geometry is based on the post-irradiation examination (PIE) data of a single bundle from the Bruce B Nuclear Generating Station, Bruce B bundle G85159W, which was irradiated for more than two years in the core during reactor commissioning. The subchannel code ASSERT-PV IST is used to assess changes in CHF and dryout power due to bundle deformation, compared to the reference, undeformed bundle. (author)

  1. Effects of divertor geometry and pumping on plasma performance on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Allen, S.L.; Hill, D.N.; Porter, G.D. [and others

    1997-06-01

    This paper reports the status of an ongoing investigation to discern the influence of the divertor and plasma geometry on the confinement of both ELM-free and ELMing discharges in DIII-D. The ultimate goal is to achieve a high-performance core plasma which coexists with an advanced divertor plasma. The divertor plasma must reduce the heat flux to acceptable levels; the current technique disperses the heat flux over a wide area by radiation (a radiative divertor). To date, we have obtained our best performance in double-null (DN) high-triangularity ({delta} {approximately} 0.8) ELM-free discharges. As discussed in detail elsewhere, there are several advantages for both the core and divertor plasma with highly-shaped DN operation. Previous radiative-divertor experiments with D{sub 2} injection in DN high-{delta} ELMing H-mode have shown that this configuration is more sensitive to gas puffing ({tau} decreases). Moving the X-point away from the target plate (to {approximately}15 cm above the plate) decreases this sensitivity. Preliminary measurements also indicate that gas puffing reduces the divertor heat flux but does not reduce the plasma pressure along the field line. The up/down heat flux balance can be varied magnetically (by changing the distance between the separatrices), with a slight magnetic imbalance required to balance the heat flux. The overall mission of the Radiative Divertor Project (RDP) is to install a fully pumped and baffled high-{delta} DN divertor. To date, however, both the DIII-D divertor diagnostics and pump were optimized for lower single-null (LSN) low-{delta} ({delta}{approximately} 0.4) plasmas, so much of the divertor physics has been performed in LSN; these results are discussed in Section 2. As part of the first phase of the RDP, we have installed a new high-{delta} USN divertor baffle and pump; these results are discussed in Section 3. Both divertor and core parameters are discussed in each case.

  2. An exploration of advanced X-divertor scenarios on ITER

    International Nuclear Information System (INIS)

    It is found that the X-divertor (XD) configuration (Kotschenreuther et al 2004 Proc. 20th Int. Conf. on Fusion Energy (Vilamoura, Portugal, 2004) (Vienna: IAEA) CD-ROM file [IC/P6-43] www-naweb.iaea.org/napc/physics/fec/fec2004/datasets/index.html, Kotschenreuther et al 2006 Proc. 21st Int. Conf. on Fusion Energy 2006 (Chengdu, China, 2006) (Vienna: IAEA), CD-ROM file [IC/P7-12] www-naweb.iaea.org/napc/physics/FEC/FEC2006/html/index.htm, Kotschenreuther et al 2007 Phys. Plasmas 14 072502) can be made with the conventional poloidal field (PF) coil set on ITER (Tomabechi et al and Team 1991 Nucl. Fusion 31 1135), where all PF coils are outside the TF coils. Starting from the standard divertor, a sequence of desirable XD configurations are possible where the PF currents are below the present maximum design limits on ITER, and where the baseline divertor cassette is used. This opens the possibility that the XD could be tested and used to assist in high-power operation on ITER, but some further issues need examination. Note that the increased major radius of the super-X-divertor (Kotschenreuther et al 2007 Bull. Am. Phys. Soc. 53 11, Valanju et al 2009 Phys. Plasmas 16 5, Kotschenreuther et al 2010 Nucl. Fusion 50 035003, Valanju et al 2010 Fusion Eng. Des. 85 46) is not a feature of the XD geometry. In addition, we present an XD configuration for K-DEMO (Kim et al 2013 Fusion Eng. Des. 88 123) to demonstrate that it is also possible to attain the XD configuration in advanced tokamak reactors with all PF coils outside the TF coils. The results given here for the XD are far more encouraging than recent calculations by Lackner and Zohm (2012 Fusion Sci. Technol. 63 43) for the Snowflake (Ryutov 2007 Phys. Plasmas 14 064502, Ryutov et al 2008 Phys. Plasmas 15 092501), where the required high PF currents represent a major technological challenge. The magnetic field structure in the outboard divertor SOL (Kotschenreuther 2013 Phys. Plasmas 20 102507) in the recently created

  3. Abelian conformal field theory and determinant bundles

    DEFF Research Database (Denmark)

    Andersen, Jørgen Ellegaard; Ueno, K.

    2007-01-01

    Following [10], we study a so-called bc-ghost system of zero conformal dimension from the viewpoint of [14, 16]. We show that the ghost vacua construction results in holomorphic line bundles with connections over holomorphic families of curves. We prove that the curvature of these connections are...

  4. Capacity efficiency of recovery request bundling

    DEFF Research Database (Denmark)

    Ruepp, Sarah Renée; Dittmann, Lars; Berger, Michael Stübert; Stidsen, Thomas Riis; Lagakos, Stephen; Perlovsky, Leonid; Jha, Manoi; Covaci, Brindusa; Zaharim, Azarni; Mastorakis, Nikos

    2010-01-01

    This paper presents a comparison of recovery methods in terms of capacity efficiency. In particular, a method where recovery requests are bundled towards the destination (Shortcut Span Protection) is evaluated against traditional recovery methods. Our simulation results show that Shortcut Span Pr...... Protection uses more capacity than the unbundled related methods, but this is compensated by easier control and management of the recovery actions....

  5. Line bundles on moduli and related spaces

    CERN Document Server

    Huebschmann, Johannes

    2009-01-01

    Let G be a Lie goup, let M and N be smooth connected G-manifolds, let f be a smooth G-map from M to N, and let P denote the fiber of f. Given a closed and equivariantly closed relative 2-form for f with integral periods, we construct the principal G-circle bundles with connection on P having the given relative 2-form as curvature. Given a compact Lie group K, a biinvariant Riemannian metric on K, and a closed Riemann surface S of genus s, when we apply the construction to the particular case where f is the familiar relator map from a product of 2s copies of K to K we obtain the principal K-circle bundles on the associated extended moduli spaces which, via reduction, then yield the corresponding line bundles on possibly twisted moduli spaces of representations of the fundamental group of S in K, in particular, on moduli spaces of semistable holomorphic vector bundles or, more precisely, on a smooth open stratum when the moduli space is not smooth. The construction also yields an alternative geometric object, d...

  6. Bundle Gerbes Applied to Quantum Field Theory

    CERN Document Server

    Carey, A L; Murray, M; Carey, Alan; Mickelsson, Jouko; Murray, Michael

    2000-01-01

    This paper reviews recent work on a new geometric object called a bundle gerbe and discusses some new examples arising in quantum field theory. One application is to an Atiyah-Patodi-Singer index theory construction of the bundle of fermionic Fock spaces parametrized by vector potentials in odd space dimensions and a proof that this leads in a simple manner to the known Schwinger terms (Mickelsson-Faddeev cocycle) for the gauge group action. This gives an explicit computation of the Dixmier-Douady class of the associated bundle gerbe. The method works also in other cases of fermions in external fields (external gravitational field, for example) provided that the APS theorem can be applied; however, we have worked out the details only in the case of vector potentials. Another example, in which the bundle gerbe curvature plays a role, arises from the WZW model on Riemann surfaces. A further example is the `existence of string structures' question. We conclude by showing how global Hamiltonian anomalies fit with...

  7. Quantum field theories on Hilbert bundles

    International Nuclear Information System (INIS)

    We investigate whether it is possible to maintain the computational features of QED while avoiding some of its mathematical difficulties by formulating QFTs on Hilber bundles. This encounters two problems: 1) Haag's theorem persists, and 2) admissible fields do not generate motions on the base space. To do the latter, the coupling constant has to be a vector field upon the base space. (orig.)

  8. Assembly mechanism for nuclear fuel bundles

    International Nuclear Information System (INIS)

    This invention relates to an assembly mechanism for nuclear power reactor fuel bundles using a novel, simple and inexpensive means. The mechanism is readily operable remotely, avoids separable parts and is applicable to fuel assemblies in which the upper tie plate is rigidly mounted on the tie rods which hold it in place. (UK)

  9. Capacity efficiency of recovery request bundling

    DEFF Research Database (Denmark)

    Ruepp, Sarah Renée; Dittmann, Lars; Berger, Michael Stübert; Stidsen, Thomas Riis; Lagakos, Stephen; Perlovsky, Leonid; Jha, Manoi; Covaci, Brindusa; Zaharim, Azarni; Mastorakis, Nikos

    2010-01-01

    This paper presents a comparison of recovery methods in terms of capacity efficiency. In particular, a method where recovery requests are bundled towards the destination (Shortcut Span Protection) is evaluated against traditional recovery methods. Our simulation results show that Shortcut Span...

  10. Riemann Surfaces: Vector Bundles, Physics, and Dynamics

    DEFF Research Database (Denmark)

    Sikander, Shehryar

    the monodromy with respect to the pulled back connection. The formula for the representation includes a series with coefficients as iterated integrals. This series is closely related to the cyclotomic version of the Drinfel'd associator. The geodesic flow in the unit the tangent bundle of this Teichmueller...

  11. Investigation of conventional and Super-X divertor configurations of MAST Upgrade using SOLPS

    CERN Document Server

    Havlickova, E; Wischmeier, M; Fishpool, G; Morris, A W

    2014-01-01

    One of the first studies of MAST Upgrade divertor configurations with SOLPS5.0 are presented. We focus on understanding main prospects associated with the novel geometry of the Super-X divertor (SXD). This includes a discussion of the effect of magnetic flux expansion and volumetric power losses on the reduction of target power loads, the effect of divertor geometry on the divertor closure and distribution of neutral species and radiation in the divertor, the role of the connection length in broadening the target wetted area. A comparison in conditions typical for MAST inter-ELM H-mode plasmas confirms improved performance of the Super-X topology resulting in significantly better divertor closure with respect to neutrals (the atomic flux from the target increased by a factor of 6, but the atomic flux from the divertor to the upper SOL reduced by a factor of 2), increased radiation volume and increased total power loss (a factor of 2) and a reduction of target power loads through both magnetic flux expansion a...

  12. Direct measurement of divertor exhaust neo enrichment in DIII-D

    International Nuclear Information System (INIS)

    We report first direct measurements of divertor exhaust gas impurity enrichment, ηexh=(exhaust impurity concentration)divided-by(core impurity concentration), for both unpumped and D2 puff-with-divertor-pump conditions. The experiment was performed with neutral beam heated, ELMing H-mode, single-null diverted deuterium plasmas with matched core and exhaust parameters in the DIII-D tokamak. Neon gas impurity was puffed into the divertor. Neon density was measured in the exhaust by a specially modified Penning gauge and in the core by absolute charge exchange recombination spectroscopy. Neon particle accounting indicates that much of the puffed neon entered a temporary unmeasured reservoir, inferred to be the graphite divertor target, which makes direct measurements necessary to calculate divertor enrichments. D2 puff into the SOL (scrape-off layer) with pumping increased ηexh threefold over either unpumped conditions or D2 puff directly into the divertor with pumping. These results show that SOL flow plays an important role in divertor exhaust impurity enrichment

  13. A comprehensive 2-D divertor data set from DIII-D for edge theory validation

    International Nuclear Information System (INIS)

    A comprehensive set of experiments has been carried out on the DIII-D tokamak to measure the 2-D (R,Z) structure of the divertor plasma in a systematic way using new diagnostics. Measurements cover the divertor radially from inside the X-point to the outer target plate and vertically from the target plate to above the X-point. Identical, repeatable shots were made, each having radial sweeps of the X-point and divertor strike points, to allow complete plasma and radiation profile measurements. Data have been obtained in ohmic, L-mode, ELMing H-mode, and reversed BT operation (∇B drift away from the X-point). In addition, complete measurements were made of radiative divertor plasmas with a Partially Detached Divertor (PDD) induced by D2 injection and with a Radiating Mantle induced by Impurity injection (RMI) using neon and nitrogen. The data set includes first observations of the radial and poloidal profiles of the X-point, inner and outer leg plasmas in PDD and RMI radiative divertor operation. Preliminary data analysis shows that intrinsic impurities play a critical role in determining the SOL and divertor conditions

  14. Heat loads to divertor nearby components from secondary radiation evolved during plasma instabilities

    Science.gov (United States)

    Sizyuk, V.; Hassanein, A.

    2015-01-01

    A fundamental issue in tokamak operation related to power exhaust during plasma instabilities is the understanding of heat and particle transport from the core plasma into the scrape-off layer and to plasma-facing materials. During abnormal and disruptive operation in tokamaks, radiation transport processes play a critical role in divertor/edge-generated plasma dynamics and are very important in determining overall lifetimes of the divertor and nearby components. This is equivalent to or greater than the effect of the direct impact of escaped core plasma on the divertor plate. We have developed and implemented comprehensive enhanced physical and numerical models in the upgraded HEIGHTS package for simulating detailed photon and particle transport in the evolved edge plasma during various instabilities. The paper describes details of a newly developed 3D Monte Carlo radiation transport model, including optimization methods of generated plasma opacities in the full range of expected photon spectra. Response of the ITER divertor's nearby surfaces due to radiation from the divertor-developed plasma was simulated by using actual full 3D reactor design and magnetic configurations. We analyzed in detail the radiation emission spectra and compared the emission of both carbon and tungsten as divertor plate materials. The integrated 3D simulation predicted unexpectedly high damage risk to the open stainless steel legs of the dome structure in the current ITER design from the intense radiation during a disruption on the tungsten divertor plate.

  15. Divertor Experiments with MBI and Strong Gas Puffing on HL-2A

    Science.gov (United States)

    Duan, Xuru; Ding, Xuantong; Yang, Qingwei; Yan, Longwen; Yao, Lianghua; Hong, Wenyu; Xuan, Weimin; Liu, Dequan; Chen, Liaoyuan; Song, Xianming; Zhang, Jinhua; Cao, Zeng; Cui, Zhengying; Li, Wei; Liu, Yi; Pan, Yudong; Pan, Li; Zheng, Yinjia; Zhou, Yan; Mao, Weicheng; Liu, Yong; HL-2A Team

    2006-01-01

    In the HL-2A 2004 experiment campaign, pulsed molecular beam injection (MBI) and strong hydrogen gas puffing under the divertor configuration were used for gas fueling. The experimental results show that the MBI of hydrogen can reduce the heat flux to the divertor target plate. The electron temperature measured by the Langmuir probe array decreases significantly during the injection of the molecular beam whereas the electron density increases. This indicates that the plasma pressure near the target plates tends to be constant at a new equilibrium level. In the divertor plasmas with strong hydrogen gas puffing a high plasma density up to 4.4 × 1019 m-3 was achieved. In addition, a phenomenon similar to the partially detached divertor regime was observed, which is being studied in open divertor tokamaks such as DIII-D to reduce the peak heat flux on the target plates near the separatrix. After a strong gas puffing the electron temperature measured on the outer divertor target plate near the separatrix decreases till below 5 eV or even lower, but that of the farther outer divertor target plate does not change obviously; and the CIII and the Hα emissions at the plasma edge decrease as expected, but the Hα emission near the X-point increases. These results reflects some interesting characteristics, which needs to be studied by further modeling and experiments.

  16. Improvement of the divertor bolometer diagnostic in the ASDEX Upgrade tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Sehmer, Till; Meister, Hans; Bernert, Matthias; Koll, Juergen; Reimold, Felix; Wischmeier, Marco; Fantz, Ursel [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Collaboration: ASDEX Upgrade Team

    2015-05-01

    For future fusion devices such as ITER, the radiation balance in the divertor region will have a significant impact on the power exhaust balance. Therefore, scenarios with strongly localized radiation, like radiation in the high field side high density (HFSHD) region, X-Point radiation or radiation in the divertor legs during detachment, will be investigated in the next ASDEX Upgrade (AUG) operation campaign 2015. To obtain accurately the absolute divertor radiation out of these measurements, the AUG foil bolometer diagnostic system in the divertor region has been enhanced; two new cameras have been designed and manufactured. One will be mounted below the roof baffle and contains 28 lines of sight (LOS), which will observe the mentioned regions of particular physical interest. The second camera consists of 4 LOS and will be mounted at the high field side above the inner divertor nose. It will observe radiation arising from the X-Point region and from the outer divertor. The data will be analysed with a tomographic reconstruction algorithm to localize and quantify the divertor radiation.

  17. The WEST project: Current status of the ITER-like tungsten divertor

    International Nuclear Information System (INIS)

    Highlights: • We presented the ITER-like W components occurred for the WEST divertor. • The main features including key elements of the design were detailed. • The main results of studies investigating the integration constraints or issues were reported. • The WEST ITER-like divertor design reached a mature stage to enable the launching of the procurement phase. - Abstract: The WEST (W – for tungsten – Environment in Steady-state Tokamak) project is an upgrade of Tore Supra from a limiter based tokamak with carbon PFCs into an X-point divertor tokamak with full-tungsten armour while keeping its long discharge capability. The WEST project will primarily offer the key capability of testing for the first time the ITER technology in real plasma environment. In particular, the main divertor (i.e. the lower divertor) of the WEST project will be based on actively cooled tungsten monoblock components and will follow as closely as possible the design and the assembling technology, foreseen for the ITER divertor units. The current design of WEST ITER-like tungsten divertor has now reached a mature stage following the 2013 WEST Final Design Review. This paper presents the key elements of the design, reports the technological requirements and reviews the main design and integration issues

  18. In-pile test of Qinshan PWR fuel bundle

    International Nuclear Information System (INIS)

    In-pile test of Qinshan Nuclear Power Plant PWR fuel bundle has been conducted in HWRR HTHP Test loop at CIAE. The test fuel bundle was irradiated to an average burnup of 25000 Mwd/tU. The authors describe the structure of (3 x 3-2) test fuel bundle, structure of irradiation rig, fuel fabrication, irradiation conditions, power and fuel burnup. Some comments on the in-pile performance for fuel bundle, fuel rod and irradiation rig were made

  19. Holomorphic Vector Bundle on Hopf Manifolds with Abelian Fundamental Groups

    Institute of Scientific and Technical Information of China (English)

    Xiang Yu ZHOU; Wei Ming LIU

    2004-01-01

    Let X be a Hopf manifolds with an Abelian fundamental group. E is a holomorphic vector bundle of rank r with trivial pull-back to W = Cn - {0}. We prove the existence of a non-vanishing section of L(×) E for some line bundle on X and study the vector bundles filtration structure of E. These generalize the results of D. Mall about structure theorem of such a vector bundle E.

  20. Anatomic Double-Bundle Posterior Cruciate Ligament Reconstruction

    OpenAIRE

    Chahla, Jorge; Nitri, Marco; Civitarese, David; Dean, Chase S.; Moulton, Samuel G.; LaPrade, Robert F.

    2016-01-01

    The posterior cruciate ligament (PCL) is known to be the main posterior stabilizer of the knee. Anatomic single-bundle PCL reconstruction, focusing on reconstruction of the larger anterolateral bundle, is the most commonly performed procedure. Because of the residual posterior and rotational tibial instability after the single-bundle procedure and the inability to restore the normal knee kinematics, an anatomic double-bundle PCL reconstruction has been proposed in an effort to re-create the n...

  1. Existence of vector bundles and global resolutions for singular surfaces

    OpenAIRE

    Vezzosi, G; S. SCHROER

    2002-01-01

    Abstract- We prove two results about vector bundles on singular algebraic surfaces. First, on proper surfaces there are vector bundles of rank two with arbitrarily large second Chern number and fixed determinant. Second, on separated normal surfaces any coherent sheaf is the quotient of a vector bundle. As a consequence, for such surfaces the Quillen K-theory of vector bundles coincides with the Waldhausen K-theory of perfect complexes. Examples show that, on non-separated schemes, usually...

  2. CANFLEX - an advanced fuel bundle for CANDU

    International Nuclear Information System (INIS)

    The performance of CANDU pressurized heavy-water reactors, in terms of lifetime load factors, is excellent. More than 600 000 bundles containing natural-uranium fuel have been irradiated, with a low defect rate; reactor unavailability due to fuel incidents is typically zero. To maintain and improve CANDU's competitive position, Atomic Energy of Canada Limited (AECL) has an ongoing program comprising design, safety and availability improvements, advanced fuel concepts and schemes to reduce construction time. One key finding is that the introduction of slightly-enriched uranium (SEU, less than 1.5 wt% U-235 in U) offers immediate benefits for CANDU, in terms of fuelling and back-end disposal costs. The use of SEU places more demands on the fuel because of extended burnup, and an anticipated capability to load-follow also adds to the performance requirements. To ensure that the duty-cycle targets for SEU and load-following are achieved, AECL is developing a new fuel bundle, termed CANFLEX (CANdu FLEXible), where flexible refers to the versatility of the bundle with respect to operational and fuel-cycle options. Though the initial purpose of the new 43-element bundle is to introduce SEU into CANDU, CANFLEX is extremely versatile in its application, and is compatible with other fuel cycles of interest: natural uranium in existing CANDU reactors, recycled uranium and mixed-oxides from light-water reactors, and thoria-based fuels. Capability with a variety of fuel cycles is the key to future CANDU success in the international market. The improved performance of CANFLEX, particularly at high burnups, will ensure that the full economic benefits of advanced fuels cycles are achieved. A proof-tested CANFLEX bundle design will be available in 1993 for large-scale commercial-reactor demonstration

  3. Interplanetary Overlay Network Bundle Protocol Implementation

    Science.gov (United States)

    Burleigh, Scott C.

    2011-01-01

    The Interplanetary Overlay Network (ION) system's BP package, an implementation of the Delay-Tolerant Networking (DTN) Bundle Protocol (BP) and supporting services, has been specifically designed to be suitable for use on deep-space robotic vehicles. Although the ION BP implementation is unique in its use of zero-copy objects for high performance, and in its use of resource-sensitive rate control, it is fully interoperable with other implementations of the BP specification (Internet RFC 5050). The ION BP implementation is built using the same software infrastructure that underlies the implementation of the CCSDS (Consultative Committee for Space Data Systems) File Delivery Protocol (CFDP) built into the flight software of Deep Impact. It is designed to minimize resource consumption, while maximizing operational robustness. For example, no dynamic allocation of system memory is required. Like all the other ION packages, ION's BP implementation is designed to port readily between Linux and Solaris (for easy development and for ground system operations) and VxWorks (for flight systems operations). The exact same source code is exercised in both environments. Initially included in the ION BP implementations are the following: libraries of functions used in constructing bundle forwarders and convergence-layer (CL) input and output adapters; a simple prototype bundle forwarder and associated CL adapters designed to run over an IPbased local area network; administrative tools for managing a simple DTN infrastructure built from these components; a background daemon process that silently destroys bundles whose time-to-live intervals have expired; a library of functions exposed to applications, enabling them to issue and receive data encapsulated in DTN bundles; and some simple applications that can be used for system checkout and benchmarking.

  4. Experimental study of the topological aspect of the ergodic divertor in Tore-supra tokamak; Etude experimentale des aspects topologiques du divertor ergodique de Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Costanzo, L

    2001-10-01

    The control of power deposition onto plasma facing components in tokamaks is a determining factor for future thermonuclear fusion reactors. Plasma surface interaction can be performed using limiters or divertors. The ergodic divertor installed on Tore Supra is an atypical example of a magnetic divertor. It consists in applying a magnetic perturbation which establishes a particular topology of the plasma in contact with the wall (edge plasma). We carried out dedicated experiments in order to study parallel heat flux which strike the divertor neutralizers. This quantitative and qualitative analysis of heat flux as a function of experimental conditions allows to determine the profiles of power deposition along the neutralizers. The influence of plasma electron density, additional heating, impurities and injected gas was established. An experimental study of the sheath heat transmission factor {gamma} was carried out by correlating measurements made with Langmuir probes and infrared imaging. This study gave rise to a major conclusion: for ohmic discharges with deuterium injection and most of the time with helium, it was experimentally confirmed that {gamma}=7 in agreement with classical sheath theory. However, an increase of this factor with additional power has been shown. Detached plasma, which is an attractive regime in order to reduce the power deposition, requires an optimized control. A new measurement of the detachment onset has been developed. It is based on the variation of heat flux onto the plates derived from infrared measurements. A detachment cartography with the determination of a new 2D 'IR' Degree of Detachment was carried out allowing to locate the zone where the detachment starts. We can apply this concept both to other tokamaks such as JET and ITER. A comparison between the axisymmetric divertor and the ergodic divertor is also presented concerning the power deposition in the two configurations. Low heat flux with the ergodic divertor is a

  5. Compactifications of reductive groups as moduli stacks of bundles

    DEFF Research Database (Denmark)

    Martens, Johan; Thaddeus, Michael

    Let G be a reductive group. We introduce the moduli problem of "bundle chains" parametrizing framed principal G-bundles on chains of lines. Any fan supported in a Weyl chamber determines a stability condition on bundle chains. Its moduli stack provides an equivariant toroidal compactification of ...

  6. VECTOR BUNDLE, KILLING VECTOR FIELD AND PONTRYAGIN NUMBERS

    Institute of Scientific and Technical Information of China (English)

    周建伟

    1991-01-01

    Let E be a vector bundle over a compact Riemannian manifold M. We construct a natural metric on the bundle space E and discuss the relationship between the killing vector fields of E and M. Then we give a proof of the Bott-Baum-Cheeger Theorem for vector bundle E.

  7. Noncommutative principal torus bundles via parametrised strict deformation quantization

    OpenAIRE

    Hannabuss, Keith; Mathai, Varghese

    2009-01-01

    In this paper, we initiate the study of a parametrised version of Rieffel's strict deformation quantization. We apply it to give a classification of noncommutative principal torus bundles, in terms of parametrised strict deformation quantization of ordinary principal torus bundles. The paper also contains a putative definition of noncommutative non-principal torus bundles.

  8. Laser cutting for dismantling of PHWR fuel bundles

    International Nuclear Information System (INIS)

    Detailed investigation was carried out on laser cutting of zircaloy-2 PHWR fuel pin bundles. Initially, trials were done to standardize ten parameters for cutting of tie plates to which individual fuel pins are welded in a bundle. Using these parameters, the tie plates were cut into several pieces so that each fuel pin is individually separated out from the bundle. (author)

  9. Geometry of torus bundles in integrable Hamiltonian systems

    NARCIS (Netherlands)

    Lukina, Olga

    2008-01-01

    Thesis is concerned with global properties of Lagrangian bundles, i.e. symplectic n-torus bundles, as these occur in integrable Hamiltonian systems. It treats obstructions to triviality and concerns with classification of such bundles, as well as with manifestations of global invariants in real-worl

  10. Stability of Picard Bundle Over Moduli Space of Stable Vector Bundles of Rank Two Over a Curve

    Indian Academy of Sciences (India)

    Indranil Biswas; Tomás L Gómez

    2001-08-01

    Answering a question of [BV] it is proved that the Picard bundle on the moduli space of stable vector bundles of rank two, on a Riemann surface of genus at least three, with fixed determinant of odd degree is stable.

  11. Feasibility study of inside automatic welding system of cooling pipe of divertors for FER

    International Nuclear Information System (INIS)

    In order to replace divertors for FER, cooling pipes of divertors should be cut and welded since they are too long to be replaced with divertors via horizontal maintenance ports. An inside cutting and welding system is also required because of an accessibility to pipes. A combination of an inside disc-cutting machine and an inside TIG-welding machine has been proposed as a candidate of the systems. We have made tests to confirm possibility to weld pipes which were cut with the disc-cutting machine. Possibility of welding has been proven. The tests result is described in the paper. (orig.)

  12. Present status of linear plasma devices and issues on DEMO divertor design

    International Nuclear Information System (INIS)

    Construction of ITER has inspired discussion on the design of DEMO reactor. In the country, a team for constructing technological base of the fusion prototype reactor (a joint core team) has been organized in the nuclear science and technology committee thermonuclear-research working group of MEXT. In IAEA, the DEMO program workshop has been held. With these situations, we consider design issues of divertor in DEMO reactor and the role of linear plasma devices. First, the status of researches in linear plasma devices is reviewed. Next we explain R and D issues in the design of DEMO divertors and in the modeling and numerical simulation of divertors. (author)

  13. DESIGN OPTIMIZATION OF A FUSION REACTOR DIVERTOR COOLING FINGER DESIGN FOR DECREASED THERMO-MECHANICAL STRESSES

    OpenAIRE

    MATTEOLI CAMILLA; Martin, Oliver; SIMONOVSKI IGOR

    2012-01-01

    In this paper a design of a divertor cooling finger for a fusion reactor is looked at with the aim of reducing the thermo-mechanical stresses. The function of a divertor in a fusion reactor is to reduce the dilution of the plasma by removing alpha particles, helium and other impurities. In addition, this component has to remove approximately 15% of the total thermal power. The divertor is therefore exposed to a significant thermal load, and during the operation it has to be actively cooled, w...

  14. Plasma density control with ergodic divertor on Tore Supra

    International Nuclear Information System (INIS)

    Plasma density control on the tokamak Tore Supra is important for the optimization of every experimental scenario dealing with the improvement of plasma performances. Specific conditions are required both in the plasma bulk and at the edge. Within the framework of the present study, a magnetic configuration is used in the e plasma edge of Tore Supra: the ergodic divertor configuration. A magnetic perturbation which is resonant with the permanent field destroys the plasma confinement locally, opening the field lines onto the material components. They aim of the study is the characterization of the edge density in every relevant scenario for Tore Supra. The first part of this work is dedicated to density and temperature measurements by a series of fixed Langmuir probes located at the very edge of the plasma. Thanks to them, density regimes have been put in evidence during experiments where the volume averaged density e>, an usual control parameter of the plasma, was varied. The analysis of heat and particle transport through the plasma edge region explains the mechanisms leading to those regimes. The essential factor in our analysis is the dependence of the electron conductivity and ionization depth on temperature. While heat conduction governs the heat transport, the edge density varies linearly according to e>. Below a critical temperature, reached when the ion flux amplification at constant power density is large enough, a parallel temperature gradient appears leading to a density gradient in the opposite direction in order to maintain the pressure constant along the field lines. A high recycling regime is obtained and the edge density varies like e>3. The pressure conservation is no more satisfied during the detachment of the plasma, which is characterized by a high neutral density at low temperatures leading to a ion momentum loss by friction against the neutrals. The edge density drops in those conditions. These regimes are similar to those encountered with the

  15. Productivity and costs of slash bundling in Nordic conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kaerhae, K.; Vartiamaeki, T. [Metsaeteho Oy, P.O. Box 101, FI-00171 Helsinki (Finland)

    2006-12-15

    The number of slash bundlers and the volume of slash bundling have been rapidly increasing during the last few years in Finland. However, no comprehensive time or follow-up studies have been carried out on slash bundling technology in Finland or in any other country. Metsateho Oy carried out studies on the productivity and costs of slash bundling in different Nordic recovering conditions. The study methods included both time and follow-up studies. Data were collected during the summer and winter period primarily in Norway spruce (Picea abies L. Karst.) dominated clear cutting sites. The bundling techniques performed by different types of bundler (Fiberpac 370, Timberjack 1490D, Pika RS 2000, Valmet WoodPac) were studied. The average productivity of slash bundling was 18.1 bundles per operating (E{sub 15}, including delays shorter than 15min) hour with the Timberjack 1490D and Fiberpac 370 bundlers in the follow-up study. The operator of the slash bundler had the greatest effect on the productivity of bundling. The prerequisite for increased bundling volumes is a reduction in the costs of the most expensive sub-stage of the bundling supply chain, i.e. bundling itself. This requires improved recovery conditions at bundling sites, increased bundling productivity, larger sized bundles, and the execution of bundling operations in two work shifts using an efficient bundler and effective operator working methods. Implementation of these development measures will bring the bundling supply chain up to a speed that makes it the most competitive supply chain for forest chips in terms of total supply costs for long-distance transportation distances of more than 60km. (author)

  16. Bundling of harvesting residues and whole-trees and the treatment of bundles; Hakkuutaehteiden ja kokopuiden niputus ja nippujen kaesittely

    Energy Technology Data Exchange (ETDEWEB)

    Kaipainen, H.; Seppaenen, V.; Rinne, S.

    1996-12-31

    The conditions on which the bundling of the harvesting residues from spruce regeneration fellings would become profitable were studied. The calculations showed that one of the most important features was sufficient compaction of the bundle, so that the portion of the wood in the unit volume of the bundle has to be more than 40 %. The tests showed that the timber grab loader of farm tractor was insufficient for production of dense bundles. The feeding and compression device of the prototype bundler was constructed in the research and with this device the required density was obtained.The rate of compaction of the dry spruce felling residues was about 40 % and that of the fresh residues was more than 50 %. The comparison between the bundles showed that the calorific value of the fresh bundle per unit volume was nearly 30 % higher than that of the dry bundle. This means that the treatment of the bundles should be done of fresh felling residues. Drying of the bundles succeeded well, and the crushing and chipping tests showed that the processing of the bundles at the plant is possible. The treatability of the bundles was also excellent. By using the prototype, developed in the research, it was possible to produce a bundle of the fresh spruce harvesting residues, the diameter of which was about 50 cm and the length about 3 m, and the rate of compaction over 50 %. By these values the reduction target of the costs is obtainable

  17. Plasma performance control during ergodic divertor experiments in Tore Supra

    International Nuclear Information System (INIS)

    Ohmic plasma particle confinement times are controlled during magnetic perturbation and stochastic boundary layer experiments in TORE SUPRA with small currents in the ergodic divertor coils. Particle confinement may be improved or degraded depending on the plasma configuration and base parameters used. The magnitude of these steady state confinement changes are controlled by changing led and the base plasma parameters. Plasma confinement changes manifest either density increase with a reduction in the wall fueling flux or density decreases with an increase in the fueling flux depending on the geometric configuration. In addition, the effective thermal insulation of the boundary layer is controlled. Impurity and radiated power profiles are readily modified in the boundary layer

  18. Erosion/redeposition analysis of the DIII-D divertor

    International Nuclear Information System (INIS)

    Carbon and tungsten sputtering and transport in the DIII-D divertor is analyzed with the impurity transport codes REDEP and WBC. Analysis is carried out for a recent DiMES experiment in which a carbon sample with a tungsten marker in the center was exposed to six well controlled ELM-free plasma discharges. WBC analysis predicts a high rate of ionization of tungsten neutrals within the sheath and subsequent redeposition on the DiMES sample. Qualitative comparison of the tungsten redeposited flux agrees well with measurements. REDEP analysis of net carbon erosion shows a factor of 2-3 agreement with measured data on the outboard side of DiMES and poor agreement on the inboard side

  19. Manufacturing and testing of a HETS module for DEMO divertor

    Energy Technology Data Exchange (ETDEWEB)

    Visca, Eliseo, E-mail: eliseo.visca@enea.it [ENEA - Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, 00044 Frascati (Italy); Agostini, Pietro; Crescenzi, Fabio; Malavasi, A.; Pizzuto, Aldo; Rossi, Paolo; Storai, Sandro; Utili, Marco [ENEA - Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, 00044 Frascati (Italy)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer The HETS (high-efficiency thermal shield) concept, initially developed by ENEA for water, has been adapted for use with He as coolant. Black-Right-Pointing-Pointer This DEMO divertor concept is based on elements joined in series and protected by a hemispheric dome. Black-Right-Pointing-Pointer It has been calculated to be capable of sustaining an incident heat flux of 10 MW/m{sup 2} when operating at 10 MPa, an inlet He temperature of 600 Degree-Sign C, and an outlet temperature of 800 Degree-Sign C. Black-Right-Pointing-Pointer The activity is focused on the manufacturing of a single HETS module with W armor and on its thermal-hydraulic testing. Black-Right-Pointing-Pointer A CFD analysis by ANSYS-CFX was performed in order to predict the thermal-mechanical behavior of the module and a final comparison with the experimental data is required to validate the CFD results. - Abstract: The development of a divertor concept for fusion power plants that is able to grant efficient recovery and conversion of the considerable fraction ({approx}15%) of the total fusion thermal power incident is deemed to be an urgent task to meet in the EU Fast Track scenario. The He-cooled conceptual divertor design is one of the possible candidates. Helium cooling offers several advantages including chemical and neutronic inertness and the ability to operate at higher temperatures and lower pressures than those required for water cooling. The HETS (high-efficiency thermal shield) concept, initially developed by ENEA for water, has been adapted for use with He as coolant. This DEMO divertor concept is based on elements joined in series and protected by a hemispheric dome; it allows an increase of thermal exchange coefficient both for high speed of gas and for 'jet impingement' effects of gas coming out from the internal side of hemispheric dome. It has been calculated to be capable of sustaining an incident heat flux of 10 MW/m{sup 2} when

  20. Development of plasma control system for divertor configuration on QUEST

    International Nuclear Information System (INIS)

    A plasma control system to sustain divertor configurations is developed on QUEST (Q-shu university experiment with steady-state spherical tokamak). Magnetic fluxes are numerically integrated at 100 kHz using FPGA (Field-Programmable Gate Array) modules and transferred to a main calculation loop at 4 kHz. With these signals, plasma shapes are identified in real time at 2 kHz under the assumption that the plasma current can be represented as one filament current. This calculation is done in another calculation loop in parallel by taking advantage of a multi-core processor of the plasma control system. The inside and outside plasma edge positions are controlled to their target positions using PID (proportional-integral-derivative) control loops. Whereas the outside edge position can not be controlled by the outer PF coil current, the inside edge position can be controlled by the inner PF coil current

  1. Performance of JT-60SA divertor Thomson scattering diagnostics

    International Nuclear Information System (INIS)

    For the satellite tokamak JT-60 Super Advanced (JT-60SA), a divertor Thomson scattering measurement system is planning to be installed. In this study, we improved the design of the collection optics based on the previous one, in which it was found that the solid angle of the collection optics became very small, mainly because of poor accessibility to the measurement region. By improvement, the solid angle was increased by up to approximately five times. To accurately assess the measurement performance, background noise was assessed using the plasma parameters in two typical discharges in JT-60SA calculated from the SONIC code. Moreover, the influence of the reflection of bremsstrahlung radiation by the wall is simulated by using a ray tracing simulation. The errors in the temperature and the density are assessed based on the simulation results for three typical field of views

  2. Performance of JT-60SA divertor Thomson scattering diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Kajita, Shin, E-mail: kajita.shin@nagoya-u.jp [Nagoya University, Nagoya 464-8603 (Japan); Hatae, Takaki; Tojo, Hiroshi; Hamano, Takashi; Shimizu, Katsuhiro; Kawashima, Hisato [Japan Atomic Energy Agency, Naka, Ibaraki 801-1 (Japan); Enokuchi, Akito [Genesia Co., Mitaka, Tokyo 181-0013 (Japan)

    2015-08-15

    For the satellite tokamak JT-60 Super Advanced (JT-60SA), a divertor Thomson scattering measurement system is planning to be installed. In this study, we improved the design of the collection optics based on the previous one, in which it was found that the solid angle of the collection optics became very small, mainly because of poor accessibility to the measurement region. By improvement, the solid angle was increased by up to approximately five times. To accurately assess the measurement performance, background noise was assessed using the plasma parameters in two typical discharges in JT-60SA calculated from the SONIC code. Moreover, the influence of the reflection of bremsstrahlung radiation by the wall is simulated by using a ray tracing simulation. The errors in the temperature and the density are assessed based on the simulation results for three typical field of views.

  3. Simulations of carbon sputtering in fusion reactor divertor plates

    International Nuclear Information System (INIS)

    The interaction of edge plasma with material surfaces raises key issues for the viability of the International Thermonuclear Reactor (ITER) and future fusion reactors, including heat-flux limits, net material erosion, and impurity production. After exposure of the graphite divertor plate to the plasma in a fusion device, an amorphous C/H layer forms. This layer contains 20-30 atomic percent D/T bonded to C. Subsequent D/T impingement on this layer produces a variety of hydrocarbons that are sputtered back into the sheath region. We present molecular dynamics (MD) simulations of D/T impacts on amorphous carbon layer as a function of ion energy and orientation, using the AIREBO potential. In particular, energies are varied between 10 and 150 eV to transition from chemical to physical sputtering. These results are used to quantify yield, hydrocarbon composition and eventual plasma contamination

  4. Comparison of ASSERT subchannel code with Marviken bundle data

    International Nuclear Information System (INIS)

    In this paper ASSERT predictions are compared with the Marviken 6-rod bundle and 36+1 rod bundle. The predictions are presented for two experiments in the 6-rod bundle and four experiments in the 36+1 rod bundle. For low inlet subcooling, the void predictions are in good agreement with the experimental data. For high inlet subcooling, however, the agreement is not as good. This is attributed to the fact that in the high inlet subcooling experiments, single phase turbulent mixing plays a more important role in determining flow conditions in the bundle

  5. Multiwalled carbon nanotube reinforced biomimetic bundled gel fibres.

    Science.gov (United States)

    Kim, Young-Jin; Yamamoto, Seiichiro; Takahashi, Haruko; Sasaki, Naruo; Matsunaga, Yukiko T

    2016-08-19

    This work describes the fabrication and characterization of hydroxypropyl cellulose (HPC)-based biomimetic bundled gel fibres. The bundled gel fibres were reinforced with multiwalled carbon nanotubes (MWCNTs). A phase-separated aqueous solution with MWCNT and HPC was transformed into a bundled fibrous structure after being injected into a co-flow microfluidic device and applying the sheath flow. The resulting MWCNT-bundled gel fibres consist of multiple parallel microfibres. The mechanical and electrical properties of MWCNT-bundled gel fibres were improved and their potential for tissue engineering applications as a cell scaffold was demonstrated. PMID:27200527

  6. Effectiveness of Hair Bundle Motility as the Cochlear Amplifier

    OpenAIRE

    Sul, Bora; Iwasa, Kuni H.

    2009-01-01

    The effectiveness of hair bundle motility in mammalian and avian ears is studied by examining energy balance for a small sinusoidal displacement of the hair bundle. The condition that the energy generated by a hair bundle must be greater than energy loss due to the shear in the subtectorial gap per hair bundle leads to a limiting frequency that can be supported by hair-bundle motility. Limiting frequencies are obtained for two motile mechanisms for fast adaptation, the channel re-closure mode...

  7. Anatomic Double-Bundle Posterior Cruciate Ligament Reconstruction.

    Science.gov (United States)

    Chahla, Jorge; Nitri, Marco; Civitarese, David; Dean, Chase S; Moulton, Samuel G; LaPrade, Robert F

    2016-02-01

    The posterior cruciate ligament (PCL) is known to be the main posterior stabilizer of the knee. Anatomic single-bundle PCL reconstruction, focusing on reconstruction of the larger anterolateral bundle, is the most commonly performed procedure. Because of the residual posterior and rotational tibial instability after the single-bundle procedure and the inability to restore the normal knee kinematics, an anatomic double-bundle PCL reconstruction has been proposed in an effort to re-create the native PCL footprint more closely and to restore normal knee kinematics. We detail our technique for an anatomic double-bundle PCL reconstruction using Achilles and anterior tibialis tendon allografts. PMID:27284530

  8. The turbulent flow in rod bundles

    International Nuclear Information System (INIS)

    Experimental studies have shown that the axial and azimuthal turbulence intensities in the gap regions of rod bundles increase strongly with decreasing rod spacing; the fluctuating velocities in the axial and azimuthal directions have a quasi-periodic behaviour. To determine the origin of this phenomenon, an its characteristics as a function of the geometry and the Reynolds number, an experimental investigation was performed on the turbulent in several rod bundles with different aspect ratios (P/D, W/D). Hot-wires and microsphones were used for the measurements of velocity and wall pressure fluctuations. The data were evaluated to obtain spectra as well as auto and cross correlations. Based on the results, a phenomenological model is presented to explain this phenomenon. By means of the model, the mass exchange between neighbouring subchannels is explained

  9. Venereau polynomials and related fiber bundles

    OpenAIRE

    Kaliman, Shulim; ZAIDENBERG, MIKHAIL

    2003-01-01

    The Venereau polynomials v-n:=y+x^n(xz+y(yu+z^2)), n>= 1, on A4 have all fibers isomorphic to the affine space A3. Moreover, for all n>= 1 the map (v-n, x) : A4 -> A2 yields a flat family of affine planes over A2. In the present note we show that over the punctured plane A2\\0, this family is a fiber bundle. This bundle is trivial if and only if v-n is a variable of the ring C[x][y,z,u] over C[x]. It is an open question whether v1 and v2 are variables of the polynomial ring C[x,y,z,u]. S. Vene...

  10. A fibre bundle formulation of quantum geometry

    International Nuclear Information System (INIS)

    Quantum geometries whose points are stochastic and serve as seats for quantum space-time excitons are formulated as fibre bundles over base spaces of mean values with a Minkowski or general relativistic structure. The fibres contain the proper wave functions of all exciton states in a given model. The notion of covariance and propagation in quantum space-times constituting such fibre bundles is investigated. Maxwell and Yang-Mills gauge degrees of freedom are introduced by appropriately enlarging the structure group, which in all cases contains phase-space representations of the Poincare group corresponding to the exciton wave function sample space specific to a given model. It is shown that these formulations give rise in a natural manner to certain realizations of the relativistic canonical commutation relations in terms of covariant derivatives involving internal as well as external degrees of freedom of space-time excitons

  11. Heterotic String Compactification and New Vector Bundles

    Science.gov (United States)

    Lin, Hai; Wu, Baosen; Yau, Shing-Tung

    2016-07-01

    We propose a construction of Kähler and non-Kähler Calabi-Yau manifolds by branched double covers of twistor spaces. In this construction we use the twistor spaces of four-manifolds with self-dual conformal structures, with the examples of connected sum of n {mathbb{P}2}s. We also construct K3-fibered Calabi-Yau manifolds from the branched double covers of the blow-ups of the twistor spaces. These manifolds can be used in heterotic string compactifications to four dimensions. We also construct stable and polystable vector bundles. Some classes of these vector bundles can give rise to supersymmetric grand unified models with three generations of quarks and leptons in four dimensions.

  12. Rod bundle burnout data and correlation comparisons

    International Nuclear Information System (INIS)

    Rod bundle burnout data from 30 steady-state and 3 transient tests were obtained from experiments performed in the Thermal Hydraulic Test Facility at the Oak Ridge National Laboratory. The tests covered a parameter range relevant to intact core reactor accidents ranging from large break to small break loss-ofcoolant conditions. Instrumentation within the 64-rod test section indicated that burnout occurred over an axial range within the bundle. The distance from the point where the first dry rod was detected to the point where all rods were dry was up to 60 cm in some of the tests. The burnout data should prove useful in developing new correlations for use in reactor thermalhydraulic codes. Evaluation of several existing critical heat flux correlations using the data show that three correlations, the Barnett, Bowring, and Katto correlations, perform similarly and correlate the data better than the Biasi correlation

  13. Client Provider Collaboration for Service Bundling

    Directory of Open Access Journals (Sweden)

    LETIA, I. A.

    2008-04-01

    Full Text Available The key requirement for a service industry organization to reach competitive advantages through product diversification is the existence of a well defined method for building service bundles. Based on the idea that the quality of a service or its value is given by the difference between expectations and perceptions, we draw the main components of a frame that aims to support the client and the provider agent in an active collaboration meant to co-create service bundles. Following e3-value model, we structure the supporting knowledge around the relation between needs and satisfying services. We deal with different perspectives about quality through an ontological extension of Value Based Argumentation. The dialog between the client and the provider takes the form of a persuasion whose dynamic object is the current best configuration. Our approach for building service packages is a demand driven approach, allowing progressive disclosure of private knowledge.

  14. Radiological evidence for the triple bundle anterior cruciate ligament.

    Science.gov (United States)

    MacKay, James W; Whitehead, Harry; Toms, Andoni P

    2014-10-01

    The anterior cruciate ligament (ACL) has traditionally been described as having two bundles--one anteromedial and one posterolateral. This has been challenged by studies proposing the existence of a third, intermediate, bundle with distinct functional significance, an arrangement that has been described in a number of domesticated animal species. No radiological evidence for the triple bundle ACL has previously been described. A prevalence study was carried out on 73 consecutive human knee magnetic resonance (MR) studies to determine the number of visible bundles, excluding individuals with a history of ACL injury or mucoid degeneration. A triple bundle ACL was demonstrated in 15 out of 73 human knees (20.5%, 95% confidence interval 12.9-31.2%). This is the first radiological description of the human triple bundle ACL. There was MR imaging evidence of a triple bundle ACL in approximately one fifth of human knees in this study. PMID:24890455

  15. Correlation between hydrogen isotope profiles and surface structure of divertor tiles in JT-60U

    International Nuclear Information System (INIS)

    The present paper is devoted to depth profiles by secondary ion mass spectroscopy (SIMS) of hydrogen/deuterium in tiles taken from the dome unit area of JT-60U. This information is correlated with surface features, particularly from the aspect of erosion and deposition, determined by scanning electron microscope (SEM) and X-ray photoelectron spectroscopy (XPS). The outer divertor-facing surface was mostly covered by re-deposited layers a maximum of 10 μm thick, while the inner divertor-facing side was eroded. The deposition profile is opposite to the observation for the divertor area in most tokamaks that the outer divertor side is eroded, while the inner deposited. However, H + D retention was higher for the deposited layers than that for the eroded area. Nevertheless, hydrogen retention seems very small and showed no appreciable effects on C1s spectra of XPS compared to the constituent elements boron and oxygen

  16. Erosion and deposition on JET divertor and limiter tiles during the experimental campaigns 2005–2009

    Energy Technology Data Exchange (ETDEWEB)

    Krat, S., E-mail: stepan.krat@gmail.com [National Research Nuclear University “MEPhI”, Kashirskoe Road 31, 115409 Moscow (Russian Federation); Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany); Coad, J.P. [Culham Science Centre, EURATOM/UKAEA – Fusion Association, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Gasparyan, Yu. [National Research Nuclear University “MEPhI”, Kashirskoe Road 31, 115409 Moscow (Russian Federation); Hakola, A.; Likonen, J. [Association EURATOM-Tekes, Technical Research Centre of Finland, PO Box 1000, FI-02044 VTT (Finland); Mayer, M. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany); Pisarev, A. [National Research Nuclear University “MEPhI”, Kashirskoe Road 31, 115409 Moscow (Russian Federation); Widdowson, A. [Culham Science Centre, EURATOM/UKAEA – Fusion Association, Abingdon, Oxfordshire OX14 3DB (United Kingdom)

    2013-07-15

    Erosion from and deposition on JET divertor tiles used during the 2007–2009 campaign and on inner wall guard limiter (IWGL) tiles used during 2005–2009 are studied. The tungsten coating on the divertor tiles was mostly intact with the largest erosion ∼30% in a small local area. Locally high erosion areas were observed on the load bearing divertor tile 5 and on the horizontal surface of the divertor tile 8. The IWGL tiles show a complicated distribution of erosion and deposition areas. The total amount of carbon deposited on the all IWGL tiles during the campaign 2005–2009 is estimated to be 65 g. The density of carbon deposits is estimated to be 0.67–0.83 g/cm{sup 3}.

  17. Comparison of JET main chamber erosion with dust collected in the divertor

    Energy Technology Data Exchange (ETDEWEB)

    Widdowson, A., E-mail: anna.widdowson@ccfe.ac.uk [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Ayres, C.F.; Booth, S. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Coad, J.P.; Hakola, A. [Association EURATOM-TEKES, VTT, PO Box 1000, 02044 VTT, Espoo (Finland); Heinola, K. [Association EURATOM-TEKES, University of Helsinki, PO Box 64, 00560 Helsinki (Finland); Ivanova, D. [Laboratory, Royal Institute of Technology, Association EURATOM-VR, 100 44 Stockholm (Sweden); Koivuranta, S.; Likonen, J. [Association EURATOM-TEKES, VTT, PO Box 1000, 02044 VTT, Espoo (Finland); Mayer, M. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Stamp, M. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)

    2013-07-15

    A complete global balance for carbon in JET requires knowledge of the net erosion in the main chamber, net deposition in the divertor and the amount of dust and flakes collecting in the divertor region. This paper describes a number of measurements on aspects of this global picture. Profiler measurements and cross section microscopy on tiles that were removed in the 2009 JET intervention are used to evaluate the net erosion in the main chamber and net deposition in the divertor. In addition the mass of dust and flakes collected from the JET divertor during the same intervention is also reported and included as part of the balance. Spectroscopic measurements of carbon erosion from the main chamber are presented and compared with the erosion measurements for the main chamber.

  18. A full tungsten divertor for ITER: Physics issues and design status

    Energy Technology Data Exchange (ETDEWEB)

    Pitts, R.A., E-mail: richard.pitts@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Carpentier, S.; Escourbiac, F.; Hirai, T.; Komarov, V.; Lisgo, S.; Kukushkin, A.S.; Loarte, A.; Merola, M.; Sashala Naik, A.; Mitteau, R.; Sugihara, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Bazylev, B. [Karlsruhe Institute of Technology, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany); Stangeby, P.C. [University of Toronto Institute Aerospace Studies, Ontario, Canada M3H 5T6 (Canada)

    2013-07-15

    Budget restrictions have forced the ITER Organization to reconsider the baseline divertor strategy, in which operations would begin with carbon (C) in the high heat flux regions, changing out to a full-tungsten (W) variant before the first nuclear campaigns. Substantial cost reductions can be achieved if one of these two divertors is eliminated. The new strategy implies not only that ITER would start-up on a full-W divertor, but that this component should survive until well into the nuclear phase. This paper considers the risks engendered by such an approach with regard to known W plasma-material interaction issues and briefly presents the current status of a possible full-W divertor design.

  19. A full tungsten divertor for ITER: Physics issues and design status

    International Nuclear Information System (INIS)

    Budget restrictions have forced the ITER Organization to reconsider the baseline divertor strategy, in which operations would begin with carbon (C) in the high heat flux regions, changing out to a full-tungsten (W) variant before the first nuclear campaigns. Substantial cost reductions can be achieved if one of these two divertors is eliminated. The new strategy implies not only that ITER would start-up on a full-W divertor, but that this component should survive until well into the nuclear phase. This paper considers the risks engendered by such an approach with regard to known W plasma-material interaction issues and briefly presents the current status of a possible full-W divertor design

  20. Design of a diagnostic residual gas analyzer for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Klepper, C Christopher [ORNL; Biewer, T. M. [Oak Ridge National Laboratory (ORNL); Graves, Van B [ORNL; Andrew, P. [EURATOM / UKAEA, UK; Lukens, P. C. [United States ITER Project Office; Marcus, Chris [ORNL; Shimada, M. [ITER Organization, Saint Paul Lez Durance, France; Hughes, S. [ITER Organization, Cadarache, France; Boussier, B. [ITER Organization, Saint Paul Lez Durance, France; Johnson, D. W. [Princeton Plasma Physics Laboratory (PPPL); Gardner, W. L. [United States ITER Project Office; Hillis, D. L. [Oak Ridge National Laboratory (ORNL); Vayakis, G. [ITER Organization, Cadarache, France; Vayakis, G. [ITER Organization, Saint Paul Lez Durance, France; Walsh, M. [ITER Organization, Saint Paul Lez Durance, France

    2015-01-01

    One of the ITER diagnostics having reached an advanced design stage is a diagnostic RGA for the divertor, i.e. residual gas analysis system for the ITER divertor, which is intended to sample the divertor pumping duct region during the plasma pulse and to have a response time compatible with plasma particle and impurity lifetimes in the divertor region. Main emphasis is placed on helium (He) concentration in the ducts, as well as the relative concentration between the hydrogen isotopes (H2, D2, T2). Measurement of the concentration of radiative gases, such as neon (Ne) and nitrogen (N2), is also intended. Numerical modeling of the gas flow from the sampled region to the cluster of analysis sensors, through a long (~8m long, ~110mm diameter) sampling pipe terminating in a pressure reducing orifice, confirm that the desired response time (~1s for He or D2) is achieved with the present design.

  1. Development of microwave interferometer system for divertor simulation experiments in GAMMA 10/PDX

    Science.gov (United States)

    Kohagura, J.; Wang, X.; Kanno, S.; Yoshikawa, M.; Kuwahara, D.; Nagayama, Y.; Shima, Y.; Chikatsu, M.; Nojiri, K.; Sakamoto, M.; Imai, T.; Nakashima, Y.; Mase, A.

    2015-12-01

    Microwave interferometer has newly been installed on GAMMA 10/PDX for divertor simulation study. A divertor simulation experimental module (D-module) is used to investigate the physics of divertor in the end-cell of GAMMA 10/PDX where an open magnetic field configuration is formed. D-module has a rectangular chamber with an inlet aperture. Two tungsten target plates are mounted in V-shape inside the chamber. In order to develop understandings of divertor simulation experiments the microwave interferometer using heterodyne scheme and a 1D horn-antenna mixer array (HMA) is applied to obtain electron density and density distribution inside the V-shaped target plates. Line-averaged electron density distributions inside D-module are first observed in H2 gas injection experiments.

  2. Current state-of-the-art manufacturing technology for He-cooled divertor finger

    International Nuclear Information System (INIS)

    A divertor concept for DEMO has been investigated at Karlsruhe Institute of Technology (KIT) which has to withstand a heat flux of 10 MW/m2. The design utilizes small finger module composed of a small tungsten tile brazed on a thimble made from tungsten alloy. The divertor finger is cooled by helium jet impingement at 10 MPa and 600 deg. C. The subject of this paper is technological studies on machining and braze joining the divertor components. Goal of this task, which is considered an important R and D issue, is to find out appropriate manufacturing methods to ensure high functionality and high reliability of the divertor as well as to meet the economic aspect. One of the major requirements for manufacturing is micro-crack-free surface of tungsten parts, since crack propagations in tungsten were observed in the previous high-heat-flux tests at Efremov. Different manufacturing methods and the corresponding results are discussed in the following report.

  3. Effects of magnetic configuration on divertor power and particle deposition for long pulse operation in EAST

    International Nuclear Information System (INIS)

    The magnetic configuration exhibits a strong influence on the dynamics of Edge Localized Modes (ELMs), as demonstrated in the EAST superconducting tokamak. We find that poloidal drifts play an important role in particle deposition during the ELMs, leading to a strong up/down asymmetry in the double null divertor configuration, favoring the upper divertor for normal toroidal field, Bt, i.e., with the ion ∇B drift towards the bottom, while the heat flux distribution appears to be rather uniform during ELMs. These observations are well reproduced by the boundary plasma turbulence code, BOUT++. As divertor pumping was only available at the bottom, the preferential particle flow towards the bottom divertor associated with reverse Bt led to a preferred scenario for long pulse operation in EAST

  4. ATHENA calculation model for the ITER-FEAT divertor cooling system. Final report with updates

    International Nuclear Information System (INIS)

    An ATHENA model of the ITER-FEAT divertor cooling system has been developed for the purpose of calculating and evaluating consequences of different thermal-hydraulic accidents as specified in the Accident Analysis Specifications for the ITER-FEAT Generic Site Safety Report. The model is able to assess situations for a variety of conceivable operational transients from small flow disturbances to more critical conditions such as total blackout caused by a loss of offsite and emergency power. The main objective for analyzing this type of scenarios is to determine margins against jeopardizing the integrity of the divertor cooling system components and pipings. The model of the divertor primary heat transport system encompasses the divertor cassettes, the port limiter systems, the pressurizer, the heat exchanger and all feed and return pipes of these components. The development was pursued according to practices and procedures outlined in the ATHENA code manuals using available modelling components such as volumes, junctions, heat structures and process controls

  5. Comparison of JET main chamber erosion with dust collected in the divertor

    CERN Document Server

    Widdowson, A; Booth, S; Coad, J P; Hakola, A; Heinola, K; Ivanova, S; Koivuranta, S; Likonen, J; Mayer, M; Stamp, M; Contributors, JET-EFDA

    2013-01-01

    A complete global balance for carbon in JET requires knowledge of the net erosion in the main chamber, net deposition in the divertor and the amount of dust and flakes collecting in the divertor region. This paper describes a number of measurements on aspects of this global picture. Profiler measurements and cross section microscopy on tiles that were removed in the 2009 JET intervention are used to evaluate the net erosion in the main chamber and net deposition in the divertor. In addition the mass of dust and flakes collected from the JET divertor during the same intervention is also reported and included as part of the balance. Spectroscopic measurements of carbon erosion from the main chamber are presented and compared with the erosion measurements for the main chamber.

  6. CHF and flow instability in rod bundles

    International Nuclear Information System (INIS)

    Data for two very different rod bundles have been analyzed using a new CHF correlation and a crude, but simple, subchannel analysis. The CHF correlation was developed for round uniform tubes and has been shown to accurately predict CHF in nonuniform tubes. The first set of data was for a KWU rod bundle (37 rods) with a heated length of 3.00 m and an O.D. (outside diameter) of 12.9 mm over a range of pressure 70 to 150 bar in upflow. The second set of data was for a 5 x 5 TRIGA rod bundle with a heated length of 0.559 m and 13.75 mm O.D. over a range of pressure of 0.945 to 1.372 bar in downflow. In contrast to the KWU data, the correlation greatly over estimates the CHF values for the TRIGA data. The TRIGA CHF data correlate very well with the variable qsat assuming no mixing, qc,exp = 0.955qsat (stdev = 9.87%). This result strongly suggests that these instabilities, which resulted immediately in CHF, are triggered by the Onset of Flow Instability (OFI) rather than CHF. The wide spread in rod power factors, the low pressure, and the downflow condition all contribute to promoting this type of instability (Ledinegg). The crude subchannel analysis has been compared with calculations of exit conditions of the hot channel using COBRA code. The agreement is fair when the homogeneous equilibrium model is used in the COBRA code. This is expected since the exit of the hot channel is always subcooled. Using Zuber's, along with other, void fraction relations in COBRA yields much lower exit velocities and high positive exit qualities, and, in some cases, convergence difficulties arise. The facts indicate that the bundle has already past the OFI point: which is possible since no CHF calculation was made in these COBRA analyses. (J.P.N)

  7. Interstitial He and Ne in Nanotube Bundles

    OpenAIRE

    Stan, G.; Crespi, V. H.; Cole, M. W.; Boninsegni, M.

    1998-01-01

    We explore the properties of atoms confined to the interstitial regions within a carbon nanotube bundle. We find that He and Ne atoms are of ideal size for physisorption interactions, so that their binding energies are much greater there than on planar surfaces of any known material. Hence high density phases exist at even small vapor pressure. There can result extraordinary anisotropic liquids or crystalline phases, depending on the magnitude of the corrugation within the interstitial channels.

  8. Effective freeness of adjoint line bundles

    OpenAIRE

    Heier, Gordon

    2001-01-01

    In this note we establish a new Fujita-type effective bound for the base point freeness of adjoint line bundles on a compact complex projective manifold of complex dimension $n$. The bound we obtain (approximately) differs from the linear bound conjectured by Fujita only by a factor of the cube root of $n$. As an application, a new effective statement for pluricanonical embeddings is derived.

  9. On Complex Supermanifolds with Trivial Canonical Bundle

    CERN Document Server

    Groeger, Josua

    2016-01-01

    We give an algebraic characterisation for the triviality of the canonical bundle of a complex supermanifold in terms of a certain Batalin-Vilkovisky superalgebra structure. As an application, we study the Calabi-Yau case, in which an explicit formula in terms of the Levi-Civita connection is achieved. Our methods include the use of complex integral forms and the recently developed theory of superholonomy.

  10. Telescope sipping - pinpointing leaking fuel bundles

    International Nuclear Information System (INIS)

    Given the top priority operators of nuclear power plants assign to safety, even the slightest sign of damage to the fuel assemblies has to be carefully monitored and analyzed. The detection of leaking fuel bundles also plays an important role in ensuring good availability and economy for the plants. ABB Atom has developed a new, highly accurate method, called 'telescope sipping', for identifying defective fuel assemblies. (orig.)

  11. Imperfect Bundling In Public-Private Partnerships

    OpenAIRE

    Luciano Greco

    2012-01-01

    The economic literature on PPPs has generally overlooked agency problems within private consortia. We provide a first contribution in this direction, relying on a simple incomplete contracts framework where a Builder and an Operator set up a Special Purpose Vehicle (SPV) to carry out a contract with the government. Because of incomplete contracts, the bundling of tasks is imperfect, and the SPV ownership structure is the main tool to regulate the power of private incentives. The scope for wel...

  12. Using Advanced Fuel Bundles in CANDU Reactors

    International Nuclear Information System (INIS)

    Improving the exit fuel burnup in CANDU reactors was a long-time challenge for both bundle designers and performance analysts. Therefore, the 43-element design together with several fuel compositions was studied, in the aim of assessing new reliable, economic and proliferation-resistant solutions. Recovered Uranium (RU) fuel is intended to be used in CANDU reactors, given the important amount of slightly enriched Uranium (~0.96% w/o U235) that might be provided by the spent LWR fuel recovery plants. Though this fuel has a far too small U235 enrichment to be used in LWR's, it can be still used to fuel CANDU reactors. Plutonium based mixtures are also considered, with both natural and depleted Uranium, either for peacefully using the military grade dispositioned Plutonium or for better using Plutonium from LWR reprocessing plants. The proposed Thorium-LEU mixtures are intended to reduce the Uranium consumption per produced MW. The positive void reactivity is a major concern of any CANDU safety assessment, therefore reducing it was also a task for the present analysis. Using the 43-element bundle with a certain amount of burnable poison (e.g. Dysprosium) dissolved in the 8 innermost elements may lead to significantly reducing the void reactivity. The expected outcomes of these design improvements are: higher exit burnup, smooth/uniform radial bundle power distribution and reduced void reactivity. Since the improved fuel bundles are intended to be loaded in existing CANDU reactors, we found interesting to estimate the local reactivity effects of a mechanical control absorber (MCA) on the surrounding fuel cells. Cell parameters and neutron flux distributions, as well as macroscopic cross-sections were estimated using the transport code DRAGON and a 172-group updated nuclear data library. (author)

  13. Uncovering ecosystem service bundles through social preferences.

    Directory of Open Access Journals (Sweden)

    Berta Martín-López

    Full Text Available Ecosystem service assessments have increasingly been used to support environmental management policies, mainly based on biophysical and economic indicators. However, few studies have coped with the social-cultural dimension of ecosystem services, despite being considered a research priority. We examined how ecosystem service bundles and trade-offs emerge from diverging social preferences toward ecosystem services delivered by various types of ecosystems in Spain. We conducted 3,379 direct face-to-face questionnaires in eight different case study sites from 2007 to 2011. Overall, 90.5% of the sampled population recognized the ecosystem's capacity to deliver services. Formal studies, environmental behavior, and gender variables influenced the probability of people recognizing the ecosystem's capacity to provide services. The ecosystem services most frequently perceived by people were regulating services; of those, air purification held the greatest importance. However, statistical analysis showed that socio-cultural factors and the conservation management strategy of ecosystems (i.e., National Park, Natural Park, or a non-protected area have an effect on social preferences toward ecosystem services. Ecosystem service trade-offs and bundles were identified by analyzing social preferences through multivariate analysis (redundancy analysis and hierarchical cluster analysis. We found a clear trade-off among provisioning services (and recreational hunting versus regulating services and almost all cultural services. We identified three ecosystem service bundles associated with the conservation management strategy and the rural-urban gradient. We conclude that socio-cultural preferences toward ecosystem services can serve as a tool to identify relevant services for people, the factors underlying these social preferences, and emerging ecosystem service bundles and trade-offs.

  14. Noncommutative line bundle and Morita equivalence

    OpenAIRE

    Jurco, Branislav; Schupp, Peter; Wess, Julius

    2001-01-01

    Global properties of abelian noncommutative gauge theories based on $\\star$-products which are deformation quantizations of arbitrary Poisson structures are studied. The consistency condition for finite noncommutative gauge transformations and its explicit solution in the abelian case are given. It is shown that the local existence of invertible covariantizing maps (which are closely related to the Seiberg-Witten map) leads naturally to the notion of a noncommutative line bundle with noncommu...

  15. Bundling harvester; Harvennuspuun automaattisen nippukorjausharvesterin kehittaeminen

    Energy Technology Data Exchange (ETDEWEB)

    Koponen, K. [Eko-Log Oy, Kuopio (Finland)

    1997-12-01

    The starting point of the project was to design and construct, by taking the silvicultural point of view into account, a harvesting and processing system especially for energy-wood, containing manually driven bundling harvester, automating of the harvester, and automated loading. The equipment forms an ideal method for entrepreneur`s-line harvesting. The target is to apply the system also for owner`s-line harvesting. The profitability of the system promotes the utilisation of the system in both cases. The objectives of the project were: to construct a test equipment and prototypes for all the project stages, to carry out terrain and strain tests in order to examine the usability and durability, as well as the capacity of the machine, to test the applicability of the Eko-Log system in simultaneous harvesting of energy and pulp woods, and to start the marketing and manufacturing of the products. The basic problems of the construction of the bundling harvester have been solved using terrain-tests. The prototype machine has been shown to be operable. Loading of the bundles to form sufficiently economically transportable loads has been studied, and simultaneously, the branch-biomass has been tried to be utilised without loosing the profitability of transportation. The results have been promising, and will promote the profitable utilisation of wood-energy. (orig.)

  16. Nuclear reactor control bundle guide system

    International Nuclear Information System (INIS)

    Each bundle is formed by several absorbent rods, which are vertically movable and are connected together by a spider to a common axial operating rod, and guide means for the control bundles in their displacement, out of the core; the said means comprise guide boxes containing horizontal plates for discontinuous guiding, at the upper part of the boxes, of absorbent rods positioned in pairs on a radius and individual peripheral absorbent rods of the control bundle. At the lower part of the boxes in a continuous guiding zone, guiding of the absorbent rods positioned in pairs on a radius is effected by association of the horizontal plates for mechanical guiding of the rods, with housings which minimise hydraulic effects by smoothing the coolant flow in the radial direction around the absorbent rods. The hydraulic housings are mounted between the horizontal plates as discontinuous spacers. Pressure differences around each rod are minimised or eliminated and continuous guiding is achieved without affecting the design of the guide boxes, the internal equipment or the pressure vessel. The invention can be applied to PWRs

  17. Transport studies in boundary and divertor plasmas of JT-60U

    International Nuclear Information System (INIS)

    This thesis describes an investigation on transport of plasma, neutral particle and impurity in the boundary and divertor of the JT-60U tokamak to provide a better understanding of plasma-surface interactions and divertor physics. The asymmetry between the inboard and outboard divertor on plasma parameters (in-out asymmetry) are usually observed in tokamaks with the divertor. In this study, the in-out asymmetry was investigated under various plasma conditions and discharge parameters. The observed results were discussed with several mechanisms that can produce the in-out asymmetry. It was confirmed experimentally that the importance of each mechanism depends on the plasma parameters and discharge conditions. The current flowing in the scrape-off layer (SOL) due to the in-out asymmetry was observed. The SOL currents in the high density plasma with the occurrence of the plasma detachment were investigated for the first time in this study. The ion temperature in the divertor region is one of the most important factors for both generation and transport of impurity. However, the background ion temperature in the divertor region has not been measured in any tokamak so far. The ion temperature in the divertor region has been measured for the first time with the Doppler broading of the C3+ ion emission line. The measured temperature was analyzed by an impurity particle transport code. The code calculation showed that the measured temperature reflects the low temperature at the outside of the separatrix in the inboard region. The spectral profile of Balmer-α (Dα) line emitted from the deuterium atoms reflects the velocity distribution of neutral particles by the Doppler effect and is effective for investigating the detailed neutral behavior and recycling process. The spatial variation of the Dα line spectral profile in the divertor region has been measured for the first time in this study. The observed results were compared with the calculated one by a neutral particle

  18. Optimization design study of an innovative divertor concept for future experimental tokamak-type fusion reactors

    International Nuclear Information System (INIS)

    The design optimization study of an innovative divertor concept for future experimental tokamak-type fusion devices is both an answer to the actual problems encountered in the multilayer divertor proposals and an illustration of a rational modelling philosophy and optimization strategy for the development of a new divertor structure. Instead of using mechanical attachment or metallurgical bonding of the protective material to the heat sink as in most actual divertor concepts, the so-called brush divertor in this study uses an array of unidirectional fibers penetrating in both the protective armor and the underling composite heat sink. Although the approach is fully concentrated on the divertor performance, including both a description of its function from the theoretical point of view and an overview of the problems related to the materials choice and evaluation, both the approach followed in the numerical modelling and the judgment of the results are thought to be valid also for other applications. Therefore the spin-off of the study must be situated in both the technological progress towards a feasible divertor solution, which introduces no additional physical uncertainties, and in the general area of the thermo-mechanical finite-element modelling on both macro-and microscale. The brush divertor itself embodies the use, and thus the modelling, of advanced materials such as tailor-made metal matrix composites and dispersion strengthened metals, and is shown to offer large potential advantages, demanding however and experimental validation under working conditions. It is clearly indicated where the need originates for an integrated experimental program which must allow to verify the basic modelling assumptions in order to arrive at the use of numerical computation as a powerful and realistic tool of structural testing and life-time prediction

  19. Design and analysis of the DII-D radiative divertor water-cooled structures

    International Nuclear Information System (INIS)

    The Radiative Divertor is a major modification to the divertor of DIII-D and is being designed and fabricated for installation in late 1996. The Radiative Divertor Program (RDP) will enhance the dissipative processes in the edge and divertor plasmas to reduce the heat flux and plasma erosion at the divertor target. This approach will have major implications for the heat removal methods used in future devices. The divertor is of slot-type configuration designed to minimize the flow of sputtered and injected impurities back to the core plasma. The new divertor will be composed of toroidally continuous, Inconel 625 water-cooled rings of sandwich construction with an internal water channel, incorporating seam welding to provide the water-to-vacuum seal as well as structural integrity. The divertor structure is designed to withstand electromagnetic loads as a result of halo currents and induced toroidal currents. It also accommodates the thermal differences experienced during the 400 degrees C bake used on DIII-D. A low Z plasma-facing surface is provided by mechanically attached graphite tiles. Water flow through the rings will inertially cool these tiles which will be subjected to 38 MW, 10 second pulses. Current schedules call for detailed design in 1996 with installation completed in March 1997. A full size prototype, one-quarter of one ring, is being built to validate manufacturing techniques, machining, roll-forming, and seam welding. The experience and knowledge gained through the fabrication of the prototype is discussed. The design of the electrically isolated (5 kV) vacuum-to-air water feedthroughs supplying the water-cooled rings is also discussed

  20. Feasibility study for a multi-channel pulsed radar reflectometer for the jet divertor region

    International Nuclear Information System (INIS)

    In this report, the feasibility of a pulsed radar system for measuring the electron density profile in the divertor region of JET is studied. Some dedicated experiments are performed with a four-channel system, which was designed for the Rijnhuizen Tokamak Project. To simulate divertor plasmas the measurements are performed in ECRH induced plasmas without current. The parameters of these kinds of plasmas are: ne19 m-3, Te<100 eV, and a diameter of ∼30 cm. (HSI)

  1. Materials issues in the design of the ITER first wall, blanket, and divertor

    International Nuclear Information System (INIS)

    During the ITER conceptual design study, a property data base was assembled, the key issues were identified, and a comprehensive R ampersand D plan was formulated to resolve these issues. The desired properties of candidate ITER divertor, first wall, and blanket materials are briefly reviewed, and the major materials issues are presented. Estimates of the influence of materials properties on the performance limits of the first wall, blanket, and divertor are presented

  2. Materials issues in the design of the ITER first wall, blanket, and divertor

    Energy Technology Data Exchange (ETDEWEB)

    Mattas, R.F.; Smith, D.L. [Argonne National Lab., IL (United States); Wu, C.H. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany). NET Team; Koroda, T. [Japan Atomic Energy Research Inst., Ibaraki-ken (Japan); Shatalov, G. [Kurchatov Inst. of Atomic Energy, Moscow (USSR)

    1992-01-01

    During the ITER conceptual design study, a property data base was assembled, the key issues were identified, and a comprehensive R&D plan was formulated to resolve these issues. The desired properties of candidate ITER divertor, first wall, and blanket materials are briefly reviewed, and the major materials issues are presented. Estimates of the influence of materials properties on the performance limits of the first wall, blanket, and divertor are presented.

  3. Materials issues in the design of the ITER first wall, blanket, and divertor

    Energy Technology Data Exchange (ETDEWEB)

    Mattas, R.F.; Smith, D.L. (Argonne National Lab., IL (United States)); Wu, C.H. (Max-Planck-Institut fuer Plasmaphysik, Garching (Germany). NET Team); Koroda, T. (Japan Atomic Energy Research Inst., Ibaraki-ken (Japan)); Shatalov, G. (Kurchatov Inst. of Atomic Energy, Moscow (USSR))

    1992-01-01

    During the ITER conceptual design study, a property data base was assembled, the key issues were identified, and a comprehensive R D plan was formulated to resolve these issues. The desired properties of candidate ITER divertor, first wall, and blanket materials are briefly reviewed, and the major materials issues are presented. Estimates of the influence of materials properties on the performance limits of the first wall, blanket, and divertor are presented.

  4. Two-dimensional analysis for a scrapeoff and divertor regions with an MHD model

    International Nuclear Information System (INIS)

    With a two-dimensional time dependent fluid code for transport processes in the edge plasma in a tokamak, coupled with Monte-Carlo method for neutral gas behavior, preliminary numerical study has been carried out for the FER divertor. Design base data such as energy flux, particle flux and so on which are essentially important to make an divertor design reliable have been obtained. (author)

  5. Numerical simulations of resistive magnetohydrodynamic instabilities in a poloidal divertor tokamak

    International Nuclear Information System (INIS)

    A new 3-D resistive MHD initial value code RPD has been successfully developed from scratch to study the linear and nonlinear evolution of long wavelength resistive MHD instabilities in a square cross-section tokamak with or without a poloidal divertor. The code numerically advances the full set of compressible resistive MHD equations in a toroidal geometry, with an important option of permitting the divertor separatrix and the region outside it to be in the computational domain. A severe temporal step size restriction for numerical stability imposed by the fast compressional waves was removed by developing and implementing a new, efficient semi-implicit scheme extending one first proposed by Harned and Kerner. As a result, the code typically runs faster than that with a mostly explicit scheme by a factor of about the aspect ratio. The equilibrium input for RPD is generated by a new 2-D code EQPD that is based on the Chodura-Schluter method. The RPD code, as well as the new semi-implicit scheme, has passed very extensive numerical tests in both divertor and divertorless geometries. Linear and nonlinear simulations in a divertorless geometry have reproduced the standard, previously known results. In a geometry with a four-node divertor the m = 2,n = 1 (2/1) tearing mode tends to be linearly stabilized as the q = 2 surface approaches the divertor separatrix. However, the m = 1,n = 1 (1/1) resistive kink mode remains relatively unaffected by the nearness of the q = 1 surface to the divertor separatrix. When plasma current is added to the region outside the divertor separatrix, the 2/1 tearing mode is linearly stabilized not by this current, but by the profile modifications induced near the q = 2 surface and the divertor separatrix. A similar stabilization effect is seen for the 1/1 resistive kink mode, but to a lesser extent. 77 refs., 91 figs

  6. Possible divertor solutions for a fusion reactor. Pt. I. Physical aspects based on present day divertor operation

    International Nuclear Information System (INIS)

    For pt.II see ibid., p.109-117 (1997). With an anticipated power flux across the separatrix of up to 300 MW of an ITER-like fusion reactor, conventional measures of power spread lead to a peak power load at the target plates in the order of 30 MW m-2, far beyond the technically feasible limit for stationary operation. Radiative cooling by seed impurities appears to be the most promising plasma-physical option to reduce the target power load, but extrapolations of present experiments predict an only marginally tolerable increase of the plasma effective charge Zeff. Key points will be the achievement of very high electron densities, leading to more effective radiative cooling by δPrad/δZeff∝ne2 while keeping the edge temperature within its optimum range. This range is bounded from below by the H→L mode temperature threshold due to confinement requirements, whereas the upper boundary is given by the ideal ballooning stability limit which is connected to type-I ELM activity which may cause non-tolerable divertor heat loads. The completely detached H-mode (CDH) in ASDEX Upgrade demonstrates radiative H-mode operation within this operational range exhibiting high-frequent type-III ELMs and target power load in the order of 10% of the heating power. At present, open questions on high density reactor operation are related to radiative instabilities as well as edge transport enhancement and H-mode impairment observed in several tokamaks under high density conditions. Measures to overcome these detrimental effects will be investigated with improved divertor concepts in the near future. The possible problems connected to high density reactor operation can be relaxed, if the design of plasma facing components with higher heat flux endurance is successful. (orig.)

  7. Exploration of magnetic perturbation effects on advanced divertor configurations in NSTX-U

    Science.gov (United States)

    Frerichs, H.; Schmitz, O.; Waters, I.; Canal, G. P.; Evans, T. E.; Feng, Y.; Soukhanovskii, V. A.

    2016-06-01

    The control of divertor heat loads - both steady state and transient - remains a key challenge for the successful operation of ITER and FNSF. Magnetic perturbations provide a promising technique to control ELMs (Edge Localized Modes) (transients), but understanding their detailed impact is difficult due to their symmetry breaking nature. One approach for reducing steady state heat loads is so called "advanced divertors" which aim at optimizing the magnetic field configuration: the snowflake and the (super-)X-divertor. It is likely that both concepts - magnetic perturbations and advanced divertors - will have to work together, and we explore their interaction based on the NSTX-U setup. An overview of different divertor configurations under the impact of magnetic perturbations is presented, and the resulting impact on plasma edge transport is investigated with the EMC3-EIRENE code. Variations in size of the magnetic footprint of the perturbed separatrix are found, which are related to the level of flux expansion on the divertor target. Non-axisymmetric peaking of the heat flux related to the perturbed separatrix is found at the outer strike point, but only in locations where flux expansion is not too large.

  8. Spectroscopic imaging system for quantitative analysis of the divertor plasma of the Tokamak de Varennes

    International Nuclear Information System (INIS)

    A toroidally viewing spectroscopic imaging system has been developed for the Tokamak de Varennes providing measurements of the poloidal distribution of the absolute radiated power of deuterium and impurity species in the upper divertor region. Real time digitization is achieved using a low cost PC based digital imaging system. This system is used to obtain measurements of the divertor strike point as well as the shape of the flux surfaces in the divertor. The diagnostic close-quote s excellent spatial resolution and toroidal view provides an opportunity to quantitatively compare the measured two dimensional (2D) radiated power distribution to that calculated from 2D Monte Carlo transport codes. These 2D images provide unique and valuable information on the physics of local plasma interactions with divertor components and particle transport in a closed divertor. Additionally, by using two cameras simultaneously, the line ratio technique can be applied to the images to estimate plasma parameters in the divertor. copyright 1997 American Institute of Physics

  9. Spectroscopic imaging system for quantitative analysis of the divertor plasma of the Tokamak de Varennes

    Energy Technology Data Exchange (ETDEWEB)

    Meo, F.; Stansfield, B.L.; Chartre, M.; de Villers, P.; Marchand, R.; Ratel, G. [Centre Canadien de Fusion Magnetique, 1804 Boulevard Lionel-Boulet, Varennes, Quebec, J3X 1S1 (CANADA)

    1997-09-01

    A toroidally viewing spectroscopic imaging system has been developed for the Tokamak de Varennes providing measurements of the poloidal distribution of the absolute radiated power of deuterium and impurity species in the upper divertor region. Real time digitization is achieved using a low cost PC based digital imaging system. This system is used to obtain measurements of the divertor strike point as well as the shape of the flux surfaces in the divertor. The diagnostic{close_quote}s excellent spatial resolution and toroidal view provides an opportunity to quantitatively compare the measured two dimensional (2D) radiated power distribution to that calculated from 2D Monte Carlo transport codes. These 2D images provide unique and valuable information on the physics of local plasma interactions with divertor components and particle transport in a closed divertor. Additionally, by using two cameras simultaneously, the line ratio technique can be applied to the images to estimate plasma parameters in the divertor. {copyright} {ital 1997 American Institute of Physics. }

  10. Linear peeling–ballooning mode simulations in snowflake-like divertor configuration using BOUT++ code

    International Nuclear Information System (INIS)

    We present linear characteristics of peeling–ballooning (P–B) modes in the pedestal region of DIII-D tokamak with snowflake (SF) plus divertor configuration using edge two-fluid code BOUT++. A set of reduced magnetohydrodynamics (MHD) equations is found to simulate the linear P–B mode in both snowflake plus and standard (STD) single-null divertor configurations. Further analysis shows that the implementation of snowflake geometry changes the local magnetic shear in the pedestal region, which leads to different linear behaviours of the P–B mode in STD and SF divertor configuration. Primary linear simulation results are the following. (1) The growth rate of the coupled P–B mode in SF-plus divertor geometry is larger than that in STD divertor geometry. (2) The global linear mode structures are more radially extended yet less poloidally extended in SF-plus divertor geometry, especially for moderate and high toroidal mode numbers. (3) The current-gradient drive (the kink term) dominates the P–B mode for low n, while the pressure gradient drive (ballooning) dominates for n > 25. In addition, constraints on poloidal field and central solenoid coils for snowflake geometry are briefly discussed based on conclusions in this paper. (paper)

  11. Thermofluid analysis of free surface liquid divertor in tokamak fusion reactor

    International Nuclear Information System (INIS)

    To attain high fusion power density, the divertor must suffer a high heat flux from the fusion plasma. It is very difficult to remove the high heat flux from the fusion plasma more than 20 MW/m2 using the only solid divertor plate due to the severe mechanical condition such as thermal stress and crack growth. Therefore, the concept of a liquid divertor is proposed to remove the high heat flux and neutron flux from the plasma by liquid films flowing on a solid wall. Feasibility study on the liquid divertor is being examined what kind of necessary condition should be satisfied if it was applied to the tokamak fusion reactor. There are many uncertain physics and techniques to apply the liquid divertor to the tokamak fusion reactor. This paper mainly descries a preliminary thermofluid analysis of a free surface liquid, made of FLiBe molten salt, flow suffering the high heat flux using the finite element analysis code ADINA-F. To realize the liquid divertor, two techniques of thermal hydraulics promotion using a secondary flow and liquid-solid multi-phase flow are proposed in this paper

  12. Development of a compact W-shaped pumped divertor in JT-60U

    International Nuclear Information System (INIS)

    In JT-60U, the modification to a W-shaped pumped divertor will be completed in May 1997, aiming to realize sufficient reduction in heat flux to the targets and good H-mode confinement simultaneously. W-shaped geometry is optimized not only for forming radiative divertor plasmas and reducing the back flow of neutral particles but also for allowing various experimental configurations. Toroidally and poloidally segmented divertor plates, dome and baffles are arranged in a W-shaped poloidal configuration. The pumping speed can be changed during a shot by variable shutter valves in the three pumping ports under the outer baffle. The net throughput is enough for particle control in the steady radiative operations with high power NBI heating. Carbon fiber composite (CFC) tiles are used for the divertor targets and the divertor throat where large heat flux is expected. Gaps between two adjacent segments are carefully sealed to suppress the leak of neutral gas from the exhaust duct below the divertor and baffles. The strength of the whole structure is confirmed by an electromagnetic force analysis and structural analysis carried out for disruptions of 3 MA discharges with a halo current. (orig.)

  13. Flow reversal, convection, and modeling in the DIII-D divertor

    International Nuclear Information System (INIS)

    Measurements of the parallel Mach number of background plasma in the DIII-D tokamak divertor [M. A. Mahdavi et al. in Proceedings, 16th International Conference, Montreal, 1996 (International Atomic Energy Agency, Vienna, 1997) Vol. I, p. 397] were performed using a fast scanning Mach probe. The parallel particle flow shows evidence of complex behavior such as reverse flow, i.e., flow away from the target plate, stagnant flow, and large scale convection. For detached discharges, measurements confirm predictions of convective flow towards the divertor target plate at near sound speed over large regions in the divertor. The resulting convected heat flux is a dominant heat transport mechanism in the divertor. For attached discharges with high recycling, particle flow reversal in a thin region at or near the outer separatrix, thereby confirming the existence of a mechanism by which impurities can be transported away from the divertor target plates. Modeling results from the two-dimensional fluid code UEDGE [G. D. Porter and the DIII-D Team, open-quotes Divertor characterization experiments and modelling in DIII-D,close quotes in Proceedings of the 23rd European Conference on Controlled Fusion and Plasma Physics, 24 endash 28 June 1996, Kiev, Ukraine (European Physical Society, Petit-Lancy, Switzerland, 1996), Vol. 20C, Part II, p. 699] can reproduce the main features of the experimental observations. copyright 1998 American Institute of Physics

  14. Sensitivity analysis of upstream plasma condition for SST-1 X-divertor configuration with SOLPS

    International Nuclear Information System (INIS)

    The extensive power exhausts and target heat loads are anticipated in reactor grade fusion devices. Prototyping of an X-Divertor based power exhaust scheme is being attempted by means of simulations of Scrape-off Layer plasma transport in the diverted plasma equilibria of SST-1 tokamak using SOLPS5.1. Evaluation of the relative advantages of an X-Divertor configuration involves simulating the SST-1 standard divertor scheme plasma transport for the reference and then achieving equivalent upstream plasma conditions in the X-divertor equilibrium to ensure an equivalent core plasma in both the cases. The first optimization is to be achieved by simulating effects of an external gas puff in the SOL region for controlling separatrix density in the X-divertor configuration with visible modifications in the downstream plasma conditions. The present work analyzes sensitivity of the upstream SOL plasma conditions to the gas puff intensity and its effect on the plasma neutral transport in the divertor region. (author)

  15. Critical need for MFE: the Alcator DX advanced divertor test facility

    Science.gov (United States)

    Vieira, R.; Labombard, B.; Marmar, E.; Irby, J.; Wolf, S.; Bonoli, P.; Fiore, C.; Granetz, R.; Greenwald, M.; Hutchinson, I.; Hubbard, A.; Hughes, J.; Lin, Y.; Lipschultz, B.; Parker, R.; Porkolab, M.; Reinke, M.; Rice, J.; Shiraiwa, S.; Terry, J.; Theiler, C.; Wallace, G.; White, A.; Whyte, D.; Wukitch, S.

    2013-10-01

    Three critical challenges must be met before a steady-state, power-producing fusion reactor can be realized: how to (1) safely handle extreme plasma exhaust power, (2) completely suppress material erosion at divertor targets and (3) do this while maintaining a burning plasma core. Advanced divertors such as ``Super X'' and ``X-point target'' may allow a fully detached, low temperature plasma to be produced in the divertor while maintaining a hot boundary layer around a clean plasma core - a potential game-changer for magnetic fusion. No facility currently exists to test these ideas at the required parallel heat flux densities. Alcator DX will be a national facility, employing the high magnetic field technology of Alcator combined with high-power ICRH and LHCD to test advanced divertor concepts at FNSF/DEMO power exhaust densities and plasma pressures. Its extended vacuum vessel contains divertor cassettes with poloidal field coils for conventional, snowflake, super-X and X-point target geometries. Divertor and core plasma performance will be explored in regimes inaccessible in conventional devices. Reactor relevant ICRF and LH drivers will be developed, utilizing high-field side launch platforms for low PMI. Alcator DX will inform the conceptual development and accelerate the readiness-for-deployment of next-step fusion facilities.

  16. Divertor design and its integration into the ITER-FEAT machine

    International Nuclear Information System (INIS)

    The physics of the edge and divertor plasma is strongly coupled with the divertor and the fuel cycle design. Due to the limited space available the design as well as the remote maintenance approach for the ITER divertor are highly optimized to allow maximum space for the divertor plasma. Several auxiliary systems (e.g. in vessel viewing, glow discharge electrodes...) as well as a part of the pumping and fuelling system have to be integrated together with the divertor into the lower level of the ITER machine. Two main options exist for the choice of the plasma-facing material in the divertor, i.e. W and CFC. Based on already existing R and D results one can be optimistic that the material choice will be mainly based on physics considerations and material issues (e.g. C-T co-deposition). The requirements for the ITER fuel cycle arise from plasma physics as well as from the envisaged operation scenarios. Due to the complex dynamic relationship of the fuel cycle subsystems among themselves and with the plasma, codes are employed for their optimization. This paper elaborates these interacting issues and gives the latest design status. (author)

  17. Adsorption of Argon on Carbon nanotube bundles and its influence on the bundle lattice parameter

    International Nuclear Information System (INIS)

    We report experimental studies of the adsorption characteristics and structure of both Ar36 and Ar40 on single-wall carbon nanotube bundles. The structural studies make use of the large difference in coherent neutron scattering cross section for the two Ar isotopes to explore the influence of the adsorbate on the nanotube lattice parameter. We observe no dilation of the nanotube lattice with Ar40, and explain the apparent expansion of this lattice upon Ar36 adsorption by the location of the adsorbed Ar atoms on the outer bundle surface

  18. The impact of ELMs on the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Leonard, A.W.; Osborne, T.H.; Suttrop, W. [General Atomics, San Diego, CA (United States); Hermann, A. [Max Planck Inst. fuer Plasmaphysics, Garching (Germany); Itami, K. [Japan Atomic Energy Research Inst. (Japan); Lingertat, J. [JET Joint Undertaking, Abingdon (United Kingdom); Loarte, A. [Next European Torus, Garching (Germany)

    1998-07-01

    Edge-Localized-Modes (ELMs) are expected to present a significant transient flux of energy and particles to the ITER divertor. The threshold for ablation of the graphite target will be reached if the ELM transient exceeds Q/t{sup 1/2} {approximately} 45 MJ-m{sup {minus}2}-s{sup {minus}1/2} where Q is the ELM deposition energy density and t is the ELM deposition time. The ablation parameter in ITER can be determined by scaling four factors from present experiments: the ELM energy loss from the core plasma, the fraction of ELM energy deposited on the divertor target, the area of the ELM profile onto the target, and finally the time for the ELM deposition. Review of the ELM energy loss of Type 1 ELM data suggests an ITER ELM energy loss of 2--6% of the stored energy or 25--80 MJ. The fraction of heating power crossing the separatrix due to ELMs is nearly constant (20--40%) resulting in an inverse relationship between ELM amplitude and frequency. Measurements on DIII-D and ASDEX-Upgrade indicate that 50--80% of the ELM energy is deposited on the target. There is currently no evidence for a large fraction of the ELM energy being dissipated through radiation. Profiles of the ELM heat flux are typically 1--2 times the width of steady heat flux between ELMs, with the ELM amplitude usually larger on the inboard target. The ELM deposition time varies from about 0.1 ms in JET to as high as 1.0 ms in ASDEX-Upgrade and DIII-D. The ELM deposition time for ITER will depend upon the level of conductive versus convective transport determined by the ratio of energy to particles released by the ELM. Preliminary analysis suggests that large Type 1 ELMs for low recycling H-mode may exceed the ablation parameter by a factor of 5. Promising regimes with smaller ELMS have been found at other edge operational regimes, including high density with gas puffing, use of rf heating and operation with Type 3 ELMs.

  19. The impact of ELMS on the ITER divertor

    International Nuclear Information System (INIS)

    Edge-Localized-Modes (ELMs) are expected to present a significant transient flux of energy and particles to the ITER divertor. The threshold for ablation of the graphite target will be reached if the ELM transient exceeds Q/t1/2 ∼ 45 MJ-m-2-s-1/2 where Q is the ELM deposition energy density and t is the ELM deposition time. The ablation parameter in ITER can be determined by scaling four factors from present experiments: the ELM energy loss from the core plasma, the fraction of ELM energy deposited on the divertor target, the area of the ELM profile onto the target, and finally the time for the ELM deposition. Review of the ELM energy loss of Type 1 ELM data suggests an ITER ELM energy loss of 2--6% of the stored energy or 25--80 MJ. The fraction of heating power crossing the separatrix due to ELMs is nearly constant (20--40%) resulting in an inverse relationship between ELM amplitude and frequency. Measurements on DIII-D and ASDEX-Upgrade indicate that 50--80% of the ELM energy is deposited on the target. There is currently no evidence for a large fraction of the ELM energy being dissipated through radiation. Profiles of the ELM heat flux are typically 1--2 times the width of steady heat flux between ELMs, with the ELM amplitude usually larger on the inboard target. The ELM deposition time varies from about 0.1 ms in JET to as high as 1.0 ms in ASDEX-Upgrade and DIII-D. The ELM deposition time for ITER will depend upon the level of conductive versus convective transport determined by the ratio of energy to particles released by the ELM. Preliminary analysis suggests that large Type 1 ELMs for low recycling H-mode may exceed the ablation parameter by a factor of 5. Promising regimes with smaller ELMS have been found at other edge operational regimes, including high density with gas puffing, use of rf heating and operation with Type 3 ELMs

  20. Effects of radial losses of particle and energy on the stability of detachment front in a divertor plasma

    International Nuclear Information System (INIS)

    Operation under partially detached divertor (PDD) plasmas is a hopeful way in order to reduce the divertor heat load in the next generation tokamaks. The physical mechanism of PDD plasmas, however, has not fully been understood yet. We have studied them with a multi-layer one-dimensional divertor model. The PDD plasmas are successfully reproduced by introducing a neutral gas puffing model. Effect of the cross-field heat transport on the PDD plasmas is investigated. It is found that cross-field heat transport both in the SOL region and in the divertor region prevents detachment fronts from moving upstream in a detached flux tube. (author)

  1. Development of a Method for Local Electron Temperature and Density Measurements in the Divertor of the JET Tokamak

    Science.gov (United States)

    Jupen, C.; Meigs, A.; Bhatia, A. K.; Brezinsek, S.; OMullane, M.

    2004-01-01

    Plasma volume recombination in the divertor, a process in which charged particles recombine to neutral atoms, contributes to plasma detachment and hence cooling at the divertor target region. Detachment has been observed at JET and other tokamaks and is known to occur at low electron temperatures (T(sub e)10(exp 20)/m(exp 3)). The ability to measure such low temperatures is therefore of interest for modelling the divertor. In present work we report development of a new spectroscopic technique for investigation of local electron density (n(sub e)) and temperature (T,) in the outer divertor at JET.

  2. Amplitude death of coupled hair bundles with stochastic channel noise

    CERN Document Server

    Kim, Kyung-Joong

    2014-01-01

    Hair cells conduct auditory transduction in vertebrates. In lower vertebrates such as frogs and turtles, due to the active mechanism in hair cells, hair bundles(stereocilia) can be spontaneously oscillating or quiescent. Recently, the amplitude death phenomenon has been proposed [K.-H. Ahn, J. R. Soc. Interface, {\\bf 10}, 20130525 (2013)] as a mechanism for auditory transduction in frog hair-cell bundles, where sudden cessation of the oscillations arises due to the coupling between non-identical hair bundles. The gating of the ion channel is intrinsically stochastic due to the stochastic nature of the configuration change of the channel. The strength of the noise due to the channel gating can be comparable to the thermal Brownian noise of hair bundles. Thus, we perform stochastic simulations of the elastically coupled hair bundles. In spite of stray noisy fluctuations due to its stochastic dynamics, our simulation shows the transition from collective oscillation to amplitude death as inter-bundle coupling str...

  3. A Tannakian approach to dimensional reduction of principal bundles

    CERN Document Server

    Álvarez-Cónsul, Luis; García-Prada, Oscar

    2016-01-01

    Let $P$ be a parabolic subgroup of a connected simply connected complex semisimple Lie group $G$. Given a compact K\\"ahler manifold $X$, the dimensional reduction of $G$-equivariant holomorphic vector bundles over $X\\times G/P$ was carried out by the first and third authors. This raises the question of dimensional reduction of holomorphic principal bundles over $X\\times G/P$. The method used for equivariant vector bundles does not generalize to principal bundles. In this paper, we adapt to equivariant principal bundles the Tannakian approach of Nori, to describe the dimensional reduction of $G$-equivariant principal bundles over $X\\times G/P$, and to establish a Hitchin--Kobayashi type correspondence. In order to be able to apply the Tannakian theory, we need to assume that $X$ is a complex projective manifold.

  4. Hydrodynamic behavior of a bare rod bundle

    International Nuclear Information System (INIS)

    The temperature distribution within the rod bundle of a nuclear reactor is of major importance in nuclear reactor design. However temperature information presupposes knowledge of the hydrodynamic behavior of the coolant which is the most difficult part of the problem due to complexity of the turbulence phenomena. In the present work a 2-equation turbulence model--a strong candidate for analyzing actual three dimensional turbulent flows--has been used to predict fully developed flow of infinite bare rod bundle of various aspect ratios (P/D). The model has been modified to take into account anisotropic effects of eddy viscosity. Secondary flow calculations have been also performed although the model seems to be too rough to predict the secondary flow correctly. Heat transfer calculations have been performed to confirm the importance of anisotropic viscosity in temperature predictions. All numerical calculations for flow and heat have been performed by two computer codes based on the TEACH code. Experimental measurements of the distribution of axial velocity, turbulent axial velocity, turbulent kinetic energy and radial Reynolds stresses were performed in the developing and fully developed regions. A 2-channel Laser Doppler Anemometer working on the Reference mode with forward scattering was used to perform the measurements in a simulated interior subchannel of a triangular rod array with P/D = 1.124. Comparisons between the analytical results and the results of this experiment as well as other experimental data in rod bundle array available in literature are presented. The predictions are in good agreement with the results for the high Reynolds numbers

  5. Historical dynamics in ecosystem service bundles.

    Science.gov (United States)

    Renard, Delphine; Rhemtulla, Jeanine M; Bennett, Elena M

    2015-10-27

    Managing multiple ecosystem services (ES), including addressing trade-offs between services and preventing ecological surprises, is among the most pressing areas for sustainability research. These challenges require ES research to go beyond the currently common approach of snapshot studies limited to one or two services at a single point in time. We used a spatiotemporal approach to examine changes in nine ES and their relationships from 1971 to 2006 across 131 municipalities in a mixed-use landscape in Quebec, Canada. We show how an approach that incorporates time and space can improve our understanding of ES dynamics. We found an increase in the provision of most services through time; however, provision of ES was not uniformly enhanced at all locations. Instead, each municipality specialized in providing a bundle (set of positively correlated ES) dominated by just a few services. The trajectory of bundle formation was related to changes in agricultural policy and global trends; local biophysical and socioeconomic characteristics explained the bundles' increasing spatial clustering. Relationships between services varied through time, with some provisioning and cultural services shifting from a trade-off or no relationship in 1971 to an apparent synergistic relationship by 2006. By implementing a spatiotemporal perspective on multiple services, we provide clear evidence of the dynamic nature of ES interactions and contribute to identifying processes and drivers behind these changing relationships. Our study raises questions about using snapshots of ES provision at a single point in time to build our understanding of ES relationships in complex and dynamic social-ecological systems. PMID:26460005

  6. Numerical simulations of square arrayed rod bundles

    International Nuclear Information System (INIS)

    Highlights: ► CFD simulations with square arrayed rod bundles. ► Mesh dependency and turbulence model study by comparison with experiments. ► Gibson and Launder Reynolds stress model shows good agreement with experiments. ► Effect of pitch to diameter ratio and Reynolds number is correctly captured. - Abstract: Computational fluid dynamics (CFD) simulations were performed with square arrayed rod bundles featuring pitch to diameter (P/D) ratio of 1.194 and 1.326 in order to find an optimal mesh and turbulence model for simulations with more complex geometries in the future. With the tighter lattice a mesh sensitivity and turbulence model study were accomplished and the post processed turbulence quantities, velocity field and wall shear stress were compared with experimental data ( Developed single phase turbulent flow through a square-pitch rod cluster. Nuclear Engineering and Design 60, 365–379.). The comparisons show that Reynolds-Averaged Navier–Stokes method with the Reynolds stress model of Gibson and Launder in conjunction with an appropriate mesh can provide reasonable agreement with the experiment for this lattice. For pure bundle simulations the body fitted structured meshes are suggested, since slightly better agreement can be captured considering all quantities with the same number of cells. Based on the drawn conclusions the procession was repeated for P/D = 1.326, where, due to lack of experiment, just the correct tendencies of the turbulence quantities and velocity field were established. The results show Reynolds number independency correctly and the increase of P/D issues in more similar flow to axisymmetric pipe flow.

  7. Systematic Bundle Adjustment of HRSC Image Data

    Science.gov (United States)

    Bostelmann, J.; Schmidt, R.; Heipke, C.

    2012-07-01

    The European Mars Express mission was launched in June 2003 and sent into orbit around Mars. On board the orbiter is the German High Resolution Stereo Camera (HRSC). This multi-line sensor images the Martian surface with a resolution of up to 12m per pixel in three dimensions and provides RGB and infra-red color information. The usage of the stereoscopic image information for the improvement of the observed position and attitude information via bundle adjustment is important to derive high quality 3D surface models, color orthoimages and other data products. In many cases overlapping image strips of different orbits can be used to form photogrammetric blocks, thus allowing the simultaneous adjustment of the exterior orientation data. This reduces not only local, but also regional inconsistencies in the data. With the growing number of HRSC image strips in this ongoing mission, the size and complexity of potential blocks is increasing. Therefore, a workflow has been built up for the systematic improvement of the exterior orientation using single orbit strips and regional blocks. For a successful bundle adjustment of blocks using multiple image strips a sufficient number of tie points in the overlapping area is needed. The number of tie points depends mainly on the geometric and radiometric quality of the images. This is considered by detailed analysis of the tie point accuracy and distribution. The combination of methods for image pre-processing, tie point matching, bundle adjustment and evaluation of the results in an automated workflow allows for all HRSC images a global assessment of the quality and a systematic selection of data for larger blocks.

  8. Tiling spaces are Cantor set fiber bundles

    OpenAIRE

    Sadun, Lorenzo; Williams, R F

    2001-01-01

    We prove that fairly general spaces of tilings of R^d are fiber bundles over the torus T^d, with totally disconnected fiber. This was conjectured (in a weaker form) in [W3], and proved in certain cases. In fact, we show that each such space is homeomorphic to the d-fold suspension of a Z^d subshift (or equivalently, a tiling space whose tiles are marked unit d-cubes). The only restrictions on our tiling spaces are that 1) the tiles are assumed to be polygons (polyhedra if d>2) that meet full-...

  9. Higher order mechanics on graded bundles

    International Nuclear Information System (INIS)

    In this paper we develop a geometric approach to higher order mechanics on graded bundles in both, the Lagrangian and Hamiltonian formalism, via the recently discovered weighted algebroids. We present the corresponding Tulczyjew triple for this higher order situation and derive in this framework the phase equations from an arbitrary (also singular) Lagrangian or Hamiltonian, as well as the Euler–Lagrange equations. As important examples, we geometrically derive the classical higher order Euler–Lagrange equations and analogous reduced equations for invariant higher order Lagrangians on Lie groupoids. (paper)

  10. Assembly mechanism for nuclear fuel bundles

    International Nuclear Information System (INIS)

    The invention relates to a nuclear power reactor fuel bundle of the type wherein several rods are mounted in parallel array between two tie plates which secure the fuel rods in place and are maintained in assembled position by means of a number of tie rods secured to both of the end plates. Improved apparatus is provided for attaching the tie rods to the upper tie plate by the use of locking lugs fixed to rotatable sleeves which engage the upper tie plate. (auth)

  11. Compression of a bundle of light rays.

    Science.gov (United States)

    Marcuse, D

    1971-03-01

    The performance of ray compression devices is discussed on the basis of a phase space treatment using Liouville's theorem. It is concluded that the area in phase space of the input bundle of rays is determined solely by the required compression ratio and possible limitations on the maximum ray angle at the output of the device. The efficiency of tapers and lenses as ray compressors is approximately equal. For linear tapers and lenses the input angle of the useful rays must not exceed the compression ratio. The performance of linear tapers and lenses is compared to a particular ray compressor using a graded refractive index distribution. PMID:20094478

  12. Differential geometry of complex vector bundles

    CERN Document Server

    Kobayashi, Shoshichi

    2014-01-01

    Holomorphic vector bundles have become objects of interest not only to algebraic and differential geometers and complex analysts but also to low dimensional topologists and mathematical physicists working on gauge theory. This book, which grew out of the author's lectures and seminars in Berkeley and Japan, is written for researchers and graduate students in these various fields of mathematics. Originally published in 1987. The Princeton Legacy Library uses the latest print-on-demand technology to again make available previously out-of-print books from the distinguished backlist of Princeto

  13. Experimental study of the topological aspect of the ergodic divertor in Tore-supra tokamak

    International Nuclear Information System (INIS)

    The control of power deposition onto plasma facing components in tokamaks is a determining factor for future thermonuclear fusion reactors. Plasma surface interaction can be performed using limiters or divertors. The ergodic divertor installed on Tore Supra is an atypical example of a magnetic divertor. It consists in applying a magnetic perturbation which establishes a particular topology of the plasma in contact with the wall (edge plasma). We carried out dedicated experiments in order to study parallel heat flux which strike the divertor neutralizers. This quantitative and qualitative analysis of heat flux as a function of experimental conditions allows to determine the profiles of power deposition along the neutralizers. The influence of plasma electron density, additional heating, impurities and injected gas was established. An experimental study of the sheath heat transmission factor γ was carried out by correlating measurements made with Langmuir probes and infrared imaging. This study gave rise to a major conclusion: for ohmic discharges with deuterium injection and most of the time with helium, it was experimentally confirmed that γ=7 in agreement with classical sheath theory. However, an increase of this factor with additional power has been shown. Detached plasma, which is an attractive regime in order to reduce the power deposition, requires an optimized control. A new measurement of the detachment onset has been developed. It is based on the variation of heat flux onto the plates derived from infrared measurements. A detachment cartography with the determination of a new 2D 'IR' Degree of Detachment was carried out allowing to locate the zone where the detachment starts. We can apply this concept both to other tokamaks such as JET and ITER. A comparison between the axisymmetric divertor and the ergodic divertor is also presented concerning the power deposition in the two configurations. Low heat flux with the ergodic divertor is a major advantage

  14. The two-dimensional structure of radiative divertor plasmas in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Recent measurements of the two-dimensional (2-D) spatial profiles of divertor plasma density, temperature, and emissivity in the DIII-D tokamak [J. Luxon et al., in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion (International Atomic Energy Agency, Vienna, 1987), p. 159] under highly radiating conditions are presented. Data are obtained using a divertor Thomson scattering system and other diagnostics optimized for measuring the high electron densities and low temperatures in these detached divertor plasmas (ne≤1021m-3, 0.5eV≤Te). D2 gas injection in the divertor increases the plasma radiation and lowers Te to less than 2 eV in most of the divertor volume. Modeling shows that this temperature is low enough to allow ion endash neutral collisions, charge exchange, and volume recombination to play significant roles in reducing the plasma pressure along the magnetic separatrix by a factor of 3 endash 5, consistent with the measurements. Absolutely calibrated vacuum ultraviolet spectroscopy and 2-D images of impurity emission show that carbon radiation near the X-point, and deuterium radiation near the target plates contribute to the reduction in Te. Uniformity of radiated power (Prad) (within a factor of 2) along the outer divertor leg, with peak heat flux on the divertor target reduced fourfold, was obtained. A comparison with 2-D fluid simulations shows good agreement when physical sputtering and an ad hoc chemical sputtering source (0.5%) from the private flux region surface are used. copyright 1997 American Institute of Physics

  15. A Unified Framework for Quasi-Linear Bundle Adjustment

    OpenAIRE

    Bartoli, Adrien

    2002-01-01

    Obtaining 3D models from long image sequences is a major issue in computer vision. One of the main tools used to obtain accurate structure and motion estimates is bundle adjustment. Bundle adjustment is usually performed using nonlinear Newton-type optimizers such as Levenberg-Marquardt which might be quite slow when handling a large number of points or views. We investigate an algorithm for bundle adjustment based on quasi-linear optimization. The method is straightforward to implement and r...

  16. Non-commutative P-1-bundles over commutative schemes

    OpenAIRE

    Van den Bergh, Michel

    2012-01-01

    In this paper we develop the theory of non-commutative P-1-bundles over commutative (smooth) schemes. Such non-commutative P-1-bundles occur in the theory of D-modules but our definition is more general. We can show that every non-commutative deformation of a Hirzebruch surface is given by a non-commutative P-1-bundle over P-1 in our sense.

  17. Contacting single bundles of carbon nanotubes with alternating electric fields

    OpenAIRE

    Krupke, R.; Hennrich, F.; Weber, H. B.; Beckmann, D.; Hampe, O.; Malik, S.; Kappes, M. M.; Löhneysen, H. v.

    2002-01-01

    Single bundles of carbon nanotubes have been selectively deposited from suspensions onto sub-micron electrodes with alternating electric fields. We explore the resulting contacts using several solvents and delineate the differences between Au and Ag as electrode materials. Alignment of the bundles between electrodes occurs at frequencies above 1 kHz. Control over the number of trapped bundles is achieved by choosing an electrode material which interacts strongly with the chemical functional g...

  18. Dark-field illuminated reflectance fiber bundle endoscopic microscope

    OpenAIRE

    Liu, Xuan; Huang, Yong; Kang, Jin U.

    2011-01-01

    We propose a reflectance fiber bundle microscope using a dark-field illumination configuration for applications in endoscopic medical imaging and diagnostics. Our experiment results show that dark-field illumination can effectively suppress strong specular reflection from the proximal end of the fiber bundle. We realized a lateral resolution of 4.4 μm using the dark-field illuminated fiber bundle configuration. To demonstrate the feasibility of using the system to study cell morphology, we ob...

  19. Physical Engineering Test and First Divertor Plasma Configuration in EAST

    Institute of Scientific and Technical Information of China (English)

    WAN Baonian

    2007-01-01

    Physical engineering capability on the superconducting magnetic system of EAST was tested and first divertor plasma configuration in EAST was obtained.The extrapolation of the safety limit has verified the reliability of the system for long pulse operation.A stably controlled diverted plasmas configuration with an elongation κ in excess of 1.8 and plasma current of up to 500 kA,by using the (copper) internal coils to control the vertical displacement instability was obtained by an optimized plasma control algorithm.Highly shaped plasma at various configura-tions,which almost covers all designed configurations for EAST,was generated stably.A number of operational issues,such as plasma initiation,ramp up and configuration control with constraints of superconducting coils,were successfully investigated.All of the results obtained proved both the capability of the superconducting poloidal magnets for operation under steady-state condition and effectiveness of the plasma control algorithm for EAST.

  20. Fabrication and high heat flux test of divertor cooling elements

    International Nuclear Information System (INIS)

    The plasma facing components in ITER are subjected to a high heat flux from fusion plasma. The heat flux is not only higher than that of existing tokamaks but also has a longer pulse duration (burn time). To minimize a peaking of the heat flux, the thermal deformation towards the plasma should be restrained. One-meter-long monoblock divertor modules with a sliding support structure were fabricated and tested at JAERI. Two kinds of support mechanisms were provided to minimize the thermal deformation of the modules in the upward and downward directions ; one is a pin type sliding structure and the other is a rail type support structure. Both modules were tested on the electron beam HHF test facility, JEBIS (JAERI Electron Beam Irradiation System), in JAERI. The steady-state heat flux of 15 MW/m2 was applied to the surface of the modules to simulate the design condition of ITER CDA. As a result of the HHF test, the performance of the sliding support structures was successfully demonstrated. Three dimensional elastic stress analyses were conducted using a finite element method. The result shows that the relatively high thermal stress is observed at the cooling tube ; and that the maximum thermal stress at the cooling tube exceeds its yield strength. It is necessary to perform the lifetime evaluation of the copper cooling tube against cyclic thermal stresses. (author)