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Sample records for iter heating neutral

  1. Status of the ITER heating neutral beam system

    Science.gov (United States)

    Hemsworth, R.; Decamps, H.; Graceffa, J.; Schunke, B.; Tanaka, M.; Dremel, M.; Tanga, A.; DeEsch, H. P. L.; Geli, F.; Milnes, J.; Inoue, T.; Marcuzzi, D.; Sonato, P.; Zaccaria, P.

    2009-04-01

    The ITER neutral beam (NB) injectors are the first injectors that will have to operate under conditions and constraints similar to those that will be encountered in a fusion reactor. These injectors will have to operate in a hostile radiation environment and they will become highly radioactive due to the neutron flux from ITER. The injectors will use a single large ion source and accelerator that will produce 40 A 1 MeV D- beams for pulse lengths of up to 3600 s. Significant design changes have been made to the ITER heating NB (HNB) injector over the past 4 years. The main changes are: Modifications to allow installation and maintenance of the beamline components with an overhead crane. The beam source vessel shape has been changed and the beam source moved to allow more space for the connections between the 1 MV bushing and the beam source. The RF driven negative ion source has replaced the filamented ion source as the reference design. The ion source and extractor power supplies will be located in an air insulated high voltage (-1 MV) deck located outside the tokamak building instead of inside an SF6 insulated HV deck located above the injector. Introduction of an all metal absolute valve to prevent any tritium in the machine to escape into the NB cell during maintenance. This paper describes the status of the design as of December 2008 including the above mentioned changes. The very important power supply system of the neutral beam injectors is not described in any detail as that merits a paper beyond the competence of the present authors. The R&D required to realize the injectors described in this paper must be carried out on a dedicated neutral beam test facility, which is not described here.

  2. Overview of the design of the ITER heating neutral beam injectors

    Science.gov (United States)

    Hemsworth, R. S.; Boilson, D.; Blatchford, P.; Dalla Palma, M.; Chitarin, G.; de Esch, H. P. L.; Geli, F.; Dremel, M.; Graceffa, J.; Marcuzzi, D.; Serianni, G.; Shah, D.; Singh, M.; Urbani, M.; Zaccaria, P.

    2017-02-01

    The heating neutral beam injectors (HNBs) of ITER are designed to deliver 16.7 MW of 1 MeV D0 or 0.87 MeV H0 to the ITER plasma for up to 3600 s. They will be the most powerful neutral beam (NB) injectors ever, delivering higher energy NBs to the plasma in a tokamak for longer than any previous systems have done. The design of the HNBs is based on the acceleration and neutralisation of negative ions as the efficiency of conversion of accelerated positive ions is so low at the required energy that a realistic design is not possible, whereas the neutralisation of H‑ and D‑ remains acceptable (≈56%). The design of a long pulse negative ion based injector is inherently more complicated than that of short pulse positive ion based injectors because: • negative ions are harder to create so that they can be extracted and accelerated from the ion source; • electrons can be co-extracted from the ion source along with the negative ions, and their acceleration must be minimised to maintain an acceptable overall accelerator efficiency; • negative ions are easily lost by collisions with the background gas in the accelerator; • electrons created in the extractor and accelerator can impinge on the extraction and acceleration grids, leading to high power loads on the grids; • positive ions are created in the accelerator by ionisation of the background gas by the accelerated negative ions and the positive ions are back-accelerated into the ion source creating a massive power load to the ion source; • electrons that are co-accelerated with the negative ions can exit the accelerator and deposit power on various downstream beamline components. The design of the ITER HNBs is further complicated because ITER is a nuclear installation which will generate very large fluxes of neutrons and gamma rays. Consequently all the injector components have to survive in that harsh environment. Additionally the beamline components and the NB cell, where the beams are housed, will be

  3. Beyond ITER: Neutral beams for DEMO

    CERN Document Server

    McAdams, R

    2013-01-01

    In the development of magnetically confined fusion as an economically sustainable power source, ITER is currently under construction. Beyond ITER is the DEMO programme in which the physics and engineering aspects of a future fusion power plant will be demonstrated. DEMO will produce net electrical power. The DEMO programme will be outlined and the role of neutral beams for heating and current drive will be described. In particular, the importance of the efficiency of neutral beam systems in terms of injected neutral beam power compared to wallplug power will be discussed. Options for improving this efficiency including advanced neutralisers and energy recovery are discussed.

  4. Powerloads on the front end components and the duct of the heating and diagnostic neutral beam lines at ITER

    Energy Technology Data Exchange (ETDEWEB)

    Singh, M. J.; Boilson, D.; Hemsworth, R. S.; Geli, F.; Graceffa, J.; Urbani, M.; Schunke, B.; Chareyre, J. [ITER Organisation, 13607 St. Paul-Lez-Durance Cedex (France); Dlougach, E.; Krylov, A. [RRC Kurchatov institute, 1, Kurchatov Sq, Moscow, 123182 (Russian Federation)

    2015-04-08

    The heating and current drive beam lines (HNB) at ITER are expected to deliver ∼16.7 MW power per beam line for H beams at 870 keV and D beams at 1 MeV during the H-He and the DD/DT phases of ITER operation respectively. On the other hand the diagnostic neutral beam (DNB) line shall deliver ∼2 MW power for H beams at 100 keV during both the phases. The path lengths over which the beams from the HNB and DNB beam lines need to be transported are 25.6 m and 20.7 m respectively. The transport of the beams over these path lengths results in beam losses, mainly by the direct interception of the beam with the beam line components and reionisation. The lost power is deposited on the surfaces of the various components of the beam line. In order to ensure the survival of these components over the operational life time of ITER, it is important to determine to the best possible extent the operational power loads and power densities on the various surfaces which are impacted by the beam in one way or the other during its transport. The main factors contributing to these are the divergence of the beamlets and the halo fraction in the beam, the beam aiming, the horizontal and vertical misalignment of the beam, and the gas profile along the beam path, which determines the re-ionisation loss, and the re-ionisation cross sections. The estimations have been made using a combination of the modified version of the Monte Carlo Gas Flow code (MCGF) and the BTR code. The MCGF is used to determine the gas profile in the beam line and takes into account the active gas feed into the ion source and neutraliser, the HNB-DNB cross over, the gas entering the beamline from the ITER machine, the additional gas atoms generated in the beam line due to impacting ions and the pumping speed of the cryopumps. The BTR code has been used to obtain the power loads and the power densities on the various surfaces of the front end components and the duct modules for different scenarios of ITER

  5. Ion beam transport: modelling and experimental measurements on a large negative ion source in view of the ITER heating neutral beam

    Science.gov (United States)

    Veltri, P.; Sartori, E.; Agostinetti, P.; Aprile, D.; Brombin, M.; Chitarin, G.; Fonnesu, N.; Ikeda, K.; Kisaki, M.; Nakano, H.; Pimazzoni, A.; Tsumori, K.; Serianni, G.

    2017-01-01

    Neutral beam injectors are among the most important methods of plasma heating in magnetic confinement fusion devices. The propagation of the negative ions, prior to their conversion into neutrals, is of fundamental importance in determining the properties of the beam, such as its aiming and focusing at long-distances, so as to deposit the beam power in the proper position inside the confined plasma, as well as to avoid interaction with the material surfaces along the beam path. The final design of the ITER Heating Neutral Beam prototype has been completed at Consorzio RFX (Padova, Italy), in the framework of a close collaboration with European, Japanese and Indian fusion research institutes. The physical and technical rationales on which the design is based were essentially driven by numerical modelling of the relevant physical processes, and the same models and codes will be useful to design the DEMO neutral beam injector in the near future. This contribution presents a benchmark study of the codes used for this purpose, by comparing their results against the measures performed in an existing large-power device, hosted at the National Institute for Fusion Science, Japan. In particular, the negative ion formation and acceleration are investigated. A satisfactory agreement was found between codes and experiments, leading to an improved understanding of beam transport dynamics. The interpretation of the discrepancies identified in previous works, possibly related to the non-uniformity of the extracted negative ion current, is also presented.

  6. The ITER Neutral Beam Test Facility towards SPIDER operation

    Science.gov (United States)

    Toigo, V.; Dal Bello, S.; Gaio, E.; Luchetta, A.; Pasqualotto, R.; Zaccaria, P.; Bigi, M.; Chitarin, G.; Marcuzzi, D.; Pomaro, N.; Serianni, G.; Agostinetti, P.; Agostini, M.; Antoni, V.; Aprile, D.; Baltador, C.; Barbisan, M.; Battistella, M.; Boldrin, M.; Brombin, M.; Dalla Palma, M.; De Lorenzi, A.; Delogu, R.; De Muri, M.; Fellin, F.; Ferro, A.; Gambetta, G.; Grando, L.; Jain, P.; Maistrello, A.; Manduchi, G.; Marconato, N.; Pavei, M.; Peruzzo, S.; Pilan, N.; Pimazzoni, A.; Piovan, R.; Recchia, M.; Rizzolo, A.; Sartori, E.; Siragusa, M.; Spada, E.; Spagnolo, S.; Spolaore, M.; Taliercio, C.; Valente, M.; Veltri, P.; Zamengo, A.; Zaniol, B.; Zanotto, L.; Zaupa, M.; Boilson, D.; Graceffa, J.; Svensson, L.; Schunke, B.; Decamps, H.; Urbani, M.; Kushwah, M.; Chareyre, J.; Singh, M.; Bonicelli, T.; Agarici, G.; Garbuglia, A.; Masiello, A.; Paolucci, F.; Simon, M.; Bailly-Maitre, L.; Bragulat, E.; Gomez, G.; Gutierrez, D.; Mico, G.; Moreno, J.-F.; Pilard, V.; Chakraborty, A.; Baruah, U.; Rotti, C.; Patel, H.; Nagaraju, M. V.; Singh, N. P.; Patel, A.; Dhola, H.; Raval, B.; Fantz, U.; Fröschle, M.; Heinemann, B.; Kraus, W.; Nocentini, R.; Riedl, R.; Schiesko, L.; Wimmer, C.; Wünderlich, D.; Cavenago, M.; Croci, G.; Gorini, G.; Rebai, M.; Muraro, A.; Tardocchi, M.; Hemsworth, R.

    2017-08-01

    SPIDER is one of two projects of the ITER Neutral Beam Test Facility under construction in Padova, Italy, at the Consorzio RFX premises. It will have a 100 keV beam source with a full-size prototype of the radiofrequency ion source for the ITER neutral beam injector (NBI) and also, similar to the ITER diagnostic neutral beam, it is designed to operate with a pulse length of up to 3600 s, featuring an ITER-like magnetic filter field configuration (for high extraction of negative ions) and caesium oven (for high production of negative ions) layout as well as a wide set of diagnostics. These features will allow a reproduction of the ion source operation in ITER, which cannot be done in any other existing test facility. SPIDER realization is well advanced and the first operation is expected at the beginning of 2018, with the mission of achieving the ITER heating and diagnostic NBI ion source requirements and of improving its performance in terms of reliability and availability. This paper mainly focuses on the preparation of the first SPIDER operations—integration and testing of SPIDER components, completion and implementation of diagnostics and control and formulation of operation and research plan, based on a staged strategy.

  7. Status of ITER neutral beam cell remote handling system

    Energy Technology Data Exchange (ETDEWEB)

    Sykes, N., E-mail: nick.sykes@ccfe.ac.uk [CCFE. Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Belcher, C. [Oxford Technologies Ltd, Abingdon OX14 1RJ (United Kingdom); Choi, C.-H. [ITER Organisation, CS90 046, 13067 St. Paul les Durance Cedex (France); Crofts, O. [CCFE. Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Crowe, R. [Oxford Technologies Ltd, Abingdon OX14 1RJ (United Kingdom); Damiani, C. [Fusion for Energy, C/Josep Pla 2, Torres Diagonal Litoral-B3, E-08019 Barcelona (Spain); Delavalle, S.; Meredith, L. [Oxford Technologies Ltd, Abingdon OX14 1RJ (United Kingdom); Mindham, T.; Raimbach, J. [CCFE. Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Tesini, A. [ITER Organisation, CS90 046, 13067 St. Paul les Durance Cedex (France); Van Uffelen, M. [Fusion for Energy, C/Josep Pla 2, Torres Diagonal Litoral-B3, E-08019 Barcelona (Spain)

    2013-10-15

    The ITER neutral beam cell will contain up to three heating neutral beams and one diagnostic neutral beam, and four upper ports. Though manual maintenance work is envisaged within the cell, when containment is breached, or the radiological protection is removed the maintenance must be conducted remotely. This maintenance constitutes the removal and replacement of line replaceable units, and their transport to and from a cask docked to the cell. A design of the remote handling system has been prepared to concept level which this paper describes including the development of a beam line transporter, beam source remote handling equipment, upper port remote handling equipment and equipment for the maintenance of the neutral shield. This equipment has been developed complete the planned maintenance tasks for the components of the neutral beam cell and to have inherent flexibility to enable as yet unforeseen tasks and recovery operations to be performed.

  8. Status of ITER neutral beam cell remote handling system

    CERN Document Server

    Sykes, N; Choi, C-H; Crofts, O; Crowe, R; Damiani, C; Delavalle, S; Meredith, L; Mindham, T; Raimbach, J; Tesini, A; Van Uffelen, M

    2013-01-01

    The ITER neutral beam cell will contain up to three heating neutral beams and one diagnostic neutral beam, and four upper ports. Though manual maintenance work is envisaged within the cell, when containment is breached, or the radiological protection is removed the maintenance must be conducted remotely. This maintenance constitutes the removal and replacement of line replaceable units, and their transport to and from a cask docked to the cell. A design of the remote handling system has been prepared to concept level which this paper describes including the development of a beam line transporter, beam source remote handling equipment, upper port remote handling equipment and equipment for the maintenance of the neutral shield. This equipment has been developed complete the planned maintenance tasks for the components of the neutral beam cell and to have inherent flexibility to enable as yet unforeseen tasks and recovery operations to be performed.

  9. Diagnostics of the ITER neutral beam test facility.

    Science.gov (United States)

    Pasqualotto, R; Serianni, G; Sonato, P; Agostini, M; Brombin, M; Croci, G; Dalla Palma, M; De Muri, M; Gazza, E; Gorini, G; Pomaro, N; Rizzolo, A; Spolaore, M; Zaniol, B

    2012-02-01

    The ITER heating neutral beam (HNB) injector, based on negative ions accelerated at 1 MV, will be tested and optimized in the SPIDER source and MITICA full injector prototypes, using a set of diagnostics not available on the ITER HNB. The RF source, where the H(-)∕D(-) production is enhanced by cesium evaporation, will be monitored with thermocouples, electrostatic probes, optical emission spectroscopy, cavity ring down, and laser absorption spectroscopy. The beam is analyzed by cooling water calorimetry, a short pulse instrumented calorimeter, beam emission spectroscopy, visible tomography, and neutron imaging. Design of the diagnostic systems is presented.

  10. Predictive Simulations of ITER Including Neutral Beam Driven Toroidal Rotation

    Energy Technology Data Exchange (ETDEWEB)

    Halpern, Federico D.; Kritz, Arnold H.; Bateman, Glenn; Pankin, Alexei Y.; Budny, Robert V.; McCune, Douglas C.

    2008-06-16

    Predictive simulations of ITER [R. Aymar et al., Plasma Phys. Control. Fusion 44, 519 2002] discharges are carried out for the 15 MA high confinement mode (H-mode) scenario using PTRANSP, the predictive version of the TRANSP code. The thermal and toroidal momentum transport equations are evolved using turbulent and neoclassical transport models. A predictive model is used to compute the temperature and width of the H-mode pedestal. The ITER simulations are carried out for neutral beam injection (NBI) heated plasmas, for ion cyclotron resonant frequency (ICRF) heated plasmas, and for plasmas heated with a mix of NBI and ICRF. It is shown that neutral beam injection drives toroidal rotation that improves the confinement and fusion power production in ITER. The scaling of fusion power with respect to the input power and to the pedestal temperature is studied. It is observed that, in simulations carried out using the momentum transport diffusivity computed using the GLF23 model [R.Waltz et al., Phys. Plasmas 4, 2482 (1997)], the fusion power increases with increasing injected beam power and central rotation frequency. It is found that the ITER target fusion power of 500 MW is produced with 20 MW of NBI power when the pedesta temperature is 3.5 keV. 2008 American Institute of Physics. [DOI: 10.1063/1.2931037

  11. ITER neutral beam system US conceptual design

    Energy Technology Data Exchange (ETDEWEB)

    Purgalis, P.

    1990-09-01

    In this document we present the US conceptual design of a neutral beam system for International Thermonuclear Experimental Reactor (ITER). The design incorporates a barium surface conversion D{sup {minus}} source feeding a linear array of accelerator channels. The system uses a dc accelerator with electrostatic quadrupoles for strong focusing. A high voltage power supply that is integrated with the accelerator is presented as an attractive option. A gas neutralizer is used and residual ions exiting the neutralizer are deflected to water-cooled dumps. Cryopanels are located at the accelerator exit to pump excess gas from the source and the neutralizer, and in the ion dump cavity to pump re-neutralized ions and neutralizer gas. All the above components are packaged in compact identical, independent modules which can be removed for remote maintenance. The neutral beam system delivers 75 MW of DO at 1.3 MeV, into three ports with a total of 9 modules arranged in stacks of three modules per port . To increase reliability each module is designed to deliver up to 10 MW; this allows eight modules operating at partial capacity to deliver the required power in the event one module is out of service, and provides 20% excess capacity to improve availability. Radiation protection is provided by shielding and by locating critical components in the source and accelerator 46.5 m from the torus centerline. Neutron shielding in the drift duct and neutralizer provides the added feature of limiting conductance and thus reducing gas flow to and from the torus.

  12. Assembly process of the ITER neutral beam injectors

    Energy Technology Data Exchange (ETDEWEB)

    Graceffa, J., E-mail: joseph.graceffa@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul lez Durance (France); Boilson, D.; Hemsworth, R.; Petrov, V.; Schunke, B.; Urbani, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul lez Durance (France); Pilard, V. [Fusion for Energy, C/ Josep Pla, n°2, Torres Diagonal Litoral, Edificio B3, 08019 Barcelona (Spain)

    2013-10-15

    The ITER neutral beam (NB) injectors are used for heating and diagnostics operations. There are 4 injectors in total, 3 heating neutral beam injectors (HNBs) and one diagnostic neutral beam injector (DNB). Two HNBs and the DNB will start injection into ITER during the hydrogen/helium phase of ITER operations. A third HNB is considered as an upgrade to the ITER heating systems, and the impact of the later installation and use of that injector have to be taken into account when considering the installation and assembly of the whole NB system. It is assumed that if a third HNB is to be installed, it will be installed before the nuclear phase of the ITER project. The total weight of one injector is around 1200 t and it is composed of 18 main components and 36 sets of shielding plates. The overall dimensions are length 20 m, height 10 m and width 5 m. Assembly of the first two HNBs and the DNB will start before the first plasma is produced in ITER, but as the time required to assemble one injector is estimated at around 1.5 year, the assembly will be divided into 2 steps, one prior to first plasma, and the second during the machine second assembly phase. To comply with this challenging schedule the assembly sequence has been defined to allow assembly of three first injectors in parallel. Due to the similar design between the DNB and HNBs it has been decided to use the same tools, which will be designed to accommodate the differences between the two sets of components. This reduces the global cost of the assembly and the overall assembly time for the injector system. The alignment and positioning of the injectors is a major consideration for the injector assembly as the alignment of the beamline components and the beam source are critical if good injector performance is to be achieved. The theoretical axes of the beams are defined relative to the duct liners which are installed in the NB ports. The concept adopted to achieve the required alignment accuracy is to use the

  13. The ITER neutral beam front end components integration

    Energy Technology Data Exchange (ETDEWEB)

    Urbani, M., E-mail: marc.urbani@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Hemsworth, R.; Schunke, B.; Graceffa, J.; Delmas, E.; Svensson, L.; Boilson, D. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Krylov, A.; Panasenkov, A. [RRC Kurchatov Institute, 1, Kurchatov Square, Moscow 123182 (Russian Federation); Agarici, G. [Fusion For Energy, C/Josep Pla 2, Torres Diagonal Litoral-B3, E-08019 Barcelona (Spain); Stafford Allen, R.; Jones, C.; Kalsey, M.; Muir, A.; Milnes, J. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Geli, F. [FGI Consulting, Le Garde d’Estienne, 4565 route du Puy Sainte Reparade, 13540 Puyricard (France); Sherlock, P. [AMEC Limited, Booths Park Chelford Road, Knutsford Cheshire WA16 8QZ (United Kingdom)

    2013-10-15

    The neutral beam (NB) system for ITER is composed of two heating neutral beam injectors (HNBs) and a diagnostic neutral beam injector (DNB). A third HNB can be installed as a future up-grade. This paper will present the design development of the components between the injectors and the tokamak; the so-called ‘front end components’: the drift duct consists of the NB bellows and the drift duct liner, the vacuum vessel pressure suppression system box (VVPSS box), the absolute valve, and the fast shutter. These components represent the key links between the ITER tokamak and the vessels of the NB injectors. The design of these components is demanding due to the different loads that these components will have to stand. The paper will describe the different design solutions which have to be implemented regarding the primary vacuum confinement, the power handling capability and the remote maintenance operations. The sizes of the components are determined by the large cross section of the neutral beam. The power handling capability is driven by the anticipated re-ionization of the neutral beam and the electromagnetic fields in this region. The drift duct bellows (with an inner diameter of 2.5 m) shall guarantee a leak tight vacuum enclosure during the vertical and radial displacements of the ITER vacuum vessel. The conductance of the VVPSS box must be maximized in the available space. The absolute valve remains a challenging development. The total leak rate through the valve must be ≤1 × 10{sup −8} Pa m{sup 3}/s when the valve is closed. Due to the radiation environment, the seals of the gate valve will be metallic. An R and D program has been launched to develop a suitable metallic seal solution with the required dimensions. The maximum allowed closing time for the fast shutter shall be less than 1 s. For all these components the leak tightness will be guaranteed by a welded lip seal and the mechanical stability by bolted structures.

  14. Beyond ITER: neutral beams for a demonstration fusion reactor (DEMO) (invited).

    Science.gov (United States)

    McAdams, R

    2014-02-01

    In the development of magnetically confined fusion as an economically sustainable power source, International Tokamak Experimental Reactor (ITER) is currently under construction. Beyond ITER is the demonstration fusion reactor (DEMO) programme in which the physics and engineering aspects of a future fusion power plant will be demonstrated. DEMO will produce net electrical power. The DEMO programme will be outlined and the role of neutral beams for heating and current drive will be described. In particular, the importance of the efficiency of neutral beam systems in terms of injected neutral beam power compared to wallplug power will be discussed. Options for improving this efficiency including advanced neutralisers and energy recovery are discussed.

  15. Beyond ITER: Neutral beams for a demonstration fusion reactor (DEMO) (invited)

    Energy Technology Data Exchange (ETDEWEB)

    McAdams, R., E-mail: roy.mcadams@ccfe.ac.uk [EURATOM/CCFE Association, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom)

    2014-02-15

    In the development of magnetically confined fusion as an economically sustainable power source, International Tokamak Experimental Reactor (ITER) is currently under construction. Beyond ITER is the demonstration fusion reactor (DEMO) programme in which the physics and engineering aspects of a future fusion power plant will be demonstrated. DEMO will produce net electrical power. The DEMO programme will be outlined and the role of neutral beams for heating and current drive will be described. In particular, the importance of the efficiency of neutral beam systems in terms of injected neutral beam power compared to wallplug power will be discussed. Options for improving this efficiency including advanced neutralisers and energy recovery are discussed.

  16. Recent improvements to the ITER neutral beam system design

    Energy Technology Data Exchange (ETDEWEB)

    Grisham, L.R., E-mail: lgrisham@pppl.gov [Princeton University, Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Agostinetti, P. [Consorzio RFX, Euratom-ENEA Association, C.so Stati Uniti 4, I-35127 Padova (Italy); Barrera, G. [EURATOM-CIEMAT Association, Avda. Complutense 40, 28040 Madrid (Spain); Blatchford, P. [Culham Center for Fusion Energy, Abingdon, Oxon. OX14 3DB (United Kingdom); Boilson, D.; Chareyre, J. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Chitarin, G. [Consorzio RFX, Euratom-ENEA Association, C.so Stati Uniti 4, I-35127 Padova (Italy); Esch, H.P.L. de [CEA-Cadarache, IRFM, F-13108 Saint-Paul-lez-Durance (France); De Lorenzi, A. [Consorzio RFX, Euratom-ENEA Association, C.so Stati Uniti 4, I-35127 Padova (Italy); Franzen, P.; Fantz, U. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, D-85748 Garching (Germany); Gagliardi, M. [Culham Center for Fusion Energy, Abingdon, Oxon. OX14 3DB (United Kingdom); Hemsworth, R.S. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Kashiwagi, M. [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); King, D. [Culham Center for Fusion Energy, Abingdon, Oxon. OX14 3DB (United Kingdom); Krylov, A. [Russian Research Centre, Kurchatov Institute, Moscow (Russian Federation); Kuriyama, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Marconato, N.; Marcuzzi, D. [Consorzio RFX, Euratom-ENEA Association, C.so Stati Uniti 4, I-35127 Padova (Italy); Roccella, M. [L.T. Calcoli SaS, Via C. Baslini 13, 23807 Merate (Italy); and others

    2012-11-15

    Highlights: Black-Right-Pointing-Pointer Improvements to ITER accelerator voltage holding. Black-Right-Pointing-Pointer Improvements to ITER negative ion source design. Black-Right-Pointing-Pointer Improvements to ITER megavolt bushing. Black-Right-Pointing-Pointer Improvements to beamline components. Black-Right-Pointing-Pointer Accelerator design improvements. - Abstract: The ITER [1] fusion device is expected to demonstrate the feasibility of magnetically confined deuterium-tritium plasma as an energy source which might one day lead to practical power plants. Injection of energetic beams of neutral atoms (up to 1 MeV D{sup 0} or up to 870 keV H{sup 0}) will be one of the primary methods used for heating the plasma, and for driving toroidal electrical current within it, the latter being essential in producing the required magnetic confinement field configuration. The design calls for each beamline to inject up to 16.5 MW of power through the duct into the tokamak, with an initial complement of two beamlines injecting parallel to the direction of the current arising from the tokamak transformer effect, and with the possibility of eventually adding a third beamline, also in the co-current direction. The general design of the beamlines has taken shape over the past 17 years [2], and is now predicated upon an RF-driven negative ion source based upon the line of sources developed by the Institute for Plasma Physics (IPP) at Garching during recent decades [3-5], and a multiple-aperture multiple-grid electrostatic accelerator derived from negative ion accelerators developed by the Japan Atomic Energy Agency (JAEA) across a similar span of time [6-8]. During the past years, the basic concept of the beam system has been further refined and developed, and assessment of suitable fabrication techniques has begun. While many design details which will be important to the installation and implementation of the ITER beams have been worked out during this time, this paper focuses

  17. Status of PRIMA, the test facility for ITER neutral beam injectors

    Science.gov (United States)

    Sonato, P.; Antoni, V.; Bigi, M.; Chitarin, G.; Luchetta, A.; Marcuzzi, D.; Pasqualotto, R.; Pomaro, N.; Serianni, G.; Toigo, V.; Zaccaria, P.; ITER International Team

    2013-02-01

    The ITER project requires additional heating by two neutral beam injectors, each accelerating to 1MV a 40A beam of negative deuterons, delivering to the plasma about 17MW up to one hour. As these requirements have never been experimentally met, it was decided to build a test facility, PRIMA (Padova Research on ITER Megavolt Accelerator), in Italy, including a full-size negative ion source, SPIDER, and a prototype of the whole ITER injector, MITICA, aiming to develop the heating injectors to be installed in ITER. The Japan and the India Domestic Agencies participate in the PRIMA enterprise; European laboratories, such as KIT-Karlsruhe, IPP-Garching, CCFE-Culham, CEA-Cadarache and others are also cooperating. In the paper the main requirements are discussed and the design of the main components and systems are described.

  18. The PRIMA Test Facility: SPIDER and MITICA test-beds for ITER neutral beam injectors

    Science.gov (United States)

    Toigo, V.; Piovan, R.; Dal Bello, S.; Gaio, E.; Luchetta, A.; Pasqualotto, R.; Zaccaria, P.; Bigi, M.; Chitarin, G.; Marcuzzi, D.; Pomaro, N.; Serianni, G.; Agostinetti, P.; Agostini, M.; Antoni, V.; Aprile, D.; Baltador, C.; Barbisan, M.; Battistella, M.; Boldrin, M.; Brombin, M.; Dalla Palma, M.; De Lorenzi, A.; Delogu, R.; De Muri, M.; Fellin, F.; Ferro, A.; Fiorentin, A.; Gambetta, G.; Gnesotto, F.; Grando, L.; Jain, P.; Maistrello, A.; Manduchi, G.; Marconato, N.; Moresco, M.; Ocello, E.; Pavei, M.; Peruzzo, S.; Pilan, N.; Pimazzoni, A.; Recchia, M.; Rizzolo, A.; Rostagni, G.; Sartori, E.; Siragusa, M.; Sonato, P.; Sottocornola, A.; Spada, E.; Spagnolo, S.; Spolaore, M.; Taliercio, C.; Valente, M.; Veltri, P.; Zamengo, A.; Zaniol, B.; Zanotto, L.; Zaupa, M.; Boilson, D.; Graceffa, J.; Svensson, L.; Schunke, B.; Decamps, H.; Urbani, M.; Kushwah, M.; Chareyre, J.; Singh, M.; Bonicelli, T.; Agarici, G.; Garbuglia, A.; Masiello, A.; Paolucci, F.; Simon, M.; Bailly-Maitre, L.; Bragulat, E.; Gomez, G.; Gutierrez, D.; Mico, G.; Moreno, J.-F.; Pilard, V.; Kashiwagi, M.; Hanada, M.; Tobari, H.; Watanabe, K.; Maejima, T.; Kojima, A.; Umeda, N.; Yamanaka, H.; Chakraborty, A.; Baruah, U.; Rotti, C.; Patel, H.; Nagaraju, M. V.; Singh, N. P.; Patel, A.; Dhola, H.; Raval, B.; Fantz, U.; Heinemann, B.; Kraus, W.; Hanke, S.; Hauer, V.; Ochoa, S.; Blatchford, P.; Chuilon, B.; Xue, Y.; De Esch, H. P. L.; Hemsworth, R.; Croci, G.; Gorini, G.; Rebai, M.; Muraro, A.; Tardocchi, M.; Cavenago, M.; D'Arienzo, M.; Sandri, S.; Tonti, A.

    2017-08-01

    The ITER Neutral Beam Test Facility (NBTF), called PRIMA (Padova Research on ITER Megavolt Accelerator), is hosted in Padova, Italy and includes two experiments: MITICA, the full-scale prototype of the ITER heating neutral beam injector, and SPIDER, the full-size radio frequency negative-ions source. The NBTF realization and the exploitation of SPIDER and MITICA have been recognized as necessary to make the future operation of the ITER heating neutral beam injectors efficient and reliable, fundamental to the achievement of thermonuclear-relevant plasma parameters in ITER. This paper reports on design and R&D carried out to construct PRIMA, SPIDER and MITICA, and highlights the huge progress made in just a few years, from the signature of the agreement for the NBTF realization in 2011, up to now—when the buildings and relevant infrastructures have been completed, SPIDER is entering the integrated commissioning phase and the procurements of several MITICA components are at a well advanced stage.

  19. Maintenance schemes for the ITER neutral beam test facility

    Energy Technology Data Exchange (ETDEWEB)

    Zaccaria, P. [Consorzio RFX, Association EURATOM-ENEA, I-35127 Padova (Italy)]. E-mail: pierluigi.zaccaria@igi.cnr.it; Dal Bello, S. [Consorzio RFX, Association EURATOM-ENEA, I-35127 Padova (Italy); Marcuzzi, D. [Consorzio RFX, Association EURATOM-ENEA, I-35127 Padova (Italy); Masiello, A. [Consorzio RFX, Association EURATOM-ENEA, I-35127 Padova (Italy); Cordier, J.J. [Association EURATOM-CEA, DSM/Departement Recherche Fusion Controlee, CEA/Cadarache, F-13108 Saint Paul Lez Durance Cedex (France); Hemsworth, R. [Association EURATOM-CEA, DSM/Departement Recherche Fusion Controlee, CEA/Cadarache, F-13108 Saint Paul Lez Durance Cedex (France); Antipenkov, A. [Forschungszentrum Karlsruhe, Institut fuer Technische Physik, 76021 Karlsruhe (Germany); Day, C. [Forschungszentrum Karlsruhe, Institut fuer Technische Physik, 76021 Karlsruhe (Germany); Dremel, M. [Forschungszentrum Karlsruhe, Institut fuer Technische Physik, 76021 Karlsruhe (Germany); Mack, A. [Forschungszentrum Karlsruhe, Institut fuer Technische Physik, 76021 Karlsruhe (Germany); Jones, T. [UKAEA Culham EURATOM/UKAEA Fusion Association Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Coniglio, A. [Consorzio RFX, Association EURATOM-ENEA, I-35127 Padova (Italy); Pillon, M. [ENEA, Centro Ricerche Frascati, I-00044 Frascati, Rome (Italy); Sandri, S. [ENEA, Centro Ricerche Frascati, I-00044 Frascati, Rome (Italy); Speth, E. [IPP CSU-Max-Planck-Institut fuer Plasma Physik, D-85748 Garching (Germany); Tanga, A. [IPP CSU-Max-Planck-Institut fuer Plasma Physik, D-85748 Garching (Germany); Antoni, V. [Consorzio RFX, Association EURATOM-ENEA, I-35127 Padova (Italy); Pietro, E. Di [EFDA CSU, D-85748 Garching (Germany); Mondino, P.L. [EFDA CSU, D-85748 Garching (Germany)

    2005-11-15

    The ITER neutral beam test facility (NBTF) is planned to be built, after the approval of the ITER construction and the choice of the ITER site, with the agreement of the ITER international team and of the JA and RF participant teams. The key purpose is to progressively increase the performance of the first ITER injector and to demonstrate its reliability at the maximum operation parameters: power delivered to the plasma 16.5 MW, beam energy 1 MeV, accelerated D{sup -} ion current 40 A, pulse length 3600 s. Several interventions for possible modifications and for maintenance are expected during the early operation of the ITER injector in order to optimise the beam generation, aiming and steering. The maintenance scheme and the related design solutions are therefore a very important aspect to be considered for the NBTF design. The paper describes consistently the many interrelated aspects of the design, such as the optimisation of the vessel and cryopump geometry, in order to get a better maintenance flexibility, an easier man access and a larger access for diagnostic and monitoring.

  20. Maintenance schemes for the ITER neutral beam test facility

    Energy Technology Data Exchange (ETDEWEB)

    Zaccaria, P.; Dal Bello, S.; Marcuzzi, D.; Masiello, A.; Coniglio, A.; Antoni, V. [Consorzio RFX Association Euratom-ENEA, Padova (Italy); Cordier, J.J.; Hemsworth, R. [Association Euratom-CEA Cadarache (DSM/DRFC), 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Antipenkov, A.; Day, C.; Dremel, M.; Mack, A. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Technische Physik; Pillon, M.; Sandri, S. [ENEA, Frascati (Italy). Centro Ricerche Energia; Speth, E.; Tanga, A. [Max-Planck-Institut fuer Plasmaphysik, IPP CSU, Garching (Germany); Jones, T. [UKAEA Culham Euratom/Ukaea Fusion Association Culham Science Centre, Abingdom OX (United Kingdom); Di Pietro, E.; Mondino, P.L. [EFDA CSU, Garching (Germany)

    2004-07-01

    The ITER neutral beam test facility (NBTF) is planned to be built, after the approval of the ITER construction and the choice of the ITER site, with the agreement of the ITER International Team and of the JA and RF participant teams. The key purpose is to progressively increase the performance of the first ITER injector and to demonstrate its reliability at the maximum operation parameters: power delivered to the plasma 16.5 MW, beam energy 1 MeV, accelerated D{sup -} ion current 40 A, pulse length 3600 s. Several interventions for possible modifications and for maintenance are expected during the early operation of the ITER injector in order to optimize the beam generation, aiming and steering. The maintenance scheme and the related design solutions are therefore a very important aspect to be considered for the NBTF design. The paper describes consistently the many interrelated aspects of the design, such as the optimisation of the vessel and cryopump geometry, in order to get a better maintenance flexibility, an easier man access and a larger access for diagnostic and monitoring. (authors)

  1. Magnetic analysis of the magnetic field reduction system of the ITER neutral beam injector

    Energy Technology Data Exchange (ETDEWEB)

    Barrera, Germán, E-mail: german.barrera@ciemat.es [CIEMAT, Laboratorio Nacional de Fusión, Avda. Complutense 22, 28040 Madrid (Spain); Ahedo, Begoña; Alonso, Javier; Ríos, Luis [CIEMAT, Laboratorio Nacional de Fusión, Avda. Complutense 22, 28040 Madrid (Spain); Chareyre, Julien; El-Ouazzani, Anass [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Agarici, Gilbert [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 07/08, 08019 Barcelona (Spain)

    2015-10-15

    The neutral beam system for ITER consists of two heating and current drive neutral beam injectors (HNB) and a diagnostic neutral beam (DNB) injector. The proposed physical plant layout allows a possible third HNB injector to be installed later. For the correct operation of the beam, the ion source and the ion path until it is neutralized must operate under a very low magnetic field environment. To prevent the stray ITER field from penetrating inside those mentioned critical areas, a magnetic field reduction system (MFRS) will envelop the beam vessels and the high voltage transmission lines to ion source. This system comprises the passive magnetic shield (PMS), a box like assembly of thick low carbon steel plates, and the Active Correction and Compensation Coils (ACCC), a set of coils carrying a current which depends on the tokamak stray field. This paper describes the magnetic model and analysis results presented at the PMS and ACCC preliminary design review held in ITER organization in April 2013. The paper focuses on the magnetic model description and on the description of the analysis results. The iterative process for obtaining optimized currents in the coils is presented. The set of coils currents chosen among the many possible solutions, the magnetic field results in the interest regions and the fulfillment of the magnetic field requirements are described.

  2. ITER neutral beam system US conceptual design. Final vesion

    Energy Technology Data Exchange (ETDEWEB)

    Purgalis, P.

    1990-09-01

    In this document we present the US conceptual design of a neutral beam system for International Thermonuclear Experimental Reactor (ITER). The design incorporates a barium surface conversion D{sup {minus}} source feeding a linear array of accelerator channels. The system uses a dc accelerator with electrostatic quadrupoles for strong focusing. A high voltage power supply that is integrated with the accelerator is presented as an attractive option. A gas neutralizer is used and residual ions exiting the neutralizer are deflected to water-cooled dumps. Cryopanels are located at the accelerator exit to pump excess gas from the source and the neutralizer, and in the ion dump cavity to pump re-neutralized ions and neutralizer gas. All the above components are packaged in compact identical, independent modules which can be removed for remote maintenance. The neutral beam system delivers 75 MW of DO at 1.3 MeV, into three ports with a total of 9 modules arranged in stacks of three modules per port . To increase reliability each module is designed to deliver up to 10 MW; this allows eight modules operating at partial capacity to deliver the required power in the event one module is out of service, and provides 20% excess capacity to improve availability. Radiation protection is provided by shielding and by locating critical components in the source and accelerator 46.5 m from the torus centerline. Neutron shielding in the drift duct and neutralizer provides the added feature of limiting conductance and thus reducing gas flow to and from the torus.

  3. Overview of the negative ion based neutral beam injectors for ITER.

    Science.gov (United States)

    Schunke, B; Boilson, D; Chareyre, J; Choi, C-H; Decamps, H; El-Ouazzani, A; Geli, F; Graceffa, J; Hemsworth, R; Kushwah, M; Roux, K; Shah, D; Singh, M; Svensson, L; Urbani, M

    2016-02-01

    The ITER baseline foresees 2 Heating Neutral Beams (HNB's) based on 1 MeV 40 A D(-) negative ion accelerators, each capable of delivering 16.7 MW of deuterium atoms to the DT plasma, with an optional 3rd HNB injector foreseen as a possible upgrade. In addition, a dedicated diagnostic neutral beam will be injecting ≈22 A of H(0) at 100 keV as the probe beam for charge exchange recombination spectroscopy. The integration of the injectors into the ITER plant is nearly finished necessitating only refinements. A large number of components have passed the final design stage, manufacturing has started, and the essential test beds-for the prototype route chosen-will soon be ready to start.

  4. Overview of the negative ion based neutral beam injectors for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Schunke, B., E-mail: email@none.edu; Boilson, D.; Chareyre, J.; Choi, C.-H.; Decamps, H.; El-Ouazzani, A.; Geli, F.; Graceffa, J.; Hemsworth, R.; Kushwah, M.; Roux, K.; Shah, D.; Singh, M.; Svensson, L.; Urbani, M. [ITER Organization, Route de Vinon-sur-Verdon, 13115 St Paul lez Durance (France)

    2016-02-15

    The ITER baseline foresees 2 Heating Neutral Beams (HNB’s) based on 1 MeV 40 A D{sup −} negative ion accelerators, each capable of delivering 16.7 MW of deuterium atoms to the DT plasma, with an optional 3rd HNB injector foreseen as a possible upgrade. In addition, a dedicated diagnostic neutral beam will be injecting ≈22 A of H{sup 0} at 100 keV as the probe beam for charge exchange recombination spectroscopy. The integration of the injectors into the ITER plant is nearly finished necessitating only refinements. A large number of components have passed the final design stage, manufacturing has started, and the essential test beds—for the prototype route chosen—will soon be ready to start.

  5. Progress on the heating and current drive systems for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Jacquinot, J. [CEA, Cadarache, France; Beaumont, Bertrand [ITER Joint Work Site, Cadarache; Bora, D. [ITER Joint Work Site, Cadarache; Campbell, D. [ITER Joint Work Site, Cadarache; Darbos, Caroline [ITER Joint Work Site, Cadarache; Decamps, H. [ITER Organization, Saint Paul Lez Durance, France; Graceffa, J. [ITER Joint Work Site, Cadarache; Gassmann, T. [ITER Joint Work Site, Cadarache; Hemsworth, R. [ITER Joint Work Site, Cadarache; Henderson, Mark [ITER Joint Work Site, Cadarache; Kobayashi, N. [ITER Joint Work Site, Cadarache; Lamalle, Philippe [ITER Joint Work Site, Cadarache; Schunke, B. [ITER Joint Work Site, Cadarache; Tanaka, M. [ITER Joint Work Site, Cadarache; Tanga, A. [ITER Joint Work Site, Cadarache; Albajar, F. [Fusion for Energy (F4E), Barcelona, Spain; Bonicelli, T. [Fusion for Energy (F4E), Barcelona, Spain; Saibene, G. [Fusion for Energy (F4E), Barcelona, Spain; Sartori, R. [Fusion for Energy (F4E), Barcelona, Spain; Becoulet, A. [CEA, Cadarache, France; Hoang, G. T. [CEA, Cadarache, France; Inoue, T. [Japan Atomic Energy Agency (JAEA), Naka; Sakamoto, K. [Japan Atomic Energy Agency (JAEA), Naka; Takahashi, K. [Japan Atomic Energy Agency (JAEA), Naka; Watanabe, K. [Japan Atomic Energy Agency (JAEA), Naka; Goulding, Richard Howell [ORNL; Rasmussen, David A [ORNL; Swain, David W [ORNL; Chakraborty, A. [ITER India - Bhat, Gandhinagar, Gujarat; Mukherjee, A. [ITER India - Bhat, Gandhinagar, Gujarat; Rao, S. L. [ITER India - Bhat, Gandhinagar, Gujarat; Denisov, G. [Russian Academy of Science, Novgorod, Russia; Nightingale, M. [EURATOM / UKAEA, Abingdon, UK; Sonato, P. [EURATOM / ENEA, Italy

    2009-06-01

    The electron cyclotron (EC), ion cyclotron (IC), heating-neutral beam (H-NB) and, although not in the day 1 baseline, lower hybrid (LH) systems intended for ITER have been reviewed in 2007/2008 in light of progress of physics and technology in the field. Although the overall specifications are unchanged, notable changes have been approved. Firstly, it has been emphasized that the H&CD systems are vital for the ITER programme. Consequently, the full 73 MW should be commissioned and available on a routine basis before the D/T phase. Secondly, significant changes have been approved at system level, most notably: the possibility to operate the heating beams at full power during the hydrogen phase requiring new shine through protection; the possibility to operate IC with 2 antennas with increased robustness (no moving parts); the possible increase to 2 MW of key components of the EC transmission systems in order to provide an easier upgrading of the EC power as may be required by the project; the addition of a building dedicated to the RF power sources and to a testing facility for acceptance of diagnostics and heating port plugs. Thirdly, the need of a plan for developing, in time for the active phase, a CD system such as LH suitable for very long pulse operation of ITER was recognised. The review describes these changes and their rationale.

  6. Modelling and shielding analysis of the neutral beam injector ports in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Pereslavtsev, P., E-mail: pavel.pereslavtsev@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Fischer, U. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Loughlin, M. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Lu, Lei [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Polunovskiy, E. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Vielhaber, S. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2015-10-15

    Highlights: • The engineering CAD models of the NBI ports were simplified on the CATIA platform. • CAD to MCNP model convesion was done making use of McCAD converting tool. • The new NBI port model was integrated into 80° A-lite ITER torus sector model. • The nuclear responces important for the safety issues were assessed. - Abstract: A new MCNP geometry model of the ITER Neutral Beam Injection (NBI) ports was developed starting from the latest engineering CAD models provided by ITER. The model includes 3 heating (HNBI) ports and one diagnostic port (DNBI), and extends up to the bio-shield. The engineering CAD models were simplified on the CATIA platform according to the neutronic requirements and then converted into MCNP geometry making use of the McCad conversion tool. Finally, the new NBI port model was integrated into an available 80° A-lite ITER torus sector model. The nuclear analysis performed on the basis of this model provides the following nuclear responses: the neutron flux distribution in all NBI ports, the nuclear heating distribution in all NBI ducts; the nuclear heating and radiation loads to the TFC magnets; the radiation damage and gas production in the VV; and the distribution of the shutdown dose rate inside the cryostat.

  7. An alpha particle measurement system using an energetic neutral helium beam in ITER (invited).

    Science.gov (United States)

    Sasao, M; Kisaki, M; Kobuchi, T; Tsumori, K; Tanaka, N; Terai, K; Okamoto, A; Kitajima, S; Kaneko, O; Shinto, K; Wada, M

    2012-02-01

    An energetic helium neutral beam is involved in the beam neutralization measurement system of alpha particles confined in a DT fusion plasma. A full size strong-focusing He(+) ion source (2 A, the beam radius of 11.3 mm, the beam energy less than 20 keV). Present strong-focusing He(+) ion source shows an emittance diagram separated for each beamlet of multiple apertures without phase space mixing, despite the space charge of a beamlet is asymmetric and the beam flow is non-laminar. The emittance of beamlets in the peripheral region was larger than that of center. The heat load to the plasma electrode was studied to estimate the duty factor for the ITER application.

  8. The ITER neutral beam test facility: Designs of the general infrastructure, cryosystem and cooling plant

    Energy Technology Data Exchange (ETDEWEB)

    Cordier, J.J. [Association EURATOM-CEA, DSM, Departement Recherche Fusion Controlee, CEA/Cadarache, bat 506, F-13108 Saint Paul Lez Durance Cedex (France)]. E-mail: jean-jacques.cordier@cea.fr; Hemsworth, R. [Association EURATOM-CEA, DSM, Departement Recherche Fusion Controlee, CEA/Cadarache, bat 506, F-13108 Saint Paul Lez Durance Cedex (France); Chantant, M. [Association EURATOM-CEA, DSM, Departement Recherche Fusion Controlee, CEA/Cadarache, bat 506, F-13108 Saint Paul Lez Durance Cedex (France); Gravil, B. [Association EURATOM-CEA, DSM, Departement Recherche Fusion Controlee, CEA/Cadarache, bat 506, F-13108 Saint Paul Lez Durance Cedex (France); Henry, D. [Association EURATOM-CEA, DSM, Departement Recherche Fusion Controlee, CEA/Cadarache, bat 506, F-13108 Saint Paul Lez Durance Cedex (France); Sabathier, F. [Association EURATOM-CEA, DSM, Departement Recherche Fusion Controlee, CEA/Cadarache, bat 506, F-13108 Saint Paul Lez Durance Cedex (France); Doceul, L. [Association EURATOM-CEA, DSM, Departement Recherche Fusion Controlee, CEA/Cadarache, bat 506, F-13108 Saint Paul Lez Durance Cedex (France); Thomas, E. [Association EURATOM-CEA, DSM, Departement Recherche Fusion Controlee, CEA/Cadarache, bat 506, F-13108 Saint Paul Lez Durance Cedex (France); Houtte, D. van [Association EURATOM-CEA, DSM, Departement Recherche Fusion Controlee, CEA/Cadarache, bat 506, F-13108 Saint Paul Lez Durance Cedex (France); Zaccaria, P. [CONSORZIO RFX Association EURATOM-ENEA, Corso Stati Uniti 4, I-35127 Padova (Italy); Antoni, V. [CONSORZIO RFX Association EURATOM-ENEA, Corso Stati Uniti 4, I-35127 Padova (Italy); Bello, S. Dal; Marcuzzi, D. [CONSORZIO RFX Association EURATOM-ENEA, Corso Stati Uniti 4, I-35127 Padova (Italy); Antipenkov, A.; Day, C.; Dremel, M. [FZK, Institut fuer Technische Physik, Karlsruhe 76021 (Germany); Mondino, P.L. [EFDA CSU, Max-Planck-Institut fuer Plasma Physik Boltzmannstr. 2, D-85748 Garching (Germany)

    2005-11-15

    The CEA Association is involved, in close collaboration with ENEA, FZK, IPP and UKAEA European Associations, in the first ITER neutral beam (NB) injector and the ITER neutral beam test facility design (EFDA task ref. TW3-THHN-IITF1). A total power of about 50 MW will have to be removed in steady state on the neutral beam test facility (NBTF). The main purpose of this task is to make progress with the detailed design of the first ITER NB injector and to start the conceptual design of the ITER NBTF. The general infrastructure layout of a generic site for the NBTF includes the test facility itself equipped with a dedicated beamline vessel [P.L. Zaccaria, et al., Maintenance schemes for the ITER neutral beam test facility, this conference] and integration studies of associated auxiliaries such as cooling plant, cryoplant and forepumping system.

  9. ITER (International Thermonuclear Experimental Reactor) current drive and heating physics

    Energy Technology Data Exchange (ETDEWEB)

    Nevins, W.M.; Lindquist, W. (Lawrence Livermore National Lab., CA (USA)); Fujisawa, N.; Kimura, H. (Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan)); Hopman, H.; Rebuffi, L.; Wegrowe, J.G. (Max-Planck-Institut fuer Plasmaphysik, Garching (Germany, F.R.). NET Design Team); Parail, V.; Vdovin, V. (Gosudarstvennyj Komitet po Ispol' zovaniyu Atomnoj Ehnergii SSSR, Moscow (USSR). Inst. Atomnoj Ehn

    1990-01-01

    The ITER Current Drive and Heating (CD H) systems are required for: Ionization and current initiation; Non-inductive current ramp-up assist; Heating of the plasma; Steady-state operation with full non-inductive current drive; Current profile control; and Burn control by modulation of the auxiliary power. Steady-state current drive is the most demanding requirement, so this has driven the choice of the ITER current drive and heating systems.

  10. Physics design of the injector source for ITER neutral beam injector (invited).

    Science.gov (United States)

    Antoni, V; Agostinetti, P; Aprile, D; Cavenago, M; Chitarin, G; Fonnesu, N; Marconato, N; Pilan, N; Sartori, E; Serianni, G; Veltri, P

    2014-02-01

    Two Neutral Beam Injectors (NBI) are foreseen to provide a substantial fraction of the heating power necessary to ignite thermonuclear fusion reactions in ITER. The development of the NBI system at unprecedented parameters (40 A of negative ion current accelerated up to 1 MV) requires the realization of a full scale prototype, to be tested and optimized at the Test Facility under construction in Padova (Italy). The beam source is the key component of the system and the design of the multi-grid accelerator is the goal of a multi-national collaborative effort. In particular, beam steering is a challenging aspect, being a tradeoff between requirements of the optics and real grids with finite thickness and thermo-mechanical constraints due to the cooling needs and the presence of permanent magnets. In the paper, a review of the accelerator physics and an overview of the whole R&D physics program aimed to the development of the injector source are presented.

  11. Physics design of the injector source for ITER neutral beam injector (invited)

    Energy Technology Data Exchange (ETDEWEB)

    Antoni, V.; Agostinetti, P.; Aprile, D.; Chitarin, G.; Fonnesu, N.; Marconato, N.; Pilan, N.; Sartori, E.; Serianni, G., E-mail: gianluigi.serianni@igi.cnr.it; Veltri, P. [Consorzio RFX, Associazione EURATOM-ENEA sulla fusione, c.so Stati Uniti 4, 35127 Padova (Italy); Cavenago, M. [INFN-LNL, viale dell’Università n. 2, 35020 Legnaro (Italy)

    2014-02-15

    Two Neutral Beam Injectors (NBI) are foreseen to provide a substantial fraction of the heating power necessary to ignite thermonuclear fusion reactions in ITER. The development of the NBI system at unprecedented parameters (40 A of negative ion current accelerated up to 1 MV) requires the realization of a full scale prototype, to be tested and optimized at the Test Facility under construction in Padova (Italy). The beam source is the key component of the system and the design of the multi-grid accelerator is the goal of a multi-national collaborative effort. In particular, beam steering is a challenging aspect, being a tradeoff between requirements of the optics and real grids with finite thickness and thermo-mechanical constraints due to the cooling needs and the presence of permanent magnets. In the paper, a review of the accelerator physics and an overview of the whole R and D physics program aimed to the development of the injector source are presented.

  12. The influence of grid positioning on the beam optics in the neutral beam injectors for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Veltri, Pierluigi, E-mail: pierluigi.veltri@igi.cnr.it [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), Corso Stati Uniti 4, Padova (Italy); INFN—Laboratori Nazionali di Legnaro, Viale dell’Università 2, 35020 Legnaro, Padova (Italy); Agostinetti, Piero; Marcuzzi, Diego; Sartori, Emanuele; Serianni, Gianluigi [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), Corso Stati Uniti 4, Padova (Italy)

    2016-06-15

    Neutral beam injectors are routinely used to increase the ion temperature in magnetically confined plasmas. Typically, the beam is produced by neutralizing a bundle of hundreds of ion beamlets, energized in a multi-grid multi-stage accelerator. Precise aiming of each beamlet is required in order to focus the full beam to the plasma, avoiding any interception with beamline surfaces and with the beam duct. This paper describes the effects of grid in-plane and out-of-plane displacements (mispositioning, thermal expansion, grid tilting, etc…) in the case of the MITICA electrostatic accelerator, which is the full scale prototype of the ITER heating neutral beam injector. Various simulations have been carried out with the OPERA 3D code, by self-consistently simulating the beam charged particles travelling in an externally applied electric and magnetic field. The accelerator grids act like a series of electrostatic lenses, and produce a net deflection of the particles when one or more grids are offset. The numerical simulations were used to evaluate the “steering constant” of each grid and also showed that the linear superposition of effects was applicable, multiple causes of mispositioning are combined and used to quantify the overall effect in terms of beam misalignment.

  13. Alpha Heating in ITER L-mode and H-mode Plasma

    Energy Technology Data Exchange (ETDEWEB)

    R.V. Budny

    2011-07-18

    There are many uses of predictions of ITER plasma performance. One is assessing requirements of different plasma regimes. For instance, what current drive and control are needed for steady state. The heating, current drive, and torque systems planned for initial DT operation are negative ion neutral beam injection (NB), ion cyclotron resonance (IC), and electron cyclotron resonance (EC). Which combinations of heating are optimal. What are benefits of the torques, current drive, and fueling using NB. What are the shine-through power and optimum voltage for the NB? What are optimal locations and aiming of the EC launchers? Another application is nuclear licensing (e.g. System integrity, how many neutrons).

  14. Conceptual design of a compact absolute valve for the ITER neutral beam injectors

    Energy Technology Data Exchange (ETDEWEB)

    Jones, Chris [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom)], E-mail: chris.m.jones@jet.uk; Waldon, Chris; Martin, David; Watson, Mike [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Sonderegger, Kurt; Lenherr, Bruno [VAT Vakuumventile AG, CH-9469 Haag (Switzerland); Andrews, Ian; Mansbridge, Simon [VAT Vacuum Products Ltd., Edmund House, Rugby Road, Leamington Spa, Warwickshire CV32 6EL (United Kingdom)

    2009-06-15

    The reference design for the ITER neutral beam injectors incorporated a fast shutter to limit tritium migration to the injector vacuum enclosures. In 2005, a need for an 'absolute' isolation valve was identified to facilitate injector maintenance procedures and protect the system from an in-vessel ingress of coolant event (ICE). An outline concept for an all-metal seal valve was developed during 2006, in close cooperation with the Swiss valve manufacturer VAT. During the following year, it became apparent that the length of beamline available for the valve was significantly less than originally envisaged, resulting in a radical revision of the design concept. A casing length of 760 mm has been achieved by means of major changes to the casing structure, plate dimensions, pendulum mechanism and seal actuators. A concept for a seal protection system has been developed to prevent beam line contamination reaching the valve components and to protect the valve plate from surface heating by plasma radiation. The new design concept has been extensively validated by analysis, including a whole-system FE model of the valve.

  15. Low pressure and high power rf sources for negative hydrogen ions for fusion applications (ITER neutral beam injection).

    Science.gov (United States)

    Fantz, U; Franzen, P; Kraus, W; Falter, H D; Berger, M; Christ-Koch, S; Fröschle, M; Gutser, R; Heinemann, B; Martens, C; McNeely, P; Riedl, R; Speth, E; Wünderlich, D

    2008-02-01

    The international fusion experiment ITER requires for the plasma heating and current drive a neutral beam injection system based on negative hydrogen ion sources at 0.3 Pa. The ion source must deliver a current of 40 A D(-) for up to 1 h with an accelerated current density of 200 Am/(2) and a ratio of coextracted electrons to ions below 1. The extraction area is 0.2 m(2) from an aperture array with an envelope of 1.5 x 0.6 m(2). A high power rf-driven negative ion source has been successfully developed at the Max-Planck Institute for Plasma Physics (IPP) at three test facilities in parallel. Current densities of 330 and 230 Am/(2) have been achieved for hydrogen and deuterium, respectively, at a pressure of 0.3 Pa and an electron/ion ratio below 1 for a small extraction area (0.007 m(2)) and short pulses (ITER source but without extraction system, is intended to demonstrate the size scaling and plasma homogeneity of rf ion sources. The source operates routinely now. First results on plasma homogeneity obtained from optical emission spectroscopy and Langmuir probes are very promising. Based on the success of the IPP development program, the high power rf-driven negative ion source has been chosen recently for the ITER beam systems in the ITER design review process.

  16. In-vacuum sensors for the beamline components of the ITER neutral beam test facility

    Science.gov (United States)

    Dalla Palma, M.; Pasqualotto, R.; Sartori, E.; Spagnolo, S.; Spolaore, M.; Veltri, P.

    2016-11-01

    Embedded sensors have been designed for installation on the components of the MITICA beamline, the prototype ITER neutral beam injector (Megavolt ITER Injector and Concept Advancement), to derive characteristics of the particle beam and to monitor the component conditions during operation for protection and thermal control. Along the beamline, the components interacting with the particle beam are the neutralizer, the residual ion dump, and the calorimeter. The design and the positioning of sensors on each component have been developed considering the expected beam-surface interaction including non-ideal and off-normal conditions. The arrangement of the following instrumentation is presented: thermal sensors, strain gages, electrostatic probes including secondary emission detectors, grounding shunt for electrical currents, and accelerometers.

  17. Surface heat loads on the ITER divertor vertical targets

    Science.gov (United States)

    Gunn, J. P.; Carpentier-Chouchana, S.; Escourbiac, F.; Hirai, T.; Panayotis, S.; Pitts, R. A.; Corre, Y.; Dejarnac, R.; Firdaouss, M.; Kočan, M.; Komm, M.; Kukushkin, A.; Languille, P.; Missirlian, M.; Zhao, W.; Zhong, G.

    2017-04-01

    The heating of tungsten monoblocks at the ITER divertor vertical targets is calculated using the heat flux predicted by three-dimensional ion orbit modelling. The monoblocks are beveled to a depth of 0.5 mm in the toroidal direction to provide magnetic shadowing of the poloidal leading edges within the range of specified assembly tolerances, but this increases the magnetic field incidence angle resulting in a reduction of toroidal wetted fraction and concentration of the local heat flux to the unshadowed surfaces. This shaping solution successfully protects the leading edges from inter-ELM heat loads, but at the expense of (1) temperatures on the main loaded surface that could exceed the tungsten recrystallization temperature in the nominal partially detached regime, and (2) melting and loss of margin against critical heat flux during transient loss of detachment control. During ELMs, the risk of monoblock edge melting is found to be greater than the risk of full surface melting on the plasma-wetted zone. Full surface and edge melting will be triggered by uncontrolled ELMs in the burning plasma phase of ITER operation if current models of the likely ELM ion impact energies at the divertor targets are correct. During uncontrolled ELMs in pre-nuclear deuterium or helium plasmas at half the nominal plasma current and magnetic field, full surface melting should be avoided, but edge melting is predicted.

  18. Heat exchanger identification by using iterative fuzzy observers

    Science.gov (United States)

    Lalot, Sylvain; Guðmundsson, Oddgeir; Pálsson, Halldór; Pálsson, Ólafur Pétur

    2016-05-01

    The principle of fuzzy observers is first illustrated on a general example: the determination of the two parameters of second order systems using a step response. The set of equations describing the system are presented and it is shown that accurate results are obtained, even for a high level of noise. The heat exchanger model is then introduced. It is based on a spatial division of a counter flow heat exchanger into multiple sections. The governing equations are rewritten as a state space representation. The number of sections needed to get accurate results is determined by comparing estimated values to experimental data. Based on the mean value of the root mean squared errors, it is shown that 80 sections is an appropriate value for this heat exchanger. It is then shown that the iterative fuzzy observers can be used to determine the main parameters of the counter flow heat exchanger, i.e. the convection heat transfer coefficients, when in transient state. The final values of these parameters are heat transfer coefficient corresponds to a ±0.5 % variation of the estimated overall heat transfer coefficient. This study also shows that the fuzzy observers are equally efficient when the heat exchanger is in steady state.

  19. Sawtooth control in JET with ITER relevant low field side resonance ion cyclotron resonance heating and ITER-like wall

    NARCIS (Netherlands)

    Graves, J. P.; Lennholm, M.; Chapman, I.T.; Lerche, E.; Reich, M.; Alper, B.; Bobkov, V.; Dumont, R.; Faustin, J. M.; Jacquet, P.; Jaulmes, F.; Johnson, T.; Keeling, D. L.; Liu, Y. Q.; Nicolas, T.; Tholerus, S.; Blackman, T.; Carvalho, I. S.; Coelho, R.; Van Eester, D.; Felton, R.; Goniche, M.; Kiptily, V.; Monakhov, I.; Nave, M. F. F.; von Thun, Perez; Sabot, R.; Sozzi, C.; Tsalas, M.

    2015-01-01

    New experiments at JET with the ITER-like wall show for the first time that ITER-relevant low field side resonance first harmonic ion cyclotron resonance heating (ICRH) can be used to control sawteeth that have been initially lengthened by fast particles. In contrast to previous (Graves et al 2012

  20. Current Control in ITER Steady State Plasmas With Neutral Beam Steering

    Energy Technology Data Exchange (ETDEWEB)

    R.V. Budny

    2009-09-10

    Predictions of quasi steady state DT plasmas in ITER are generated using the PTRANSP code. The plasma temperatures, densities, boundary shape, and total current (9 - 10 MA) anticipated for ITER steady state plasmas are specified. Current drive by negative ion neutral beam injection, lower-hybrid, and electron cyclotron resonance are calculated. Four modes of operation with different combinations of current drive are studied. For each mode, scans with the NNBI aimed at differing heights in the plasma are performed to study effects of current control on the q profile. The timeevolution of the currents and q are calculated to evaluate long duration transients. Quasi steady state, strongly reversed q profiles are predicted for some beam injection angles if the current drive and bootstrap currents are sufficiently large.

  1. Applying Remote Handling Attributes to the ITER Neutral Beam Cell Monorail Crane

    CERN Document Server

    Crofts, O; Raimbach, J; Tesini, A; Choi, C-H; Damiani, C; Van Uffelen, M

    2013-01-01

    The maintenance requirements for the equipment in the ITER Neutral Beam Cell requires components to be lifted and transported within the cell by remote means. To meet this requirement, the provision of an overhead crane with remote handling capabilities has been initiated. The layout of the cell has driven the design to consist of a monorail crane that travels on a branched monorail track attached to the cell ceiling. This paper describes the principle design constraints and how the remote handling attributes were applied to the concept design of the monorail crane, concentrating on areas where novel design solutions have been required and on the remote recovery requirements and solutions.

  2. The ITER neutral beam test facility: designs of the general infrastructure, cryo-system and cooling plant

    Energy Technology Data Exchange (ETDEWEB)

    Cordier, J.J.; Hemsworth, R.; Chantant, M.; Gravil, B.; Henry, D.; Sabathier, F.; Doceul, L.; Thomas, E.; Van Houtte, D. [Association Euratom-CEA Cadarache (DSM/DRFC), 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Zaccaria, P.; Antoni, V.; Dal Bello, S.; Masiello, A.; Marcuzzi, D. [Consorzio RFX Association Euratom-ENEA, Padova (Italy); Antipenkov, A.; Dremel, M.; Day, C. [Institut fur Technische Physik, FZK, Karlsruhe (Germany); Mondino, P.L. [Max-Planck-Institut fuer Plasmaphysik, EFDA CSU, Garching (Germany)

    2004-07-01

    The CEA Association is involved, in close collaboration with ENEA, FZK, IPP and UKEA European Associations, in the first ITER neutral beam injector and the ITER neutral beam test facility design (NBTF). A total power of about 50 MW will have to be removed in steady state on the neutral beam test facility (NBTF). The main purpose of this task is to make progress with the detailed design of the first ITER NB injector and to start the conceptual design of the ITER NBTF. The general infrastructure layout of a generic site for the NBTF, includes the test facility itself equipped of a dedicated beamline vessel and integration studies of associated auxiliaries as cooling plant, cryo-plant and fore-pumping system. The general infrastructure and auxiliaries layout of the NBTF are described. (authors)

  3. Development of KSTAR Neutral Beam Heating System

    Energy Technology Data Exchange (ETDEWEB)

    Oh, B. H.; Song, W. S.; Yoon, B. J. (and others)

    2007-10-15

    The prototype components of a neutral beam injection (NBI) system have been developed for the KSTAR, and a capability of the manufactured components has been tested. High power ion source, acceleration power supply, other ion source power supplies, neutralizer, bending magnet for ion beam separation, calorimeter, and cryo-sorption pump have been developed by using the domestic technologies and tested for a neutral beam injection of 8 MW per beamline with a pulse duration of 300 seconds. The developed components have been continuously upgraded to achieve the design requirements. The development technology of high power and long pulse neutral beam injection system has been proved with the achievement of 5.2 MW output for a short pulse length and 1.6 MW output for a pulse length of 300 seconds. Using these development technologies, the domestic NB technology has been stabilized under the development of high power ion source, NB beamline components, high voltage and current power supplies, NB diagnostics, NB system operation and control.

  4. Predictions of Alpha Heating in ITER L-mode and H-mode Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    R.V. Budny

    2011-01-06

    Predictions of alpha heating in L-mode and H-mode DT plasmas in ITER are generated using the PTRANSP code. The baseline toroidal field of 5.3 T, plasma current ramped to 15 MA and a flat electron density profile ramped to Greenwald fraction 0.85 are assumed. Various combinations of external heating by negative ion neutral beam injection, ion cyclotron resonance, and electron cyclotron resonance are assumed to start half-way up the density ramp. The time evolution of plasma temperatures and, for some cases, toroidal rotation are predicted assuming GLF23 and boundary parameters. Significant toroidal rotation and flow-shearing rates are predicted by GLF23 even in the L-mode phase with low boundary temperatures, and the alpha heating power is predicted to be significant if the power threshold for the transition to H-mode is higher than the planned total heating power. The alpha heating is predicted to be 8-76 MW in L-mode at full density. External heating mixes with higher beam injection power have higher alpha heating power. Alternatively if the toroidal rotation is predicted assuming that the ratio of the momentum to thermal ion energy conductivity is 0.5, the flow-shearing rate is predicted to have insignificant effects on the GLF23- predicted temperatures, and alpha heating is predicted to be 8-20 MW. In H-mode plasmas the alpha heating is predicted to depend sensitively on the assumed pedestal temperatures. Cases with fusion gain greater than 10 are predicted to have alpha heating greater than 80 MW.

  5. Vessel design and interfaces development for the 1 MV ITER Neutral Beam Injector and Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Rigato, Wladi [Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, I-35127 Padova (Italy)], E-mail: wladi.rigato@igi.cnr.it; Dal Bello, Samuele; Marcuzzi, Diego; Rizzolo, Andrea [Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, I-35127 Padova (Italy)

    2009-06-15

    In the framework of the design activities for the ITER Neutral Beam Injector (NBI) and full power neutral beam injector prototype, the vacuum vessel has been designed concurrently with the whole other components, and in particular with the Beam Source (BS) and the large Cryopumps, that strongly conditioned the design. The definition of the interfaces has been focused on the design for the 1 MV neutral beam injector prototype, anyway keeping to the absolute minimum the differences with respect to the ITER NBI Vessel. The Vacuum Vessel is composed of two separate parts which shall be welded on site: the Beam Line Vessel (BLV) and the Beam Source Vessel (BSV). Three main bolted lids are foreseen for horizontal and vertical access to the internal components. The vessel is composed of double wall and ribs in critical areas to minimize deformations and stresses under the atmospheric pressure load. New concepts for the Beam Source Support, Positioning and Tilting Systems have been developed and an engineering design has been carried out, able to satisfy precise requirements on stiffness, accuracy of regulation, vacuum compatibility, electric insulation and Remote Handling operation. These components and the BS have been fully integrated inside the BSV by means of support structures and vacuum feedthroughs for mechanical links allowing the transmission of motion and forces. The interfaces between the BLV and the Beam Line Components (BLCs) have been revised to be compatible with the new vessel design and the BLCs support frames. Further interfaces with the high voltage bushing, the vacuum pumping and the diagnostic systems have been considered. The number and the position of the diagnostic viewports were identified taking into account both diagnostics and structural requirements. Static, buckling and seismic analyses, based on EN 13445, have been performed considering operative and exceptional load cases. Requirements, criteria and design details are presented in the paper

  6. Plasma heating with multi-MeV neutral atom beams

    Energy Technology Data Exchange (ETDEWEB)

    Grisham, L.R.; Post, D.E.; Mikkelsen, D.R.; Eubank, H.P.

    1981-10-01

    We explore the utility and feasibility of neutral beams of greater than or equal to 6 AMU formed from negative ions, and also of D/sup 0/ formed from D/sup -/. The negative ions would be accelerated to approx. 1 to 2 MeV/AMU and neutralized, whereupon the neutral atoms would be used to heat and, perhaps, to drive current in magnetically confined plasmas. Such beams appear feasible and offer the promise of significant advantages relative to conventional neutral beams based on positive deuterium ions at approx. 150 keV.

  7. ICRF heating in JET during initial operations with the ITER-like wall

    Energy Technology Data Exchange (ETDEWEB)

    Jacquet, P.; Brix, M.; Graham, M.; Mayoral, M.-L.; Meigs, A.; Monakhov, I.; Sirinelli, A. [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Bobkov, V.; Drewelow, P.; Pütterich, T. [Max-Planck-Institut für Plasmaphysik, EURATOM-Assoziation, Garching (Germany); Brezinsek, S. [IEK-4, Forschungszentrum Jülich, Association EURATOM-FZJ (Germany); Campergue, A-L. [Ecole Nationale des Ponts et Chaussées, F77455 Marne-la-Vallée (France); Colas, L. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Czarnecka, A. [Association Euratom-IPPLM, Hery 23, 01-497 Warsaw (Poland); Klepper, C. C. [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6169 (United States); Lerche, E.; Van-Eester, D. [Association EURATOM-Belgian State, ERM-KMS, Brussels (Belgium); Milanesio, D. [Politecnico di Torino, Department of Electronics, Torino (Italy); Mlynar, J. [Association EURATOM-IPP.CR, Za Slovankou 3, 182 21 Praha 8 (Czech Republic); Collaboration: JET-EFDA Contributors

    2014-02-12

    In 2011/12, JET started operation with its new ITER-Like Wall (ILW) made of a tungsten (W) divertor and a beryllium (Be) main chamber wall. The impact of the new wall material on the JET Ion Cyclotron Resonance Frequency (ICRF) operation was assessed and also the properties of JET plasmas heated with ICRF were studied. No substantial change of the antenna coupling resistance was observed with the ILW as compared with the carbon wall. Heat-fluxes on the protecting limiters close the antennas quantified using Infra-Red (IR) thermography (maximum 4.5 MW/m{sup 2} in current drive phasing) are within the wall power load handling capabilities. A simple RF sheath rectification model using the antenna near-fields calculated with the TOPICA code can well reproduce the heat-flux pattern around the antennas. ICRF heating results in larger tungsten and nickel (Ni) contents in the plasma and in a larger core radiation when compared to Neutral Beam Injection (NBI) heating. Some experimental facts indicate that main-chamber W components could be an important impurity source: the divertor W influx deduced from spectroscopy is comparable when using RF or NBI at same power and comparable divertor conditions; the W content is also increased in ICRF-heated limiter plasmas; and Be evaporation in the main chamber results in a strong and long lasting reduction of the impurity level. The ICRF specific high-Z impurity content decreased when operating at higher plasma density and when increasing the hydrogen concentration from 5% to 20%. Despite the higher plasma bulk radiation, ICRF exhibited overall good plasma heating efficiency; The ICRF power can be deposited at plasma centre and the radiation is mainly from the outer part of the plasma. Application of ICRF heating in H-mode plasmas started, and the beneficial effect of ICRF central electron heating to prevent W accumulation in the plasma core could be observed.

  8. Iterative Feedback Tuning in district heating systems; Iterative Feedback Tuning i vaermeproduktionsanlaeggningar

    Energy Technology Data Exchange (ETDEWEB)

    Raaberg, Martin; Velut, Stephane; Bari, Siavosh Amanat

    2010-10-15

    The project goal is to evaluate and describe how Iterative Feedback Tuning (IFT) can be used to tune controllers in the typical control loops in heat- and power plants. There are only a few practical studies carried out for IFT and they are not really relevant for power and heat processes. It is the practical problems in implementing the IFT and the result of trimming that is the focus of this project. The project will start with theoretical studies of the IFT-method, then realization and simple simulations in scilab. The IFT equations are then implemented in Freelance 2000, an ABB control system, for practical tests on a SISO- and a MIMO-process. By performing reproducible experiments on the process and analyze the results IFT can adjust the controller parameters to minimize a cost function that represents the control goal. The project selected for SISO experiments a pressure controller in an oil transportation system. By controlling the valve position of a control valve for the reversal to the supply tank, the pressure in the oil transport system is regulated. A disturbance in oil pressure can be achieved by changing the position of a valve that lets oil through to the day tank. The selected MIMO-process is a pre-heater in a degassing process. In this process, a valve on the secondary side is utilized to control the flow in the secondary system. A valve on the primary side is utilized to control the district heating water flow through the heat exchanger to control the temperature on the secondary side. An increased secondary flow increases the heat demand and thus requiring an increase in primary flow to maintain the secondary side outlet temperature. This is the cross-coupling responsible for why it is an advantage to consider the process as multi-variable. Using the IFT method, the two original PID-controllers and a feed-forward controller is tuned simultaneously. IFT-method was difficult to implement but worked well in both simulations and in real processes

  9. Design of the 'half-size' ITER neutral beam source for the test facility ELISE

    Energy Technology Data Exchange (ETDEWEB)

    Heinemann, B. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Postfach 1533, D-85740 Garching (Germany)], E-mail: bernd.heinemann@ipp.mpg.de; Falter, H.; Fantz, U.; Franzen, P.; Froeschle, M.; Gutser, R.; Kraus, W.; Nocentini, R.; Riedl, R.; Speth, E.; Staebler, A.; Wuenderlich, D. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Postfach 1533, D-85740 Garching (Germany); Agostinetti, P. [Consorzio RFX, EURATOM Association, Corso Stati Uniti 4, I-35127 Padova (Italy); Jiang, T. [Southwestern Institute of Physics, ChengDu (China)

    2009-06-15

    In 2007 the radio frequency driven negative hydrogen ion source developed at IPP in Garching was chosen by the ITER board as the new reference source for the ITER neutral beam system. In order to support the design and the commissioning and operating phases of the ITER test facilities ISTF and NBTF in Padua, IPP is presently constructing a new test facility ELISE (Extraction from a Large Ion Source Experiment). ELISE will be operated with the so-called 'half-size ITER source' which is an intermediate step between the present small IPP RF sources (1/8 ITER size) and the full size ITER source. The source will have approximately the width but only half the height of the ITER source. The modular concept with 4 drivers will allow an easy extrapolation to the full ITER size with 8 drivers. Pulsed beam extraction and acceleration up to 60 kV (corresponding to pre-acceleration voltage of SINGAP) is foreseen. The aim of the design of the ELISE source and extraction system was to be as close as possible to the ITER design; it has however some modifications allowing a better diagnostic access as well as more flexibility for exploring open questions. Therefore one major difference compared to the source of ITER, NBTF or ISTF is the possible operation in air. Specific requirements for RF sources as found on IPP test facilities BATMAN and MANITU are implemented [A. Staebler, et al., Development of a RF-driven ion source for the ITER NBI system, SOFT Conference 2008, Fusion Engineering and Design, 84 (2009) 265-268].

  10. ITER-W monoblocks under high pulse number transient heat loads at high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Loewenhoff, Th., E-mail: T.Loewenhoff@fz-juelich.de [Forschungszentrum Jülich, 52428 Jülich (Germany); Linke, J., E-mail: J.Linke@fz-juelich.de [Forschungszentrum Jülich, 52428 Jülich (Germany); Pintsuk, G., E-mail: G.Pintsuk@fz-juelich.de [Forschungszentrum Jülich, 52428 Jülich (Germany); Pitts, R.A., E-mail: Richard.Pitts@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-Lez-Durance (France); Riccardi, B., E-mail: Bruno.Riccardi@f4e.europa.eu [Fusion for Energy Joint Undertaking, Josep Pla No. 2 – T B3 7/01, Barcelona 08019 (Spain)

    2015-08-15

    In the context of using a full-tungsten (W) divertor for ITER, thermal shock resistance has become even more important as an issue that may potentially influence the long term performance. To address this issue a unique series of experiments has been performed on ITER-W monoblock mock ups in three EU high heat flux facilities: GLADIS (neutral beam), JUDITH 2 (electron beam) and Magnum-PSI (plasma beam). This paper discusses the JUDITH 2 experiments. Two different base temperatures, 1200 °C and 1500 °C, were chosen superimposed by ∼18,000/100,000 transient events (Δt = 0.48 ms) of 0.2 and 0.6 GW/m{sup 2}. Results showed a stronger surface deterioration at higher base temperature, quantified by an increase in roughening. This is intensified if the same test is done after preloading (exposure to high temperature without transients), especially at higher base temperature when the material recrystallizes.

  11. Iter

    Science.gov (United States)

    Iotti, Robert

    2015-04-01

    ITER is an international experimental facility being built by seven Parties to demonstrate the long term potential of fusion energy. The ITER Joint Implementation Agreement (JIA) defines the structure and governance model of such cooperation. There are a number of necessary conditions for such international projects to be successful: a complete design, strong systems engineering working with an agreed set of requirements, an experienced organization with systems and plans in place to manage the project, a cost estimate backed by industry, and someone in charge. Unfortunately for ITER many of these conditions were not present. The paper discusses the priorities in the JIA which led to setting up the project with a Central Integrating Organization (IO) in Cadarache, France as the ITER HQ, and seven Domestic Agencies (DAs) located in the countries of the Parties, responsible for delivering 90%+ of the project hardware as Contributions-in-Kind and also financial contributions to the IO, as ``Contributions-in-Cash.'' Theoretically the Director General (DG) is responsible for everything. In practice the DG does not have the power to control the work of the DAs, and there is not an effective management structure enabling the IO and the DAs to arbitrate disputes, so the project is not really managed, but is a loose collaboration of competing interests. Any DA can effectively block a decision reached by the DG. Inefficiencies in completing design while setting up a competent organization from scratch contributed to the delays and cost increases during the initial few years. So did the fact that the original estimate was not developed from industry input. Unforeseen inflation and market demand on certain commodities/materials further exacerbated the cost increases. Since then, improvements are debatable. Does this mean that the governance model of ITER is a wrong model for international scientific cooperation? I do not believe so. Had the necessary conditions for success

  12. Core fusion power gain and alpha heating in JET, TFTR, and ITER

    Science.gov (United States)

    Budny, R. V.; Cordey, J. G.; TFTR Team; Contributors, JET

    2016-05-01

    Profiles of the ratio of fusion power and the auxiliary heating power q DT are calculated for the TFTR and JET discharges with the highest neutron emission rates, and are predicted for ITER. Core values above 1.3 for JET and 0.8 for TFTR are obtained. Values above 20 are predicted for ITER baseline plasmas.

  13. Fabrication study on the cooling module of the ITER neutral beam duct liner

    Energy Technology Data Exchange (ETDEWEB)

    Sa, J.W. [National Fusion Research Institute, Yuseong-gu, Daejeon (Korea, Republic of); Kim, H.S., E-mail: hskim@nfri.re.k [National Fusion Research Institute, Yuseong-gu, Daejeon (Korea, Republic of); Kim, B.Y.; Kim, B.C.; Ahn, H.J.; Bak, J.S. [National Fusion Research Institute, Yuseong-gu, Daejeon (Korea, Republic of); Jung, H.J. [Korean Intellectual Property Office, Seo-gu, Daejeon (Korea, Republic of); Han, M.H.; Hong, C.D.; Lee, J.S.; Kim, Y.K. [Hyundai Heavy Industries Co. Ltd., Dong-gu, Ulsan (Korea, Republic of); Urbani, M. [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Geli, F. [UKAEA Culham Division, Oxfordshire OX14 3DB, Abingdon (United Kingdom)

    2010-12-15

    Recently the new concept of the ITER neutral beam (NB) duct liner has been developed to improve thermo-mechanical performance and satisfy the requirements for remote handling and maintenance. The design concept of cooling module located inside neutron shield structure is to use deep-drilled panels instead of the original design concept of the casting-modularized component with tubes. In this study, the manufacturing feasibility has been investigated through the fabrication of small size coupons and full scale mock-up. Firstly, the small size coupons are for developing the electron beam welding processes. Secondly, the full scale mock-up which has 6 holes for cooling passage has been fabricated in order to develop the main fabrication processes such as deep drilling, bending and machining. In addition, the pressure and the leak tests have been carried out to check the required performance for completed cooling panel. Although some improvement is required, but the Electron Beam Welding (EBW) has been successfully achieved and generally the deep drilling and bending process also shown good results in dimensional control.

  14. Formation and Sustainment of ITPs in ITER with the Baseline Heating Mix

    Energy Technology Data Exchange (ETDEWEB)

    Francesca M. Poli and Charles Kessel

    2012-12-03

    Plasmas with internal transport barriers (ITBs) are a potential and attractive route to steady-state operation in ITER. These plasmas exhibit radially localized regions of improved con nement with steep pressure gradients in the plasma core, which drive large bootstrap current and generate hollow current pro les and negative shear. This work examines the formation and sustainment of ITBs in ITER with electron cyclotron heating and current drive. It is shown that, with a trade-o of the power delivered to the equatorial and to the upper launcher, the sustainment of steady-state ITBs can be demonstrated in ITER with the baseline heating con guration.

  15. Ion cyclotron resonance frequency heating in JET during initial operations with the ITER-like wall

    Energy Technology Data Exchange (ETDEWEB)

    Jacquet, P., E-mail: philippe.jacquet@ccfe.ac.uk; Monakhov, I.; Arnoux, G.; Brix, M.; Graham, M.; Meigs, A.; Sirinelli, A. [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Bobkov, V.; Devaux, S.; Drewelow, P.; Pütterich, T. [Max-Planck-Institut für Plasmaphysik, EURATOM-Assoziation, Garching (Germany); Colas, L. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Czarnecka, A. [Association Euratom-IPPLM, Hery 23, 01-497 Warsaw (Poland); Lerche, E.; Van-Eester, D. [Association EURATOM-Belgian State, ERM-KMS, Brussels (Belgium); Mayoral, M.-L. [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); EFDA Close Support Unit, Garching (Germany); Brezinsek, S. [IEK-4, Forschungszentrum Jülich, Association EURATOM-FZJ, Jülich (Germany); Campergue, A.-L. [Ecole Nationale des Ponts et Chaussées, F77455 Marne-la-Vallée (France); Klepper, C. C. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831-6169 (United States); Milanesio, D. [Politecnico di Torino, Department of Electronics, Torino (Italy); and others

    2014-06-15

    In 2011/12, JET started operation with its new ITER-Like Wall (ILW) made of a tungsten (W) divertor and a beryllium (Be) main chamber wall. The impact of the new wall materials on the JET Ion Cyclotron Resonance Frequency (ICRF) operation is assessed and some important properties of JET plasmas heated with ICRF are highlighted. A ∼ 20% reduction of the antenna coupling resistance is observed with the ILW as compared with the JET carbon (JET-C) wall. Heat-fluxes on the protecting limiters close the antennas, quantified using Infra-Red thermography (maximum 4.5 MW/m{sup 2} in current drive phasing), are within the wall power load handling capabilities. A simple RF sheath rectification model using the antenna near-fields calculated with the TOPICA code can reproduce the heat-flux pattern around the antennas. ICRF heating results in larger tungsten and nickel (Ni) contents in the plasma and in a larger core radiation when compared to Neutral Beam Injection (NBI) heating. The location of the tungsten ICRF specific source could not be identified but some experimental observations indicate that main-chamber W components could be an important impurity source: for example, the divertor W influx deduced from spectroscopy is comparable when using RF or NBI at same power and comparable divertor conditions, and Be evaporation in the main chamber results in a strong reduction of the impurity level. In L-mode plasmas, the ICRF specific high-Z impurity content decreased when operating at higher plasma density and when increasing the hydrogen concentration from 5% to 15%. Despite the higher plasma bulk radiation, ICRF exhibited overall good plasma heating performance; the power is typically deposited at the plasma centre while the radiation is mainly from the outer part of the plasma bulk. Application of ICRF heating in H-mode plasmas has started, and the beneficial effect of ICRF central electron heating to prevent W accumulation in the plasma core has been observed.

  16. Iterative inversion of global magnetospheric information from energy neutral atom (ENA) images recorded by the TC-2/NUADU instrument

    Institute of Scientific and Technical Information of China (English)

    S.; MCKENNA-LAWLOR; S.; BARABASH; J.; BALAZ

    2008-01-01

    A method is presented for retrieving the magnetospheric ion distribution from En-ergetic Neutral Atom (ENA) measurements made by the NUADU instrument on the TC-2 spacecraft. Based on the already well-established method of constrained lin-ear inversion, an iterance technique suitable for the low count ENA measurements has been developed which is tolerant of the noise background. By the iterance technique, it is possible for the first time to simultaneously retrieve the magneto-spheric ion distribution and the exospheric neutral density, and further to recover global ENA emissions in three dimensions. The technique is applied to a repre-sentative ENA image recorded in energy channel 2 (protons: 50―81 keV) of the NUADU instrument during a major geomagnetic storm and it is, thereby, shown that the retrieval method developed provides a useful tool for extracting ion distribution information from ENA data.

  17. Iterative inversion of global magnetospheric information from energy neutral atom (ENA)images recorded by the TC-2/NUADU instrument

    Institute of Scientific and Technical Information of China (English)

    LU Li; S. MCKENNA-LAWLOR; S. BARABASH; J. BALAZ; LIU ZhenXing; SHEN Chao; CAO JinBin; ANG ChaoLing

    2008-01-01

    A method is presented for retrieving the magnetospheric ion distribution from En-ergetic Neutral Atom (ENA) measurements made by the NUADU instrument on the TC-2 spacecraft. Based on the already well-established method of constrained lin-ear inversion, an iterance technique suitable for the low count ENA measurements has been developed which is tolerant of the noise background. By the iterance technique, it is possible for the first time to simultaneously retrieve the magneto-spheric ion distribution and the exospheric neutral density, and further to recover global ENA emissions in three dimensions. The technique is applied to a repre-sentative ENA image recorded in energy channel 2 (protons: 50-81 keV) of the NUADU instrument during a major geomagnetic storm and it is, thereby, shown that the retrieval method developed provides a useful tool for extracting ion distribution information from ENA data.

  18. An analysis of JET fast-wave heating and current drive experiments directly related to ITER

    Energy Technology Data Exchange (ETDEWEB)

    Bhatnagar, V.P.; Eriksson, L.; Gormezano, C.; Jacquinot, J.; Kaye, A.; Start, D.F.H. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

    1994-07-01

    The ITER fast-wave system is required to serve a variety of purposes, in particular, plasma heating to ignition, current profile and burn control and eventually, in conjunction with other schemes, a central non-inductive current drive (CD) for the steady-state operation of ITER. The ICRF heating and current drive data that has been obtained in JET are analyzed in terms of dimensionless parameters, with a view to ascertaining its direct relevance to key ITER requirements. The analysis is then used to identify areas both in physics and technological aspects of ion-cyclotron resonance heating (ICRH) and CD that require further experimentation in ITER-relevant devices such as JET to establish the required data base. (authors). 12 refs., 8 figs.

  19. Voltage holding study of 1 MeV accelerator for ITER neutral beam injector.

    Science.gov (United States)

    Taniguchi, M; Kashiwagi, M; Umeda, N; Dairaku, M; Takemoto, J; Tobari, H; Tsuchida, K; Yamanaka, H; Watanabe, K; Kojima, A; Hanada, M; Sakamoto, K; Inoue, T

    2012-02-01

    Voltage holding test on MeV accelerator indicated that sustainable voltage was a half of that of ideal quasi-Rogowski electrode. It was suggested that the emission of the clumps is enhanced by a local electric field concentration, which leads to discharge initiation at lower voltage. To reduce the electric field concentration in the MeV accelerator, gaps between the grid supports were expanded and curvature radii at the support corners were increased. After the modifications, the accelerator succeeded in sustaining -1 MV in vacuum without beam acceleration. However, the beam energy was still limited at a level of 900 keV with a beam current density of 150 A∕m(2) (346 mA) where the 3 × 5 apertures were used. Measurement of the beam profile revealed that deflection of the H(-) ions was large and a part of the H(-) ions was intercepted at the acceleration grid. This causes high heat load on the grids and the breakdowns during beam acceleration. To suppress the direct interception, new grid system was designed with proper aperture displacement based on a 3D beam trajectory analysis. As the result, the beam deflection was compensated and the voltage holding during the beam acceleration was improved. Beam parameter of the MeV accelerator was increased to 980 keV, 185 A∕m(2) (427 mA), which is close to the requirement of ITER accelerator (1 MeV, 200 A∕m(2)).

  20. Manufacturing of the full size prototype of the ion source for the ITER neutral beam injector – The SPIDER beam source

    Energy Technology Data Exchange (ETDEWEB)

    Pavei, Mauro, E-mail: mauro.pavei@igi.cnr.it [Consorzio RFX, C.so Stati Uniti 4, I-35127, Padova (Italy); Boilson, Deirdre [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Bonicelli, Tullio [Fusion for Energy, C/Joseph Pla 2, 08019 Barcelona (Spain); Boury, Jacques [Thales Electron Devices, Velizy Villacoublay (France); Bush, Michael [Galvano-T GmbH, T, Raiffeisenstraße 8, 51570 Windeck (Germany); Ceracchi, Andrea; Faso, Diego [CECOM S.r.l., Via Tiburtina – Guidonia Montecelio, Roma (Italy); Graceffa, Joseph [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Heinemann, Bernd [Max-Planck-Institut für Plasmaphysik, D-85740 Garching (Germany); Hemsworth, Ronald [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Lievin, Christophe [Thales Electron Devices, Velizy Villacoublay (France); Marcuzzi, Diego [Consorzio RFX, C.so Stati Uniti 4, I-35127, Padova (Italy); Masiello, Antonio [Fusion for Energy, C/Joseph Pla 2, 08019 Barcelona (Spain); Sczepaniak, Bernd [Galvano-T GmbH, T, Raiffeisenstraße 8, 51570 Windeck (Germany); Singh, Mahendrajit [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Toigo, Vanni; Zaccaria, Pierluigi [Consorzio RFX, C.so Stati Uniti 4, I-35127, Padova (Italy)

    2015-10-15

    Highlights: • Negative ion sources are key components of neutral beam injectors for nuclear fusion. • The SPIDER experiment aims to optimize the negative ion source of MITICA and HNB. • The SPIDER Beam Source manufacturing is currently on-going. • Manufacturing and assembling technological issues encountered are presented. - Abstract: In ITER, each heating neutral beam injector (HNB) will deliver about 16.5 MW heating power by accelerating a 40 A deuterium negative ion beam up to the energy of 1 MeV. The ions are generated inside a caesiated negative ion source, where the injected H{sub 2}/D{sub 2} is ionized by a radio frequency electromagnetic field. The SPIDER test bed, currently being manufactured, is going to be the ion source test facility for the full size ion source of the HNBs and of the diagnostic neutral beam injector of ITER. The SPIDER beam source comprises an ion source with 8 radio-frequency drivers and a three-grid system, providing an overall acceleration up to energies of about 100 keV [1]. SPIDER represents a substantial step forward between the half ITER size ion source, which is currently being tested at the ELISE test bed in IPP-Garching, and the negative ion sources to be used on ITER, in terms of layout, dimensions and operating parameters. The SPIDER beam source will be housed inside a vacuum vessel which will be equipped with a beam dump and a graphite diagnostic calorimeter. The manufacturing design of the main parts of the SPIDER beam source has been completed and many of the tests on the prototypes have been successfully passed. The most complex parts, from the manufacturing point of view, of the ion source and the accelerator, developed by galvanic deposition of copper are being manufactured. The manufacturing phase will be completed within 2015, when the assembly of the device will start at the PRIMA site, in Padova (I). The paper describes the status of the procurement, the adaptations operated on the design of the beam

  1. Conceptual design and integration of a diagnostic neutral beam in ITER

    NARCIS (Netherlands)

    Di Pietro, E.; Costley, A.; Inoue, T.; Krylov, A.; Panasenkov, A.; Utin, Y.; Vayakis, G.; Von Hellerman, M.; Yamada, M.

    2001-01-01

    In ITER, one of the key issues to achieve 400 s driven-burn operation at Q about 10 (Technical Basis for the ITER-FEAT Outline Design-Section I-3.2.2.IAEA) is helium ash accumulation. As a result, the real-time measurement of the thermalised helium density profile in the confinement region is of fun

  2. Heating, current drive and energetic particle studies on JET in preparation of ITER operation

    NARCIS (Netherlands)

    Noterdaeme, J. M.; Budny, R.; Cardinali, A.; Castaldo, C.; Cesario, R.; Crisanti, F.; DeGrassie, J.; D' Ippolito, D. A.; Durodie, F.; Ekedahl, A.; Figueiredo, A.; Ingesson, C.; Joffrin, E.; Hartmann, D.; Heikkinen, J.; Hellsten, T.; Jones, T.; Kiptily, V.; Lamalle, P.; Litaudon, X.; Nguyen, F.; Mailloux, J.; Mantsinen, M.; Mayoral, M.; Mazon, D.; Meo, F.; Monakhov, I.; Myra, J. R.; Pamela, J.; Pericoli, V.; Petrov, Y.; Sauter, O.; Sarazin, Y.; Sharapov, S. E.; Tuccillo, A. A.; Van Eester, D.

    2003-01-01

    This paper summarizes the recent work on JET in the three areas of heating, current drive and energetic particles. The achievements have extended the possibilities of JET, have a direct connection to ITER operation and provide new and interesting physics. Toroidal rotation profiles of plasmas heated

  3. Capabilities of the ITER Electron Cyclotron Equatorial Launcher for Heating and Current Drive

    Directory of Open Access Journals (Sweden)

    Ramponi G.

    2012-09-01

    Full Text Available The ITER Electron Cyclotron Equatorial Launcher is designed to be one of the heating systems to assist and sustain the development of various ITER plasma scenarios starting with the very first plasma operation. Here the capabilities for Heating and Current Drive of this system are reviewed. In particular, the optimum launching conditions are investigated for two scenarios at burn, comparing toroidal and poloidal steering options. Then, the EC capabilities are investigated for different plasma parameters corresponding to various phases of the ITER plasma discharge, from current ramp-up up to burn, and for a wide range of magnetic field, focusing in particular on the EC potential for heating and for L to H-mode assist. It is found that the EC system can contribute to a wide range of heating scenarios during the ramp-up of the magnetic field, significantly increasing the applicable range as a function of magnetic field.

  4. A Newton type iterative method for heat-conduction inverse problems

    Institute of Scientific and Technical Information of China (English)

    HE Guo-qiang; MENG Ze-hong

    2007-01-01

    An inverse problem for identification of the coefficient in heat-conduction equation is considered. After reducing the problem to a nonlinear ill-posed operator equation, Newton type iterative methods are considered. The implicit iterative method is applied to the linearized Newton equation, and the key step in the process is that a new reasonable a posteriori stopping rule for the inner iteration is presented. Numerical experiments for the new method as well as for Tikhonov method and Bakushikskii method are given, and these results show the obvious advantages of the new method over the other ones.

  5. Work function measurements during plasma exposition at conditions relevant in negative ion sources for the ITER neutral beam injection.

    Science.gov (United States)

    Gutser, R; Wimmer, C; Fantz, U

    2011-02-01

    Cesium seeded sources for surface generated negative hydrogen ions are major components of neutral beam injection systems in future large-scale fusion experiments such as ITER. The stability and delivered current density depend highly on the work function during vacuum and plasma phases of the ion source. One of the most important quantities that affect the source performance is the work function. A modified photocurrent method was developed to measure the temporal behavior of the work function during and after cesium evaporation. The investigation of cesium exposed Mo and MoLa samples under ITER negative hydrogen ion based neutral beam injection relevant surface and plasma conditions showed the influence of impurities which result in a fast degradation when the plasma exposure or the cesium flux onto the sample is stopped. A minimum work function close to that of bulk cesium was obtained under the influence of the plasma exposition, while a significantly higher work function was observed under ITER-like vacuum conditions.

  6. Fast wave direct electron heating in advanced inductive and ITER baseline scenario discharges in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Pinsker, R. I.; Jackson, G. L.; Luce, T. C.; Politzer, P. A. [General Atomics, PO Box 85608, San Diego, California 92186-5608 (United States); Austin, M. E. [University of Texas at Austin, Austin, Texas 78712 (United States); Diem, S. J.; Kaufman, M. C.; Ryan, P. M. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Doyle, E. J.; Zeng, L. [University of California Los Angeles, Los Angeles, California 90095 (United States); Grierson, B. A.; Hosea, J. C.; Nagy, A.; Perkins, R.; Solomon, W. M.; Taylor, G. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Maggiora, R.; Milanesio, D. [Politecnico di Torino, Dipartimento di Elettronica, Torino (Italy); Porkolab, M. [Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Turco, F. [Columbia University, New York, New York 10027 (United States)

    2014-02-12

    Fast Wave (FW) heating and electron cyclotron heating (ECH) are used in the DIII-D tokamak to study plasmas with low applied torque and dominant electron heating characteristic of burning plasmas. FW heating via direct electron damping has reached the 2.5 MW level in high performance ELMy H-mode plasmas. In Advanced Inductive (AI) plasmas, core FW heating was found to be comparable to that of ECH, consistent with the excellent first-pass absorption of FWs predicted by ray-tracing models at high electron beta. FW heating at the ∼2 MW level to ELMy H-mode discharges in the ITER Baseline Scenario (IBS) showed unexpectedly strong absorption of FW power by injected neutral beam (NB) ions, indicated by significant enhancement of the D-D neutron rate, while the intended absorption on core electrons appeared rather weak. The AI and IBS discharges are compared in an effort to identify the causes of the different response to FWs.

  7. Non-ideal operating conditions of the ion source prototype for the ITER neutral beam injector due to thermal deformation of the support structure.

    Science.gov (United States)

    Sartori, E; Pavei, M; Marcuzzi, D; Zaccaria, P

    2014-02-01

    The beam formation and acceleration of the ITER neutral beam injector will be studied in the full-scale ion source, Source for Production of Ions of Deuterium Extracted from a RF plasma (SPIDER). It will be able to sustain 40 A deuterium ion beam during 1-h pulses. The operating conditions of its multi-aperture electrodes will diverge from ideality, as a consequence of inhomogeneous heating and thermally induced deformations in the support structure of the extraction and acceleration grids, which operate at different temperatures. Meeting the requirements on the aperture alignment and distance between the grids with such a large number of apertures (1280) and the huge support structures constitute a challenge. Examination of the structure thermal deformation in transient and steady conditions has been carried out, evaluating their effect on the beam performance: the paper describes the analyses and the solutions proposed to mitigate detrimental effects.

  8. Non-ideal operating conditions of the ion source prototype for the ITER neutral beam injector due to thermal deformation of the support structure

    Science.gov (United States)

    Sartori, E.; Pavei, M.; Marcuzzi, D.; Zaccaria, P.

    2014-02-01

    The beam formation and acceleration of the ITER neutral beam injector will be studied in the full-scale ion source, Source for Production of Ions of Deuterium Extracted from a RF plasma (SPIDER). It will be able to sustain 40 A deuterium ion beam during 1-h pulses. The operating conditions of its multi-aperture electrodes will diverge from ideality, as a consequence of inhomogeneous heating and thermally induced deformations in the support structure of the extraction and acceleration grids, which operate at different temperatures. Meeting the requirements on the aperture alignment and distance between the grids with such a large number of apertures (1280) and the huge support structures constitute a challenge. Examination of the structure thermal deformation in transient and steady conditions has been carried out, evaluating their effect on the beam performance: the paper describes the analyses and the solutions proposed to mitigate detrimental effects.

  9. Non-ideal operating conditions of the ion source prototype for the ITER neutral beam injector due to thermal deformation of the support structure

    Energy Technology Data Exchange (ETDEWEB)

    Sartori, E., E-mail: emanuele.sartori@igi.cnr.it; Pavei, M.; Marcuzzi, D.; Zaccaria, P. [Consorzio RFX, Euratom-ENEA Association, Corso Stati Uniti 4, 35127 Padova (Italy)

    2014-02-15

    The beam formation and acceleration of the ITER neutral beam injector will be studied in the full-scale ion source, Source for Production of Ions of Deuterium Extracted from a RF plasma (SPIDER). It will be able to sustain 40 A deuterium ion beam during 1-h pulses. The operating conditions of its multi-aperture electrodes will diverge from ideality, as a consequence of inhomogeneous heating and thermally induced deformations in the support structure of the extraction and acceleration grids, which operate at different temperatures. Meeting the requirements on the aperture alignment and distance between the grids with such a large number of apertures (1280) and the huge support structures constitute a challenge. Examination of the structure thermal deformation in transient and steady conditions has been carried out, evaluating their effect on the beam performance: the paper describes the analyses and the solutions proposed to mitigate detrimental effects.

  10. Transient, compressible heat and mass transfer in porous media using the strongly implicit iteration procedure.

    Science.gov (United States)

    Curry, D. M.; Cox, J. E.

    1972-01-01

    Coupled nonlinear partial differential equations describing heat and mass transfer in a porous matrix are solved in finite difference form with the aid of a new iterative technique (the strongly implicit procedure). Example numerical results demonstrate the characteristics of heat and mass transport in a porous matrix such as a charring ablator. It is emphasized that multidimensional flow must be considered when predicting the thermal response of a porous material subjected to nonuniform boundary conditions.

  11. Transient, compressible heat and mass transfer in porous media using the strongly implicit iteration procedure.

    Science.gov (United States)

    Curry, D. M.; Cox, J. E.

    1972-01-01

    Coupled nonlinear partial differential equations describing heat and mass transfer in a porous matrix are solved in finite difference form with the aid of a new iterative technique (the strongly implicit procedure). Example numerical results demonstrate the characteristics of heat and mass transport in a porous matrix such as a charring ablator. It is emphasized that multidimensional flow must be considered when predicting the thermal response of a porous material subjected to nonuniform boundary conditions.

  12. Impact of heating and current drive mix on the ITER hybrid scenario

    NARCIS (Netherlands)

    Citrin, J.; Artaud, J. F.; Garcia, J.; Hogeweij, G. M. D.; Imbeaux, F.

    2010-01-01

    Hybrid scenario performance in ITER is studied with the CRONOS integrated modelling suite, using the GLF23 anomalous transport model for heat transport prediction. GLF23 predicted core confinement is optimized through tailoring the q-profile shape by a careful choice of current drive actuators, affe

  13. Neutral gas density depletion due to neutral gas heating and pressure balance in an inductively coupled plasma

    Science.gov (United States)

    Shimada, Masashi; Tynan, George R.; Cattolica, Robert

    2007-02-01

    The spatial distribution of neutral gas temperature and total pressure have been measured for pure N2, He/5%N2 and Ar/5%N2 in an inductively coupled plasma (ICP) reactor, and a significant rise in the neutral gas temperature has been observed. When thermal transpiration is used to correct total pressure measurements, the total pressure remains constant regardless of the plasma condition. Neutral pressure is depleted due to the pressure balance when the plasma pressure (mainly electron pressure) becomes comparable to the neutral pressure in high density plasma. Since the neutral gas follows the ideal gas law, the neutral gas density profile was obtained from the neutral gas temperature and the corrected neutral pressure measurements. The results show that the neutral gas density at the centre of the plasma chamber (factor of 2-4 ×) decreases significantly in the presence of a plasma discharge. Significant spatial variation in neutral gas uniformity occurs in such plasmas due to neutral gas heating and pressure balance.

  14. Critical heat flux performance of hypervapotrons proposed for use in the ITER divertor vertical target

    Energy Technology Data Exchange (ETDEWEB)

    Youchison, D.L.; Marshall, T.D.; McDonald, J.M.; Lutz, T.J.; Watson, R.D. [Sandia National Labs., Albuquerque, NM (United States); Driemeyer, D.E. Kubik, D.L.; Slattery, K.T.; Hellwig, T.H. [McDonnell Douglas Aerospace, St. Louis, MO (United States)

    1997-09-01

    Task T-222 of the International Thermonuclear Experimental Reactor (ITER) program addresses the manufacturing and testing of permanent components for use in the ITER divertor. Thermalhydraulic and critical heat flux performance of the heat sinks proposed for use in the divertor vertical target are part of subtask T-222.4. As part of this effort, two single channel, medium scale, bare copper alloy, hypervapotron mockups were designed, fabricated, and tested using the EB-1200 electron beam system. The objectives of the effort were to develop the design and manufacturing procedures required for construction of robust high heat flux (HHF) components, verify thermalhydraulic, thermomechanical and critical heat flux (CHF) performance under ITER relevant conditions, and perform analyses of HHF data to identify design guidelines and failure criteria and possibly modify any applicable CHF correlations. The design, fabrication, and finite element modeling of two types of hypervapotrons are described; a common version already in use at the Joint European Torus (JET) and a new attached fin design. HHF test data on the attached fin hypervapotron will be used to compare the CHF performance under uniform heating profiles on long heated lengths with that of localized, highly peaked, off nominal profiles.

  15. Investigation of first mirror heating for the collective Thomson scattering diagnostic in ITER

    DEFF Research Database (Denmark)

    Salewski, Mirko; Meo, Fernando; Bindslev, Henrik;

    2008-01-01

    Collective Thomson scattering (CTS) has the capabilities to measure phase space densities of fast ion populations in ITER resolved in configuration space, in velocity space, and in time. In the CTS system proposed for ITER, probing radiation at 60 GHz generated by two 1 MW gyrotrons is scattered...... modeling of a first mirror on the high field side indicates that the mirror curvature may warp due to heating. This may alter the beam quality, and therefore, thermal effects have to be accounted for during the design of the mirror. The modeling further demonstrates that thin mirrors are superior to thick...

  16. Qualification of Fin-Type Heat Exchangers for the ITER Current Leads

    CERN Document Server

    Ballarino, A; Bordini, B; Devred, A; Ding, K; Niu, E; Sitko, M; Taylor, T; Yang, Y; Zhou, T

    2015-01-01

    The ITER current leads will transfer large currents of up to 68 kA into the biggest superconducting magnets ever built. Following the development of prototypes and targeted trials of specific manufacturing processes through mock-ups, the ASIPP (Chinese Institute of Plasma Physics) is preparing for the series fabrication. A key component of the ITER HTS current leads are the resistive heat exchangers. Special R&D was conducted for these components at CERN and ASIPP in support of their designs. In particular several mock-ups were built and tested in room temperature gas to measure the dynamic pressure drop and compare to 3D CFD models.

  17. Modelling ELM heat flux deposition on the ITER main chamber wall

    Energy Technology Data Exchange (ETDEWEB)

    Kočan, M., E-mail: martin.kocan@iter.org [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, F-13067 St Paul lez Durance Cedex (France); Pitts, R.A.; Lisgo, S.W.; Loarte, A. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, F-13067 St Paul lez Durance Cedex (France); Gunn, J.P. [Association Euratom-CEA, CEA/DSM/IRFM, Cadarache, 13108 Saint-Paul-lez-Durance (France); Fuchs, V. [Institute of Plasma Physics, Association EURATOM/IPP.CR, Praha 18200 (Czech Republic)

    2015-08-15

    The interaction of ELM filaments with the ITER beryllium first wall panels (FWPs) is studied using a simple ad-hoc fluid model of the filament parallel transport, taking into account the full, three-dimensional structure of the FWPs, including magnetic shadowing effects. The calculated ELM surface heat loads are used as input to the RACLETTE heat transfer code to estimate the FWP surface temperature rise. The results indicate that controlled ELMs in ITER during burning plasma operation (ΔW{sub ELM} ≈ 0.6 M J) will not lead to melting or significant evaporation of the beryllium surfaces, even in the case of high ELM broadening and the minimum allowable distance between the primary and secondary separatrices. The ELM-averaged steady-state heat load also stays below the maximum power handling capability of the FWPs.

  18. Status of the ITER Electron Cyclotron Heating and Current Drive System

    Science.gov (United States)

    Darbos, Caroline; Albajar, Ferran; Bonicelli, Tullio; Carannante, Giuseppe; Cavinato, Mario; Cismondi, Fabio; Denisov, Grigory; Farina, Daniela; Gagliardi, Mario; Gandini, Franco; Gassmann, Thibault; Goodman, Timothy; Hanson, Gregory; Henderson, Mark A.; Kajiwara, Ken; McElhaney, Karen; Nousiainen, Risto; Oda, Yasuhisa; Omori, Toshimichi; Oustinov, Alexander; Parmar, Darshankumar; Popov, Vladimir L.; Purohit, Dharmesh; Rao, Shambhu Laxmikanth; Rasmussen, David; Rathod, Vipal; Ronden, Dennis M. S.; Saibene, Gabriella; Sakamoto, Keishi; Sartori, Filippo; Scherer, Theo; Singh, Narinder Pal; Strauß, Dirk; Takahashi, Koji

    2016-01-01

    The electron cyclotron (EC) heating and current drive (H&CD) system developed for the ITER is made of 12 sets of high-voltage power supplies feeding 24 gyrotrons connected through 24 transmission lines (TL), to five launchers, four located in upper ports and one at the equatorial level. Nearly all procurements are in-kind, following general ITER philosophy, and will come from Europe, India, Japan, Russia and the USA. The full system is designed to couple to the plasma 20 MW among the 24 MW generated power, at the frequency of 170 GHz, for various physics applications such as plasma start-up, central H&CD and magnetohydrodynamic (MHD) activity control. The design takes present day technology and extends toward high-power continuous operation, which represents a large step forward as compared to the present state of the art. The ITER EC system will be a stepping stone to future EC systems for DEMO and beyond.

  19. Policies and initiatives for carbon neutrality in nordic heating and transport systems

    DEFF Research Database (Denmark)

    Muller, Jakob Glarbo; Wu, Qiuwei; Ostergaard, Jacob;

    2012-01-01

    to heat pumps in the Nordic region rely on both private economic and national economic incentives. Initiatives toward carbon neutrality in the transport system are mostly concentrated on research, development and demonstration for deployment of a large number of EVs. All Nordic countries have plans......Policies and initiatives promoting carbon neutrality in the Nordic heating and transport systems are presented. The focus within heating systems is the propagation of heat pumps while the focus within transport systems is initiatives regarding electric vehicles (EVs). It is found that conversion...... for the future heating and transport systems with the ambition of realizing carbon neutrality....

  20. Critical heat flux performance of hypervapotrons proposed for use in the ITER divertor vertical target

    Science.gov (United States)

    Youchison, Dennis L.; Marshall, Theron D.; McDonald, Jimmie M.; Lutz, Thomas J.; Watson, Robert D.; Driemeyer, Daniel E.; Kubik, David L.; Slattery, Kevin T.; Hellwig, Theodore H.

    1997-12-01

    Task T-222 of the International Thermonuclear Experimental Reactor (ITER) program addresses the manufacturing and testing of permanent components for use in the ITER divertor. Thermal-hydraulic and critical heat flux performance of the heat sinks proposed for use in the divertor vertical target are part of subtask T-222.4. As part of this effort, two single channel, medium-scale, bare copper alloy, hypervapotron mock-ups were designed by Sandia National Laboratories and McDonnell Douglas Aerospace (MDA), fabricated at MDA and tested at Sandia' Plasma Materials Test Facility using the EB-1200 electron beam system. The objectives of our effort were to develop the design and manufacturing procedures required for construction of robust HHF components, verify thermal-hydraulic, thermomechanical and CHF performance under ITER relevant conditions, and perform analyses of HHF data to identify design guidelines, failure criteria and possibly modify any applicable CHF correlations. This paper describes the design, fabrication and finite elements modeling of two types of hypervapotrons, a common version already in use at JET and a new attached- fin design. HHF test data on the attached-fin hypervapotron will be used to compare the CHF performance under uniform heating profiles on long heated lengths to that of localized, highly peaked, off-nominal profiles.

  1. Thermo-mechanical study of high heat flux component mock-ups for ITER TBM

    Energy Technology Data Exchange (ETDEWEB)

    Bonelli, Flavia [Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (Germany); Dipartimento Energia, Politecnico di Torino (Italy); Boccaccini, Lorenzo Virgilio, E-mail: lorenzo.boccaccini@kit.edu [Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (Germany); Kunze, André; Maione, Ivan Alessio [Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (Germany); Savoldi, Laura; Zanino, Roberto [Dipartimento Energia, Politecnico di Torino (Italy)

    2015-10-15

    Highlights: • Infrared radiation heaters for test of plasma facing component available at KIT. • Numerical model developed and validated to check uniformity of heat flux. • Thermo-mechanical calculations performed on a mock-up of the HCPB TBM FW. • Assessment done of representativity of stress conditions for the ITER TBMs. - Abstract: Commercial infrared heaters have been proposed to be used in the HELOKA facility under construction at Karlsruhe Institute of Technology (KIT) to test a mock-up of the first wall (FW), called thermo-cycle mock-up (TCM) plate, under stress loading comparable to those experienced by the test blanket modules (TBMs) in ITER. Two related issues are analyzed in this paper, in relation to the ongoing European project aimed at the design of the two EU TBMs: (1) the possibility to reproduce, by means of those heaters, high heat flux loading conditions on the TCM plate similar to those expected on the ITER TBMs, and (2) the thermo-mechanical analysis of the TCM itself, in order to define a suitable choice of experimental parameters and mechanical constraints leading to a relevant stress condition. A suitable heater model is developed and validated against experimental data from an ad-hoc test campaign. A thermo-mechanical study of the TCM plate is presented, showing that the structure is able to withstand high thermal loads, even in the most constrained case, reaching stress levels comparable to the ITER TBM.

  2. Manufacturing, assembly and tests of SPIDER Vacuum Vessel to develop and test a prototype of ITER neutral beam ion source

    Energy Technology Data Exchange (ETDEWEB)

    Zaccaria, Pierluigi, E-mail: pierluigi.zaccaria@igi.cnr.it [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete S.p.A.), Padova (Italy); Valente, Matteo; Rigato, Wladi; Dal Bello, Samuele; Marcuzzi, Diego; Agostini, Fabio Degli; Rossetto, Federico; Tollin, Marco [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete S.p.A.), Padova (Italy); Masiello, Antonio [Fusion for Energy F4E, Barcelona (Spain); Corniani, Giorgio; Badalocchi, Matteo; Bettero, Riccardo; Rizzetto, Dario [Ettore Zanon S.p.A., Schio (VI) (Italy)

    2015-10-15

    Highlights: • The SPIDER experiment aims to qualify and optimize the ion source for ITER injectors. • The large SPIDER Vacuum Vessel was built and it is under testing at the supplier. • The main working and assembly steps for production are presented in the paper. - Abstract: The SPIDER experiment (Source for the Production of Ions of Deuterium Extracted from an RF plasma) aims to qualify and optimize the full size prototype of the negative ion source foreseen for MITICA (full size ITER injector prototype) and the ITER Heating and Current Drive Injectors. Both SPIDER and MITICA experiments are presently under construction at Consorzio RFX in Padova (I), with the financial support from IO (ITER Organization), Fusion for Energy, Italian research institutions and contributions from Japan and India Domestic Agencies. The vacuum vessel hosting the SPIDER in-vessel components (Beam Source and calorimeters) has been manufactured, assembled and tested during the last two years 2013–2014. The cylindrical vessel, about 6 m long and 4 m in diameter, is composed of two cylindrical modules and two torispherical lids at the ends. All the parts are made by AISI 304 L stainless steel. The possibility of opening/closing the vessel for monitoring, maintenance or modifications of internal components is guaranteed by bolted junctions and suitable movable support structures running on rails fixed to the building floor. A large number of ports, about one hundred, are present on the vessel walls for diagnostic and service purposes. The main working steps for construction and specific technological issues encountered and solved for production are presented in the paper. Assembly sequences and tests on site are furthermore described in detail, highlighting all the criteria and requirements for correct positioning and testing of performances.

  3. Design status and procurement activities of the High Voltage Deck 1 and Bushing for the ITER Neutral Beam Injector

    Energy Technology Data Exchange (ETDEWEB)

    Boldrin, Marco, E-mail: marco.boldrin@igi.cnr.it [Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, I-35127 Padova (Italy); De Lorenzi, Antonio [Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, I-35127 Padova (Italy); Decamps, Hans [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Grando, Luca [Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, I-35127 Padova (Italy); Simon, Muriel [Fusion For Energy, c/ Josep Pla 2, 08019 Barcelona (Spain); Toigo, Vanni [Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, I-35127 Padova (Italy)

    2013-10-15

    Highlights: ► ITER Neutral Beam Injector includes several non-standard components. ► The design status of the −1 MV{sub dc} HVD1 and Bushing is described. ► The paper reports also on the integrated layout of the two components. ► Preliminary electrostatic and thermal analyses are presented. ► Procurement activities are outlined. -- Abstract: The ITER Neutral Beam Injector (NBI) power supply system includes several non-standard components, whose ratings go beyond the present industrial practice. Two of these items, to be procured by Fusion for Energy, are: 1.A −1 MV{sub dc} air-insulated Faraday cage, called High Voltage Deck 1 (HVD1), hosting the Ion Source and Extractor Power Supplies (ISEPS) and the associated diagnostics. 2.A −1 MV{sub dc} feedthrough, called HVD1-TL Bushing, aimed at connecting the HVD1 to the gas (SF{sub 6}) insulated Transmission Line (TL), containing inside its High Voltage (HV) conductor all ISEPS power and control cables coming from the HVD1 to be connected to the NBI Ion Source services. The paper deals with the status of the design of the HVD1 and HVD1-TL Bushing, focusing on insulation, mechanical and thermal issues as well as on their integration with the other components of the power supply system. In particular, the insulation issue of the integrated system has been addressed by means of an electrostatic Finite Element (FE) analysis whilst a FE thermal simulation has been carried out to assess the impact of the dissipation of the proposed design of the inner conductors (ISEPS conductors) not actively cooled. Finally, the paper describes the status of procurement strategy and execution.

  4. Effect of boundary conditions on the neutral gas temperatures and densities in the ITER divertor and pump duct

    Science.gov (United States)

    Ruzic, D. N.; Juliano, D. R.

    1992-12-01

    The DEGAS neutral atom transport code was used to simulate helium pumping and D/T throughput in ITER. The sensitivity of the simulation to two different reflection models, four transmission probabilities from the exit of the simulation to the pump (0.0625, 0.125, 0.1875 and 0.250), and a 2-D model versus a 3-D model were analyzed. The variation in reflection model changes the densities in the duct and the recycling of D/T by a factor of 1.6. The variation in the transmission probabilities affects these same quantities by a factor of 2.5. The dimensionality of the simulation affects the density profile in the duct. A transmission probability from the exit of the DEGAS simulation to the pump of 0.110 to 0.125 was calculated from the ITER reference drawings. Using this quantity and the DEGAS results, an exhaust rate of 112 to 127 moles/h is predicted, implying that the reference pumping systems may be larger than necessary by a factor of 2.

  5. The heat removal capability of actively cooled plasma-facing components for the ITER divertor

    Science.gov (United States)

    Missirlian, M.; Richou, M.; Riccardi, B.; Gavila, P.; Loarer, T.; Constans, S.

    2011-12-01

    Non-destructive examination followed by high-heat-flux testing was performed for different small- and medium-scale mock-ups; this included the most recent developments related to actively cooled tungsten (W) or carbon fibre composite (CFC) armoured plasma-facing components. In particular, the heat-removal capability of these mock-ups manufactured by European companies with all the main features of the ITER divertor design was investigated both after manufacturing and after thermal cycling up to 20 MW m-2. Compliance with ITER requirements was explored in terms of bonding quality, heat flux performances and operational compatibility. The main results show an overall good heat-removal capability after the manufacturing process independent of the armour-to-heat sink bonding technology and promising behaviour with respect to thermal fatigue lifetime under heat flux up to 20 MW m-2 for the CFC-armoured tiles and 15 MW m-2 for the W-armoured tiles, respectively.

  6. Lower hybrid heating and current drive design for ITER and application for present tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Froissard, P.; Rey, G.; Bibet, P.; Goniche, M.; Kazarian, F.; Portafaix, C.; Tonon, G. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Bosia, G.; Bruno, L. [ITER Joint Work Site, Garching (Germany); Kuzikov, S. [Inst. of Applied Physics, Nizhny Novgorod (Russian Federation); Wasastjerna, F. [VTT Energy (Finland)

    1998-07-01

    The lower Hybrid Heating and Current Drive (LHH and CD) System shall provide on ITER off-axis current profile control during burn, main contribution to the non-inductive current generation in the advanced Tokamak scenario, current profile tailoring during ramp up phase, heating and current drive during plasma shut-down, extension of the pulse duration during commissioning phase. The LHH and CD system operates at 5 GHz, this frequency being a trade-off between power absorption by alpha particles and klystron technology and couples a minimum of 50 MW using two ITER ports. This article describes the launcher plug and the transmission lines. Specific converters, such as the mode converters, RF windows and the hyper-guide have now been successfully tested at high power and long pulse duration.

  7. The Influence of Neutral Beam Injection on the Heating and Current Drive with Electron Cyclotron Wave on EAST

    Science.gov (United States)

    Chang, Pengxiang; Wu, Bin; Wang, Jinfang; Li, Yingying; Wang, Xiaoguang; Xu, Handong; Wang, Xiaojie; Liu, Yong; Zhao, Hailin; Hao, Baolong; Yang, Zhen; Zheng, Ting; Hu, Chundong

    2016-11-01

    Both neutral beam injection (NBI) and electron cyclotron resonance heating (ECRH) have been applied on the Experimental Advanced Superconducting Tokamak (EAST) in the 2015 campaign. In order to achieve more effective heating and current drive, the effects of NBI on the heating and current drive with electron cyclotron wave (ECW) are analyzed utilizing the code TORAY and experimental data in the shot #54411 and #54417. According to the experimental and simulated results, for the heating with ECW, NBI can improve the heating efficiency and move the power deposition place towards the inside of the plasma. On the other hand, for the electron cyclotron current drive (ECCD), NBI can also improve the efficiency of ECCD and move the place of ECCD inward. These results will be valuable for the center heating, the achievement of fully non-inductive current drive operation and the suppression of magnetohydrodynamic (MHD) instabilities with ECW on EAST or ITER with many auxiliary heating methods. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB101001 and 2014DFG61950) and National Natural Science Foundation of China (Nos. 11405212 and 11175211)

  8. High heat flux testing of EU tungsten monoblock mock-ups for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Gavila, P., E-mail: pierre.gavila@f4e.europa.eu [Fusion for Energy, 08019 Barcelona (Spain); Riccardi, B. [Fusion for Energy, 08019 Barcelona (Spain); Pintsuk, G. [Forschungszentrum Juelich, 52425 Juelich (Germany); Ritz, G. [AREVA NP, Centre Technique France, 71205 Le Creusot (France); Kuznetsov, V. [JCS “Efremov Institute”, Doroga na Metallostroy 3, Metallostroy, Saint-Petersburg 196641 (Russian Federation); Durocher, A. [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 Saint Paul-lez-Durance (France)

    2015-10-15

    Highlights: • All the tested items sustained the ITER Full W divertor qualification program requirements. This confirms that the technology for the manufacturing of the first set of the ITER Divertor is available in Europe. • The surface roughening and local melting of the W surface under high heat flux was proven to be significantly reduced for an armour thickness lower or equal to 6 mm. • However, this campaign highlighted some specific areas of improvement to be implemented ideally before the upcoming ITER Divertor IVT serial production. • The issue of the self-castellation of the W monoblocks, which typically appears after a few tenths of cycles at 20 MW/m{sup 2}, is critical because it generates some uncontrolled defects at the amour to heat sink joints. Besides, they create a gap which exposure is almost perpendicular to the magnetic field lines and which might lead to local W melting in the strike point region. • This campaign also evidenced that the minimum IO requirements on the CuCrZr ductility could be revised to avoid the occurrence of rather early fatigue failures. Although the W material characterization program has been set up by the IO, the strategy on the CuCrZr still needs to be defined. - Abstract: With the aim to assess the option to start the ITER operation with a full tungsten divertor, an R&D program was launched in order to evaluate the performances of tungsten (W) armoured plasma facing components (PFCs) under high heat flux. The F4E program consisted in the manufacturing and high heat flux (HHF) testing of W monoblock mock-ups and medium scale prototypes up to 20 MW/m{sup 2}. During the test campaign, 26 W mock-ups and two medium scale prototypes manufactured by Plansee SE (Austria) and by Ansaldo Nucleare (Italy) have been tested at the FE200 (AREVA, Le Creusot, France) and ITER Divertor Test Facility (IDTF) (Efremov Institute Saint Petersburg, Russian Federation) electron beam test facilities. The high heat flux (HHF) testing

  9. High heat flux engineering for the upgraded neutral beam injection systems of MAST-U

    Energy Technology Data Exchange (ETDEWEB)

    Dhalla, F., E-mail: Fahim.dhalla@ccfe.ac.uk; Mistry, S.; Turner, I.; Barrett, T.R.; Day, I.; McAdams, R.

    2015-10-15

    Highlights: • A new Residual Ion Dump (RID) and bend magnet system for the upgraded NBI systems have been designed for the 5 s MAST-U pulse requirements. • Design scoping was performed using numerical ion-tracing analysis software (MAGNET and OPERA codes). • A more powerful bending magnet will separate the residual ions into full, half and third energy components. • Three separate CuCrZr dumps spread the power loading resulting in acceptable power footprints. • FE thermo-mechanical analyses using ANSYS to validate the designs against the ITER SDC-IC code. • New bend magnet coils, yoke and CuCrZr water-cooled plates are in the procurement phase. - Abstract: For the initial phase of MAST-U operation the two existing neutral beam injection systems will be used, but must be substantially upgraded to fulfil expected operational requirements. The major elements are the design, manufacture and installation of a bespoke bending magnet and Residual Ion Dump (RID) system. The MAST-design full energy dump is being replaced with new actively-cooled full, half and third energy dumps, designed to receive 2.4 MW of ion power deflected by an iron-cored electromagnet. The main design challenge is limited space available in the vacuum vessel, requiring ion-deflection calculations to ensure acceptable heat flux distribution on the dump panels. This paper presents engineering and physics analysis of the upgraded MAST beamlines and reports the current status of manufacture.

  10. High voltage power supplies for ITER RF heating and current drive systems

    Energy Technology Data Exchange (ETDEWEB)

    Gassmann, T., E-mail: thibault.gassmann@iter.org [ITER Organization, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Arambhadiya, B.; Beaumont, B. [ITER Organization, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Baruah, U.K. [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Bonicelli, T. [Fusion For Energy, C/3 Josep Pla 2, Torres Diagonal Litoral-B3, E-08019 Barcelona (Spain); Darbos, C.; Purohit, D.; Decamps, H. [ITER Organization, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Albajar, F. [Fusion For Energy, C/3 Josep Pla 2, Torres Diagonal Litoral-B3, E-08019 Barcelona (Spain); Gandini, F.; Henderson, M.; Kazarian, F.; Lamalle, P.U.; Omori, T. [ITER Organization, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Parmar, D.; Patel, A. [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Rathi, D. [ITER Organization, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Singh, N.P. [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India)

    2011-10-15

    The RF heating and current drive (H and CD) systems to be installed for the ITER fusion machine are the electron cyclotron (EC), ion cyclotron (IC) and, although not in the first phase of the project, lower hybrid (LH). These systems require high voltage, high current power supplies (HVPS) in CW operation. These HVPS should deliver around 50 MW electrical power to each of the RF H and CD systems with stringent requirements in terms of accuracy, voltage ripple, response time, turn off time and fault energy. The PSM (Pulse Step Modulation) technology has demonstrated over the past 20 years its ability to fulfill these requirements in many industrial facilities and other fusion reactors and has therefore been chosen as reference design for the IC and EC HVPS systems. This paper describes the technical specifications, including interfaces, the resulting constraints on the design, the conceptual design proposed for ITER EC and IC HVPS systems and the current status.

  11. Erosion simulation of first wall beryllium armour under ITER transient heat loads

    Science.gov (United States)

    Bazylev, B.; Janeschitz, G.; Landman, I.; Pestchanyi, S.; Loarte, A.

    2009-04-01

    The beryllium is foreseen as plasma facing armour for the first wall in the ITER in form of Be-clad blanket modules in macrobrush design with brush size about 8-10 cm. In ITER significant heat loads during transient events (TE) are expected at the main chamber wall that may leads to the essential damage of the Be armour. The main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. Melting thresholds and melt layer depth of the Be armour under transient loads are estimated for different temperatures of the bulk Be and different shapes of transient loads. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the Lorentz force are analyzed for bulk Be and different sizes of Be-brushes. The damage of FW under radiative loads arising during mitigated disruptions is numerically simulated.

  12. On the meniscus formation and the negative hydrogen ion extraction from ITER neutral beam injection relevant ion source

    Science.gov (United States)

    Mochalskyy, S.; Wünderlich, D.; Ruf, B.; Fantz, U.; Franzen, P.; Minea, T.

    2014-10-01

    The development of a large area (Asource,ITER = 0.9 × 2 m2) hydrogen negative ion (NI) source constitutes a crucial step in construction of the neutral beam injectors of the international fusion reactor ITER. To understand the plasma behaviour in the boundary layer close to the extraction system the 3D PIC MCC code ONIX is exploited. Direct cross checked analysis of the simulation and experimental results from the ITER-relevant BATMAN source testbed with a smaller area (Asource,BATMAN ≈ 0.32 × 0.59 m2) has been conducted for a low perveance beam, but for a full set of plasma parameters available. ONIX has been partially benchmarked by comparison to the results obtained using the commercial particle tracing code for positive ion extraction KOBRA3D. Very good agreement has been found in terms of meniscus position and its shape for simulations of different plasma densities. The influence of the initial plasma composition on the final meniscus structure was then investigated for NIs. As expected from the Child-Langmuir law, the results show that not only does the extraction potential play a crucial role on the meniscus formation, but also the initial plasma density and its electronegativity. For the given parameters, the calculated meniscus locates a few mm downstream of the plasma grid aperture provoking a direct NI extraction. Most of the surface produced NIs do not reach the plasma bulk, but move directly towards the extraction grid guided by the extraction field. Even for artificially increased electronegativity of the bulk plasma the extracted NI current from this region is low. This observation indicates a high relevance of the direct NI extraction. These calculations show that the extracted NI current from the bulk region is low even if a complete ion-ion plasma is assumed, meaning that direct extraction from surface produced ions should be present in order to obtain sufficiently high extracted NI current density. The calculated extracted currents, both ions

  13. Neutron emission in neutral beam heated KSTAR plasmas and its application to neutron radiography

    Energy Technology Data Exchange (ETDEWEB)

    Kwak, Jong-Gu, E-mail: jgkwak@nfri.re.kr; Kim, H.S.; Cheon, M.S.; Oh, S.T.; Lee, Y.S.; Terzolo, L.

    2016-11-01

    Highlights: • We measured the neutron emission from KSTAR plasmas quantitatively. • We confirmed that neutron emission is coming from neutral beam-plasma interactions. • The feasibility study shows that the fast neutron from KSTAR could be used for fast neutron radiography. - Abstract: The main mission of Korea Superconducting Tokamak Advanced Research (KSTAR) program is exploring the physics and technologies of high performance steady state Tokamak operation that are essential for ITER and fusion reactor. Since the successful first operation in 2008, the plasma performance is enhanced and duration of H-mode is extended to around 50 s which corresponds to a few times of current diffusion time and surpassing the current conventional Tokamak operation. In addition to long-pulse operation, the operational boundary of the H-mode discharge is further extended over MHD no-wall limit(β{sub N} ∼ 4) transiently and higher stored energy region is obtained by increased total heating power (∼6 MW) and plasma current (I{sub p} up to 1 MA for ∼10 s). Heating system consists of various mixtures (NB, ECH, LHCD, ICRF) but the major horse heating resource is the neutral beam(NB) of 100 keV with 4.5 MW and most of experiments are conducted with NB. So there is a lot of production of fast neutrons coming from via D(d,n){sup 3}He reaction and it is found that most of neutrons are coming from deuterium beam plasma interaction. Nominal neutron yield and the area of beam port is about 10{sup 13}–10{sup 14}/s and 1 m{sup 2} at the closest access position of the sample respectively and neutron emission could be modulated for application to the neutron radiography by varying NB power. This work reports on the results of quantitative analysis of neutron emission measurements and results are discussed in terms of beam-plasma interaction and plasma confinement. It also includes the feasibility study of neutron radiography using KSTAR.

  14. High power millimeter wave experiment of ITER relevant electron cyclotron heating and current drive system.

    Science.gov (United States)

    Takahashi, K; Kajiwara, K; Oda, Y; Kasugai, A; Kobayashi, N; Sakamoto, K; Doane, J; Olstad, R; Henderson, M

    2011-06-01

    High power, long pulse millimeter (mm) wave experiments of the RF test stand (RFTS) of Japan Atomic Energy Agency (JAEA) were performed. The system consists of a 1 MW/170 GHz gyrotron, a long and short distance transmission line (TL), and an equatorial launcher (EL) mock-up. The RFTS has an ITER-relevant configuration, i.e., consisted by a 1 MW-170 GHz gyrotron, a mm wave TL, and an EL mock-up. The TL is composed of a matching optics unit, evacuated circular corrugated waveguides, 6-miter bends, an in-line waveguide switch, and an isolation valve. The EL-mock-up is fabricated according to the current design of the ITER launcher. The Gaussian-like beam radiation with the steering capability of 20°-40° from the EL mock-up was also successfully proved. The high power, long pulse power transmission test was conducted with the metallic load replaced by the EL mock-up, and the transmission of 1 MW/800 s and 0.5 MW/1000 s was successfully demonstrated with no arcing and no damages. The transmission efficiency of the TL was 96%. The results prove the feasibility of the ITER electron cyclotron heating and current drive system.

  15. Investigating the performance of simplified neutral-ion collisional heating rate in a global IT model

    Science.gov (United States)

    Zhu, Jie; Ridley, Aaron J.

    2016-01-01

    The Joule heating rate has usually been used as an approximate form of the neutral-ion collisional heating rate in the thermospheric energy equation in global thermosphere-ionosphere models. This means that the energy coupling has ignored the energy gained by the ions from collisions with electrons. It was found that the globally averaged thermospheric temperature (Tn) was underestimated in simulations using the Joule heating rate, by about 11% when F10.7=110 solar flux unit (sfu, 1 sfu = 10-22 W m-2 Hz-1) in a quiet geomagnetic condition. The underestimation of Tn was higher at low latitudes than high latitudes, and higher at F region altitudes than at E region altitudes. It was found that adding additional neutral photoelectron heating in a global IT model compensated for the underestimation of Tn using the Joule heating approximation. Adding direct photoelectron heating to the neutrals compensated for the indirect path for the energy that flows from the electrons to the ions then to the neutrals naturally and therefore was an adequate compensation over the dayside. There was a slight dependence of the underestimation of Tn on F10.7, such that larger activity levels resulted in a need for more compensation in direct photoelectron heating to the neutrals to make up for the neglected indirect heating through ions and electrons.

  16. Evaluation of induced activity, decay heat and dose rate distribution after shutdown in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Maki, Koichi [Hitachi Ltd., Ibaraki (Japan). Hitachi Research Lab.; Satoh, Satoshi; Hayashi, Katsumi; Yamada, Koubun; Takatsu, Hideyuki; Iida, Hiromasa

    1997-03-01

    Induced activity, decay heat and dose rate distributions after shutdown were estimated for 1MWa/m{sup 2} operation in ITER. The activity in the inboard blanket one day after shutdown is 1.5x10{sup 11}Bq/cm{sup 3}, and the average decay heating rate 0.01w/cm{sup 3}. The dose rate outside the 120cm thick concrete biological shield is two order higher than the design criterion of 5{mu}Sv/h. This indicates that the biological shield thickness should be enhanced by 50cm in concrete, that is, total thickness 170cm for workers to enter the reactor room and to perform maintenance. (author)

  17. Study on mitigation of pulsed heat load for ITER cryogenic system

    Science.gov (United States)

    Peng, N.; Xiong, L. Y.; Jiang, Y. C.; Tang, J. C.; Liu, L. Q.

    2015-03-01

    One of the key requirements for ITER cryogenic system is the mitigation of the pulsed heat load deposited in the magnet system due to magnetic field variation and pulsed DT neutron production. As one of the control strategies, bypass valves of Toroidal Field (TF) case helium loop would be adjusted to mitigate the pulsed heat load to the LHe plant. A quasi-3D time-dependent thermal-hydraulic analysis of the TF winding packs and TF case has been performed to study the behaviors of TF magnets during the reference plasma scenario with the pulses of 400 s burn and repetition time of 1800 s. The model is based on a 1D helium flow and quasi-3D solid heat conduction model. The whole TF magnet is simulated taking into account thermal conduction between winding pack and case which are cooled separately. The heat loads are given as input information, which include AC losses in the conductor, eddy current losses in the structure, thermal radiation, thermal conduction and nuclear heating. The simulation results indicate that the temperature variation of TF magnet stays within the allowable range when the smooth control strategy is active.

  18. Evaporation and Vapor Shielding of CFC Targets Exposed to Plasma Heat Fluxes Relevant to ITER ELMs

    Energy Technology Data Exchange (ETDEWEB)

    Safronov, V.; Arkhipov, N.I.; Toporkov, D.A.; Zhitlukhin, A.M. [Troitsk Inst. for Innovation and Fusion Research, TRINITI, Kostromskaya, 12A, 79, RU-142092 Troitsk, Moscow Region (Russian Federation); Landman, I. [FZK-Forschungszentrum Karlsruhe, Association Euratom-FZK, Technik und Umwelt, Postfach 3640, D-7602l Karlsruhe (Germany)

    2007-07-01

    Full text of publication follows: Carbon-fibre composite (CFC) is foreseen presently as armour material for the divertor target in ITER. During the transient processes such as instabilities of Edge Localized Modes (ELMs) the target as anticipated will be exposed to the plasma heat loads of a few MJ/m{sup 2} on the time scale of a fraction of ms, which causes an intense evaporation at the target surface and contaminates tokamak plasma by evaporated carbon. The ITER transient loads are not achievable at existing tokamaks therefore for testing divertor armour materials other facilities, in particular plasma guns are employed. In the present work the CFC targets have been tested for ITER at the plasma gun facility MK- 200 UG in Troitsk by ELM relevant heat fluxes. The targets in the applied magnetic field up to 2 T were irradiated by hydrogen plasma streams of diameter 6 - 8 cm, impact ion energy 2 - 3 keV, pulse duration 0.05 ms and energy density varying in the range 0.05 - 1 MJ/m{sup 2}. Primary attention has been focused on the measurement of evaporation threshold and investigation of carbon vapor properties. Fast infrared pyrometer, optical and VUV spectrometers, framing cameras and plasma calorimeters were applied as diagnostics. The paper reports the results obtained on the evaporation threshold of CFC, the evaporation rate of the carbon fibers oriented parallel and perpendicular to the exposed target surface, the velocity of carbon vapor motion along and across the magnetic field lines, and the parameters of carbon plasma such as temperature, density and ionization state measured up to the distance 15 cm at varying plasma load. First experimental results on investigation of the vapor shield onset conditions are presented also. (authors)

  19. The targeted heating and current drive applications for the ITER electron cyclotron system

    Energy Technology Data Exchange (ETDEWEB)

    Henderson, M.; Darbos, C.; Gandini, F.; Gassmann, T.; Loarte, A.; Omori, T.; Purohit, D. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Saibene, G.; Gagliardi, M. [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Farina, D.; Figini, L. [Istituto di Fisica del Plasma CNR, 20125 Milano (Italy); Hanson, G. [US ITER Project Office, ORNL, 1055 Commerce Park, PO Box 2008, Oak Ridge, Tennessee 37831 (United States); Poli, E. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Takahashi, K. [Japan Atomic Energy Agency (JAEA), Naka, Ibaraki 311-0193 (Japan)

    2015-02-15

    A 24 MW Electron Cyclotron (EC) system operating at 170 GHz and 3600 s pulse length is to be installed on ITER. The EC plant shall deliver 20 MW of this power to the plasma for Heating and Current Drive (H and CD) applications. The EC system is designed for plasma initiation, central heating, current drive, current profile tailoring, and Magneto-hydrodynamic control (in particular, sawteeth and Neo-classical Tearing Mode) in the flat-top phase of the plasma. A preliminary design review was performed in 2012, which identified a need for extended application of the EC system to the plasma ramp-up, flattop, and ramp down phases of ITER plasma pulse. The various functionalities are prioritized based on those applications, which can be uniquely addressed with the EC system in contrast to other H and CD systems. An initial attempt has been developed at prioritizing the allocated H and CD applications for the three scenarios envisioned: ELMy H-mode (15 MA), Hybrid (∼12 MA), and Advanced (∼9 MA) scenarios. This leads to the finalization of the design requirements for the EC sub-systems.

  20. Visualization of heat transfer in material for varians of boundary value with Relaxation Iteration Gauss-Seidel method

    Science.gov (United States)

    Basuki, Imam; Cari; Suparmi

    2017-01-01

    The research was aimed to know the effect of initial boundary value to the heat propagation rate pattern using iterations over Gauss-Seidel relaxation method and to analyze the exact value of each node descritization profile of test material. This study was an analytical study to fine analytical an numerical solution. Result from this study is that the pattern of variation of the boundary or initial conditions of a material with regard conductivity value remains at steady state the exact value of the smallest are in the same iteration value. The indicates that the value of the thermal equilibrium tend to be at the same iteration. Result from study showed that the pattern of initial boundary values that causes steady state of heat propagation of test material that has smallest exact similar to the iteration value.

  1. Neutral beam heating of the TFTR vacuum vessel protective plates

    Energy Technology Data Exchange (ETDEWEB)

    Sink, D.A.

    1976-04-01

    The transmission of neutral beams through plasmas expected in the Tokamak Fusion Test Reactor (TFTR) has been investigated. An analytical expression for the transmission coefficient of a 120 keV deuterium beam through a tritium plasma was used and a model for the flux profile of the TFTR Neutral Beam System designed by LBL/LLL was developed and incorporated. The plasma is assumed to have a parabolic profile and is characterized by a major radius of 310 cm, a minor radius of 57 cm, and a central plasma density of greater than or equal to 0.4 x 10/sup 14/ cm-/sup 3/. To protect the stainless steel vacuum vessel walls of the TFTR device, tungsten plates are located inside the vessel. The loading of the tungsten protective plates during normal operation is well below the neutral beam fluxes which would melt the tungsten. The TFTR Neutral Beam System will inject a total of 20 MW of 120 keV deuterium atoms from twelve sources, as well as approximately 12 MW of 60 keV deuterium atoms. The fluxes anticipated on the tungsten plates due to an unattenuated beam which would be incident at an angle of 45/sup 0/ are less than or equal to 6.5 kW/cm/sup 2/. The fluxes due to an attenuated beam are calculated to be less than or equal to 0.35 kW/cm/sup 2/. For the maximum injection time of 0.5 second, a fault condition in which the plasma was not formed at the time of injection could induce a surface temperature very near the melting point of tungsten. For the standard 0.1 second injection time anticipated for TFTR, a similar fault condition would not cause the temperature to rise to more than 2000 K which is well below the melting point (3640 K) of tungsten.

  2. Local fractional variational iteration algorithm II for non-homogeneous model associated with the non-differentiable heat flow

    Directory of Open Access Journals (Sweden)

    Yu Zhang

    2015-10-01

    Full Text Available In this article, we begin with the non-homogeneous model for the non-differentiable heat flow, which is described using the local fractional vector calculus, from the first law of thermodynamics in fractal media point view. We employ the local fractional variational iteration algorithm II to solve the fractal heat equations. The obtained results show the non-differentiable behaviors of temperature fields of fractal heat flow defined on Cantor sets.

  3. Joule heating of the ITER TF cold structure: Effects of vertical control coil currents and ELMS

    Energy Technology Data Exchange (ETDEWEB)

    Radovinsky, A.; Pillsbury, R.D. Jr.

    1993-11-09

    The toroidal field coil and support structures for ITER are maintained at cryogenic temperatures. The time-varying currents in the poloidal field coil system will induce eddy currents in these structures. The associated Joule dissipation will cause local heating and require heat removal which will show up as a load on the cryogenic system. Studies of Joule heating of the ITER TF cold structure (TFCS) due to the currents in the poloidal field coil system are presented. The two regimes considered in this study are the plasma vertical stability control and the Edge Loss Mode (ELM) events. The 3-D, thin-shell, eddy current program, EDDYCUFF was used to analyze the eddy currents and Joule losses in the cold structure. The current versus time scenarios were defined. Four control coil options were studied. All schemes use coils external to the TF cold structure. Analyses of power depositions during the plasma vertical stability control were performed for each of the four options. For each of these options three different recovery times were assumed. The times were 3, 1, and 1/3 seconds. Sets of four sequential ELMs, as well as isolated ELMs have been studied for various sets of active PF coils. The results showed that the lowest average power dissipation in the TF cold structure occurs when a subset of PF2 and PF7 are active, and all the other PF coils are passive. The general conclusion is that to minimize power dissipation in the TF cold structure it is preferable that only coils PF2 and PF7 are active. The other coils (PF3-PF6) should be passive and driven by a condition of constant flux. It is recommended in particular, that coils PF3 and PF5 be allowed to change currents to conserve flux, since they provide the maximum shielding of the TFCS from the fields caused by the active coils.

  4. Uncertainty quantification of bacterial aerosol neutralization in shock heated gases

    Science.gov (United States)

    Schulz, J. C.; Gottiparthi, K. C.; Menon, S.

    2015-01-01

    A potential method for the neutralization of bacterial endospores is the use of explosive charges since the high thermal and mechanical stresses in the post-detonation flow are thought to be sufficient in reducing the endospore survivability to levels that pose no significant health threat. While several experiments have attempted to quantify endospore survivability by emulating such environments in shock tube configurations, numerical simulations are necessary to provide information in scenarios where experimental data are difficult to obtain. Since such numerical predictions require complex, multi-physics models, significant uncertainties could be present. This work investigates the uncertainty in determining the endospore survivability from using a reduced order model based on a critical endospore temperature. Understanding the uncertainty in such a model is necessary in quantifying the variability in predictions using large-scale, realistic simulations of bacterial endospore neutralization by explosive charges. This work extends the analysis of previous large-scale simulations of endospore neutralization [Gottiparthi et al. in (Shock Waves, 2014. doi:10.1007/s00193-014-0504-9)] by focusing on the uncertainty quantification of predicting endospore neutralization. For a given initial mass distribution of the bacterial endospore aerosol, predictions of the intact endospore percentage using nominal values of the input parameters match the experimental data well. The uncertainty in these predictions are then investigated using the Dempster-Shafer theory of evidence and polynomial chaos expansion. The studies show that the endospore survivability is governed largely by the endospore's mass distribution and their exposure or residence time at the elevated temperatures and pressures. Deviations from the nominal predictions can be as much as 20-30 % in the intermediate temperature ranges. At high temperatures, i.e., strong shocks, which are of the most interest, the

  5. Comparative decomposition kinetics of neutral monosaccharides by microwave and induction heating treatments.

    Science.gov (United States)

    Tsubaki, Shuntaro; Oono, Kiriyo; Onda, Ayumu; Yanagisawa, Kazumichi; Azuma, Jun-ichi

    2013-06-28

    The stabilities of five neutral monosaccharides (glucose, galactose, mannose, arabinose, and xylose) were kinetically compared after the molecules were submitted to microwave heating (internal heating) and induction heating (external heating) under completely identical thermal histories by employing PID (proportional, integral, and derivative) temperature controlled ovens and homogeneous mixing. By heating in water at 200°C, the rate constants for the decomposition reactions varied from 2.13×10(-4) to 3.87×10(-4)s(-1) for microwave heating; however, the values increased by 1.1- to 1.5-fold for induction heating. Similarly, in a dilute (0.8%) sulfuric acid solution, the decomposition rate constants varied from 0.61×10(-3) to 2.00×10(-3)s(-1) for microwave heating; however, the values increased by 1.5- to 2.2-fold for induction heating. The results show that microwave heating imparts greater stability to neutral monosaccharides than does induction heating. The undesirable decomposition of monosaccharides at the surface boundary of reactor walls may have increased the probability of monosaccharide decomposition during induction heating.

  6. Low energy, high power hydrogen neutral beam for plasma heating

    Energy Technology Data Exchange (ETDEWEB)

    Deichuli, P.; Davydenko, V.; Ivanov, A., E-mail: ivanov@inp.nsk.su; Mishagin, V.; Sorokin, A.; Stupishin, N. [Budker Institute of Nuclear Physics, Prospect Lavrentieva 11, 630090 Novosibirsk (Russian Federation); Korepanov, S.; Smirnov, A. [Tri Alpha Energy, Inc., Foothill Ranch, California 92610 (United States)

    2015-11-15

    A high power, relatively low energy neutral beam injector was developed to upgrade of the neutral beam system of the gas dynamic trap device and C2-U experiment. The ion source of the injector produces a proton beam with the particle energy of 15 keV, current of up to 175 A, and pulse duration of a few milliseconds. The plasma emitter of the ion source is produced by superimposing highly ionized plasma jets from an array of four arc-discharge plasma generators. A multipole magnetic field produced with permanent magnets at the periphery of the plasma box is used to increase the efficiency and improve the uniformity of the plasma emitter. Multi-slit grids with 48% transparency are fabricated from bronze plates, which are spherically shaped to provide geometrical beam focusing. The focal length of the Ion Optical System (IOS) is 3.5 m and the initial beam diameter is 34 cm. The IOS geometry and grid potentials were optimized numerically to ensure accurate beam formation. The measured angular divergences of the beam are ±0.01 rad parallel to the slits and ±0.03 rad in the transverse direction.

  7. Low energy, high power hydrogen neutral beam for plasma heating

    Science.gov (United States)

    Deichuli, P.; Davydenko, V.; Ivanov, A.; Korepanov, S.; Mishagin, V.; Smirnov, A.; Sorokin, A.; Stupishin, N.

    2015-11-01

    A high power, relatively low energy neutral beam injector was developed to upgrade of the neutral beam system of the gas dynamic trap device and C2-U experiment. The ion source of the injector produces a proton beam with the particle energy of 15 keV, current of up to 175 A, and pulse duration of a few milliseconds. The plasma emitter of the ion source is produced by superimposing highly ionized plasma jets from an array of four arc-discharge plasma generators. A multipole magnetic field produced with permanent magnets at the periphery of the plasma box is used to increase the efficiency and improve the uniformity of the plasma emitter. Multi-slit grids with 48% transparency are fabricated from bronze plates, which are spherically shaped to provide geometrical beam focusing. The focal length of the Ion Optical System (IOS) is 3.5 m and the initial beam diameter is 34 cm. The IOS geometry and grid potentials were optimized numerically to ensure accurate beam formation. The measured angular divergences of the beam are ±0.01 rad parallel to the slits and ±0.03 rad in the transverse direction.

  8. Investigation of damages induced by ITER-relevant heat loads during massive gas injections on Beryllium

    Directory of Open Access Journals (Sweden)

    B. Spilker

    2016-12-01

    Full Text Available Massive gas injections (MGIs will be used in ITER to mitigate the strong damaging effect of full performance plasma disruptions on the plasma facing components. The MGI method transforms the stored plasma energy to radiation that is spread across the vacuum vessel with poloidal and toroidal asymmetries. This work investigated the impact of MGI like heat loading on the first wall armor material beryllium. ITER-relevant power densities of 90-260MWm−2in combination with pulse durations of 5-10ms were exerted onto the S-65 grade beryllium specimens in the electron beam facility JUDITH 1. All tested loading conditions led to noticeable surface morphology changes and in the expected worst case scenario, a crater with thermally induced cracks with a depth of up to ∼340µm formed in the loaded area. The level of destruction in the loaded area was strongly dependent on the pulse number but also on the formation of beryllium oxide. The cyclic melting of beryllium could lead to an armor thinning mechanism under the presence of melt motion driving forces such as surface tension, magnetic forces, and plasma pressure.

  9. Qualification Program of Korea Heat Load Test Facility KoHLT-EB for ITER Plasma Facing Components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Suk-Kwon; Park, Seoung Dae; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The qualification tests were performed to evaluate the high heat flux test facility for the PFCs and fusion reactor materials. For the thermal fatigue test, two types of tungsten mock-ups were fabricated. The cooling performance was tested under the similar operation condition of ITER and fusion reactor. After the completion of the preliminary mockup test and facility qualification, the high heat flux test facility will assess the performance test for the various plasma facing components in fusion reactor materials. Preliminary thermo-hydraulic and performance tests were conducted using various test mockups for the plasma facing components in the high heat flux test facilities of the world. The previous heat flux tests were performed by using the graphite heater facilities in Korea. Several facilities which equipped with an electron beam as the uniform heat source were fabricated for the tokamak PFCs in the EU, Russia and US. These heat flux test facilities are utilized for a cyclic heat flux test of the PFCs. Each facility working for their own purpose in EU FZJ, US SNL, and Russia Efremov institute. For this purpose, KoHLTEB was constructed and this facility will be used for ITER TBM performance test with the small-scale and large-scale mockups, and prototype. Also, it has been used for other fusion application for developing plasma facing component (PFC) for ITER FW, tungsten divertor, and heat transfer experiment and so on under the domestic R and D program. Korea heat load test facility by using electron beam KoHLT-EB was constructed for the high heat flux test to verify the plasma facing components, including ITER TBM first wall.

  10. The WEST project: Testing ITER divertor high heat flux component technology in a steady state tokamak environment

    Energy Technology Data Exchange (ETDEWEB)

    Bucalossi, J., E-mail: jerome.bucalossi@cea.fr; Missirlian, M.; Moreau, P.; Samaille, F.; Tsitrone, E.; Houtte, D. van; Batal, T.; Bourdelle, C.; Chantant, M.; Corre, Y.; Courtois, X.; Delpech, L.; Doceul, L.; Douai, D.; Dougnac, H.; Faïsse, F.; Fenzi, C.; Ferlay, F.; Firdaouss, M.; Gargiulo, L.; and others

    2014-10-15

    The WEST project recently launched at Cadarache consists in transforming Tore Supra in an X-point divertor configuration while extending its long pulse capability, in order to test the ITER divertor technology. The implementation of a full tungsten actively cooled divertor with plasma facing unit representative of ITER divertor targets will allow addressing risks both in terms of industrial-scale manufacturing and operation of such components. Relevant plasma scenarios are foreseen for extensive testing under high heat load in the 10–20 MW/m{sup 2} range and ITER-like fluences (1000 s pulses). Plasma facing unit monitoring and development of protection strategies will be key elements of the WEST program. WEST is scheduled to enter into operation in 2016, and will provide a key facility to prepare and be prepared for ITER.

  11. The future already today. Autarkic and CO{sub 2} neutral heating and cooling; Die Zukunft schon heute. Autark und CO{sub 2}-neutral Heizen und Kuehlen

    Energy Technology Data Exchange (ETDEWEB)

    Melzer, Martin [Melzer Kaelte + Klima GmbH, Bornich (Germany); Burgunder, Jan [Daikin Airconditioning Germany GmbH, Duesseldorf (Germany). Regionalbuero

    2010-12-15

    The heat concept of the central office of the Volksbank Rhein-Lahn in Diez (Federal Republic of Germany) enables the complete heating and cooling of existing office accomodation CO{sub 2} neutral with a heat pump. In this year, the planning company was priced for this with the Golden-Refnet of the company Daikin.

  12. Transmission line component testing for the ITER Ion Cyclotron Heating and Current Drive System

    Science.gov (United States)

    Goulding, Richard; Bell, G. L.; Deibele, C. E.; McCarthy, M. P.; Rasmussen, D. A.; Swain, D. W.; Barber, G. C.; Barbier, C. N.; Cambell, I. H.; Moon, R. L.; Pesavento, P. V.; Fredd, E.; Greenough, N.; Kung, C.

    2014-10-01

    High power RF testing is underway to evaluate transmission line components for the ITER Ion Cyclotron Heating and Current Drive System. The transmission line has a characteristic impedance Z0 = 50 Ω and a nominal outer diameter of 305 mm. It is specified to carry up to 6 MW at VSWR = 1.5 for 3600 s pulses, with transient voltages up to 40 kV. The transmission line is actively cooled, with turbulent gas flow (N2) used to transfer heat from the inner to outer conductor, which is water cooled. High voltage and high current testing of components has been performed using resonant lines generating steady state voltages of 35 kV and transient voltages up to 60 kV. A resonant ring, which has operated with circulating power of 6 MW for 1 hr pulses, is being used to test high power, low VSWR operation. Components tested to date include gas barriers, straight sections of various lengths, and 90 degree elbows. Designs tested include gas barriers fabricated from quartz and aluminum nitride, and transmission lines with quartz and alumina inner conductor supports. The latest results will be presented. This manuscript has been authored by UT-Battelle, LLC, under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy.

  13. Progress in the design of Normal Heat Flux First Wall panels for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Cicero, Tindaro, E-mail: Tindaro.Cicero@f4e.europa.eu [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Jimenez, Marc; D’Amico, Gabriele; Pou, Jordi Ayneto; Dellopoulos, Georges; Alvaro, Elena; Cardenes, Sabas; Banetta, Stefano; Bellin, Boris; Zacchia, Francesco [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Calcagno, Barbara; Chappuis, Philippe; Gicquel, Stefan; Mitteau, Raphael; Raffray, Rene [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St., Paul Lez Durance Cedex (France)

    2015-10-15

    Highlights: • Improved detail design of several NHF FW panels for ITER. • Implemented design solutions to improve the manufacturing of NHF FW panels. • Performed FEM simulations for the overall assessment of NHF FW panels. • Performed detailed analyses for integration of diagnostics in the NHF FW panels. - Abstract: A typical NHF FW panel consists of a series of fingers, which represent the elementary plasma facing units and are designed to withstand 15,000 cycles at 2 MW/m{sup 2}. The fingers are mechanically joined and supported by a back structural element or “supporting beam”. The structure of a finger is made of three different materials: stainless steel for the supporting structure, copper chromium zirconium for the heat sink, and beryllium as armour material. Due to their location and to the interfaces with other systems (e.g. Diagnostics, Remote Handling), the NHF FW panels are divided in different main and minor variants. The aim of this paper is to present the design work performed towards the PA signature. CAD detailed models have been created in CATIA for main and minor variants. Examples of local design solutions, as well as design work to achieve the global configuration of specific modules are provided. Finite Element (FE) analyses have been carried out, in order to simulate the operational scenario of ITER and assess the thermo-mechanical behaviour of the most important FW panels against the required design criteria. This design and analyses activity is required to progress towards the finalization of the detailed design of the NHF FW main and minor variants.

  14. High heat flux testing of ITER ICH&CD antenna beryllium faraday screen bars mock-ups

    Energy Technology Data Exchange (ETDEWEB)

    Courtois, X., E-mail: xavier.courtois@cea.fr [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Meunier, L. [Fusion for Energy, 08019 Barcelona (Spain); Kuznetsov, V. [Efremov Institute, FSUE NIIEFA, St. Petersburg, 196641 (Russian Federation); Beaumont, B.; Lamalle, P. [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St Paul Lez Durance (France); Conchon, D. [ATMOSTAT Co, F-94815 Villejuif (France); Languille, P. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France)

    2016-11-01

    Highlights: • ITER ICH&CD antenna beryllium faraday screen bars mock-ups were manufactured. • The mock-ups are submitted to high heat loads to test their heat exhaust capabilities. • The mock-ups withstand without damage the design limit load. • Lifetime is gradually reduced when the heat load is augmented beyond the design limit. • Thermal and mechanical behavior are reproducible, and coherent with the calculation. - Abstract: The Faraday Screen (FS) is the plasma facing component of ITER ion cyclotron heating antennas shielding. The requirement for the high heat exhaust, and the limitation of the temperatures to minimize strain and thus offer sufficient resistance to fatigue, imply the need for high conductivity materials and a high cooling flow rate. The FS bars are constructed by a hipping process involving beryllium tiles, a pure copper layer, a copper chrome zirconium alloy for the cooling channel and a stainless steel backing strip. Two FS bars small scale mock-ups were manufactured and tested under high heat flux. They endured 15,000 heating cycles without degradation under nominal heat flux, and revealed growing flaws when the heat flux was progressively augmented beyond. In this case, the ultrasonic test confirms a strong delamination of the Be tiles.

  15. Numerical reconstruction of unknown Robin inclusions inside a heat conductor by a non-iterative method

    Science.gov (United States)

    Nakamura, Gen; Wang, Haibing

    2017-05-01

    Consider the problem of reconstructing unknown Robin inclusions inside a heat conductor from boundary measurements. This problem arises from active thermography and is formulated as an inverse boundary value problem for the heat equation. In our previous works, we proposed a sampling-type method for reconstructing the boundary of the Robin inclusion and gave its rigorous mathematical justification. This method is non-iterative and based on the characterization of the solution to the so-called Neumann- to-Dirichlet map gap equation. In this paper, we give a further investigation of the reconstruction method from both the theoretical and numerical points of view. First, we clarify the solvability of the Neumann-to-Dirichlet map gap equation and establish a relation of its solution to the Green function associated with an initial-boundary value problem for the heat equation inside the Robin inclusion. This naturally provides a way of computing this Green function from the Neumann-to-Dirichlet map and explains what is the input for the linear sampling method. Assuming that the Neumann-to-Dirichlet map gap equation has a unique solution, we also show the convergence of our method for noisy measurements. Second, we give the numerical implementation of the reconstruction method for two-dimensional spatial domains. The measurements for our inverse problem are simulated by solving the forward problem via the boundary integral equation method. Numerical results are presented to illustrate the efficiency and stability of the proposed method. By using a finite sequence of transient input over a time interval, we propose a new sampling method over the time interval by single measurement which is most likely to be practical.

  16. General-Purpose Heat Source Development: Safety Test Program. Postimpact evaluation, Design Iteration Test 3

    Energy Technology Data Exchange (ETDEWEB)

    Schonfeld, F.W.; George, T.G.

    1984-07-01

    The General-Purpose Heat Source(GPHS) provides power for space missions by transmitting the heat of /sup 238/PuO/sub 2/ decay to thermoelectric elements. Because of the inevitable return of certain aborted missions, the heat source must be designed and constructed to survive both re-entry and Earth impact. The Design Iteration Test (DIT) series is part of an ongoing test program. In the third test (DIT-3), a full GPHS module was impacted at 58 m/s and 930/sup 0/C. The module impacted the target at an angle of 30/sup 0/ to the pole of the large faces. The four capsules used in DIT-3 survived impact with minimal deformation; no internal cracks other than in the regions indicated by Savannah River Plant (SRP) preimpact nondestructive testing were observed in any of the capsules. The 30/sup 0/ impact orientation used in DIT-3 was considerably less severe than the flat-on impact utilized in DIT-1 and DIT-2. The four capsules used in DIT-1 survived, while two of the capsules used in DIT-2 breached; a small quantity (approx. = 50 ..mu..g) of /sup 238/PuO/sub 2/ was released from the capsules breached in the DIT-2 impact. All of the capsules used in DIT-1 and DIT-2 were severely deformed and contained large internal cracks. Postimpact analyses of the DIT-3 test components are described, with emphasis on weld structure and the behavior of defects identified by SRP nondestructive testing.

  17. Component development for the ITER Ion Cyclotron Heating and Current Drive Transmission Line and Matching System

    Science.gov (United States)

    Goulding, R. H.; McCarthy, M. P.; Rasmussen, D. A.; Swain, D. W.; Barber, G. C.; Barbier, C. N.; Cambell, I. H.; Gray, S. L.; Moon, R. L.; Pesavento, P. V.; Sanabria, R. M.; Fredd, E.; Greenough, N.

    2013-10-01

    The transmission line and matching network for the ITER Ion Cyclotron Heating and Current Drive System feeds two equatorial launchers, each with 24 phased current straps combined into groups of three, and each designed to couple up to 20 MW into ELMy H-mode plasmas in the frequency range 40-55 MHz, for pulse lengths up to 3600 s. The network includes > 1 km of 50 Ohm 300 mm diameter transmission line carrying up to 6 MW net power per line at VSWR = 1.5. In addition, there are 8 power splitters, 32 hybrid phase shifters incorporating 64 tuning stubs, 32 additional tuning stubs, and 36 vacuum capacitors, which are configured to provide pre-matching in the port cell region adjacent to the antenna, final matching, decoupling of mutual inductances between antenna elements, and passive ELM resilience. The development and design of the various system components will be discussed. High power tests of components have begun, and the latest results will be presented. This manuscript has been authored by UT-Battelle, LLC, under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy.

  18. Design of neutral particle incident heating apparatus for large scale helical apparatus

    Energy Technology Data Exchange (ETDEWEB)

    Kaneko, Osamu; Oka, Yoshihide; Osakabe, Masaki; Takeiri, Yasuhiko; Tsumori, Katsuyoshi; Akiyama, Ryuichi; Asano, Eiji; Kawamoto, Toshikazu; Kuroda, Tsutomu [National Inst. for Fusion Science, Nagoya (Japan)

    1997-02-01

    In the Institute of Nuclear Fusion Science, construction of the large scale helical apparatus has been progressed favorably, and constructions of the heating apparatus as well as of electron resonance apparatus were begun in their orders under predetermined manner since 1994 fiscal year. And, on 1995 fiscal year, construction of neutral particle incident heating apparatus, leading heat apparatus, was begun under 3 years planning. The plasma heating study system adopted the study results developed in this institute through the large scale hydrogen negative ion source and also adopted thereafter development on nuclear fusion study by modifying the original specification set at the beginning of the research plan before 7 years. As a result, system design was changed from initial 125 KeV to 180 KeV in the beam energy and to execute 15 MW incidence using two sets beam lines, to begin its manufacturing. Here is described on its new design with reason of its modifications. (G.K.)

  19. Ion heating during geomagnetic storms measured using energetic neutral atom imaging

    Science.gov (United States)

    Keesee, Amy; Elfritz, Justin; Katus, Roxanne; Scime, Earl

    2015-11-01

    Energy from the solar wind is deposited into the magnetosphere during geomagnetic storms. Much of this energy is deposited into the plasma sheet, driving phenomena that leads to heating. The plasma sheet ions are then injected to the inner magnetosphere, driving the ring current. While ions can undergo adiabatic heating during typical drift motion, collisional and wave-particle interactions can also lead to ion heating. A technique to measure ion temperatures using energetic neutral atom (ENA) data has been developed using ENA data from the Two Wide-angle Imaging Neutral-atom Spectrometers (TWINS) mission global maps of ion temperature during the evolution of geomagnetic storms are made. These maps exhibit the location and characteristics of regions of ion heating and during which storm phase they occur. Superposed epoch analyses of such maps have demonstrated typical characteristics of ion heating during storms driven by coronal mass ejections as compared to those driven by high speed solar wind streams. The temperatures have been used to establish boundary conditions for modeling of the inner magnetosphere. We will give an overview of recent studies using TWINS ion temperature maps. Work supported by NNX10AN08A and AGS-1113478.

  20. On the safety of ITER accelerators.

    Science.gov (United States)

    Li, Ge

    2013-01-01

    Three 1 MV/40A accelerators in heating neutral beams (HNB) are on track to be implemented in the International Thermonuclear Experimental Reactor (ITER). ITER may produce 500 MWt of power by 2026 and may serve as a green energy roadmap for the world. They will generate -1 MV 1 h long-pulse ion beams to be neutralised for plasma heating. Due to frequently occurring vacuum sparking in the accelerators, the snubbers are used to limit the fault arc current to improve ITER safety. However, recent analyses of its reference design have raised concerns. General nonlinear transformer theory is developed for the snubber to unify the former snubbers' different design models with a clear mechanism. Satisfactory agreement between theory and tests indicates that scaling up to a 1 MV voltage may be possible. These results confirm the nonlinear process behind transformer theory and map out a reliable snubber design for a safer ITER.

  1. Beam optics in a MeV-class multi-aperture multi-grid accelerator for the ITER neutral beam injector.

    Science.gov (United States)

    Kashiwagi, M; Taniguchi, M; Umeda, N; de Esch, H P L; Grisham, L R; Boilson, D; Hemsworth, R S; Tanaka, M; Tobari, H; Watanabe, K; Inoue, T

    2012-02-01

    In a multi-aperture multi-grid accelerator of the ITER neutral beam injector, the beamlets are deflected due to space charge repulsion between beamlets and beam groups, and also due to magnetic field. Moreover, the beamlet deflection is influenced by electric field distortion generated by grid support structure. Such complicated beamlet deflections and the compensations have been examined utilizing a three-dimensional beam analysis. The space charge repulsion and the influence by the grid support structure were studied in a 1∕4 model of the accelerator including 320 beamlets. Beamlet deflection due to the magnetic field was studied by a single beamlet model. As the results, compensation methods of the beamlet deflection were designed, so as to utilize a metal bar (so-called field shaping plate) of 1 mm thick beneath the electron suppression grid (ESG), and an aperture offset of 1 mm in the ESG.

  2. Iterative method for the numerical solution of a system of integral equations for the heat conduction initial boundary value problem

    Science.gov (United States)

    Svetushkov, N. N.

    2016-11-01

    The paper deals with a numerical algorithm to reduce the overall system of integral equations describing the heat transfer process at any geometrically complex area (both twodimensional and three-dimensional), to the iterative solution of a system of independent onedimensional integral equations. This approach has been called "string method" and has been used to solve a number of applications, including the problem of the detonation wave front for the calculation of heat loads in pulse detonation engines. In this approach "the strings" are a set of limited segments parallel to the coordinate axes, into which the whole solving area is divided (similar to the way the strings are arranged in a tennis racket). Unlike other grid methods where often for finding solutions, the values of the desired function in the region located around a specific central point here in each iteration step is determined by the solution throughout the length of the one-dimensional "string", which connects the two end points and set them values and determine the temperature distribution along all the strings in the first step of an iterative procedure.

  3. Soil as natural heat resource for very shallow geothermal application: laboratory and test site updates from ITER Project

    Science.gov (United States)

    Di Sipio, Eloisa; Bertermann, David

    2017-04-01

    Nowadays renewable energy resources for heating/cooling residential and tertiary buildings and agricultural greenhouses are becoming increasingly important. In this framework, a possible, natural and valid alternative for thermal energy supply is represented by soils. In fact, since 1980 soils have been studied and used also as heat reservoir in geothermal applications, acting as a heat source (in winter) or sink (in summer) coupled mainly with heat pumps. Therefore, the knowledge of soil thermal properties and of heat and mass transfer in the soils plays an important role in modeling the performance, reliability and environmental impact in the short and long term of engineering applications. However, the soil thermal behavior varies with soil physical characteristics such as soil texture and water content. The available data are often scattered and incomplete for geothermal applications, especially very shallow geothermal systems (up to 10 m depths), so it is worthy of interest a better comprehension of how the different soil typologies (i.e. sand, loamy sand...) affect and are affected by the heat transfer exchange with very shallow geothermal installations (i.e. horizontal collector systems and special forms). Taking into consideration these premises, the ITER Project (Improving Thermal Efficiency of horizontal ground heat exchangers, http://iter-geo.eu/), funded by European Union, is here presented. An overview of physical-thermal properties variations under different moisture and load conditions for different mixtures of natural material is shown, based on laboratory and field test data. The test site, located in Eltersdorf, near Erlangen (Germany), consists of 5 trenches, filled in each with a different material, where 5 helix have been installed in an horizontal way instead of the traditional vertical option.

  4. Investigation of the heat handling capabilities of DIII-D neutral beamline internal components

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, J.C.; Baxi, C.B.; Hong, R.

    1993-10-01

    The current DIII-D neutral beam system is a nominal five second pulse length upgrade of a previous design rated for only 500 msec operation. While the ion sources are rated for 60 sec operation, in practice pulse lengths are limited both by the beamline internal components ability to handle the fraction of the power which is scraped off, and by the power supplies ability to provide pulse lengths of greater than 5 sec. This paper examines the capability of the existing DIII-D neutral beamline heat removing components both in terms of present and desired operating parameters. To date, at 2.5 MW per ion source, pulses are limited to 3.5 sec by beamline internal components, while at lower ratings of 2.0 MW per ion source, up to 5 sec pulses have been achieved. Recent advances and demonstration of the extraction of 3.5 MW per DIII-D ion source give an even wider window of operating conditions. A full series of beamline thermocouple data has been collected to determine the heat loading and implied surface temperatures for the various DIII-D neutral beamline internal components. These data will be presented along with an analysis of the needs for specific component upgrades, given a desire for 10 sec operation.

  5. Transverse heat transfer coefficient in the dual channel ITER TF CICCs. Part III: Direct method of assessment

    Science.gov (United States)

    Lewandowska, Monika; Malinowski, Leszek

    2016-01-01

    The data resulting from the thermal-hydraulic test of the ITER TF CICC are used to determine the flow partition and the overall effective heat transfer coefficient (hBC) between bundle and central channel in a direct way, i.e. by analysis of the heat transfer between both flow channels, based on the mass and energy balance equations and the readings of thermometers located inside the cable. In cases without a local heat source in the considered cable segment the obtained hBC values were consistent with those obtained in earlier studies by analysis of experimental data using indirect methods. It was also observed that the transverse heat transfer was strongly enhanced in a cable segment heated from outside. This phenomenon results from the mass transfer from the bundle region to the central channel. The experimental hBC data obtained for the case without a heat source in the considered segment were also compared with those calculated using various heat transfer correlations.

  6. The European contribution to the development of the ITER NB injector

    Energy Technology Data Exchange (ETDEWEB)

    Masiello, A., E-mail: antonio.masiello@f4e.europa.eu [Fusion for Energy, C/Josep Pla 2, 08019 Barcelona (Spain); Agarici, G.; Bonicelli, T.; Simon, M. [Fusion for Energy, C/Josep Pla 2, 08019 Barcelona (Spain); Alonso, J. [Associacion EURATOM-CIEMAT, Av. Complutense 22, 28040, Madrid (Spain); Bigi, M. [Consorzio RFX, Euratom-ENEA Association, C.so Stati Uniti 4,I-35126, Padova (Italy); Boilson, D. [ITER, ITER Joint Work Site, CEA Cadarache, 13108 St. Paul Lez Durance (France); Chitarin, G. [Consorzio RFX, Euratom-ENEA Association, C.so Stati Uniti 4,I-35126, Padova (Italy); Day, C. [Karlsruhe Institute of Technology P.O. Box 3640, 76021, Karlsruhe (Germany); Franzen, P. [Max-Planck-Institut fuer Plasmaphysik - D-85740 Garching (Germany); Hanke, S. [Karlsruhe Institute of Technology P.O. Box 3640, 76021, Karlsruhe (Germany); Heinemann, B. [Max-Planck-Institut fuer Plasmaphysik - D-85740 Garching (Germany); Hemsworth, R. [ITER, ITER Joint Work Site, CEA Cadarache, 13108 St. Paul Lez Durance (France); Luchetta, A.; Marcuzzi, D. [Consorzio RFX, Euratom-ENEA Association, C.so Stati Uniti 4,I-35126, Padova (Italy); Milnes, J. [CCFE - Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, OX14 3DB, Oxfordshire (United Kingdom); Minea, T. [CNRS, Delegation Ile-de-France Sud - Avenue de la Terrasse, 91190 Gif-sur-Yvette (France); Pasqualotto, R.; Pomaro, N.; Serianni, G. [Consorzio RFX, Euratom-ENEA Association, C.so Stati Uniti 4,I-35126, Padova (Italy)

    2011-10-15

    This paper reviews the on-going design, R and D and procurement activities, mostly conducted within the ITER framework, on-going in Europe under the co-ordination of Fusion for Energy (F4E), in co-operation with the European Fusion Associations and aimed at the establishment of the ITER Heating Neutral Beam (HNB) system.

  7. Investigation of first mirror heating for the collective Thomson scattering diagnostic in ITER

    DEFF Research Database (Denmark)

    Salewski, Mirko; Meo, Fernando; Bindslev, Henrik

    2008-01-01

    Collective Thomson scattering (CTS) has the capabilities to measure phase space densities of fast ion populations in ITER resolved in configuration space, in velocity space, and in time. In the CTS system proposed for ITER, probing radiation at 60 GHz generated by two 1 MW gyrotrons is scattered...... in the plasma and collected by arrays of receivers. The transmission lines from the gyrotrons to the plasma and from the plasma to the receivers contain several quasioptical mirrors among other components. These are designed to produce astigmatic beam patterns in the plasma where the beam shapes will have...

  8. Dynamic disturbance rejection controllers for neutral time delay systems with application to a central heating system

    Institute of Scientific and Technical Information of China (English)

    KOUMBOULIS Fotis N.; KOUVAKAS Nikolaos D.; PARASKEVOPOULOS Paraskevas N.

    2009-01-01

    In the present paper the problem of disturbance rejection of single input-single output neutral time delay systems with multiple measurable disturbances is solved via dynamic controllers. In particular, the general form of the controller matrices is presented, while the necessary and sufficient conditions for the controller to be realizable are offered. The proposed technique is applied to a test case neutral time delay central heating system. In particular, the nonlinear model of the plant and its linearized approximation are presented. Based on the linearized model, a two-stage controller is designed in order to regulate the room temperature and the boiler effluent temperature. The performance of the closed loop system is investigated through computational experiments.

  9. A computational procedure for the investigation of whipping effect on ITER High Energy Piping and its application to the ITER divertor primary heat transfer system

    Energy Technology Data Exchange (ETDEWEB)

    Spagnuolo, G.A., E-mail: Alessandro.Spagnuolo@kraftanlagen.com [Kraftanlagen Heidelberg Gmbh, Im Breitspiel 7, D-69126 Heidelberg (Germany); Dell’Orco, G. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Di Maio, P.A. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy); Mazzei, M. [Kraftanlagen Heidelberg Gmbh, Im Breitspiel 7, D-69126 Heidelberg (Germany)

    2015-10-15

    Highlights: • High Energy Piping (HEP) are components containing water or steam with P ≥ 2.0 MPa and/or T ≥ 100 °C. • The whipping effect in HEP may cause dangerous domino effect with relative rupture propagation. • The rapture is envisaged or postulated according to the stress state of piping. • A FEM analysis is performed in order to study the dynamic of whipping effect. • Study of special support to avoid and/or mitigate the whipping effect. - Abstract: The Tokamak Cooling Water System of nuclear facility has the function to remove heat from plasma facing components maintaining coolant temperatures, pressures and flow rates as required and, depending on thermal-hydraulic requirements, its systems are defined as High Energy Piping (HEP) because they contain fluids, such as water or steam, at a pressure greater than or equal to 2.0 MPa and/or at a temperature greater than or equal to 100 °C, or even gas at pressure above the atmospheric one. The French standards contemplate the need to consider the whipping effect on HEP design. This effect happens when, after a double ended guillotine break, the reaction force could create a displacement of the piping which might affect adjacent components. A research campaign has been performed, in cooperation by ITER Organization and University of Palermo, to outline the procedure to check whether whipping effect might occur and assess its potential damage effects so to allow their mitigation. This procedure is based on the guidelines issued by U.S. Nuclear Regulatory Commission. The proposed procedure has been applied to the analysis of the whipping effect of divertor primary heat transfer system HEP, using a theoretical–computational approach based on the finite element method.

  10. Analysis of the Pipe Heat Loss of the Water Flow Calorimetry System in EAST Neutral Beam Injector

    Science.gov (United States)

    Hu, Chundong; Chen, Yu; Xu, Yongjian; Yu, Ling; Li, Xiang; Zhang, Weitang

    2016-11-01

    Neutral beam injection heating is one of the main auxiliary heating methods in controllable nuclear fusion research. In the EAST neutral beam injector, a water flow calorimetry (WFC) system is applied to measure the heat load on the electrode system of the ion source and the heat loading components of the beamline. Due to the heat loss in the return water pipe, there are some measuring errors for the current WFC system. In this paper, the errors were measured experimentally and analyzed theoretically, which lay a basis for the exact calculation of beam power deposition distribution and neutralization efficiency. supported by the National Magnetic Confinement Fusion Science Program of China (No. 2013GB101001) and the International Science & Technology Cooperation Program of China (No. 2014DFG61950)

  11. The ITER project construction status

    Science.gov (United States)

    Motojima, O.

    2015-10-01

    The pace of the ITER project in St Paul-lez-Durance, France is accelerating rapidly into its peak construction phase. With the completion of the B2 slab in August 2014, which will support about 400 000 metric tons of the tokamak complex structures and components, the construction is advancing on a daily basis. Magnet, vacuum vessel, cryostat, thermal shield, first wall and divertor structures are under construction or in prototype phase in the ITER member states of China, Europe, India, Japan, Korea, Russia, and the United States. Each of these member states has its own domestic agency (DA) to manage their procurements of components for ITER. Plant systems engineering is being transformed to fully integrate the tokamak and its auxiliary systems in preparation for the assembly and operations phase. CODAC, diagnostics, and the three main heating and current drive systems are also progressing, including the construction of the neutral beam test facility building in Padua, Italy. The conceptual design of the Chinese test blanket module system for ITER has been completed and those of the EU are well under way. Significant progress has been made addressing several outstanding physics issues including disruption load characterization, prediction, avoidance, and mitigation, first wall and divertor shaping, edge pedestal and SOL plasma stability, fuelling and plasma behaviour during confinement transients and W impurity transport. Further development of the ITER Research Plan has included a definition of the required plant configuration for 1st plasma and subsequent phases of ITER operation as well as the major plasma commissioning activities and the needs of the accompanying R&D program to ITER construction by the ITER parties.

  12. Convergence and Error Criteria of Iterative Numerical Solutions to the Transient Heat Conduction Equation.

    Science.gov (United States)

    1982-03-01

    L× x +b The Gauss- Seidel method attempts to accelerate convergence by using the (n+l)th iterative vector elements as soon as 8 they become available...Jacobi method, D x -U - L x ("’) + b D× + L x -U x ()+ (D + L) x -")= U X’ -b 9 x~"" -(DL)-’ (-U) x (D L+ for the Gauss- Seidel method , and ", ( D) / D

  13. R&D activities on RF contacts for the ITER ion cyclotron resonance heating launcher

    CERN Document Server

    Hillairet, Julien; Bamber, Rob; Beaumont, Bertrand; Bernard, Jean-Michel; Delaplanche, Jean-Marc; Durodié, Frédéric; Lamalle, Philippe; Lombard, Gilles; Nicholls, Keith; Shannon, Mark; Vulliez, Karl; Cantone, Vincent; Hatchressian, Jean-Claude; Lebourg, Philippe; Martinez, André; Mollard, Patrick; Mouyon, David; Pagano, Marco; Patterlini, Jean-Claude; Soler, Bernard; Thouvenin, Didier; Toulouse, Lionel; Verger, Jean-Marc; Vigne, Terence; Volpe, Robert

    2015-01-01

    Embedded RF contacts are integrated within the ITER ICRH launcher to allow assembling, sliding and to lower the thermo-mechanical stress. They have to withstand a peak RF current up to 2.5 kA at 55 MHz in steady-state conditions, in the vacuum environment of themachine.The contacts have to sustain a temperature up to 250{\\textdegree}Cduring several days in baking operations and have to be reliable during the whole life of the launcher without degradation. The RF contacts are critical components for the launcher performance and intensive R&D is therefore required, since no RF contactshave so far been qualified at these specifications. In order to test and validate the anticipated RF contacts in operational conditions, CEA has prepared a test platform consisting of a steady-state vacuum pumped RF resonator. In collaboration with ITER Organization and the CYCLE consortium (CYclotronCLuster for Europe), an R&D program has been conducted to develop RF contacts that meet the ITER ICRH launcher specification...

  14. Design of the ITER Electron Cyclotron Heating and Current Drive Waveguide Transmission Line

    Science.gov (United States)

    Bigelow, T. S.; Rasmussen, D. A.; Shapiro, M. A.; Sirigiri, J. R.; Temkin, R. J.; Grunloh, H.; Koliner, J.

    2007-11-01

    The ITER ECH transmission line system is designed to deliver the power, from twenty-four 1 MW 170 GHz gyrotrons and three 1 MW 127.5 GHz gyrotrons, to the equatorial and upper launchers. The performance requirements, initial design of components and layout between the gyrotrons and the launchers is underway. Similar 63.5 mm ID corrugated waveguide systems have been built and installed on several fusion experiments; however, none have operated at the high frequency and long-pulse required for ITER. Prototype components are being tested at low power to estimate ohmic and mode conversion losses. In order to develop and qualify the ITER components prior to procurement of the full set of 24 transmission lines, a 170 GHz high power test of a complete prototype transmission line is planned. Testing of the transmission line at 1-2 MW can be performed with a modest power (˜0.5 MW) tube with a low loss (10-20%) resonant ring configuration. A 140 GHz long pulse, 400 kW gyrotron will be used in the initial tests and a 170 GHz gyrotron will be used when it becomes available. Oak Ridge National Laboratory, managed by UT-Battelle, LLC, for the U.S. Dept. of Energy under contract DE-AC05-00OR22725.

  15. Artificial Neural Networks: a viable tool to design heat load smoothing strategies for the ITER Toroidal Field coils

    Science.gov (United States)

    Froio, A.; Bonifetto, R.; Carli, S.; Quartararo, A.; Savoldi, L.; Zanino, R.

    2015-12-01

    In superconducting tokamaks, cryoplants provide the helium needed to cool the superconducting magnet systems. The evaluation of the heat load from the magnets to the cryoplant is fundamental for the design of the latter and the assessment of suitable strategies to smooth the heat load pulses induced by the pulsed plasma scenarios is crucial for the operation. Here, a simplified thermal-hydraulic model of an ITER Toroidal Field (TF) magnet, based on Artificial Neural Networks (ANNs), is developed and inserted into a detailed model of the ITER TF winding and casing cooling circuits based on the state-of-the-art 4C code, which also includes active controls. The low computational effort requested by such a model allows performing a fast parametric study, to identify the best smoothing strategy during standard plasma operation. The ANNs are trained using 4C simulations, and the predictive capabilities of the simplified model are assessed against 4C simulations, both with and without active smoothing, in terms of accuracy and computational time.

  16. Active beam spectroscopy for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Hellermann, M.G. von, E-mail: mgvh@jet.u [FOM Institute Rijnhuizen, Euratom Association, 3430BE Nieuwegein (Netherlands); Barnsley, R. [ITER Organization, 13108 St.-Paul-Lez-Durance, Cadarache (France); Biel, W. [Institut fuer Energieforschung, Plasmaphysik, Forschungszentrum Juelich, Euratom Association, 52425 Juelich (Germany); Delabie, E. [FOM Institute Rijnhuizen, Euratom Association, 3430BE Nieuwegein (Netherlands); Hawkes, N. [Culham Centre for Fusion Energy, Euratom Association, Culham OX14 3DB (United Kingdom); Jaspers, R. [FOM Institute Rijnhuizen, Euratom Association, 3430BE Nieuwegein (Netherlands); Johnson, D. [Princeton Plasma Physics Laboratory, Princeton, NJ-08548 (United States); Klinkhamer, F. [TNO Science and Industry, Stieltjesweg 1, 2628CK Delft (Netherlands); Lischtschenko, O. [FOM Institute Rijnhuizen, Euratom Association, 3430BE Nieuwegein (Netherlands); Marchuk, O. [Institut fuer Energieforschung, Plasmaphysik, Forschungszentrum Juelich, Euratom Association, 52425 Juelich (Germany); Schunke, B. [ITER Organization, 13108 St.-Paul-Lez-Durance, Cadarache (France); Singh, M.J. [Institute for Plasma Research, Bhat, Gandhinagar, Gurajat 384828 (India); Snijders, B. [TNO Science and Industry, Stieltjesweg 1, 2628CK Delft (Netherlands); Summers, H.P. [Culham Centre for Fusion Energy, Euratom Association, Culham OX14 3DB (United Kingdom); Thomas, D. [ITER Organization, 13108 St.-Paul-Lez-Durance, Cadarache (France); Tugarinov, S. [TRINITI Troitsk, Moscow Region 142092 (Russian Federation); Vasu, P. [Institute for Plasma Research, Bhat, Gandhinagar, Gurajat 384828 (India)

    2010-11-11

    Since the first feasibility studies of active beam spectroscopy on ITER in 1995 the proposed diagnostic has developed into a well advanced and mature system. Substantial progress has been achieved on the physics side including comprehensive performance studies based on an advanced predictive code, which simulates active and passive features of the expected spectral ranges. The simulation has enabled detailed specifications for an optimized instrumentation and has helped to specify suitable diagnostic neutral beam parameters. Four ITER partners share presently the task of developing a suite of ITER active beam diagnostics, which make use of the two 0.5 MeV/amu 18 MW heating neutral beams and a dedicated 0.1 MeV/amu, 3.6 MW diagnostic neutral beam. The IN ITER team is responsible for the DNB development and also for beam physics related aspects of the diagnostic. The RF will be responsible for edge CXRS system covering the outer region of the plasma (1>r/a>0.4) using an equatorial observation port, and the EU will develop the core CXRS system for the very core (0ITER environment. Additionally, an essential change of the orientation of the DNB injection angle and specification of suitable blanket aperture has been made to avoid trapped particle damage to the first wall.

  17. R&D activities on RF contacts for the ITER ion cyclotron resonance heating launcher

    Energy Technology Data Exchange (ETDEWEB)

    Hillairet, Julien, E-mail: julien.hillairet@cea.fr [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Argouarch, Arnaud [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Bamber, Rob [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Beaumont, Bertrand [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Bernard, Jean-Michel; Delaplanche, Jean-Marc [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Durodié, Frédéric [Laboratory for Plasmas Physics, 1000 Brussels (Belgium); Lamalle, Philippe [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Lombard, Gilles [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Nicholls, Keith; Shannon, Mark [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Vulliez, Karl [Maestral Laboratory, Technetics Group, Pierrelatte (France); Cantone, Vincent; Hatchressian, Jean-Claude; Larroque, Sébastien; Lebourg, Philippe; Martinez, André; Mollard, Patrick; Mouyon, David; Pagano, Marco [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); and others

    2015-10-15

    Highlights: • CEA have developed a dedicated test-bed for testing RF contact in ITER relevant conditions (vacuum, temperature, RF current). • A prototype of RF contacts have been designed and manufactured, with copper lamellas brazed on a titanium holder. • This RF contact prototype failed at RF current larger than 1.8 kA. • Extensive R&D is foreseen with new RF contact designs. - Abstract: Embedded RF contacts are integrated within the ITER ICRH launcher to allow assembling, sliding and to lower the thermo-mechanical stress. They have to withstand a peak RF current up to 2.5 kA at 55 MHz in steady-state conditions, in the vacuum environment of the machine. The contacts have to sustain a temperature up to 250 °C during several days in baking operations and have to be reliable during the whole life of the launcher without degradation. The RF contacts are critical components for the launcher performance and intensive R&D is therefore required, since no RF contacts have so far been qualified at these specifications. In order to test and validate the anticipated RF contacts in operational conditions, CEA has prepared a test platform consisting of a steady-state vacuum pumped RF resonator. In collaboration with ITER Organization and the CYCLE consortium (CYclotron CLuster for Europe), an R&D program has been conducted to develop RF contacts that meet the ITER ICRH launcher specifications. A design proposed by CYCLE consortium, using brazed lamellas supported by a spring to improve thermal exchange efficiency while guaranteeing high contact force, was tested successfully in the T-resonator up to 1.7 kA during 1200 s, but failed for larger current values due to a degradation of the contacts. Details concerning the manufacturing of the brazed contacts on its titanium holder, the RF tests results performed on the resonator and the non-destructive tests analysis of the contacts are given in this paper.

  18. Iterative inversion of global magnetospheric ion distributions using energetic neutral atom (ENA images recorded by the NUADU/TC2 instrument

    Directory of Open Access Journals (Sweden)

    L. Lu

    2008-06-01

    Full Text Available A method has been developed for extracting magnetospheric ion distributions from Energetic Neutral Atom (ENA measurements made by the NUADU instrument on the TC-2 spacecraft. Based on a constrained linear inversion, this iterative technique is suitable for use in the case of an ENA image measurement, featuring a sharply peaked spatial distribution. The method allows for magnetospheric ion distributions to be extracted from a low-count ENA image recorded over a short integration time (5 min. The technique is demonstrated through its application to a set of representative ENA images recorded in energy Channel~2 (hydrogen: 50–81 keV, oxygen: 138–185 keV of the NUADU instrument during a geomagnetic storm. It is demonstrated that this inversion method provides a useful tool for extracting ion distribution information from ENA data that are characterized by high temporal and spatial resolution. The recovered ENA images obtained from inverted ion fluxes match most effectively the measurements made at maximum ENA intensity.

  19. Neutralized wettability effect of superhydrophilic Cr-layered surface on pool boiling critical heat flux

    Energy Technology Data Exchange (ETDEWEB)

    Son, Hong Hyun; Jeong, Ui Ju; Seo, Gwang Hyeok; Jeun, Gyoo Dong; Kim, Sung Joong [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The former method is deemed challenging due to longer development period and license issue. In this regard, FeCrAl, Cr, and SiC have been received positive attention as ATF coating materials because they are highly resistant to high temperature steam reaction causing massive hydrogen generation. In this study, Cr was selected as a target deposition material on the metal substrate because we found that Cr-layered surface becomes superhydrophilic, favorable to delaying the triggering of the critical heat flux (CHF). Thus in order to investigate the effect of Cr-layered superhydrophilic surfaces (under explored coating conditions) on pool boiling heat transfer, pool boiling experiment was conducted in the saturated deionized water under atmospheric pressure. As a physical vapor deposition (PVD) method, the DC magnetron sputtering technique was introduced to develop Cr-layered nanostructure. As a control variable of DC sputtering, substrate temperature was selected. Surface wettability and nanostructure were analyzed as major surface parameters on the CHF. We believe that highly dense micro/nano structure without nucleation cavities and inner pores neutralized the wettability effect on the CHF. Moreover, superhydrophilic surface with deficient cavity density rather hinders active nucleation. This emphasizes the importance of micro/nano structure surface for enhanced boiling heat transfer.

  20. Simulation of Be armour cracking under ITER-like transient heat loads

    Directory of Open Access Journals (Sweden)

    S. Pestchanyi

    2016-12-01

    Full Text Available Simulation of beryllium cracking under action of multiple severe surface heatings has been performed using the PEGASUS-3D code and verified by experiments in the JUDITH 1 facility. Analysis of the results has revealed beryllium thermo conductivity degradation under action of repetitive pulsed heat load due to accumulation of the cracks in the surface layer. Thermo conductivity degradation is found to be at least 4 times after 100 pulses in JUDITH 1 facility. An analytical model for the Be cracking threshold under action of arbitrary heat pulses has been developed.

  1. Heating and current drive requirements for ideal MHD stability and ITB sustainment in ITER steady state scenarios

    Science.gov (United States)

    Poli, Francesca

    2012-10-01

    Steady state scenarios envisaged for ITER aim at optimizing the bootstrap current, while maintaining sufficient confinement and stability to provide the necessary fusion yield. Non-inductive scenarios will need to operate with Internal Transport Barriers (ITBs) in order to reach adequate fusion gain at typical currents of 9 MA. However, the large pressure gradients associated with ITBs in regions of weak or negative magnetic shear can be conducive to ideal MHD instabilities in a wide range of βN, reducing the no-wall limit. Scenarios are established as relaxed flattop states with time-dependent transport simulations with TSC [1]. Fully non-inductive configurations with current in the range of 7-10 MA and various heating mixes (NB, EC, IC and LH) have been studied against variations of the pressure profile peaking and of the Greenwald fraction. It is found that stable equilibria have qmin> 2 and moderate ITBs at 2/3 of the minor radius [2]. The ExB flow shear from toroidal plasma rotation is expected to be low in ITER, with a major role in the ITB dynamics being played by magnetic geometry. Combinations of H&CD sources that maintain reverse or weak magnetic shear profiles throughout the discharge and ρ(qmin)>=0.5 are the focus of this work. The ITER EC upper launcher, designed for NTM control, can provide enough current drive off-axis to sustain moderate ITBs at mid-radius and maintain a non-inductive current of 8-9MA and H98>=1.5 with the day one heating mix. LH heating and current drive is effective in modifying the current profile off-axis, facilitating the formation of stronger ITBs in the rampup phase, their sustainment at larger radii and larger bootstrap fraction. The implications for steady state operation and fusion performance are discussed.[4pt] [1] Jardin S.C. et al, J. Comput. Phys. 66 (1986) 481[0pt] [2] Poli F.M. et al, Nucl. Fusion 52 (2012) 063027.

  2. Benchmarking of Decay Heat Measured Values of ITER Materials Induced by 14 MeV Neutron Activation with Calculated Results by ACAB Activation Code

    Energy Technology Data Exchange (ETDEWEB)

    Tore, C.; Ortego, P.; Rodriguez Rivada, A.

    2014-07-01

    The aim of this paper is the comparison between the calculated and measured decay heat of material samples which were irradiated at the Fusion Neutron Source of JAERI in Japan with D-T production of 14MeV neutrons. In the International Thermonuclear Experimental Reactor (ITER) neutron activation of the structural material will result in a source of heat after shutdown of the reactor. The estimation of decay heat value with qualified codes and nuclear data is an important parameter for the safety analyses of fusion reactors against lost of coolant accidents. When a loss of coolant and/or flow accident happen plasma facing components are heated up by decay heat. If the temperature of the components exceeds the allowable temperature, the accident would expand to loose the integrity of ITER. Uncertainties associated with decay prediction less than 15% are strongly requested by the ITER designers. Additionally, accurate decay heat prediction is required for making reasonable shutdown scenarios of ITER. (Author)

  3. Simulation of Be armour cracking under ITER-like transient heat loads [in press

    OpenAIRE

    Pestchanyi, S.; Spilker, B.; Bazylev, B.

    2016-01-01

    Simulation of beryllium cracking under action of multiple severe surface heatings has been performed using the PEGASUS-3D code and verified by experiments in the JUDITH 1 facility. Analysis of the results has revealed beryllium thermo conductivity degradation under action of repetitive pulsed heat load due to accumulation of the cracks in the surface layer. Thermo conductivity degradation is found to be at least 4 times after 100 pulses in JUDITH 1 facility. An analytical model for the Be cra...

  4. Simulation of Be armour cracking under ITER-like transient heat loads

    OpenAIRE

    Pestchanyi, S.; Spilker, B.; Bazylev, B

    2016-01-01

    Simulation of beryllium cracking under action of multiple severe surface heatings has been performed using the PEGASUS-3D code and verified by experiments in the JUDITH 1 facility. Analysis of the results has revealed beryllium thermo conductivity degradation under action of repetitive pulsed heat load due to accumulation of the cracks in the surface layer. Thermo conductivity degradation is found to be at least 4 times after 100 pulses in JUDITH 1 facility. An analytical model for the Be cra...

  5. Scenario-neutral Food Security Risk Assessment: A livestock Heat Stress Case Study

    Science.gov (United States)

    Broman, D.; Rajagopalan, B.; Hopson, T. M.

    2015-12-01

    Food security risk assessments can provide decision-makers with actionable information to identify critical system limitations, and alternatives to mitigate the impacts of future conditions. The majority of current risk assessments have been scenario-led and results are limited by the scenarios - selected future states of the world's climate system and socioeconomic factors. A generic scenario-neutral framework for food security risk assessments is presented here that uses plausible states of the world without initially assigning likelihoods. Measures of system vulnerabilities are identified and system risk is assessed for these states. This framework has benefited greatly by research in the water and natural resource fields to adapt their planning to provide better risk assessments. To illustrate the utility of this framework we develop a case study using livestock heat stress risk within the pastoral system of West Africa. Heat stress can have a major impact not only on livestock owners, but on the greater food production system, decreasing livestock growth, milk production, and reproduction, and in severe cases, death. A heat stress index calculated from daily weather is used as a vulnerability measure and is computed from historic daily weather data at several locations in the study region. To generate plausible states, a stochastic weather generator is developed to generate synthetic weather sequences at each location, consistent with the seasonal climate. A spatial model of monthly and seasonal heat stress provide projections of current and future livestock heat stress measures across the study region, and can incorporate in seasonal climate and other external covariates. These models, when linked with empirical thresholds of heat stress risk for specific breeds offer decision-makers with actionable information for use in near-term warning systems as well as for future planning. Future assessment can indicate under which states livestock are at greatest risk

  6. Characterization and heat flux testing of beryllium coatings on Inconel for JET ITER-like wall project

    Energy Technology Data Exchange (ETDEWEB)

    Hirai, T [Forschungszentrum Juelich, Association EURATOM-FZJ, 52425 Juelich (Germany); Linke, J [Forschungszentrum Juelich, Association EURATOM-FZJ, 52425 Juelich (Germany); Sundelin, P [Alfven Laboratory, Association EURATOM-VR, 100 44 Stockholm (Sweden); Rubel, M [Alfven Laboratory, Association EURATOM-VR, 100 44 Stockholm (Sweden); Kuehnlein, W [Forschungszentrum Juelich, Association EURATOM-FZJ, 52425 Juelich (Germany); Wessel, E [Forschungszentrum Juelich, Association EURATOM-FZJ, 52425 Juelich (Germany); Coad, J P [Culham Science Centre, EURATOM-UKAEA Fusion Association, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Lungu, C P [National Institute of Lasers, Plasma and Radiation Physics, Association EURATOM-MEdC, Bucharest (Romania); Matthews, G F [Culham Science Centre, EURATOM-UKAEA Fusion Association, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Pedrick, L [Culham Science Centre, EURATOM-UKAEA Fusion Association, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Piazza, G [EFDA, CSU, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom)

    2007-03-15

    In order to perform a fully integrated material test, JET has launched the ITER-like wall project with the aim of installing a full metal wall during the next major shutdown. The material foreseen for the main chamber wall is bulk Be at the limiters and Be coatings on inconel tiles elsewhere. R and D process comprises global characterization (structure, purity etc) of the evaporated films and testing of their performance under heat loads. The major results are (i) the layers have survived energy loads of 20 MJ m{sup -2} which is significantly above the required level of 5-10 MJ m{sup -2} (ii) melting limit of beryllium coating would be at the energy level of 30 MJ m{sup -2} (iii) cyclic thermal load of 10 MJ m{sup -2} for up to 50 cycles have not induced any noticeable damage such as flaking or detachment.

  7. Benchmarking of codes for electron cyclotron heating and electron cyclotron current drive under ITER conditions

    NARCIS (Netherlands)

    Prater, R.; Farina, D.; Gribov, Y.; Harvey, R. W.; Ram, A. K.; Lin-Liu, Y. R.; Poli, E.; Smirnov, A. P.; Volpe, F.; Westerhof, E.; Zvonkovo, A.

    2008-01-01

    Optimal design and use of electron cyclotron heating requires that accurate and relatively quick computer codes be available for prediction of wave coupling, propagation, damping and current drive at realistic levels of EC power. To this end, a number of codes have been developed in laboratories wor

  8. Evolution of transiently melt damaged tungsten under ITER-relevant divertor plasma heat loading

    Energy Technology Data Exchange (ETDEWEB)

    Bardin, S., E-mail: s.bardin@differ.nl [FOM Institute DIFFER – Dutch Institute For Fundamental Energy Research, Ass EURATOM-FOM, Trilateral Euregio Cluster, Nieuwegein (Netherlands); Morgan, T.W. [FOM Institute DIFFER – Dutch Institute For Fundamental Energy Research, Ass EURATOM-FOM, Trilateral Euregio Cluster, Nieuwegein (Netherlands); Glad, X. [Université de Lorraine, Institut Jean Lamour, Vandoeuvre-les-Nancy (France); Pitts, R.A. [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); De Temmerman, G. [FOM Institute DIFFER – Dutch Institute For Fundamental Energy Research, Ass EURATOM-FOM, Trilateral Euregio Cluster, Nieuwegein (Netherlands); ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

    2015-08-15

    A high-repetition-rate ELM simulation system was used at both the Pilot-PSI and Magnum-PSI linear plasma devices to investigate the nature of W damage under multiple shallow melt events and the subsequent surface evolution under ITER relevant plasma fluence and high ELM number. First, repetitive shallow melting of two W monoblocks separated by a 0.5 mm gap was obtained by combined pulsed/steady-state hydrogen plasma loading at normal incidence in the Pilot-PSI device. Surface modifications including melting, cracking and strong net-reshaping of the surface are obtained. During the second step, the pre-damaged W sample was exposed to a high flux plasma regime in the Magnum-PSI device with a grazing angle of 35°. SEM analysis indicates no measurable change to the surface state after the exposure in Magnum-PSI. An increase in transient-induced temperature rise of 40% is however observed, indicating a degradation of thermal properties over time.

  9. Experimental study of ELM-like heat loading on beryllium under ITER operational conditions

    Science.gov (United States)

    Spilker, B.; Linke, J.; Pintsuk, G.; Wirtz, M.

    2016-02-01

    The experimental fusion reactor ITER, currently under construction in Cadarache, France, is transferring the nuclear fusion research to the power plant scale. ITER’s first wall (FW), armoured by beryllium, is subjected to high steady state and transient power loads. Transient events like edge localized modes not only deposit power densities of up to 1.0 GW m-2 for 0.2-0.5 ms in the divertor of the machine, but also affect the FW to a considerable extent. Therefore, a detailed study was performed, in which transient power loads with absorbed power densities of up to 1.0 GW m-2 were applied by the electron beam facility JUDITH 1 on beryllium specimens at base temperatures of up to 300 °C. The induced damage was evaluated by means of scanning electron microscopy and laser profilometry. As a result, the observed damage was highly dependent on the base temperatures and absorbed power densities. In addition, five different classes of damage, ranging from ‘no damage’ to ‘crack network plus melting’, were defined and used to locate the damage, cracking, and melting thresholds within the tested parameter space.

  10. A Spectroscopic Study of Impurity Behavior in Neutral-beam and Ohmically Heated TFTR Discharges

    Science.gov (United States)

    Stratton, B. C.; Ramsey, A. T.; Boody, F. P.; Bush, C. E.; Fonck, R. J.; Groenbner, R. J.; Hulse, R. A.; Richards, R. K.; Schivell, J.

    1987-02-01

    Quantitative spectroscopic measurements of Z{sub eff}, impurity densities, and radiated power losses have been made for ohmic- and neutral-beam-heated TFTR discharges at a plasma current of 2.2 MA and toroidal field of 4.7 T. Variations in these quantities with line-average plasma density (anti n{sub e}) and beam power up to 5.6 MW are presented for discharges on a graphite movable limiter. A detailed discussion of the use of an impurity transport model to infer absolute impurity densities and radiative losses from line intensity and visible continuum measurements is given. These discharges were dominated by low-Z impurities with carbon having a considerably higher density than oxygen, except in high-anti n{sub e} ohmic discharges, where the densities of carbon and oxygen were comparable. Metallic impurity concentrations and radiative losses were small, resulting in hollow radiated power profiles and fractions of the input power radiated being 30 to 50% for ohmic heating and 30% or less with beam heating. Spectroscopic estimates of the radiated power were in good agreement with bolometrically measured values. Due to an increase in the carbon density, Z{sub eff} rose from 2.0 to 2.8 as the beam power increased from 0 to 5.6 MW, pointing to a potentially serious dilution of the neutron-producing plasma ions as the beam power increased. Both the low-Z and metallic impurity concentrations were approximately constant with minor radius, indicating no central impurity accumulation in these discharges.

  11. ITER-relevant transient heat loads on tungsten exposed to plasma and beryllium

    Science.gov (United States)

    Yu, J. H.; Doerner, R. P.; Dittmar, T.; Höschen, T.; Schwarz-Selinger, T.; Baldwin, M. J.

    2014-04-01

    Tungsten (W) is presently the most attractive plasma facing material for future fusion reactors. Off-normal transient events such as edge localized modes and disruptions are simulated with a pulsed laser system in the PISCES-B facility, providing pulses with 1-10 ms duration with absorbed heat flux factors up to ˜90 MJ m-2 s-1/2. This paper characterizes surface morphology changes and damage thresholds under transient heating on W exposed to He plasma or D plasma with and without Be coatings. W is damaged in the form of grain growth, surface roughening, melting and cracking. With a Be coating on the order of μm thick, the laser pulse produces a variety of Be surface changes including Be-W alloying, vaporization of the Be layer, melting and delamination.

  12. Nucleus-staining with biomolecule-mimicking nitrogen-doped carbon dots prepared by a fast neutralization heat strategy.

    Science.gov (United States)

    Kang, Yan-Fei; Fang, Yang-Wu; Li, Yu-Hao; Li, Wen; Yin, Xue-Bo

    2015-12-11

    Biomolecule-mimicking nitrogen-doped carbon dots (N-Cdots) were synthesized from dopamine by a neutralization heat strategy. Fluorescence imaging of various cells validated their nucleus-staining efficiency. The dopamine-mimicking N-Cdots "trick" nuclear membranes to achieve nuclear localization and imaging.

  13. Strength Evaluation of Heat Affected Zone in Electron Beam Welded ARAA for HCCR TBM in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, J. S.; Kim, S. K.; Jin, H. G.; Lee, E. H.; Lee, D. W. [KAERI, Daejeon (Korea, Republic of); Cho, S. [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The Korean helium cooled ceramic reflector (HCCR) test blanket module (TBM) has been developed for ITER, and Korean reduced activation ferritic martensitic (RAFM) steel, called advanced reduced activation alloy (ARAA), has also been developed for a structural material of the HCCR TBM. One case of limited optimized electron beam (EB) welding conditions was selected based on previous work, and the weldability of an EB weld was evaluated for TBM fabrication. The micro-hardness was measured from the base to the weld region, and the microstructures were also observed. A small punch (SP) test considering the HAZ was carried out at room and high (550 .deg. C) temperatures. The empirical mechanical properties of HAZ in the EB weld were evaluated, and the fracture behavior was investigated after the SP test. The SP results show that the estimated yield and tensile strength of the HAZ were higher than the base metal at both temperatures. Korean RAFM steel, ARAA, was developed as a TBM structural material. Using one of the program alloys in ARAA (F206), one case of a limited optimized EB welding condition was selected based on previous works, and the weldability of an EB weld using the SP test was evaluated for TBM fabrication at room and high (550 .deg. C) temperatures. From a micro-Vickers hardness evaluation, the HAZ gave the highest values compared with the other regions. The irregular grain boundaries in the HAZ were observed, but its width was narrower than the TIG weld from the previous results. The optimized welding methods such as the TIG, EB, and laser weld, and the welding procedure considering the PWHT are being established, and the weldability evaluation is also progressing according to the development of the ARAA for the fusion material application in Korea.

  14. Application of powerful quasi-steady-state plasma accelerators for simulation of ITER transient heat loads on divertor surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Tereshin, V I [Institute of Plasma Physics of the NSC KIPT, Kharkov 61108 (Ukraine); Bandura, A N [Institute of Plasma Physics of the NSC KIPT, Kharkov 61108 (Ukraine); Byrka, O V [Institute of Plasma Physics of the NSC KIPT, Kharkov 61108 (Ukraine); Chebotarev, V V [Institute of Plasma Physics of the NSC KIPT, Kharkov 61108 (Ukraine); Garkusha, I E [Institute of Plasma Physics of the NSC KIPT, Kharkov 61108 (Ukraine); Landman, I [Forschungszentrum Karlsruhe, IHM, Karlsruhe 76021 (Germany); Makhlaj, V A [Institute of Plasma Physics of the NSC KIPT, Kharkov 61108 (Ukraine); Neklyudov, I M [Institute of Plasma Physics of the NSC KIPT, Kharkov 61108 (Ukraine); Solyakov, D G [Institute of Plasma Physics of the NSC KIPT, Kharkov 61108 (Ukraine); Tsarenko, A V [Institute of Plasma Physics of the NSC KIPT, Kharkov 61108 (Ukraine)

    2007-05-15

    The paper presents the investigations of high power plasma interaction with material surfaces under conditions simulating the ITER disruptions and type I ELMs. Different materials were exposed to plasma with repetitive pulses of 250 {mu}s duration, the ion energy of up to 0.6 keV, and the heat loads varying in the 0.5-25 MJ m{sup -2} range. The plasma energy transfer to the material surface versus impact load has been analysed. The fraction of plasma energy that is absorbed by the target surface is rapidly decreased with the achievement of the evaporation onset for exposed targets. The distributions of evaporated material in front of the target surface and the thickness of the shielding layer are found to be strongly dependent on the target atomic mass. The surface analysis of tungsten targets exposed to quasi-steady-state plasma accelerators plasma streams is presented together with measurements of the melting onset load and evaporation threshold, and also of erosion patterns with increasing heat load and the number of plasma pulses.

  15. Investigation of effect of post weld heat treatment conditions on residual stress for ITER blanket shield blocks

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hun-Chea, E-mail: hcjung@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, Sa-Woong [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Yun-Hee [Division of Convergence Technology, Korea Research Institute of Standard and Science (KRISS), Daejeon (Korea, Republic of); Baek, Seung-Wook [Division of Industrial Metrology, Korea Research Institute of Standard and Science (KRISS), Daejeon (Korea, Republic of); Ha, Min-Su; Shim, Hee-Jin [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of)

    2016-11-01

    Highlights: • PWHT for ITER blanket shield block should be performed for dimensional stability. • Investigation of the effect of PWHT conditions on properties was performed. • Instrumented indentation method for evaluation of properties was used. • Residual stress and hardness decreased with increasing PWHT temperature. • Optimization of PWHT conditions would be needed for satisfaction of requirement. - Abstract: The blanket shield block (SB) shall be required the tight tolerance because SB interfaces with many components, such as flexible support keypads, First Wall (FW) support contact surfaces, FW central bolt, electrical strap contact surfaces and attachment inserts for both FW and Vacuum Vessel (VV). In order to fulfil the tight tolerance requirement, stress relieving shall be performed for dimensional stability after cover welding operation. In this paper, effect of Post Weld Heat Treatment (PWHT) conditions, temperature and holding time, was investigated on the residual stress and hardness. The 316L Stainless Steel (SS) was prepared and welded by manual TIG welding by using filler material with 2.4 mm of diameter. Welded 316L SS plate was machined to prepare the specimen for PWHT. PWHT was implemented at 250, 300, 400 °C for 2 and 3 h (400 °C only) and residual stress after relaxation were determined. The evaluation of residual stress and hardness for each specimen was carried out by instrumented indentation technique. The residual stress and hardness were decreased with increasing the heat treatment temperature and holding time.

  16. Energetic ions in ITER plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Pinches, S. D. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul-lez-Durance Cedex (France); Chapman, I. T.; Sharapov, S. E. [CCFE, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Lauber, Ph. W. [Max-Planck-Institut für Plasmaphysik, EURATOM-Association, Boltzmanstraße 2, D-85748 Garching (Germany); Oliver, H. J. C. [H H Wills Physics Laboratory, University of Bristol, Royal Fort, Tyndall Avenue, Bristol BS8 1TL (United Kingdom); CCFE, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Shinohara, K. [Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan); Tani, K. [Nippon Advanced Technology Co., Ltd, Naka, Ibaraki 311-0102 (Japan)

    2015-02-15

    This paper discusses the behaviour and consequences of the expected populations of energetic ions in ITER plasmas. It begins with a careful analytic and numerical consideration of the stability of Alfvén Eigenmodes in the ITER 15 MA baseline scenario. The stability threshold is determined by balancing the energetic ion drive against the dominant damping mechanisms and it is found that only in the outer half of the plasma (r/a>0.5) can the fast ions overcome the thermal ion Landau damping. This is in spite of the reduced numbers of alpha-particles and beam ions in this region but means that any Alfvén Eigenmode-induced redistribution is not expected to influence the fusion burn process. The influence of energetic ions upon the main global MHD phenomena expected in ITER's primary operating scenarios, including sawteeth, neoclassical tearing modes and Resistive Wall Modes, is also reviewed. Fast ion losses due to the non-axisymmetric fields arising from the finite number of toroidal field coils, the inclusion of ferromagnetic inserts, the presence of test blanket modules containing ferromagnetic material, and the fields created by the Edge Localised Mode (ELM) control coils in ITER are discussed. The greatest losses and associated heat loads onto the plasma facing components arise due to the use of the ELM control coils and come from neutral beam ions that are ionised in the plasma edge.

  17. Energetic ions in ITER plasmas

    Science.gov (United States)

    Pinches, S. D.; Chapman, I. T.; Lauber, Ph. W.; Oliver, H. J. C.; Sharapov, S. E.; Shinohara, K.; Tani, K.

    2015-02-01

    This paper discusses the behaviour and consequences of the expected populations of energetic ions in ITER plasmas. It begins with a careful analytic and numerical consideration of the stability of Alfvén Eigenmodes in the ITER 15 MA baseline scenario. The stability threshold is determined by balancing the energetic ion drive against the dominant damping mechanisms and it is found that only in the outer half of the plasma ( r / a > 0.5 ) can the fast ions overcome the thermal ion Landau damping. This is in spite of the reduced numbers of alpha-particles and beam ions in this region but means that any Alfvén Eigenmode-induced redistribution is not expected to influence the fusion burn process. The influence of energetic ions upon the main global MHD phenomena expected in ITER's primary operating scenarios, including sawteeth, neoclassical tearing modes and Resistive Wall Modes, is also reviewed. Fast ion losses due to the non-axisymmetric fields arising from the finite number of toroidal field coils, the inclusion of ferromagnetic inserts, the presence of test blanket modules containing ferromagnetic material, and the fields created by the Edge Localised Mode (ELM) control coils in ITER are discussed. The greatest losses and associated heat loads onto the plasma facing components arise due to the use of the ELM control coils and come from neutral beam ions that are ionised in the plasma edge.

  18. ITER helium ash accumulation

    Energy Technology Data Exchange (ETDEWEB)

    Hogan, J.T.; Hillis, D.L.; Galambos, J.; Uckan, N.A. (Oak Ridge National Lab., TN (USA)); Dippel, K.H.; Finken, K.H. (Forschungszentrum Juelich GmbH (Germany, F.R.). Inst. fuer Plasmaphysik); Hulse, R.A.; Budny, R.V. (Princeton Univ., NJ (USA). Plasma Physics Lab.)

    1990-01-01

    Many studies have shown the importance of the ratio {upsilon}{sub He}/{upsilon}{sub E} in determining the level of He ash accumulation in future reactor systems. Results of the first tokamak He removal experiments have been analysed, and a first estimate of the ratio {upsilon}{sub He}/{upsilon}{sub E} to be expected for future reactor systems has been made. The experiments were carried out for neutral beam heated plasmas in the TEXTOR tokamak, at KFA/Julich. Helium was injected both as a short puff and continuously, and subsequently extracted with the Advanced Limiter Test-II pump limiter. The rate at which the He density decays has been determined with absolutely calibrated charge exchange spectroscopy, and compared with theoretical models, using the Multiple Impurity Species Transport (MIST) code. An analysis of energy confinement has been made with PPPL TRANSP code, to distinguish beam from thermal confinement, especially for low density cases. The ALT-II pump limiter system is found to exhaust the He with maximum exhaust efficiency (8 pumps) of {approximately}8%. We find 1<{upsilon}{sub He}/{upsilon}{sub E}<3.3 for the database of cases analysed to date. Analysis with the ITER TETRA systems code shows that these values would be adequate to achieve the required He concentration with the present ITER divertor He extraction system.

  19. The Effect of Multipole-Enhanced Diffusion on the Joule Heating of a Cold Non-Neutral Plasma

    CERN Document Server

    Chapman, Steven Francis

    One proposed technique for trapping anti-atoms is to superimpose a Ioffe-Pritchard style magnetic-minimum neutral trap on a standard Penning trap used to trap the charged atomic constituents. Adding a magnetic multipole field in this way removes the azimuthal symmetry of the ideal Penning trap and introduces a new avenue for radial diffusion. Enhanced diffusion will lead to increased Joule heating of a non-neutral plasma, potentially adversely affecting the formation rate of anti-atoms and increasing the required trap depth. We present a model of this effect, along with an approach to minimizing it, with comparison to measurements from an intended anti-atom trap.

  20. Iterating skeletons

    DEFF Research Database (Denmark)

    Dieterle, Mischa; Horstmeyer, Thomas; Berthold, Jost;

    2012-01-01

    block inside a bigger structure. In this work, we present a general framework for skeleton iteration and discuss requirements and variations of iteration control and iteration body. Skeleton iteration is expressed by synchronising a parallel iteration body skeleton with a (likewise parallel) state...

  1. Comparative analysis of core heat transport of JET high density H-mode plasmas in carbon wall and ITER-like wall

    Science.gov (United States)

    Kim, Hyun-Tae; Romanelli, M.; Voitsekhovitch, I.; Koskela, T.; Conboy, J.; Giroud, C.; Maddison, G.; Joffrin, E.; contributors, JET

    2015-06-01

    A consistent deterioration of global confinement in H-mode experiments has been observed in JET [1] following the replacement of all carbon plasma facing components (PFCs) with an all metal (‘ITER-like’) wall (ILW). This has been correlated to the observed degradation of the pedestal confinement, as lower electron temperature (Te) values are routinely measured at the top of the edge barrier region. A comparative investigation of core heat transport in JET-ILW and JET-CW (carbon wall) discharges has been performed, to assess whether core confinement has also been affected by the wall change. The results presented here have been obtained by analysing a set of discharges consisting of high density JET-ILW H-mode plasmas and comparing them against their counterpart discharges in JET-CW having similar global operational parameters. The set contains 10 baseline ({βN}=1.5∼ 2 ) discharge-pairs with 2.7 T toroidal magnetic field, 2.5 MA plasma current, and 14 to 17 MW of neutral beam injection (NBI) heating. Based on a Te profile analysis using high resolution Thomson scattering (HRTS) data, the Te profile peaking (i.e. core Te (ρ = 0.3) / edge Te (ρ = 0.7)) is found to be similar, and weakly dependent on edge Te, for both JET-ILW and JET-CW discharges. When ILW discharges are seeded with N2, core and edge Te both increase to maintain a similar peaking factor. The change in core confinement is addressed with interpretative TRANSP simulations. It is found that JET-ILW H-mode plasmas have higher NBI power deposition to electrons and lower NBI power deposition to ions as compared to the JET-CW counterparts. This is an effect of the lower electron temperature at the top of the pedestal. As a result, the core electron energy confinement time is reduced in JET-ILW discharges, but the core ion energy confinement time is not decreased. Overall, the core energy confinement is found to be the same in the JET-ILW discharges compared to the JET-CW counterparts.

  2. Power requirements for electron cyclotron current drive and ion cyclotron resonance heating for sawtooth control in ITER

    CERN Document Server

    Chapman, I T; Sauter, O; Zucca, C; Asunta, O; Buttery, R J; Coda, S; Goodman, T; Igochine, V; Johnson, T; Jucker, M; La Haye, R J; Lennholm, M; Contributors, JET-EFDA

    2013-01-01

    13MW of electron cyclotron current drive (ECCD) power deposited inside the q = 1 surface is likely to reduce the sawtooth period in ITER baseline scenario below the level empirically predicted to trigger neo-classical tearing modes (NTMs). However, since the ECCD control scheme is solely predicated upon changing the local magnetic shear, it is prudent to plan to use a complementary scheme which directly decreases the potential energy of the kink mode in order to reduce the sawtooth period. In the event that the natural sawtooth period is longer than expected, due to enhanced alpha particle stabilisation for instance, this ancillary sawtooth control can be provided from > 10MW of ion cyclotron resonance heating (ICRH) power with a resonance just inside the q = 1 surface. Both ECCD and ICRH control schemes would benefit greatly from active feedback of the deposition with respect to the rational surface. If the q = 1 surface can be maintained closer to the magnetic axis, the efficacy of ECCD and ICRH schemes sig...

  3. Heat-Induced Gel Formation by Soy Proteins at Neutral pH

    NARCIS (Netherlands)

    Renkema, J.M.S.; Vliet, van T.

    2002-01-01

    Heat-induced gel formation by soy protein isolate at pH 7 is discussed. Different heating and cooling rates, heating times, and heating temperatures were used to elucidate the various processes that occur and to study the relative role of covalent and noncovalent protein interactions therein. Gel fo

  4. Iterating skeletons

    DEFF Research Database (Denmark)

    Dieterle, Mischa; Horstmeyer, Thomas; Berthold, Jost;

    2012-01-01

    Skeleton-based programming is an area of increasing relevance with upcoming highly parallel hardware, since it substantially facilitates parallel programming and separates concerns. When parallel algorithms expressed by skeletons involve iterations – applying the same algorithm repeatedly...... block inside a bigger structure. In this work, we present a general framework for skeleton iteration and discuss requirements and variations of iteration control and iteration body. Skeleton iteration is expressed by synchronising a parallel iteration body skeleton with a (likewise parallel) state......-based iteration control, where both skeletons offer supportive type safety by dedicated types geared towards stream communication for the iteration. The skeleton iteration framework is implemented in the parallel Haskell dialect Eden. We use example applications to assess performance and overhead....

  5. Simulation experiment of interaction of plasma facing materials and transient heat loads in ITER divertor by use of magnetized coaxial plasma gun

    Science.gov (United States)

    Nakatsuka, M.; Ando, K.; Higashi, T.; Kikuchi, Y.; Fukumoto, N.; Nagata, M.

    2009-11-01

    Interaction of plasma facing materials and transient head loads such as type I ELMs is one of the critical issues in ITER divertor. The heat load to the ITER divertor during type I ELMs is estimated to be 0.5-3 MJ/m^2 with a pulse length of 0.1-0.5 ms. We have developed a magnetized coaxial plasma gun (MCPG) for the simulation experiment of transient heat load during type I ELMs in ITER divertor. The MCPG has inner and outer electrodes made of stainless steel 304. In addition, the inner electrode is covered with molybdenum so as to suppress the release of impurities from the electrode during the discharge. The diameters of inner and outer electrodes are 0.06 m and 0.14 m, respectively. The power supply for the MCPG is a capacitor bank (7 kV, 1 mF, 25 kJ). The plasma velocity estimated by the time of flight measurement of the magnetic fields was about 50 km/s, corresponding to the ion energy of 15 eV (H) or 30 eV (D). The absorbed energy density of the plasma stream was measured a calorimeter made of graphite. It was found that the absorbed energy density was 0.9 MJ/m^2 with a pulse width of 0.5 ms at the distance of 100 mm from the inner electrode. In the conference, experimental results of plasma exposure on the plasma facing materials in ITER divertor will be shown.

  6. Thermo-Mechanical Analyses of the High Heat Flux Component for ITER Dual Functional Lithium Lead Test Blanket Module

    Institute of Scientific and Technical Information of China (English)

    CHEN Hongli; BAI Yunqing

    2009-01-01

    The finite element code ANSYS is used to calculate the temperature and stress distributions for the first wall of DFLL-TBM (dual functional lithium lead-test blanket module),for testing in ITER. Preliminary analyses indicate that not only the low temperature design rules,the well-known 3Sm rules, are satisfied for the first wall, but the additional high temperature structural design criteria for the creep damage limits and creep-ratcheting limits are met as well.

  7. Effect of post weld heat treatment on the microstructure and mechanical properties of ITER-grade 316LN austenitic stainless steel weldments

    Science.gov (United States)

    Xin, Jijun; Fang, Chao; Song, Yuntao; Wei, Jing; Xu, Shen; Wu, Jiefeng

    2017-04-01

    The effect of postweld heat treatment (PWHT) on the microstructure and mechanical properties of ITER-grade 316LN austenitic stainless steel joints with ER316LMn filler material was investigated. PWHT aging was performed for 1 h at four different temperatures of 600 °C, 760 °C, 870 °C and 920 °C, respectively. The microstructure revealed the sigma phase precipitation occurred in the weld metals heat-treated at the temperature of 870 °C and 920 °C. The PWHT temperatures have the less effect on the tensile strength, and the maximum tensile strength of the joints is about 630 MPa, reaching the 95% of the base metal, whereas the elongation is enhanced with the rise of PWHT temperatures. Meanwhile, the sigma phase precipitation in the weld metals reduces the impact toughness.

  8. Summary report for ITER Task-T19: MHD pressure drop and heat transfer study for liquid metal systems

    Science.gov (United States)

    Reed, Claude B.; Hua, Thanh Q.; Natesan, Ken; Kirillov, Igor R.; Vitkovski, Ivan V.; Anisimov, Aleksandr M.

    1995-03-01

    A key feasibility issue for the ITER Vanadium/Lithium breeding blanket is the question of insulator coatings. Design calculations show that an electrically insulating layer is necessary to maintain an acceptably low MHD pressure drop. To begin experimental investigations of the MHD performance of candidate insulator materials and the technology for putting them in place, a new test section was prepared. Aluminum oxide was chosen as the first candidate insulating material because it may be used in combination with NaK in the ITER vacuum vessel and/or the divertor. Details on the methods used to produce the aluminum oxide layer as well as the microstructures of the coating and the aluminide sublayer are presented and discussed. The overall MHD pressure drop, local MHD pressure gradient, local transverse MHD pressure difference, and surface voltage distributions in both the circumferential and the axial directions are reported and discussed. The positive results obtained here for high-temperature NaK have two beneficial implications for ITER. First, since NaK may be used in the vacuum vessel and/or the divertor, these results support the design approach of using electrically insulating coatings to substantially reduce MHD pressure drop. Secondly, while Al2O3/SS is not the same coating/base material combination which would be used in the advanced blanket, this work nonetheless shows that it is possible to produce a viable insulating coating which is stable in contact with a high temperature alkali metal coolant.

  9. Detailed 3-D nuclear analysis of ITER outboard blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Bohm, Tim, E-mail: tdbohm@wisc.edu [Fusion Technology Institute, University of Wisconsin-Madison, Madison, WI (United States); Davis, Andrew; Sawan, Mohamed; Marriott, Edward; Wilson, Paul [Fusion Technology Institute, University of Wisconsin-Madison, Madison, WI (United States); Ulrickson, Michael; Bullock, James [Formerly, Fusion Technology, Sandia National Laboratories, Albuquerque, NM (United States)

    2015-10-15

    Highlights: • Nuclear analysis was performed on detailed CAD models placed in a 40 degree model of ITER. • The regions examined include BM09, the upper ELM coil region (BM11–13), the neutral beam (NB) region (BM13–16), and BM18. • The results show that VV nuclear heating exceeds limits in the NB and upper ELM coil regions. • The results also show that the level of He production in parts of BM18 exceeds limits. • These calculations are being used to modify the design of the ITER blanket modules. - Abstract: In the ITER design, the blanket modules (BM) provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40 degree partially homogenized ITER global model. The regions analyzed include BM09, BM16 near the heating neutral beam injection (HNB) region, BM11–13 near the upper ELM coil region, and BM18. For the BM16 HNB region, the VV nuclear heating behind the NB region exceeds the design limit by up to 80%. For the BM11–13 region, the nuclear heating of the VV exceeds the design limit by up to 45%. For BM18, the results show that He production does not meet the limit necessary for re-welding. The results presented in this work are being used by the ITER Organization Blanket and Tokamak Integration groups to modify the BM design in the cases where limits are exceeded.

  10. The effects of neutral gas heating on H mode transition and maintenance currents in a 13.56 MHz planar coil inductively coupled plasma reactor

    Science.gov (United States)

    Jayapalan, Kanesh K.; Chin, Oi-Hoong

    2012-09-01

    The H mode transition and maintenance currents in a 13.56 MHz laboratory 6 turn planar coil inductively coupled plasma (ICP) reactor are simulated for low pressure argon discharge range of 0.02-0.3 mbar with neutral gas heating and at ambient temperature. An experimentally fitted 3D power evolution plot for 0.02 mbar argon pressure is also shown to visualize the effects of hysteresis in the system. Comparisons between simulation and experimental measurements show good agreement in the pressure range of 0.02-0.3 mbar for transition currents and 0.02-0.1 mbar for maintenance currents only when neutral gas heating is considered. This suggests that neutral gas heating plays a non-negligible role in determining the mode transition points of a rf ICP system.

  11. Momentum, Heat, and Neutral Mass Transport in Convective Atmospheric Pressure Plasma-Liquid Systems and Implications for Aqueous Targets

    CERN Document Server

    Lindsay, Alexander; Slikboer, Elmar; Shannon, Steven; Graves, David

    2015-01-01

    There is a growing interest in the study of plasma-liquid interactions with application to biomedicine, chemical disinfection, agriculture, and other fields. This work models the momentum, heat, and neutral species mass transfer between gas and aqueous phases in the context of a streamer discharge; the qualitative conclusions are generally applicable to plasma-liquid systems. The problem domain is discretized using the finite element method. The most interesting and relevant model result for application purposes is the steep gradients in reactive species at the interface. At the center of where the reactive gas stream impinges on the water surface, the aqueous concentrations of OH and ONOOH decrease by roughly 9 and 4 orders of magnitude respectively within 50 $\\mu$m of the interface. Recognizing the limited penetration of reactive plasma species into the aqueous phase is critical to discussions about the therapeutic mechanisms for direct plasma treatment of biological solutions. Other interesting results fro...

  12. Heating and cooling of the neutral ISM in the NGC4736 circumnuclear ring

    CERN Document Server

    van der Laan, T P R; Beirao, P; Sandstrom, K; Groves, B; Schinnerer, E; Draine, B T; Smith, J D; Galametz, M; Wolfire, M; Croxall, K; Dale, D; Camus, R Herrera; Calzetti, D; Kennicutt, R C

    2015-01-01

    The manner in which gas accretes and orbits within circumnuclear rings has direct implications for the star formation process. In particular, gas may be compressed and shocked at the inflow points, resulting in bursts of star formation at these locations. Afterwards the gas and young stars move together through the ring. In addition, star formation may occur throughout the ring, if and when the gas reaches sufficient density to collapse under gravity. These two scenarios for star formation in rings are often referred to as the `pearls on a string' and `popcorn' paradigms. In this paper, we use new Herschel PACS observations, obtained as part of the KINGFISH Open Time Key Program, along with archival Spitzer and ground-based observations from the SINGS Legacy project, to investigate the heating and cooling of the interstellar medium in the nearby star-forming ring galaxy, NGC4736. By comparing spatially resolved estimates of the stellar FUV flux available for heating, with the gas and dust cooling derived from...

  13. Detailed design optimization of the MITICA negative ion accelerator in view of the ITER NBI

    Science.gov (United States)

    Agostinetti, P.; Aprile, D.; Antoni, V.; Cavenago, M.; Chitarin, G.; de Esch, H. P. L.; De Lorenzi, A.; Fonnesu, N.; Gambetta, G.; Hemsworth, R. S.; Kashiwagi, M.; Marconato, N.; Marcuzzi, D.; Pilan, N.; Sartori, E.; Serianni, G.; Singh, M.; Sonato, P.; Spada, E.; Toigo, V.; Veltri, P.; Zaccaria, P.

    2016-01-01

    The ITER Neutral Beam Test Facility (PRIMA) is presently under construction at Consorzio RFX (Padova, Italy). PRIMA includes two experimental devices: an ITER-size ion source with low voltage extraction, called SPIDER, and the full prototype of the whole ITER Heating Neutral Beams (HNBs), called MITICA. The purpose of MITICA is to demonstrate that all operational parameters of the ITER HNB accelerator can be experimentally achieved, thus establishing a large step forward in the performances of neutral beam injectors in comparison with the present experimental devices. The design of the MITICA extractor and accelerator grids, here described in detail, was developed using an integrated approach, taking into consideration at the same time all the relevant physics and engineering aspects. Particular care was taken also to support and validate the design on the basis of the expertise and experimental data made available by the collaborating neutral beam laboratories of CEA, IPP, CCFE, NIFS and JAEA. Considering the operational requirements and the other physics constraints of the ITER HNBs, the whole design has been thoroughly optimized and improved. Furthermore, specific innovative concepts have been introduced.

  14. CO{sub 2}-neutral heat supply to residential areas. Pt. 1. Energy demand; CO{sub 2}-neutrale Waermeversorgung fuer Wohnsiedlungen. T. 1. Energiebedarf

    Energy Technology Data Exchange (ETDEWEB)

    Kuehl, L.; Schlosser, M.; Heuer, M.; Fisch, M.N. [Technische Univ. Braunschweig (Germany). Inst. fuer Gebaeude- und Solartechnik

    2006-11-15

    The contribution shows how integration of selected measures in heat supply (better thermal insulation, biomass fuel and solar technology) will result in an optimized heat supply system for the future which is characterized by low consumption and low emissions. (orig.)

  15. An Iterative Method for Solving of Coupled Equations for Conductive-Radiative Heat Transfer in Dielectric Layers

    National Research Council Canada - National Science Library

    2017-01-01

    The mathematical model for describing combined conductive-radiative heat transfer in a dielectric layer, which emits, absorbs, and scatters IR radiation both in its volume and on the boundary, has been considered...

  16. Evaluation of ITER MSE Viewing Optics

    Energy Technology Data Exchange (ETDEWEB)

    Allen, S; Lerner, S; Morris, K; Jayakumar, J; Holcomb, C; Makowski, M; Latkowski, J; Chipman, R

    2007-03-26

    The Motional Stark Effect (MSE) diagnostic on ITER determines the local plasma current density by measuring the polarization angle of light resulting from the interaction of a high energy neutral heating beam and the tokamak plasma. This light signal has to be transmitted from the edge and core of the plasma to a polarization analyzer located in the port plug. The optical system should either preserve the polarization information, or it should be possible to reliably calibrate any changes induced by the optics. This LLNL Work for Others project for the US ITER Project Office (USIPO) is focused on the design of the viewing optics for both the edge and core MSE systems. Several design constraints were considered, including: image quality, lack of polarization aberrations, ease of construction and cost of mirrors, neutron shielding, and geometric layout in the equatorial port plugs. The edge MSE optics are located in ITER equatorial port 3 and view Heating Beam 5, and the core system is located in equatorial port 1 viewing heating beam 4. The current work is an extension of previous preliminary design work completed by the ITER central team (ITER resources were not available to complete a detailed optimization of this system, and then the MSE was assigned to the US). The optimization of the optical systems at this level was done with the ZEMAX optical ray tracing code. The final LLNL designs decreased the ''blur'' in the optical system by nearly an order of magnitude, and the polarization blur was reduced by a factor of 3. The mirror sizes were reduced with an estimated cost savings of a factor of 3. The throughput of the system was greater than or equal to the previous ITER design. It was found that optical ray tracing was necessary to accurately measure the throughput. Metal mirrors, while they can introduce polarization aberrations, were used close to the plasma because of the anticipated high heat, particle, and neutron loads. These mirrors

  17. Comparison of collective Thomson scattering signals due to fast ions in ITER scenarios with fusion and auxiliary heating

    DEFF Research Database (Denmark)

    Salewski, Mirko; Asunta, O.; Eriksson, L.-G.

    2009-01-01

    to the alpha population in these frequency ranges. The exceptions are limited regions in space with some non-negligible signal due to beam ions or fast He-3 which give rise to about 30% and 10-20% of the CTS signal, respectively. In turn, the dominance of the alpha contribution implies that the effects...... scattering (CTS) signal for the proposed CTS diagnostic in ITER. It is of interest to determine the contributions of these fast ion populations to the CTS signal for large Doppler shifts of the scattered radiation since conclusions can mostly be drawn for the dominant contributor. In this study, distribution...... functions for fast deuterons, fast tritons, fast He-3 and the fusion born alphas are presented, revealing that fusion alphas dominate the measurable signal by an order of magnitude or more in the Doppler shift frequency ranges typical for fast ions. Hence the observable CTS signal can mostly be attributed...

  18. Validation of a new mixed Bohm/gyro-Bohm model for electron and ion heat transport against ITER database, tore supra and start discharges

    Energy Technology Data Exchange (ETDEWEB)

    Erba, M.; Aniel, T.; Basiuk, V.; Becoulet, A.; Litaudon, X

    1997-08-01

    A new model based on a combination of a Bohm-like term plus a gyro-Bohm-like term is proposed for the electron and ion heat diffusivity in the L-mode regime, which is the commonest regime of operation of Tokamaks. This model is derived using the dimensionless analysis technique taking into account the indications of scaling laws for the global confinement time and other experimental constraints on the diffusivity. The model has been successfully tested against data from several different experiments from the ITER database and the local Tore Supra data-base. Statistical analysis has shown it to perform better than purely Bohm or gyro-Bohm models and global scaling laws in the chosen dataset. (author) 36 refs.

  19. Numerical investigation on transverse heat transfer properties in cross section of full size Nb3Sn CICC ITER conductor

    Directory of Open Access Journals (Sweden)

    Shuming Jia

    2015-05-01

    Full Text Available The contact mechanical characteristics in the cross section of the Nb3Sn cable are sensitive to the cryogenic cooling and cyclic transverse electromagnetic loads, which may affect the cable’s performance. In this paper, based on a proposed discrete dynamic model (DEM, where the contact heat transfer among strands and the convective heat transfer in liquid helium are taken into account, the cooling process under two heat transfer mechanisms is performed. Simulation results show that the temperature variation of Poloidal Field Insert Sample (PFIS cable with time agrees well with the existing experimental results, and the role of contact heat transfer cannot be neglected during cryogenic cooling. It is obtained from the further analysis that the effect of contact heat transfer becomes more prominent with the decrease of mass flow rate of liquid helium, which leads to the stress status within cable changed significantly. With the temperature boundary condition imposed on the cable radial direction, the effective thermal conductivity (ETC of cable can be obtained. It can be found that the ETC increases with increasing the transverse loads and is sensitive to the low temperature environment, while it is not affected by load cycles basically. These results may provide the guide for the design and application of the future CICC conductors.

  20. Benchmarking of electron cyclotron heating and current drive codes on ITER scenarios within the European Integrated Tokamak Modelling framework

    Directory of Open Access Journals (Sweden)

    Peysson Y.

    2012-09-01

    Full Text Available Electron cyclotron resonance heating (ECRH and electron cyclotron current drive (ECCD are used to heat the plasma, to tailor the current profiles and to achieve different operating regimes of tokamak plasmas. Plasmas with ECRH/ECCD are characterized by non-thermal electrons, which cannot be described by a Maxwellian distribution. Non-thermal electrons are also generated during MHD activity, like sawteeth crashes. Quantifying the non-thermal electron distribution is therefore a key for understanding EC heated fusion plasmas. For this purpose a vertical electron cyclotron emission (V-ECE diagnostic is being installed at TCV. The diagnostic layout, the calibration, the analysis technique for data interpretation, the physics potentials and limitations are discussed.

  1. High heat flux components in fusion devices: from nowadays experience in Tore Supra towards the ITER challenge

    Energy Technology Data Exchange (ETDEWEB)

    Grosman, A.; Bayetti, P.; Chappuis, P.; Cordier, J.J.; Durocher, A.; Escourbiac, F.; Guilhem, D.; Lipa, M.; Marbach, G.; Mitteau, R.; Schlosser, J

    2003-07-01

    A pioneering activity has been developed by CEA and the European industry in the field of actively cooled high heat flux plasma facing components in Tore Supra operation, which is today culminating with the routine operation of an actively cooled toroidal pumped limiter (TPL) capable to sustain up to 10 MW.m{sup -2} of nominal convected heat flux. This success is the result of a long lead development and industrialization program (about 10 years) marked out with a number of technical and managerial challenges that were taken up and has allowed us to build up an unique experience feedback database. This is illustrated in this paper with the specific example of the development of high heat flux CFC-on-CuCrZr (carbon-carbon fibre composite on hardened copper alloy CuCrZr) component from design phase to tokamak operation. (authors)

  2. Active spectroscopic measurements using the ITER diagnostic system.

    Science.gov (United States)

    Thomas, D M; Counsell, G; Johnson, D; Vasu, P; Zvonkov, A

    2010-10-01

    Active (beam-based) spectroscopic measurements are intended to provide a number of crucial parameters for the ITER device being built in Cadarache, France. These measurements include the determination of impurity ion temperatures, absolute densities, and velocity profiles, as well as the determination of the plasma current density profile. Because ITER will be the first experiment to study long timescale (∼1 h) fusion burn plasmas, of particular interest is the ability to study the profile of the thermalized helium ash resulting from the slowing down and confinement of the fusion alphas. These measurements will utilize both the 1 MeV heating neutral beams and a dedicated 100 keV hydrogen diagnostic neutral beam. A number of separate instruments are being designed and built by several of the ITER partners to meet the different spectroscopic measurement needs and to provide the maximum physics information. In this paper, we describe the planned measurements, the intended diagnostic ensemble, and we will discuss specific physics and engineering challenges for these measurements in ITER.

  3. Momentum, heat, and neutral mass transport in convective atmospheric pressure plasma-liquid systems and implications for aqueous targets

    Science.gov (United States)

    Lindsay, Alexander; Anderson, Carly; Slikboer, Elmar; Shannon, Steven; Graves, David

    2015-10-01

    There is a growing interest in the study of plasma-liquid interactions with application to biomedicine, chemical disinfection, agriculture, and other fields. This work models the momentum, heat, and neutral species mass transfer between gas and aqueous phases in the context of a streamer discharge; the qualitative conclusions are generally applicable to plasma-liquid systems. The problem domain is discretized using the finite element method. The most interesting and relevant model result for application purposes is the steep gradients in reactive species at the interface. At the center of where the reactive gas stream impinges on the water surface, the aqueous concentrations of OH and ONOOH decrease by roughly 9 and 4 orders of magnitude respectively within 50 μ m of the interface. Recognizing the limited penetration of reactive plasma species into the aqueous phase is critical to discussions about the therapeutic mechanisms for direct plasma treatment of biological solutions. Other interesting results from this study include the presence of a 10 K temperature drop in the gas boundary layer adjacent to the interface that arises from convective cooling. Though the temperature magnitudes may vary among atmospheric discharge types (different amounts of plasma-gas heating), this relative difference between gas and liquid bulk temperatures is expected to be present for any system in which convection is significant. Accounting for the resulting difference between gas and liquid bulk temperatures has a significant impact on reaction kinetics; factor of two changes in terminal aqueous species concentrations like H2O2, NO2- , and NO3- are observed in this study if the effect of evaporative cooling is not included.

  4. Axial heating and temperature of RF-excited non-neutral plasmas in Penning-Malmberg traps

    Science.gov (United States)

    Maero, G.; Pozzoli, R.; Romé, M.; Chen, S.; Ikram, M.

    2016-09-01

    Electro-magnetostatic traps have been used for decades to provide long-term storage of charged particle samples or non-neutral plasmas. The dynamics and equilibrium states of these ideally simple systems can be strongly diverted from the usual working conditions (i.e. single-species, quiescent samples) in the presence of oppositely charged particles or external electric field perturbations. Both these conditions occur when the plasma is generated by means of a radio-frequency (RF) excitation continuously applied on a trap electrode. The application of RF drives of some volts over periods larger than typical collisional time scales leads to residual-gas ionization and to the accumulation of an electron plasma, a process that has previously been exploited as an alternative to thermionic or photoemission electron sources. The analysis of the axial energy distribution shows a deviation of the continuously excited final state from maxwellianity dependent on the radial position and the subsequent relaxation to equilibrium after the interruption of the drive. Systematic measurements also indicate the high sensitivity to the residual gas pressure of both the total confined charge and of the attainable densities and plasma profiles. The results are compared to the information obtained from a very simple one-dimensional electron heating model and show the validity of its most basic features together with its shortcomings.

  5. Improvement of neutral beam injection heating efficiency with magnetic field well structures in a tokamak with a low magnetic field

    Science.gov (United States)

    Kim, S. K.; Na, D. H.; Lee, J. W.; Yoo, M. G.; Kim, H.-S.; Hwang, Y. S.; Hahm, T. S.; Na, Yong-Su

    2016-10-01

    Magnetic well structures are introduced as an effective means to reduce the prompt loss of fast ions, the so-called first orbit loss from neutral beam injection (NBI), which is beneficial to tokamaks with a low magnetic field strength such as small spherical torus devices. It is found by single-particle analysis that this additional field structure can modify the gradient of the magnetic field to reduce the shift of the guiding center trajectory of the fast ion. This result is verified by a numerical calculation of following the fast ion’s trajectory. We apply this concept to the Versatile Experiment Spherical Torus [1], where NBI is under design for the purpose of achieving high-performance plasma, to evaluate the effect of the magnetic well structure on NBI efficiency. A 1D NBI analysis code and the NUBEAM code are employed for detailed NBI calculations. The simulation results show that the orbit loss can be reduced by 70%-80%, thereby improving the beam efficiency twofold compared with the reference case without the well structure. The well-shaped magnetic field structure in the low-field side can significantly decrease orbit loss by broadening the non-orbit loss region and widening the range of the velocity direction, thus improving the heating efficiency. It is found that this magnetic well can also improve orbit loss during the slowing down process.

  6. Modeling the inflammatory response in the hypothalamus ensuing heat stroke: iterative cycle of model calibration, identifiability analysis, experimental design and data collection.

    Science.gov (United States)

    Klett, Hagen; Rodriguez-Fernandez, Maria; Dineen, Shauna; Leon, Lisa R; Timmer, Jens; Doyle, Francis J

    2015-02-01

    Heat Stroke (HS) is a life-threatening illness caused by prolonged exposure to heat that causes severe hyperthermia and nervous system abnormalities. The long term consequences of HS are poorly understood and deeper insight is required to find possible treatment strategies. Elevated pro- and anti-inflammatory cytokines during HS recovery suggest to play a major role in the immune response. In this study, we developed a mathematical model to understand the interactions and dynamics of cytokines in the hypothalamus, the main thermoregulatory center in the brain. Uncertainty and identifiability analysis of the calibrated model parameters revealed non-identifiable parameters due to the limited amount of data. To overcome the lack of identifiability of the parameters, an iterative cycle of optimal experimental design, data collection, re-calibration and model reduction was applied and further informative experiments were suggested. Additionally, a new method of approximating the prior distribution of the parameters for Bayesian optimal experimental design based on the profile likelihood is presented.

  7. An investigation of pulsed phase thermography for detection of disbonds in HIP-bonded beryllium tiles in ITER normal heat flux first wall (NHF FW) components

    Energy Technology Data Exchange (ETDEWEB)

    Bushell, J., E-mail: joe.bushell@amec.com [AMEC Foster Wheeler, Booths Hall, Chelford Road, Knutsford, Cheshire WA16 8QZ, England (United Kingdom); Sherlock, P. [AMEC Foster Wheeler, Booths Hall, Chelford Road, Knutsford, Cheshire WA16 8QZ, England (United Kingdom); Mummery, P. [School of Mechanical, Aerospace and Civil Engineering, University of Manchester, England (United Kingdom); Bellin, B.; Zacchia, F. [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, Barcelona (Spain)

    2015-10-15

    Highlights: • Pulsed phase thermography was trialled on Be-tiled plasma facing components. • Two components, one with known disbonds, one intact, were inspected and compared. • Finite element analysis was used to verify experimental observations. • PPT successfully detected disbonds in the failed component. • Good agreement found with ultrasonic test, though defect geometry was uncertain. - Abstract: Pulsed phase thermography (PPT) is a non destructive examination (NDE) technique, traditionally used in the Aerospace Industry for inspection of composite structures, which combines characteristics and benefits of flash thermography and lock-in thermography into a single, rapid inspection technique. The aim of this work was to evaluate the effectiveness of PPT as a means of inspection for the bond between the beryllium (Be) tiles and the copper alloy (CuCrZr) heatsink of the ITER NHF FW components. This is a critical area dictating the functional integrity of these components, as single tile detachment in service could result in cascade failure. PPT has advantages over existing thermography techniques using heated water which stress the component, and the non-invasive, non-contact nature presents advantages over existing ultrasonic methods. The rapid and non-contact nature of PPT also gives potential for in-service inspections as well as a quality measure for as-manufactured components. The technique has been appraised via experimental trials using ITER first wall mockups with pre-existing disbonds confirmed via ultrasonic tests, partnered with finite element simulations to verify experimental observations. This paper will present the results of the investigation.

  8. European programme towards the 1 MeV ITER NB injector

    Energy Technology Data Exchange (ETDEWEB)

    Masiello, A. [Fusion for Energy, C/Josep Pla 2, 08019 Barcelona (Spain)], E-mail: antonio.masiello@f4e.europa.eu; Agarici, G.; Bonicelli, T.; Simon, M. [Fusion for Energy, C/Josep Pla 2, 08019 Barcelona (Spain); Antoni, V. [Consorzio RFX, Euratom-ENEA Association, C.so Stati Uniti 4, I-35126, Padova (Italy); De Esch, H. [Association EURATOM, CEA Cadarache, IRFM/SCCP, 13108 St. Paul Lez Durance (France); De Lorenzi, A. [Consorzio RFX, Euratom-ENEA Association, C.so Stati Uniti 4, I-35126, Padova (Italy); Dremel, M. [Forschungszentrum Karlsruhe, EURATOM Association, 76021 Karlsruhe (Germany); Franzen, P. [Max-Planck-Institut fur Plasmaphysik, EURATOM Association, D-85740, Garching (Germany); Hemsworth, R. [ITER, ITER Joint Work Site, CEA Cadarache, 13108 St. Paul Lez Durance (France); Liniers, M. [Association EURATOM-CIEMAT, Av. Complutense 22, 28040, Madrid (Spain); Marcuzzi, D. [Consorzio RFX, Euratom-ENEA Association, C.so Stati Uniti 4, I-35126, Padova (Italy); Martin, D. [UKAEA Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Piovan, R. [Consorzio RFX, Euratom-ENEA Association, C.so Stati Uniti 4, I-35126, Padova (Italy); Simonin, A. [Association EURATOM, CEA Cadarache, IRFM/SCCP, 13108 St. Paul Lez Durance (France); Sonato, P. [Consorzio RFX, Euratom-ENEA Association, C.so Stati Uniti 4, I-35126, Padova (Italy); Surrey, E. [UKAEA Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Svensson, L. [Association EURATOM, CEA Cadarache, IRFM/SCCP, 13108 St. Paul Lez Durance (France); Tanga, A. [ITER, ITER Joint Work Site, CEA Cadarache, 13108 St. Paul Lez Durance (France); Toigo, V. [Consorzio RFX, Euratom-ENEA Association, C.so Stati Uniti 4, I-35126, Padova (Italy)] (and others)

    2009-06-15

    The ITER neutral beam (NB) system presents several challenges and a robust program is necessary in order to achieve the requirements within the tight constraints due to the ITER construction plan. The establishment of full scale NB test facilities (NBTF) has therefore become a centre piece of the international NB development strategy. This paper describes the status of the design activities that have been undertaken in Europe to develop the components for the heating NB injectors along with the main plans and results of the R and D activities. A description of the programme towards the establishment of the test facilities and the planned activities is also reported.

  9. Transmission of electrons inside the cryogenic pumps of ITER injector

    Energy Technology Data Exchange (ETDEWEB)

    Veltri, P., E-mail: pierluigi.veltri@igi.cnr.it; Sartori, E. [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), Corso Stati Uniti 4, 35127 Padova (Italy)

    2016-02-15

    Large cryogenic pumps are installed in the vessel of large neutral beam injectors (NBIs) used to heat the plasma in nuclear fusion experiments. The operation of such pumps can be compromised by the presence of stray secondary electrons that are generated along the beam path. In this paper, we present a numerical model to analyze the propagation of the electrons inside the pump. The aim of the study is to quantify the power load on the active pump elements, via evaluation of the transmission probabilities across the domain of the pump. These are obtained starting from large datasets of particle trajectories, obtained by numerical means. The transmission probability of the electrons across the domain is calculated for the NBI of the ITER and for its prototype Megavolt ITer Injector and Concept Advancement (MITICA) and the results are discussed.

  10. Transmission of electrons inside the cryogenic pumps of ITER injector.

    Science.gov (United States)

    Veltri, P; Sartori, E

    2016-02-01

    Large cryogenic pumps are installed in the vessel of large neutral beam injectors (NBIs) used to heat the plasma in nuclear fusion experiments. The operation of such pumps can be compromised by the presence of stray secondary electrons that are generated along the beam path. In this paper, we present a numerical model to analyze the propagation of the electrons inside the pump. The aim of the study is to quantify the power load on the active pump elements, via evaluation of the transmission probabilities across the domain of the pump. These are obtained starting from large datasets of particle trajectories, obtained by numerical means. The transmission probability of the electrons across the domain is calculated for the NBI of the ITER and for its prototype Megavolt ITer Injector and Concept Advancement (MITICA) and the results are discussed.

  11. Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

    Energy Technology Data Exchange (ETDEWEB)

    Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N. [Institution Project center ITER, Moscow (Russian Federation)

    2014-08-21

    In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ–ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed.

  12. Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

    Science.gov (United States)

    Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N.

    2014-08-01

    In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ-ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed.

  13. Detailed 3-D nuclear analysis of ITER blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Bohm, T.D., E-mail: tdbohm@wisc.edu [University of Wisconsin-Madison, Madison, WI (United States); Sawan, M.E.; Marriott, E.P.; Wilson, P.P.H. [University of Wisconsin-Madison, Madison, WI (United States); Ulrickson, M.; Bullock, J. [Sandia National Laboratories, Albuquerque, NM (United States)

    2014-10-15

    In ITER, the blanket modules (BM) are arranged around the plasma to provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. As a part of the BM design process, nuclear analysis is required to determine the level of nuclear heating, helium production, and radiation damage in the BM. Additionally, nuclear heating in the VV is also important for assessing the BM design. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40-degree partially homogenized ITER global model. The regions analyzed include BM01, the neutral beam injection (NB) region, and the upper port region. For BM01, the results show that He production meets the limit necessary for re-welding, and the VV heating behind BM01 is acceptable. For the NBI region, the VV nuclear heating behind the NB region exceeds the design limit by a factor of two. For the upper port region, the nuclear heating of the VV exceeds the design limit by up to 20%. The results presented in this work are being used to modify the BM design in the cases where limits are exceeded.

  14. Effect of neutral collision and radiative heat-loss function on self-gravitational instability of viscous thermally conducting partially-ionized plasma

    Directory of Open Access Journals (Sweden)

    Sachin Kaothekar

    2012-12-01

    Full Text Available The problem of thermal instability and gravitational instability is investigated for a partially ionized self-gravitating plasma which has connection in astrophysical condensations. We use normal mode analysis method in this problem. The general dispersion relation is derived using linearized perturbation equations of the problem. Effects of collisions with neutrals, radiative heat-loss function, viscosity, thermal conductivity and magnetic field strength, on the instability of the system are discussed. The conditions of instability are derived for a temperature-dependent and density-dependent heat-loss function with thermal conductivity. Numerical calculations have been performed to discuss the effect of various physical parameters on the growth rate of the gravitational instability. The temperature-dependent heat-loss function, thermal conductivity, viscosity, magnetic field and neutral collision have stabilizing effect, while density-dependent heat-loss function has a destabilizing effect on the growth rate of the gravitational instability. With the help of Routh-Hurwitz's criterion, the stability of the system is discussed.

  15. Some aspects of the design of the ITER NBI Active Correction and Compensation Coils

    Energy Technology Data Exchange (ETDEWEB)

    Alonso, Javier, E-mail: javier.alonso@ciemat.es [CIEMAT, Laboratorio Nacional de Fusión, Avda. Complutense 40, 28040 Madrid (Spain); Barrera, Germán; Cabrera, Santiago; Rincón, Esther; Ríos, Luis; Soleto, Alfonso [CIEMAT, Laboratorio Nacional de Fusión, Avda. Complutense 40, 28040 Madrid (Spain); El-Ouazzani, Anass; Graceffa, Joseph; Shah, Darshan; Urbani, Marc [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Agarici, Gilbert [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3 – 07/08, 08019 Barcelona (Spain)

    2015-10-15

    Highlights: • Water cooled coil design. • Magnetic shielding of the plasma heating Neutral Beam Injection System. • Active coils for magnetic field compensation. - Abstract: The neutral beam system for ITER consists of two heating and current drive injectors plus a diagnostic neutral beam injector. The proposed physical plant layout allows for a possible third heating injector to be installed later. For correct operation of the beam source, and to avoid deflections of the charged fraction of the beam, the magnetic field along the beam path must be very low. To minimize the stray ITER field in critical areas (ion source, acceleration grids, neutralizer, residual ion dump), a Magnetic Field Reduction System will envelop the beam vessels and the high voltage transmission lines to ion source. This whole system comprises the Passive Magnetic Shield, a set of thick steel plates, and the Active Correction and Compensation Coils, a set of coils carrying currents which depend on the tokamak stray field. This paper describes the status of the coil design, terminals and support structures, as well as a description of the calculations carried out. Most coils are suitable for removal from their final position to be replaced in case of a fault. Conclusions of the chosen design highlight the strategy for the system feasibility.

  16. Feasibility of non-thermal helium measurements with charge exchange spectroscopy on ITER

    Science.gov (United States)

    Kappatou, A.; Delabie, E.; Jaspers, R. J. E.; von Hellermann, M. G.

    2012-04-01

    The use of active charge exchange recombination spectroscopy (CXRS) as a diagnostic for fusion-produced alpha particles on ITER is constrained by the signal-to-noise ratio, which is determined by the intensity of the line of interest, the optical throughput of the diagnostic, the neutral beam penetration, and the intensity of bremsstrahlung radiation. The CX spectral line for fast ions has been modelled together with the expected background emission and we present the signal-to-noise ratios calculated as a function of the diagnostic design parameters. Combining the CXRS data from both the heating and the diagnostic neutral beams on ITER, information on fast ions with energies up to 1 MeV can be obtained for the parameters of the ITER core CXRS diagnostic design. To achieve this, energy binning of the signal is used (100 keV bins or larger), in order to improve the signal-to-noise ratio, with a time resolution of 2 s. The time resolution of the measurement can be improved using a higher throughput spectrometer, but this is ultimately limited by the amount of light from the neutral beam that can be collected. Despite the challenges and the fact that the results are not as optimistic as previously assumed, it is concluded that useful information on fast helium density profiles can be obtained using CXRS on ITER.

  17. Conceptual design of the DEMO neutral beam injectors: main developments and R&D achievements

    Science.gov (United States)

    Sonato, P.; Agostinetti, P.; Bolzonella, T.; Cismondi, F.; Fantz, U.; Fassina, A.; Franke, T.; Furno, I.; Hopf, C.; Jenkins, I.; Sartori, E.; Tran, M. Q.; Varje, J.; Vincenzi, P.; Zanotto, L.

    2017-05-01

    The objectives of the nuclear fusion power plant DEMO, to be built after the ITER experimental reactor, are usually understood to lie somewhere between those of ITER and a ‘first of a kind’ commercial plant. Hence, in DEMO the issues related to efficiency and RAMI (reliability, availability, maintainability and inspectability) are among the most important drivers for the design, as the cost of the electricity produced by this power plant will strongly depend on these aspects. In the framework of the EUROfusion Work Package Heating and Current Drive within the Power Plant Physics and Development activities, a conceptual design of the neutral beam injector (NBI) for the DEMO fusion reactor has been developed by Consorzio RFX in collaboration with other European research institutes. In order to improve efficiency and RAMI aspects, several innovative solutions have been introduced in comparison to the ITER NBI, mainly regarding the beam source, neutralizer and vacuum pumping systems.

  18. Progress in the realization of the PRIMA neutral beam test facility

    Science.gov (United States)

    Toigo, V.; Boilson, D.; Bonicelli, T.; Piovan, R.; Hanada, M.; Chakraborty, A.; Agarici, G.; Antoni, V.; Baruah, U.; Bigi, M.; Chitarin, G.; Dal Bello, S.; Decamps, H.; Graceffa, J.; Kashiwagi, M.; Hemsworth, R.; Luchetta, A.; Marcuzzi, D.; Masiello, A.; Paolucci, F.; Pasqualotto, R.; Patel, H.; Pomaro, N.; Rotti, C.; Serianni, G.; Simon, M.; Singh, M.; Singh, N. P.; Svensson, L.; Tobari, H.; Watanabe, K.; Zaccaria, P.; Agostinetti, P.; Agostini, M.; Andreani, R.; Aprile, D.; Bandyopadhyay, M.; Barbisan, M.; Battistella, M.; Bettini, P.; Blatchford, P.; Boldrin, M.; Bonomo, F.; Bragulat, E.; Brombin, M.; Cavenago, M.; Chuilon, B.; Coniglio, A.; Croci, G.; Dalla Palma, M.; D'Arienzo, M.; Dave, R.; De Esch, H. P. L.; De Lorenzi, A.; De Muri, M.; Delogu, R.; Dhola, H.; Fantz, U.; Fellin, F.; Fellin, L.; Ferro, A.; Fiorentin, A.; Fonnesu, N.; Franzen, P.; Fröschle, M.; Gaio, E.; Gambetta, G.; Gomez, G.; Gnesotto, F.; Gorini, G.; Grando, L.; Gupta, V.; Gutierrez, D.; Hanke, S.; Hardie, C.; Heinemann, B.; Kojima, A.; Kraus, W.; Maeshima, T.; Maistrello, A.; Manduchi, G.; Marconato, N.; Mico, G.; Moreno, J. F.; Moresco, M.; Muraro, A.; Muvvala, V.; Nocentini, R.; Ocello, E.; Ochoa, S.; Parmar, D.; Patel, A.; Pavei, M.; Peruzzo, S.; Pilan, N.; Pilard, V.; Recchia, M.; Riedl, R.; Rizzolo, A.; Roopesh, G.; Rostagni, G.; Sandri, S.; Sartori, E.; Sonato, P.; Sottocornola, A.; Spagnolo, S.; Spolaore, M.; Taliercio, C.; Tardocchi, M.; Thakkar, A.; Umeda, N.; Valente, M.; Veltri, P.; Yadav, A.; Yamanaka, H.; Zamengo, A.; Zaniol, B.; Zanotto, L.; Zaupa, M.

    2015-08-01

    The ITER project requires additional heating by two neutral beam injectors, each accelerating to 1 MV a 40 A beam of negative deuterium ions, to deliver to the plasma a power of about 17 MW for one hour. As these requirements have never been experimentally met, it was recognized as necessary to setup a test facility, PRIMA (Padova Research on ITER Megavolt Accelerator), in Italy, including a full-size negative ion source, SPIDER, and a prototype of the whole ITER injector, MITICA, aiming to develop the heating injectors to be installed in ITER. This realization is made with the main contribution of the European Union, through the Joint Undertaking for ITER (F4E), the ITER Organization and Consorzio RFX which hosts the Test Facility. The Japanese and the Indian ITER Domestic Agencies (JADA and INDA) participate in the PRIMA enterprise; European laboratories, such as IPP-Garching, KIT-Karlsruhe, CCFE-Culham, CEA-Cadarache and others are also cooperating. Presently, the assembly of SPIDER is on-going and the MITICA design is being completed. The paper gives a general overview of the test facility and of the status of development of the MITICA and SPIDER main components at this important stage of the overall development; then it focuses on the latest and most critical issues, regarding both physics and technology, describing the identified solutions.

  19. Updated safety analysis of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, Neill, E-mail: neill.taylor@iter.org [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Baker, Dennis; Ciattaglia, Sergio; Cortes, Pierre; Elbez-Uzan, Joelle; Iseli, Markus; Reyes, Susana; Rodriguez-Rodrigo, Lina; Rosanvallon, Sandrine; Topilski, Leonid [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

    2011-10-15

    An updated version of the ITER Preliminary Safety Report has been produced and submitted to the licensing authorities. It is revised and expanded in response to requests from the authorities after their review of an earlier version in 2008, to reflect enhancements in ITER safety provisions through design changes, to incorporate new and improved safety analyses and to take into account other ITER design evolution. The updated analyses show that changes to the Tokamak cooling water system design have enhanced confinement and reduced potential radiological releases as well as removing decay heat with very high reliability. New and updated accident scenario analyses, together with fire and explosion risk analyses, have shown that design provisions are sufficient to minimize the likelihood of accidents and reduce potential consequences to a very low level. Taken together, the improvements provided a stronger demonstration of the very good safety performance of the ITER design.

  20. Detailed design of the RF source for the 1 MV neutral beam test facility

    Energy Technology Data Exchange (ETDEWEB)

    Marcuzzi, D.; Palma, M. Dalla [Consorzio RFX, Euratom-ENEA Association, Corso Stati Uniti 4, I35127 Padova (Italy); Pavei, M. [Consorzio RFX, Euratom-ENEA Association, Corso Stati Uniti 4, I35127 Padova (Italy)], E-mail: mauro.pavei@igi.cnr.it; Heinemann, B.; Kraus, W.; Riedl, R. [Max-Planck-Institut fuer Plasmaphysik, Euratom Association, Botzmannstr. 2, D-85748 Garching (Germany)

    2009-06-15

    In the framework of the EU activities for the development of the Neutral Beam Injector for ITER, the detailed design of the Radio Frequency (RF) driven negative ion source to be installed in the 1 MV ITER Neutral Beam Test Facility (NBTF) has been carried out. Results coming from ongoing R and D on IPP test beds [A. Staebler et al., Development of a RF-Driven Ion Source for the ITER NBI System, this conference] and the design of the new ELISE facility [B. Heinemann et al., Design of the Half-Size ITER Neutral Beam Source Test Facility ELISE, this conference] brought several modifications to the solution based on the previous design. An assessment was carried out regarding the Back-Streaming positive Ions (BSI+) that impinge on the back plates of the ion source and cause high and localized heat loads. This led to the redesign of most heated components to increase cooling, and to different choices for the plasma facing materials to reduce the effects of sputtering. The design of the electric circuit, gas supply and the other auxiliary systems has been optimized. Integration with other components of the beam source has been revised, with regards to the interfaces with the supporting structure, the plasma grid and the flexible connections. In the paper the design will be presented in detail, as well as the results of the analyses performed for the thermo-mechanical verification of the components.

  1. Status of the 1 MeV Accelerator Design for ITER NBI

    Science.gov (United States)

    Kuriyama, M.; Boilson, D.; Hemsworth, R.; Svensson, L.; Graceffa, J.; Schunke, B.; Decamps, H.; Tanaka, M.; Bonicelli, T.; Masiello, A.; Bigi, M.; Chitarin, G.; Luchetta, A.; Marcuzzi, D.; Pasqualotto, R.; Pomaro, N.; Serianni, G.; Sonato, P.; Toigo, V.; Zaccaria, P.; Kraus, W.; Franzen, P.; Heinemann, B.; Inoue, T.; Watanabe, K.; Kashiwagi, M.; Taniguchi, M.; Tobari, H.; De Esch, H.

    2011-09-01

    The beam source of neutral beam heating/current drive system for ITER is needed to accelerate the negative ion beam of 40A with D- at 1 MeV for 3600 sec. In order to realize the beam source, design and R&D works are being developed in many institutions under the coordination of ITER organization. The development of the key issues of the ion source including source plasma uniformity, suppression of co-extracted electron in D beam operation and also after the long beam duration time of over a few 100 sec, is progressed mainly in IPP with the facilities of BATMAN, MANITU and RADI. In the near future, ELISE, that will be tested the half size of the ITER ion source, will start the operation in 2011, and then SPIDER, which demonstrates negative ion production and extraction with the same size and same structure as the ITER ion source, will start the operation in 2014 as part of the NBTF. The development of the accelerator is progressed mainly in JAEA with the MeV test facility, and also the computer simulation of beam optics also developed in JAEA, CEA and RFX. The full ITER heating and current drive beam performance will be demonstrated in MITICA, which will start operation in 2016 as part of the NBTF.

  2. The influence of plasma horizontal position on the neutron rate and flux of neutral atoms in injection heating experiment on the TUMAN-3M tokamak

    Science.gov (United States)

    Kornev, V. A.; Chernyshev, F. V.; Melnik, A. D.; Askinazi, L. G.; Wagner, F.; Vildjunas, M. I.; Zhubr, N. A.; Krikunov, S. V.; Lebedev, S. V.; Razumenko, D. V.; Tukachinsky, A. S.

    2013-11-01

    Horizontal displacement of plasma along the major radius has been found to significantly influence the fluxes of 2.45 MeV DD neutrons and high-energy charge-exchange atoms from neutral beam injection (NBI) heated plasma of the TUMAN-3M tokamak. An inward shift by Δ R = 1 cm causes 1.2-fold increase in the neutron flux and 1.9-fold increase in the charge-exchange atom flux. The observed increase in the neutron flux is attributed to joint action of several factors-in particular, improved high-energy ion capture and confinement and, probably, decreased impurity inflow from the walls, which leads to an increase in the density of target ions. A considerable increase in the flux of charge-exchange neutrals in inward-shifted plasma is due to the increased number of captured high-energy ions and, to some extent, the increased density of the neutral target. As a result of the increase in the content of high-energy ions, the central ion temperature T i (0) increased from 250 to 350 eV. The dependence of the neutron rate on major radius R 0 should be taken into account when designing compact tokamak-based neutron sources.

  3. SciDAC Center for Simulation of Wave-Plasma Interactions - Iterated Finite-Orbit Monte Carlo Simulations with Full-Wave Fields for Modeling Tokamak ICRF Wave Heating Experiments - Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Myunghee [Retired; Chan, Vincent S. [General Atomics

    2014-02-28

    This final report describes the work performed under U.S. Department of Energy Cooperative Agreement DE-FC02-08ER54954 for the period April 1, 2011 through March 31, 2013. The goal of this project was to perform iterated finite-orbit Monte Carlo simulations with full-wall fields for modeling tokamak ICRF wave heating experiments. In year 1, the finite-orbit Monte-Carlo code ORBIT-RF and its iteration algorithms with the full-wave code AORSA were improved to enable systematical study of the factors responsible for the discrepancy in the simulated and the measured fast-ion FIDA signals in the DIII-D and NSTX ICRF fast-wave (FW) experiments. In year 2, ORBIT-RF was coupled to the TORIC full-wave code for a comparative study of ORBIT-RF/TORIC and ORBIT-RF/AORSA results in FW experiments.

  4. Transmission of the Neutral Beam Heating Beams at TJ-II; Transmision del Haz de Neutros de Calentamiento en TJ-II

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes Lopez, C.

    2007-09-27

    Neutral beam injection heating has been development for the TJ-II stellarator. The beam has a port-through power between 700-1500 kW and injection energy 40 keV. The sensibility of the injection system to the changes of several parameters is analysed. Beam transmission is limited by losses processes since beam is born into the ions source until is coming into the fusion machine. For the beam transmission optimization several beam diagnostics have been developed. A carbon fiber composite (CFC) target calorimeter has been installed at TJ-II to study in situ the power density distribution of the neutral beams. The thermographic print of the beam can be recorded and analysed in a reliable way due to the highly anisotropic thermal conductivity of the target material. With the combined thermographic and calorimetric measurements it has been possible to determine the power density distribution of the beam. It has been found that a large beam halo is present, which can be explained by the extreme misalignment of the grids. This kind of halo has a deleterious effect on beam transport and must be minimized in order to improve the plasma heating capability of the beams. (Author) 155 refs.

  5. Study of heat and synchrotron radiation transport in fusion tokamak plasmas. Application to the modelling of steady state and fast burn termination scenarios for the international experimental fusion reactor ITER

    Energy Technology Data Exchange (ETDEWEB)

    Villar Colome, J. [Association Euratom-CEA, Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee]|[Universitat Polytechnica de Catalunya (Spain)

    1997-12-01

    The aim of this thesis is to give a global scope of the problem of energy transport within a thermonuclear plasma in the context of its power balance and the implications when modelling ITER operating scenarios. This is made in two phases. First, by furnishing new elements to the existing models of heat and synchrotron radiation transport in a thermonuclear plasma. Second, by applying the improved models to plasma engineering studies of ITER operating scenarios. The scenarios modelled are the steady state operating point and the transient that appears to have the biggest technological implications: the fast burn termination. The conduction-convection losses are modelled through the energy confinement time. This parameter is empirically obtained from the existing experimental data, since the underlying mechanisms are not well understood. In chapter 2 an expression for the energy confinement time is semi-analytically deduced from the Rebut-Lallia-Watkins local transport model. The current estimates of the synchrotron radiation losses are made with expressions of the dimensionless transparency factor deduced from a 0-dimensional cylindrical model proposed by Trubnikov in 1979. In chapter 3 realistic hypothesis for the cases of cylindrical and toroidal geometry are included in the model to deduce compact explicit expressions for the fast numerical computation of the synchrotron radiation losses. Numerical applications are provided for the cylindrical case. The results are checked against the existing models. In chapter 4, the nominal operating point of ITER and its thermal stability is studied by means of a 0-dimensional burn model of the thermonuclear plasma in ignition. This model is deduced by the elements furnished by the plasma particle and power balance. Possible heat overloading on the plasma facing components may provoke severe structural damage, implying potential safety problems related to tritium inventory and metal activation. In chapter 5, the assessment

  6. Bulk Fermi Surface of Charge-Neutral Excitations in SmB_{6} or Not: A Heat-Transport Study.

    Science.gov (United States)

    Xu, Y; Cui, S; Dong, J K; Zhao, D; Wu, T; Chen, X H; Sun, Kai; Yao, Hong; Li, S Y

    2016-06-17

    Recently, there have been increasingly hot debates on whether a bulk Fermi surface of charge-neutral excitations exists in the topological Kondo insulator SmB_{6}. To unambiguously resolve this issue, we perform the low-temperature thermal conductivity measurements of a high-quality SmB_{6} single crystal down to 0.1 K and up to 14.5 T. Our experiments show that the residual linear term of thermal conductivity at the zero field is zero, within the experimental accuracy. Furthermore, the thermal conductivity is insensitive to the magnetic field up to 14.5 T. These results demonstrate the absence of fermionic charge-neutral excitations in bulk SmB_{6}, such as scalar Majorana fermions or spinons and, thus, exclude the existence of a bulk Fermi surface suggested by a recent quantum oscillation study of SmB_{6}. This puts a strong constraint on the explanation of the quantum oscillations observed in SmB_{6}.

  7. Heat-induced denaturation and aggregation of ovalbumin at neutral pH described by irreversible first-order kinetics

    NARCIS (Netherlands)

    Weijers, M.; Barneveld, P.A.; Cohen Stuart, M.A.; Visschers, R.W.

    2003-01-01

    The heat-induced denaturation kinetics of two different sources of ovalbumin at pH 7 was studied by chromatography and differential scanning calorimetry. The kinetics was found to be independent of protein concentration and salt concentration, but was strongly dependent on temperature. For highly

  8. Limited Neutrality

    DEFF Research Database (Denmark)

    Nielsen, Morten Ebbe Juul

    2006-01-01

    Article Concerning the prospect of a kind of limited neutrality in place of the standard liberal egalitarian "neutrality of justification."......Article Concerning the prospect of a kind of limited neutrality in place of the standard liberal egalitarian "neutrality of justification."...

  9. Limited Neutrality

    DEFF Research Database (Denmark)

    Nielsen, Morten Ebbe Juul

    2006-01-01

    Article Concerning the prospect of a kind of limited neutrality in place of the standard liberal egalitarian "neutrality of justification."......Article Concerning the prospect of a kind of limited neutrality in place of the standard liberal egalitarian "neutrality of justification."...

  10. Neutral particle dynamics in the Alcator C-Mod tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Niemczewski, A.P.

    1995-08-01

    This thesis presents an experimental study of neutral particle dynamics in the Alcator C-Mod tokamak. The primary diagnostic used is a set of six neutral pressure gauges, including special-purpose gauges built for in situ tokamak operation. While a low main chamber neutral pressure coincides with high plasma confinement regimes, high divertor pressure is required for heat and particle flux dispersion in future devices such as ITER. Thus we examine conditions that optimize divertor compression, defined here as a divertor-to-midplane pressure ratio. We find both pressures depend primarily on the edge plasma regimes defined by the scrape-off-layer heat transport. While the maximum divertor pressure is achieved at high core plasma densities corresponding to the detached divertor state, the maximum compression is achieved in the high-recycling regime. Variations in the divertor geometry have a weaker effect on the neutral pressures. For otherwise similar plasmas the divertor pressure and compression are maximum when the strike point is at the bottom of the vertical target plate. We introduce a simple flux balance model, which allows us to explain the divertor neutral pressure across a wide range of plasma densities. In particular, high pressure sustained in the detached divertor (despite a considerable drop in the recycling source) can be explained by scattering of neutrals off the cold plasma plugging the divertor throat. Because neutrals are confined in the divertor through scattering and ionization processes (provided the mean-free-paths are much shorter than a typical escape distance) tight mechanical baffling is unnecessary. The analysis suggests that two simple structural modifications may increase the divertor compression in Alcator C-Mod by a factor of about 5. Widening the divertor throat would increase the divertor recycling source, while closing leaks in the divertor structure would eliminate a significant neutral loss mechanism. 146 refs., 82 figs., 14 tabs.

  11. Considerations for the development of neutral beam injection for fusion reactors or DEMO

    Science.gov (United States)

    Hemsworth, R. S.; Boilson, D.

    2017-08-01

    Neutral beam injection (NBI) has been the most successful heating scheme applied to fusion devices, the majority of which have been based on the acceleration and neutralization in a gas target of accelerated positive ions. For large fusion devices such as ITER, DEMO and fusion reactors, beam energies of the order of 0.5 MeV per nucleon or higher are required to penetrate deeply into the fusing plasma, and thus to heat the plasma in the most important region, i.e. near the poloidal axis of the device, and to drive current in the plasma. Because the efficiency of neutralization of positive ions in a gas target becomes unacceptably low at energies above ≈100 keV/nucleon, future injectors will be based on the neutralization of negative ions, either in a gas target, by photons or in a plasma target. So far only two systems based on negative ions have been used on fusion devices, at JT-60U and at LHD, both based on neutralization in a gas target. The injectors for ITER will also use a gas target, but the energy and operating environment are reactor and DEMO relevant. Also the ITER injectors will have to operate for pulse lengths orders of magnitude higher than all previous NBI systems. In this paper the R&D required for an NBI system for a reactor, or DEMO, is considered against the background of the ITER NBI system development, and the main elements of the required R&D are identified.

  12. Assembly and gap management strategy for the ITER NBI vessel passive magnetic shield

    Energy Technology Data Exchange (ETDEWEB)

    Ríos, Luis, E-mail: luis.rios@ciemat.es [CIEMAT Laboratorio Nacional de Fusión, Avda. Complutense 22, 28040 Madrid (Spain); Ahedo, Begoña; Alonso, Javier; Barrera, Germán; Cabrera, Santiago; Rincón, Esther; Ramos, Francisco [CIEMAT Laboratorio Nacional de Fusión, Avda. Complutense 22, 28040 Madrid (Spain); El-Ouazzani, Anass; Graceffa, Joseph; Urbani, Marc; Shah, Darshan [ITER Organization, Route de Vinon-sur-Verdon – CS 90 046, 13067 St Paul Lez Durance Cedex (France); Agarici, Gilbert [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3 – 07/08, 08019 Barcelona (Spain)

    2015-10-15

    The neutral beam system for ITER consists of two heating and current drive neutral ion beam injectors (HNB) and a diagnostic neutral beam (DNB) injector. The proposed physical plant layout allows a possible third HNB injector to be installed later. The HNB Passive Magnetic Shield (PMS) works in conjunction with the active compensation/correction coils to limit the magnetic field inside the Beam Line Vessel (BLV), Beam Source Vessel (BSV), High Voltage Bushing (HVB) and Transmission Line (TL) elbow to acceptable levels that do not interfere with the operation of the HNB components. This paper describes the current design of the PMS, having had only minor modifications since the preliminary design review (PDR) held in IO in April 2013, and the assembly strategy for the vessel PMS.

  13. Requirements for ITER diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Young, K.M.

    1991-01-01

    The development and design of plasma diagnostics for the International Thermonuclear Experimental Reactor (ITER) present a formidable challenge for experimental plasma physicists. The large plasma size, the high central density and temperature and the very high thermal wall loadings provide new challenges for present measurement techniques and lead to a search for new methods. But the physics and control requirements for the long burn phase of the discharge, combined with very limited access to the plasma, constrained by the requirement for radiation shielding of the coils and sharing of access ports with heating and current drive power, remote manipulation, fueling and turn blanket modules, make for very difficult design choices. An initial attempt at these choices has been made by an international team of diagnostic physicists, gathering together in a series of three workshops during the ITER Conceptual Design Activity. This paper is based on that report and provides a summary of its most important points. To provide a background against which to place the diagnostic requirements and design concepts, the ITER device, its most important plasma properties and the proposed experimental program will be described. The specifications for the measurement of the plasma parameters and the proposed diagnostics for these measurements will then be addressed, followed by some examples of the design concepts that have been proposed. As a result of these design studies, it was clear that there were many uncertainties associated with these concepts, particularly because of the nuclear radiation environment, so that a Research and Development Program for diagnostic hardware was established. It will also be briefly summarized.

  14. Embeddings of Iteration Trees

    OpenAIRE

    Mitchell, William

    1992-01-01

    This paper, dating from May 1991, contains preliminary (and unpublishable) notes on investigations about iteration trees. They will be of interest only to the specialist. In the first two sections I define notions of support and embeddings for tree iterations, proving for example that every tree iteration is a direct limit of finite tree iterations. This is a generalization to models with extenders of basic ideas of iterated ultrapowers using only ultrapowers. In the final section (which is m...

  15. Transition and Interaction of Low-Frequency Magnetohydrodynamic Modes during Neutral Beam Injection Heating on HL-2A

    Science.gov (United States)

    Yu, Liming; Chen, Wei; Ding, Xuantong; Ji, Xiaoquan; Shi, Zhongbing; Yu, Deliang; Jiang, Min; Li, Dong; Li, Jiaxian; Li, Yonggao; Zhou, Yan; Ma, Rui; Li, Wei; Feng, Beibin; Huang, Yuan; Song, Xianming; Cao, Jianyong; Rao, Jun; Dong, Jiaqi; Xu, Min; Liu, Yi; Yan, Longwen; Yang, Qingwei; Xu, Yuhong; Duan, Xuru

    2017-02-01

    The strong fishbone mode (FB) and long-lived mode (LLM) have been observed during neutral beam injection (NBI) on the HL-2A tokamak. The FB and LLM can transit between each other. The LLM is identified as an internal kink mode (IKM) with the mode structure obtained using a newly developed electron cyclotron emission radiometer imaging (ECEI) system. The frequency of the LLM (fLLM) is higher than the toroidal rotation frequency (ft) near the q = 1 surface (r ˜ 10 cm). Experimental results show that the LLM is likely to be excited at a higher line-averaged electron density (bar{n}e) than that of the FB when the NBI power is fixed. It is found that the FB and its harmonic as seed magnetic islands can trigger tearing modes (TMs). The mode numbers for the low-frequency and high-frequency TMs are m/n = 2/1 and 3/2, respectively. By further investigation, it is found that there is an m/n = 1/1 IKM coexisting at the same time and with the same frequency as the m/n = 2/1 TM, and the m = 1 mode structure of the IKM in the radial cross section is obtained by the Bayesian tomography method utilizing soft X-ray arrays. The nonlinear coupling conditions are satisfied among the two TMs and IKM.

  16. Design, manufacture and initial operation of the beryllium components of the JET ITER-like wall

    CERN Document Server

    Riccardo, V; Matthews, G F; Nunes, I; Thompson, V; Villedieu, E; Contributors, JET EFDA

    2013-01-01

    The aim of the JET ITER-like Wall Project was to provide JET with the plasma facing material combination now selected for the DT phase of ITER (bulk beryllium main chamber limiters and a full tungsten divertor) and, in conjunction with the upgraded neutral beam heating system, to achieve ITER relevant conditions. The design of the bulk Be plasma facing components had to be compatible with increased heating power and pulse length, as well as to reuse the existing tile supports originally designed to cope with disruption loads from carbon based tiles and be installed by remote handling. Risk reduction measures (prototypes, jigs, etc) were implemented to maximize efficiency during the shutdown. However, a large number of clashes with existing components not fully captured by the configuration model occurred. Restarting the plasma on the ITER-like Wall proved much easier than for the carbon wall and no deconditioning by disruptions was observed. Disruptions have been more threatening than expected due to the redu...

  17. Toolkit for high performance Monte Carlo radiation transport and activation calculations for shielding applications in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Serikov, A.; Fischer, U.; Grosse, D.; Leichtle, D.; Majerle, M., E-mail: arkady.serikov@kit.edu [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany)

    2011-07-01

    The Monte Carlo (MC) method is the most suitable computational technique of radiation transport for shielding applications in fusion neutronics. This paper is intended for sharing the results of long term experience of the fusion neutronics group at Karlsruhe Institute of Technology (KIT) in radiation shielding calculations with the MCNP5 code for the ITER fusion reactor with emphasizing on the use of several ITER project-driven computer programs developed at KIT. Two of them, McCad and R2S, seem to be the most useful in radiation shielding analyses. The McCad computer graphical tool allows to perform automatic conversion of the MCNP models from the underlying CAD (CATIA) data files, while the R2S activation interface couples the MCNP radiation transport with the FISPACT activation allowing to estimate nuclear responses such as dose rate and nuclear heating after the ITER reactor shutdown. The cell-based R2S scheme was applied in shutdown photon dose analysis for the designing of the In-Vessel Viewing System (IVVS) and the Glow Discharge Cleaning (GDC) unit in ITER. Newly developed at KIT mesh-based R2S feature was successfully tested on the shutdown dose rate calculations for the upper port in the Neutral Beam (NB) cell of ITER. The merits of McCad graphical program were broadly acknowledged by the neutronic analysts and its continuous improvement at KIT has introduced its stable and more convenient run with its Graphical User Interface. Detailed 3D ITER neutronic modeling with the MCNP Monte Carlo method requires a lot of computation resources, inevitably leading to parallel calculations on clusters. Performance assessments of the MCNP5 parallel runs on the JUROPA/HPC-FF supercomputer cluster permitted to find the optimal number of processors for ITER-type runs. (author)

  18. Development of ITER non-activation phase operation scenarios

    Science.gov (United States)

    Kim, S. H.; Poli, F. M.; Koechl, F.; Militello-Asp, E.; Polevoi, A. R.; Budny, R.; Casper, T. A.; Loarte, A.; Luce, T. C.; Na, Y.-S.; Romanelli, M.; Schneider, M.; Snipes, J. A.; de Vries, P. C.; The ITPA Topical Group on Integrated Operation Scenarios

    2017-08-01

    Non-activation phase operations in ITER in hydrogen (H) and helium (He) will be important for commissioning of tokamak systems, such as diagnostics, heating and current drive (HCD) systems, coils and plasma control systems, and for validation of techniques necessary for establishing operations in DT. The assessment of feasible HCD schemes at various toroidal fields (2.65-5.3 T) has revealed that the previously applied assumptions need to be refined for the ITER non-activation phase H/He operations. A study of the ranges of plasma density and profile shape using the JINTRAC suite of codes has indicated that the hydrogen pellet fuelling into He plasmas should be utilized taking the optimization of IC power absorption, neutral beam shine-through density limit and H-mode access into account. The EPED1 estimation of the edge pedestal parameters has been extended to various H operation conditions, and the combined EPED1 and SOLPS estimation has provided guidance for modelling the edge pedestal in H/He operations. The availability of ITER HCD schemes, ranges of achievable plasma density and profile shape, and estimation of the edge pedestal parameters for H/He plasmas have been integrated into various time-dependent tokamak discharge simulations. In this work, various H/He scenarios at a wide range of plasma current (7.5-15 MA) and field (2.65-5.3 T) have been developed for the ITER non-activation phase operation, and the sensitivity of the developed scenarios to the used assumptions has been investigated to provide guidance for further development. Extended from Preprint: 2016 Int. Conf. on Fusion Energy (Kyoto, Japan, 2016) TH/P2-22.

  19. Structures and heats of formation of the neutral and ionic PNO, NOP, and NPO systems from electronic structure calculations

    Science.gov (United States)

    Grant, Daniel J.; Dixon, David A.; Kemeny, Andre E.; Francisco, Joseph S.

    2008-04-01

    High level ab initio electronic structure calculations using the coupled cluster CCSD(T) method with augmented correlation-consistent basis sets extrapolated to the complete basis set limit have been performed on the PNO, NOP, and NPO isomers and their corresponding anions and cations. Geometries for all species were optimized up through the aug-cc-pV(Q +d)Z level and vibrational frequencies were calculated with the aug-cc-pV(T +d)Z basis set. The most stable of the three isomers is NPO and it is predicted to have a heat of formation of 23.3kcal/mol. PNO is predicted to be only 1.7kcal/mol higher in energy. The calculated adiabatic ionization potential of NPO is 12.07eV and the calculated adiabatic electron affinity is 2.34eV. The calculated adiabatic ionization potential of PNO is 10.27eV and the calculated adiabatic electron affinity is only 0.24eV. NOP is predicted to be much higher in energy by 29.9kcal/mol. The calculated rotational constants for PNO and NPO should allow for these species to be spectroscopically distinguished. The adiabatic bond dissociation energies for the P N, P O, and N O bonds in NPO and PNO are the same within ˜10kcal/mol and fall in the range of 72-83kcal/mol.

  20. First results of the ITER-relevant negative ion beam test facility ELISE (invited).

    Science.gov (United States)

    Fantz, U; Franzen, P; Heinemann, B; Wünderlich, D

    2014-02-01

    An important step in the European R&D roadmap towards the neutral beam heating systems of ITER is the new test facility ELISE (Extraction from a Large Ion Source Experiment) for large-scale extraction from a half-size ITER RF source. The test facility was constructed in the last years at Max-Planck-Institut für Plasmaphysik Garching and is now operational. ELISE is gaining early experience of the performance and operation of large RF-driven negative hydrogen ion sources with plasma illumination of a source area of 1 × 0.9 m(2) and an extraction area of 0.1 m(2) using 640 apertures. First results in volume operation, i.e., without caesium seeding, are presented.

  1. A steerable ECRF launcher for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Grunloh, H.; Prater, R.; Doane, J.L.; Moeller, C.P. [General Atomics, San Diego (United States); Makowski, M. [ITER Joint Work Site, Garching (Germany)

    1998-07-01

    A design is proposed to steer the electron cyclotron heating and current drive power for ITER using rotatable, water-cooled mirrors and long-pressure hydraulic actuators, and to accommodate changes in length of the waveguide when the temperatures of the vacuum vessel and the cryostat change using waveguide bellows. An alternative concept is also introduced that requires no moving parts within the ITER cryostat and that utilizes wave reconstruction within the waveguide to effect the steering. (author)

  2. ITER components cooling: Satisfying the distinct needs of systems and components

    Energy Technology Data Exchange (ETDEWEB)

    Ployhar, Steven James, E-mail: steve.ployhar@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Gopalapillai, Babulal; Teodoros, Liliana Cristina; Dell Orco, Giovanni [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Kumar, Ajith; Gupta, Dinesh; Patel, Nirav; Jadhav, Mahesh Ashok [ITER-India, Institute for Plasma Research, A-29, GIDC Electronic Estate, Sector-25, Gandhinagar 382016, Gujarat (India)

    2014-10-15

    The ITER Tokamak requires multiple auxiliary systems to initiate, support, and monitor the fusion reaction. Heat produced by these systems, as well as the heat produced by the fusion reaction itself is collected by the ITER Cooling Water System (CWS) and rejected to the atmosphere. The CWS is composed of several systems designed for specific cooling roles. One of these systems is the Component Cooling Water System 2 (CCWS-2) whose function is to collect the heat from auxiliary client systems and components and transfer it to the Heat Rejection System. Clients are located throughout the site and have different requirements in terms of pressure, temperature, temperature variation, flow, metallurgy of wetted surfaces, and water quality. To satisfy these different requirements the CCWS-2 is divided into four separate loops, each of which has different operating parameters. For example, the CCWS-2A loop is designed to cool components with wetted surfaces of copper and primarily serves the radio-frequency heating systems, magnet power supplies, and neutral beam injector system components. This paper describes the evolution of the CCWS-2 system to match the needs of groups of compatible clients, and describes the development of the preliminary design of one of its loops, CCWS-2A, to meet individual client needs.

  3. New CO{sub 2} neutral city area with integrated district heating system of the future in Hoeje Taastrup - Phase 1: Preparation of demonstration. Final report; Denmark; Ny CO{sub 2}-neutral bydel med fremtidens integrerede fjernvarmesystem i Hoeje Taastrup - Fase 1: Forberedelse af demonstration. Slutrapport

    Energy Technology Data Exchange (ETDEWEB)

    Kaarup Olsen, P.; Hummelshoej, R.M. (Cowi A/S (Denmark))

    2011-02-15

    The project has contributed considerably to developing and maturing many climate initiatives in Hoeje Taastrup municipality - including particularly having made an effort to develop and develop the plans for a CO{sub 2}-neutral quarter called 'Vision Gammelsa'. In this context, the project has involved relevant actors in the development of an energy concept for the quarter. The concept was later used as basis for an EU application to the CONCERTO programme, for which Hoeje Taastrup achieved final grant agreement in December 2009. In this way, the municipality obtained funding for a 6-year project 'ECO-Life' which will focus on energy measures in Hedehusene, Floeng and the Gammelsoe region. The project will be carried out in cooperation with a municipality in Lithuania (Birstonas) and Belgium (Kortrijk). In addition, the project supported the development of a climate plan and policy in the municipality. During the life of the project, the council has carried out a wide range of ambitious initiatives which to a great extent stems from the project and in particularly the cooperation that the project has brought along. Based on a preliminary draft proposal for the district, a number of energy concepts have been assessed that can make the district CO{sub 2}-neutral. Primary focus has been on the heat supply. It is recommended to establish a low-temperature heating network, since such a solution provides the best possibility of using surplus heat and renewable energy sources, and in the future, perhaps already in 15 years, district heating from VEKS will be CO{sub 2}-neutral. Three energy efficient and CO{sub 2} efficient district heating solutions have been investigated in more detail: 1) Local low-temperature district heating network with ground heat and solar heat. 2) Low-temperature district heating network based on return heat from the HTF/VEKS system. 3) Low-temperature district heating network with heat driven heat pumps and solar heat. All

  4. New CO{sub 2} neutral city area with integrated district heating system of the future in Hoeje Taastrup - Phase 1: Preparation of demonstration. Final report; Denmark; Ny CO{sub 2}-neutral bydel med fremtidens integrerede fjernvarmesystem i Hoeje Taastrup - Fase 1: Forberedelse af demonstration. Slutrapport

    Energy Technology Data Exchange (ETDEWEB)

    Kaarup Olsen, P.; Hummelshoej, R.M. (Cowi A/S (Denmark))

    2011-02-15

    The project has contributed considerably to developing and maturing many climate initiatives in Hoeje Taastrup municipality - including particularly having made an effort to develop and develop the plans for a CO{sub 2}-neutral quarter called 'Vision Gammelsa'. In this context, the project has involved relevant actors in the development of an energy concept for the quarter. The concept was later used as basis for an EU application to the CONCERTO programme, for which Hoeje Taastrup achieved final grant agreement in December 2009. In this way, the municipality obtained funding for a 6-year project 'ECO-Life' which will focus on energy measures in Hedehusene, Floeng and the Gammelsoe region. The project will be carried out in cooperation with a municipality in Lithuania (Birstonas) and Belgium (Kortrijk). In addition, the project supported the development of a climate plan and policy in the municipality. During the life of the project, the council has carried out a wide range of ambitious initiatives which to a great extent stems from the project and in particularly the cooperation that the project has brought along. Based on a preliminary draft proposal for the district, a number of energy concepts have been assessed that can make the district CO{sub 2}-neutral. Primary focus has been on the heat supply. It is recommended to establish a low-temperature heating network, since such a solution provides the best possibility of using surplus heat and renewable energy sources, and in the future, perhaps already in 15 years, district heating from VEKS will be CO{sub 2}-neutral. Three energy efficient and CO{sub 2} efficient district heating solutions have been investigated in more detail: 1) Local low-temperature district heating network with ground heat and solar heat. 2) Low-temperature district heating network based on return heat from the HTF/VEKS system. 3) Low-temperature district heating network with heat driven heat pumps and solar heat. All

  5. Overview of the JET ITER-like Wall, First Results and Scientific Programme

    Science.gov (United States)

    Matthews, Guy; JET-EFDA Collaboration

    2011-10-01

    The ITER-like Wall (ILW) is the first integrated tokamak experiment with a beryllium main chamber wall and tungsten divertor as foreseen for the activated operational phase of ITER: The ILW will study plasma-wall interaction (PWI) processes (material erosion, material mixing etc.), and the compatibility of the ITER materials with low fuel retention and high power operation. Replacement of the JET CFC first wall by solid Be limiters, and a combination of bulk W and W-coated CFC divertor tiles was performed by remote handling and completed in May 2011 in parallel with a neutral beam heating upgrade to 35 MW and enhancement of diagnostic capabilities. Mitigation of the power and energy loads in the divertor to acceptable levels at high power plasma performance will require high-density plasmas and radiative cooling via impurity seeding. Experiments were carried out with the carbon wall in preparation for the ILW to operate plasmas within ILW limits and provide reference plasmas for key physics studies. Although first plasma is scheduled for mid-August, the scientific programme in support of ITER will start earlier with machine conditioning. In this paper, an overview of the ILW, first results and the outlook for the scientific programme will be presented.

  6. A path to stable low-torque plasma operation in ITER with test blanket modules

    Science.gov (United States)

    Lanctot, M. J.; Snipes, J. A.; Reimerdes, H.; Paz-Soldan, C.; Logan, N.; Hanson, J. M.; Buttery, R. J.; deGrassie, J. S.; Garofalo, A. M.; Gray, T. K.; Grierson, B. A.; King, J. D.; Kramer, G. J.; La Haye, R. J.; Pace, D. C.; Park, J.-K.; Salmi, A.; Shiraki, D.; Strait, E. J.; Solomon, W. M.; Tala, T.; Van Zeeland, M. A.

    2017-03-01

    New experiments in the low-torque ITER Q  =  10 scenario on DIII-D demonstrate that n  =  1 magnetic fields from a single row of ex-vessel control coils enable operation at ITER performance metrics in the presence of applied non-axisymmetric magnetic fields from a test blanket module (TBM) mock-up coil. With n  =  1 compensation, operation below the ITER-equivalent injected torque is successful at three times the ITER equivalent toroidal magnetic field ripple for a pair of TBMs in one equatorial port, whereas the uncompensated TBM field leads to rotation collapse, loss of H-mode and plasma current disruption. In companion experiments at high plasma beta, where the n  =  1 plasma response is enhanced, uncorrected TBM fields degrade energy confinement and the plasma angular momentum while increasing fast ion losses; however, disruptions are not routinely encountered owing to increased levels of injected neutral beam torque. In this regime, n  =  1 field compensation leads to recovery of a dominant fraction of the TBM-induced plasma pressure and rotation degradation, and an 80% reduction in the heat load to the first wall. These results show that the n  =  1 plasma response plays a dominant role in determining plasma stability, and that n  =  1 field compensation alone not only recovers most of the impact on plasma performance of the TBM, but also protects the first wall from potentially damaging heat flux. Despite these benefits, plasma rotation braking from the TBM fields cannot be fully recovered using standard error field control. Given the uncertainty in extrapolation of these results to the ITER configuration, it is prudent to design the TBMs with as low a ferromagnetic mass as possible without jeopardizing the TBM mission.

  7. Benchmarking ICRF simulations for ITER

    Energy Technology Data Exchange (ETDEWEB)

    R. V. Budny, L. Berry, R. Bilato, P. Bonoli, M. Brambilla, R.J. Dumont, A. Fukuyama, R. Harvey, E.F. Jaeger, E. Lerche, C.K. Phillips, V. Vdovin, J. Wright, and members of the ITPA-IOS

    2010-09-28

    Abstract Benchmarking of full-wave solvers for ICRF simulations is performed using plasma profiles and equilibria obtained from integrated self-consistent modeling predictions of four ITER plasmas. One is for a high performance baseline (5.3 T, 15 MA) DT H-mode plasma. The others are for half-field, half-current plasmas of interest for the pre-activation phase with bulk plasma ion species being either hydrogen or He4. The predicted profiles are used by seven groups to predict the ICRF electromagnetic fields and heating profiles. Approximate agreement is achieved for the predicted heating power partitions for the DT and He4 cases. Profiles of the heating powers and electromagnetic fields are compared.

  8. Chevron beam dump for ITER edge Thomson scattering system.

    Science.gov (United States)

    Yatsuka, E; Hatae, T; Vayakis, G; Bassan, M; Itami, K

    2013-10-01

    This paper contains the design of the beam dump for the ITER edge Thomson scattering system and mainly concerns its lifetime under the harsh thermal and electromagnetic loads as well as tight space allocation. The lifetime was estimated from the multi-pulse laser-induced damage threshold. In order to extend its lifetime, the structure of the beam dump was optimized. A number of bent sheets aligned parallel in the beam dump form a shape called a chevron which enables it to avoid the concentration of the incident laser pulse energy. The chevron beam dump is expected to withstand thermal loads due to nuclear heating, radiation from the plasma, and numerous incident laser pulses throughout the entire ITER project with a reasonable margin for the peak factor of the beam profile. Structural analysis was also carried out in case of electromagnetic loads during a disruption. Moreover, detailed issues for more accurate assessments of the beam dump's lifetime are clarified. Variation of the bi-directional reflection distribution function (BRDF) due to erosion by or contamination of neutral particles derived from the plasma is one of the most critical issues that needs to be resolved. In this paper, the BRDF was assumed, and the total amount of stray light and the absorbed laser energy profile on the beam dump were evaluated.

  9. Chevron beam dump for ITER edge Thomson scattering system

    Science.gov (United States)

    Yatsuka, E.; Hatae, T.; Vayakis, G.; Bassan, M.; Itami, K.

    2013-10-01

    This paper contains the design of the beam dump for the ITER edge Thomson scattering system and mainly concerns its lifetime under the harsh thermal and electromagnetic loads as well as tight space allocation. The lifetime was estimated from the multi-pulse laser-induced damage threshold. In order to extend its lifetime, the structure of the beam dump was optimized. A number of bent sheets aligned parallel in the beam dump form a shape called a chevron which enables it to avoid the concentration of the incident laser pulse energy. The chevron beam dump is expected to withstand thermal loads due to nuclear heating, radiation from the plasma, and numerous incident laser pulses throughout the entire ITER project with a reasonable margin for the peak factor of the beam profile. Structural analysis was also carried out in case of electromagnetic loads during a disruption. Moreover, detailed issues for more accurate assessments of the beam dump's lifetime are clarified. Variation of the bi-directional reflection distribution function (BRDF) due to erosion by or contamination of neutral particles derived from the plasma is one of the most critical issues that needs to be resolved. In this paper, the BRDF was assumed, and the total amount of stray light and the absorbed laser energy profile on the beam dump were evaluated.

  10. Approximate Modified Policy Iteration

    CERN Document Server

    Scherrer, Bruno; Ghavamzadeh, Mohammad; Geist, Matthieu

    2012-01-01

    Modified policy iteration (MPI) is a dynamic programming (DP) algorithm that contains the two celebrated policy and value iteration methods. Despite its generality, MPI has not been thoroughly studied, especially its approximation form which is used when the state and/or action spaces are large or infinite. In this paper, we propose three approximate MPI (AMPI) algorithms that are extensions of the well-known approximate DP algorithms: fitted-value iteration, fitted-Q iteration, and classification-based policy iteration. We provide an error propagation analysis for AMPI that unifies those for approximate policy and value iteration. We also provide a finite-sample analysis for the classification-based implementation of AMPI (CBMPI), which is more general (and somehow contains) than the analysis of the other presented AMPI algorithms. An interesting observation is that the MPI's parameter allows us to control the balance of errors (in value function approximation and in estimating the greedy policy) in the fina...

  11. Current Status and Performance Tests of Korea Heat Load Test Facility KoHLT-EB

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sukkwon; Jin, Hyunggon; Shin, Kyuin; Choi, Boguen; Lee, Eohwak; Yoon, Jaesung; Lee, Dongwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Duckhoi; Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    A commissioning test has been scheduled to establish the installation and preliminary performance experiments of the copper hypervapotron mockups. And a qualification test will be performed to evaluate the CuCrZr duct liner in the ITER neutral beam injection facility and the ITER first wall small-scale mockups of the semi-prototype, at up to 1.5 and 5 MW/m{sup 2} high heat flux. Also, this system will be used to test other PFCs for ITER and materials for tokamak reactors. Korean high heat flux test facility(KoHLT-EB; Korea Heat Load Test facility - Electron Beam) by using an electron beam system has been constructed in KAERI to perform the qualification test for ITER blanket FW semi-prototype mockups, hypervapotron cooling devices in fusion devices, and other ITER plasma facing components. The commissioning and performance tests with the supplier of e-gun system have been performed on November 2012. The high heat flux test for hypervapotron cooling device and calorimetry were performed to measure the surface heat flux, the temperature profile and cooling performance. Korean high heat flux test facility for the plasma facing components of nuclear fusion machines will be constructed to evaluate the performance of each component. This facility for the plasma facing materials will be equipped with an electron beam system with a 60 kV acceleration gun.

  12. Recent ASDEX Upgrade research in support of ITER and DEMO

    Science.gov (United States)

    H. Zohmthe ASDEX Upgrade Team; the EUROfusion MST1 Team

    2015-10-01

    Recent experiments on the ASDEX Upgrade tokamak aim at improving the physics base for ITER and DEMO to aid the machine design and prepare efficient operation. Type I edge localized mode (ELM) mitigation using resonant magnetic perturbations (RMPs) has been shown at low pedestal collisionality (νped\\ast PLH occurs indicates that ITER could take advantage of it to initiate H-mode at lower density than that of the final Q = 10 operational point. Core density fluctuation measurements resolved in radius and wave number show that an increase of R/LTe introduced by off-axis electron cyclotron resonance heating (ECRH) mainly increases the large scale fluctuations. The radial variation of the fluctuation level is in agreement with simulations using the GENE code. Fast particles are shown to undergo classical slowing down in the absence of large scale magnetohydrodynamic (MHD) events and for low heating power, but show signs of anomalous radial redistribution at large heating power, consistent with a broadened off-axis neutral beam current drive current profile under these conditions. Neoclassical tearing mode (NTM) suppression experiments using electron cyclotron current drive (ECCD) with feedback controlled deposition have allowed to test several control strategies for ITER, including automated control of (3,2) and (2,1) NTMs during a single discharge. Disruption mitigation studies using massive gas injection (MGI) can show an increased fuelling efficiency with high field side injection, but a saturation of the fuelling efficiency is observed at high injected mass as needed for runaway electron suppression. Large locked modes can significantly decrease the fuelling efficiency and increase the asymmetry of radiated power during MGI mitigation. Concerning power exhaust, the partially detached ITER divertor scenario has been demonstrated at Psep/R = 10 MW m-1 in ASDEX Upgrade, with a peak time averaged target load around 5 MW m-2, well consistent with the component limits

  13. Applied iterative methods

    CERN Document Server

    Hageman, Louis A

    2004-01-01

    This graduate-level text examines the practical use of iterative methods in solving large, sparse systems of linear algebraic equations and in resolving multidimensional boundary-value problems. Assuming minimal mathematical background, it profiles the relative merits of several general iterative procedures. Topics include polynomial acceleration of basic iterative methods, Chebyshev and conjugate gradient acceleration procedures applicable to partitioning the linear system into a "red/black" block form, adaptive computational algorithms for the successive overrelaxation (SOR) method, and comp

  14. ITER test programme

    Science.gov (United States)

    Abdou, M.; Baker, C.; Casini, G.

    1991-07-01

    The International Thermonuclear Experimental Reactor (ITER) was designed to operate in two phases. The first phase, which lasts for 6 years, is devoted to machine checkout and physics testing. The second phase lasts for 8 years and is devoted primarily to technology testing. This report describes the technology test program development for ITER, the ancillary equipment outside the torus necessary to support the test modules, the international collaboration aspects of conducting the test program on ITER, the requirements on the machine major parameters and the R and D program required to develop the test modules for testing in ITER.

  15. Status and verification strategy for ITER neutronics

    Energy Technology Data Exchange (ETDEWEB)

    Loughlin, Michael, E-mail: michael.loughlin@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Angelone, Maurizio [Associazione EURATOM-ENEA Sulla Fusione, Via E. Fermi 45, I-00044 Frascati, Roma (Italy); Batistoni, Paola [Associazione EURATOM-ENEA Sulla Fusione, Via E. Fermi 45, I-00044 Frascati, Roma (Italy); JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Bertalot, Luciano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Eskhult, Jonas [Studsvik Nuclear AB, SE-611 Nyköping (Sweden); Konno, Chikara [Japan Atomic Energy Agency Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Pampin, Raul [F4E Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, Barcelona 08019 (Spain); Polevoi, Alexei; Polunovskiy, Eduard [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2014-10-15

    The paper summarizes the current status of neutronics at ITER and a first set of proposals for experimental programmes to be conducted in the early operational life-time of ITER are described for the more crucial areas. These include a TF coils heating benchmark, a streaming benchmark and streaming measurements by activation on ITER itself. Also on ITER the measurement of activated water from triton burn-up should be planned and performed. This will require the measurement of triton burn-up in DD phase. Measurements of neutron flux in the tokamak building during DD operations should also be carried out. The use of JET for verification of shut down dose rate estimates is desirable. Other facilities to examine the production and behaviour of activated corrosion products and the shielding properties of concretes to high energy (6 MeV) gamma-rays are recommended.

  16. Iteration, Not Induction

    Science.gov (United States)

    Dobbs, David E.

    2009-01-01

    The main purpose of this note is to present and justify proof via iteration as an intuitive, creative and empowering method that is often available and preferable as an alternative to proofs via either mathematical induction or the well-ordering principle. The method of iteration depends only on the fact that any strictly decreasing sequence of…

  17. ITER at Cadarache; ITER a Cadarache

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-06-15

    This public information document presents the ITER project (International Thermonuclear Experimental Reactor), the definition of the fusion, the international cooperation and the advantages of the project. It presents also the site of Cadarache, an appropriate scientifical and economical environment. The last part of the documentation recalls the historical aspect of the project and the today mobilization of all partners. (A.L.B.)

  18. The ITER EC H&CD upper launcher: Structural design

    NARCIS (Netherlands)

    Spaeh, P.; Aiello, G.; de M. Baar,; Chavan, R.; Elzendoorn, B.; Goodman, T.; Henderson, M.; Kleefeldt, K.; Landis, J. D.; Meier, A.; Ronden, D.; Saibene, G.; Scherer, T.; Schreck, S.; Serikov, A.; Strauss, D.; Vaccaro, A.

    2011-01-01

    Four ITER EC H&CD (Electron Cyclotron Heating and Current Drive) Upper Launchers will be installed in the ITER Tokamak to counteract plasma instabilities by injection of up to 20 MW of millimeter-wave power at 170 GHz. Each Launcher features a structural system which is equipped with eight b

  19. THE EC H&CD TRANSMISSION LINE FOR ITER

    NARCIS (Netherlands)

    Gandini, F.; Bigelow, T. S.; Becket, B.; Caughman, J. B.; Cox, D.; Darbos, C.; Gassmann, T.; Henderson, M. A.; Jean, O.; Kajiwara, K.; Kobayashi, N.; Nazare, C.; Oda, Y.; Omori, T.; Purohit, D.; Rasmussen, D. A.; Ronden, D. M. S.; Saibene, G.; Sakamoto, K.; Shapiro, M. A.; Takahashi, K.; Temkin, R. J.

    2011-01-01

    The transmission line (TL) subsystem associated with the ITER electron cyclotron heating and current drive system has reached the conceptual design maturity. At this stage the responsibility of finalizing the design has been transferred from the ITER Organization to the U.S. Domestic Agency. The pur

  20. Status of US ITER Diagnostics

    Science.gov (United States)

    Stratton, B.; Delgado-Aparicio, L.; Hill, K.; Johnson, D.; Pablant, N.; Barnsley, R.; Bertschinger, G.; de Bock, M. F. M.; Reichle, R.; Udintsev, V. S.; Watts, C.; Austin, M.; Phillips, P.; Beiersdorfer, P.; Biewer, T. M.; Hanson, G.; Klepper, C. C.; Carlstrom, T.; van Zeeland, M. A.; Brower, D.; Doyle, E.; Peebles, A.; Ellis, R.; Levinton, F.; Yuh, H.

    2013-10-01

    The US is providing 7 diagnostics to ITER: the Upper Visible/IR cameras, the Low Field Side Reflectometer, the Motional Stark Effect diagnostic, the Electron Cyclotron Emission diagnostic, the Toroidal Interferometer/Polarimeter, the Core Imaging X-Ray Spectrometer, and the Diagnostic Residual Gas Analyzer. The front-end components of these systems must operate with high reliability in conditions of long pulse operation, high neutron and gamma fluxes, very high neutron fluence, significant neutron heating (up to 7 MW/m3) , large radiant and charge exchange heat flux (0.35 MW/m2) , and high electromagnetic loads. Opportunities for repair and maintenance of these components will be limited. These conditions lead to significant challenges for the design of the diagnostics. Space constraints, provision of adequate radiation shielding, and development of repair and maintenance strategies are challenges for diagnostic integration into the port plugs that also affect diagnostic design. The current status of design of the US ITER diagnostics is presented and R&D needs are identified. Supported by DOE contracts DE-AC02-09CH11466 (PPPL) and DE-AC05-00OR22725 (UT-Battelle, LLC).

  1. Novel aspects of plasma control in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Humphreys, D.; Jackson, G.; Walker, M.; Welander, A. [General Atomics P.O. Box 85608, San Diego, California 92186-5608 (United States); Ambrosino, G.; Pironti, A. [CREATE/University of Naples Federico II, Napoli (Italy); Vries, P. de; Kim, S. H.; Snipes, J.; Winter, A.; Zabeo, L. [ITER Organization, St. Paul Lez durance Cedex (France); Felici, F. [Eindhoven University of Technology, Eindhoven (Netherlands); Kallenbach, A.; Raupp, G.; Treutterer, W. [Max-Planck Institut für Plasmaphysik, Garching (Germany); Kolemen, E. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Lister, J.; Sauter, O. [Centre de Recherches en Physique des Plasmas, Ecole Polytechnique Federale de Lausanne, Lausanne (Switzerland); Moreau, D. [CEA, IRFM, 13108 St. Paul-lez Durance (France); Schuster, E. [Lehigh University, Bethlehem, Pennsylvania (United States)

    2015-02-15

    ITER plasma control design solutions and performance requirements are strongly driven by its nuclear mission, aggressive commissioning constraints, and limited number of operational discharges. In addition, high plasma energy content, heat fluxes, neutron fluxes, and very long pulse operation place novel demands on control performance in many areas ranging from plasma boundary and divertor regulation to plasma kinetics and stability control. Both commissioning and experimental operations schedules provide limited time for tuning of control algorithms relative to operating devices. Although many aspects of the control solutions required by ITER have been well-demonstrated in present devices and even designed satisfactorily for ITER application, many elements unique to ITER including various crucial integration issues are presently under development. We describe selected novel aspects of plasma control in ITER, identifying unique parts of the control problem and highlighting some key areas of research remaining. Novel control areas described include control physics understanding (e.g., current profile regulation, tearing mode (TM) suppression), control mathematics (e.g., algorithmic and simulation approaches to high confidence robust performance), and integration solutions (e.g., methods for management of highly subscribed control resources). We identify unique aspects of the ITER TM suppression scheme, which will pulse gyrotrons to drive current within a magnetic island, and turn the drive off following suppression in order to minimize use of auxiliary power and maximize fusion gain. The potential role of active current profile control and approaches to design in ITER are discussed. Issues and approaches to fault handling algorithms are described, along with novel aspects of actuator sharing in ITER.

  2. Novel aspects of plasma control in ITER

    Science.gov (United States)

    Humphreys, D.; Ambrosino, G.; de Vries, P.; Felici, F.; Kim, S. H.; Jackson, G.; Kallenbach, A.; Kolemen, E.; Lister, J.; Moreau, D.; Pironti, A.; Raupp, G.; Sauter, O.; Schuster, E.; Snipes, J.; Treutterer, W.; Walker, M.; Welander, A.; Winter, A.; Zabeo, L.

    2015-02-01

    ITER plasma control design solutions and performance requirements are strongly driven by its nuclear mission, aggressive commissioning constraints, and limited number of operational discharges. In addition, high plasma energy content, heat fluxes, neutron fluxes, and very long pulse operation place novel demands on control performance in many areas ranging from plasma boundary and divertor regulation to plasma kinetics and stability control. Both commissioning and experimental operations schedules provide limited time for tuning of control algorithms relative to operating devices. Although many aspects of the control solutions required by ITER have been well-demonstrated in present devices and even designed satisfactorily for ITER application, many elements unique to ITER including various crucial integration issues are presently under development. We describe selected novel aspects of plasma control in ITER, identifying unique parts of the control problem and highlighting some key areas of research remaining. Novel control areas described include control physics understanding (e.g., current profile regulation, tearing mode (TM) suppression), control mathematics (e.g., algorithmic and simulation approaches to high confidence robust performance), and integration solutions (e.g., methods for management of highly subscribed control resources). We identify unique aspects of the ITER TM suppression scheme, which will pulse gyrotrons to drive current within a magnetic island, and turn the drive off following suppression in order to minimize use of auxiliary power and maximize fusion gain. The potential role of active current profile control and approaches to design in ITER are discussed. Issues and approaches to fault handling algorithms are described, along with novel aspects of actuator sharing in ITER.

  3. Approximate iterative algorithms

    CERN Document Server

    Almudevar, Anthony Louis

    2014-01-01

    Iterative algorithms often rely on approximate evaluation techniques, which may include statistical estimation, computer simulation or functional approximation. This volume presents methods for the study of approximate iterative algorithms, providing tools for the derivation of error bounds and convergence rates, and for the optimal design of such algorithms. Techniques of functional analysis are used to derive analytical relationships between approximation methods and convergence properties for general classes of algorithms. This work provides the necessary background in functional analysis a

  4. Progressing state of design and R and D of NBI for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1997-02-01

    In the International Thermal Nuclear Fusion Experimental Reactor (ITER), Neutral Beam Injection (NBI) apparatus is thought to be a powerful means of electric current drive for heating and stabilizing of plasma, and of controlling the plasma stably. Then, design and development of 1 MeV class negative ion NBI apparatus with compactness and better consistency with reactor have been conducted. On its engineering design, numbers of ports for NBI apparatus were changed form 3 to 4 on a stage of intermediate design completion on June, 1995, and then incident power per unit port was increased from 12.5 MW to 16.7 MW, which brought severer characteristics required for negative ion source. At present, designs of beam deflector, magnetic shield, neutron shielding, remote maintenance and so forth as well as negative ion source and accelerator have been progressed. On its engineering R and D, for development of negative ion source, both deuterium negative ion current and its density established about 1/3 of the characteristics in ITER actural apparatus at an aimed operational gas pressure. And, for development of negative ion accerelator, over 80% of the negative ion acceleration energy which corresponds to an aim of ITER could be established. (G.K.)

  5. The ITER in-vessel system

    Energy Technology Data Exchange (ETDEWEB)

    Lousteau, D.C.

    1994-09-01

    The overall programmatic objective, as defined in the ITER Engineering Design Activities (EDA) Agreement, is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. The ITER EDA Phase, due to last until July 1998, will encompass the design of the device and its auxiliary systems and facilities, including the preparation of engineering drawings. The EDA also incorporates validating research and development (R&D) work, including the development and testing of key components. The purpose of this paper is to review the status of the design, as it has been developed so far, emphasizing the design and integration of those components contained within the vacuum vessel of the ITER device. The components included in the in-vessel systems are divertor and first wall; blanket and shield; plasma heating, fueling, and vacuum pumping equipment; and remote handling equipment.

  6. Electrostatic steering and beamlet aiming in large neutral beam injectors

    Science.gov (United States)

    Veltri, P.; Cavenago, M.; Chitarin, G.; Marcuzzi, D.; Sartori, E.; Serianni, G.; Sonato, P.

    2015-04-01

    Neutral beam injection is the main method for plasma heating in magnetic confinement fusion devices. In high energy injector (E>100 keV/amu), neutrals are obtained with reasonable efficiency by conversion of negative ions (H- or D-) via electron detachment reactions. In the case of ITER injectors, which shall operate at 1 MeV, a total ion current of ˜ 40 A is required to satisfy the heating power demand. Gridded electrodes are therefore used in the accelerator, so that 1280 negative ion beamlets are accelerated together. A carefully designed aiming system is required to control the beamlet trajectories, and to deliver their power on a focal point located several meters away from the beam source. In nowadays injectors, the aiming is typically obtained by aperture offset technique or by grid shaping. This paper discuss an alternative concept of beamlets aiming, based on an electrostatic "steerer" to be placed at the end of the accelerator. A feasibility study of this component is also presented, and its main advantages and drawbacks with respect to other methods are discussed.

  7. Electrostatic steering and beamlet aiming in large neutral beam injectors

    Energy Technology Data Exchange (ETDEWEB)

    Veltri, P., E-mail: pierluigi.veltri@igi.cnr.it; Chitarin, G.; Marcuzzi, D.; Sartori, E.; Serianni, G.; Sonato, P. [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), Corso Stati Uniti 4 - 35127 Padova (Italy); Cavenago, M. [INFN-LNL, viale dell' Università n. 2, 35020 Legnaro (Italy)

    2015-04-08

    Neutral beam injection is the main method for plasma heating in magnetic confinement fusion devices. In high energy injector (E>100 keV/amu), neutrals are obtained with reasonable efficiency by conversion of negative ions (H- or D-) via electron detachment reactions. In the case of ITER injectors, which shall operate at 1 MeV, a total ion current of ∼ 40 A is required to satisfy the heating power demand. Gridded electrodes are therefore used in the accelerator, so that 1280 negative ion beamlets are accelerated together. A carefully designed aiming system is required to control the beamlet trajectories, and to deliver their power on a focal point located several meters away from the beam source. In nowadays injectors, the aiming is typically obtained by aperture offset technique or by grid shaping. This paper discuss an alternative concept of beamlets aiming, based on an electrostatic ”steerer” to be placed at the end of the accelerator. A feasibility study of this component is also presented, and its main advantages and drawbacks with respect to other methods are discussed.

  8. Robust iterative methods

    Energy Technology Data Exchange (ETDEWEB)

    Saadd, Y.

    1994-12-31

    In spite of the tremendous progress achieved in recent years in the general area of iterative solution techniques, there are still a few obstacles to the acceptance of iterative methods in a number of applications. These applications give rise to very indefinite or highly ill-conditioned non Hermitian matrices. Trying to solve these systems with the simple-minded standard preconditioned Krylov subspace methods can be a frustrating experience. With the mathematical and physical models becoming more sophisticated, the typical linear systems which we encounter today are far more difficult to solve than those of just a few years ago. This trend is likely to accentuate. This workshop will discuss (1) these applications and the types of problems that they give rise to; and (2) recent progress in solving these problems with iterative methods. The workshop will end with a hopefully stimulating panel discussion with the speakers.

  9. CORSICA modelling of ITER hybrid operation scenarios

    Science.gov (United States)

    Kim, S. H.; Bulmer, R. H.; Campbell, D. J.; Casper, T. A.; LoDestro, L. L.; Meyer, W. H.; Pearlstein, L. D.; Snipes, J. A.

    2016-12-01

    The hybrid operating mode observed in several tokamaks is characterized by further enhancement over the high plasma confinement (H-mode) associated with reduced magneto-hydro-dynamic (MHD) instabilities linked to a stationary flat safety factor (q ) profile in the core region. The proposed ITER hybrid operation is currently aiming at operating for a long burn duration (>1000 s) with a moderate fusion power multiplication factor, Q , of at least 5. This paper presents candidate ITER hybrid operation scenarios developed using a free-boundary transport modelling code, CORSICA, taking all relevant physics and engineering constraints into account. The ITER hybrid operation scenarios have been developed by tailoring the 15 MA baseline ITER inductive H-mode scenario. Accessible operation conditions for ITER hybrid operation and achievable range of plasma parameters have been investigated considering uncertainties on the plasma confinement and transport. ITER operation capability for avoiding the poloidal field coil current, field and force limits has been examined by applying different current ramp rates, flat-top plasma currents and densities, and pre-magnetization of the poloidal field coils. Various combinations of heating and current drive (H&CD) schemes have been applied to study several physics issues, such as the plasma current density profile tailoring, enhancement of the plasma energy confinement and fusion power generation. A parameterized edge pedestal model based on EPED1 added to the CORSICA code has been applied to hybrid operation scenarios. Finally, fully self-consistent free-boundary transport simulations have been performed to provide information on the poloidal field coil voltage demands and to study the controllability with the ITER controllers. Extended from Proc. 24th Int. Conf. on Fusion Energy (San Diego, 2012) IT/P1-13.

  10. Quantum Iterated Function Systems

    CERN Document Server

    Lozinski, A; Slomczynski, W; Lozinski, Artur; Zyczkowski, Karol; Slomczynski, Wojciech

    2003-01-01

    Iterated functions system (IFS) is defined by specifying a set of functions in a classical phase space, which act randomly on the initial point. In an analogous way, we define quantum iterated functions system (QIFS), where functions act randomly with prescribed probabilities in the Hilbert space. In a more general setting a QIFS consists of completely positive maps acting in the space of density operators. We present exemplary classical IFSs, the invariant measure of which exhibits fractal structure, and study properties of the corresponding QIFSs and their invariant state.

  11. A mature industrial solution for ITER divertor plasma facing components: hypervapotron cooling concept adapted to Tore Supra flat tile technology

    Energy Technology Data Exchange (ETDEWEB)

    Escourbiac, F.; Missirlian, M.; Schlosser, J. [Association EURATOM-CEA Cadarache, Departement de Recherches sur la Fusion Controlee, 13 - Saint Paul lez Durance (France); Bobin-Vastra, I. [AREVA Centre Technique de Framatome, 71 - Le Creusot (France); Kuznetsov, V. [Efremov Institute, Doroga na Metallostroy, St. Petersburg (Russian Federation); Schedler, B. [Plansee AG, Reutte (Austria)

    2004-07-01

    The use of flat tile technology to handle heat fluxes in the range of 20 MW/m{sup 2} with components relevant for fusion experiment applications is technically possible with the hypervapotron cooling concept. This paper deals with recent high heat flux performances operated with success on 2 identical mock-ups, based on this concept, that were tested in 2 different electron gun facilities. Each mock-up consisted of a CuCrZr heat sink armored with 25 flat tiles of the 3D carbon fibre composite material SEPcarb NS31 assembled with pure copper by active metal casting (AMC). The AMC tiles were electron beam welded on the CuCrZr bar, fins and slots on the neutral beam JET design were machined into the bar, then the bar was closed with a thick CuCrZr rear plug including hydraulic connections then the bar was electron beam welded onto the sidewalls. The testing results show that full ITER design specifications were achieved with margins, the critical heat flux limit was even higher than 30 MW/m{sup 2}. These results highlight the high potential of this technology for ITER divertor application.

  12. Long pulse acceleration of MeV class high power density negative H{sup −} ion beam for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Umeda, N., E-mail: umeda.naotaka@jaea.go.jp; Kojima, A.; Kashiwagi, M.; Tobari, H.; Hiratsuka, J.; Watanabe, K.; Dairaku, M.; Yamanaka, H.; Hanada, M. [Japan Atomic Energy Agency, 801-1 Mukouyama, Naka-shi, Ibaraki 311-0193 Japan (Japan)

    2015-04-08

    R and D of high power density negative ion beam acceleration has been carried out at MeV test facility in JAEA to realize ITER neutral beam accelerator. The main target is H{sup −} ion beam acceleration up to 1 MeV with 200 A/m{sup 2} for 60 s whose pulse length is the present facility limit. For long pulse acceleration at high power density, new extraction grid (EXG) has been developed with high cooling capability, which electron suppression magnet is placed under cooling channel similar to ITER. In addition, aperture size of electron suppression grid (ESG) is enlarged from 14 mm to 16 mm to reduce direct interception on the ESG and emission of secondary electron which leads to high heat load on the upstream acceleration grid. By enlarging ESG aperture, beam current increased 10 % at high current beam and total acceleration grid heat load reduced from 13 % to 10 % of input power at long pulse beam. In addition, heat load by back stream positive ion into the EXG is measured for the first time and is estimated as 0.3 % of beam power, while heat load by back stream ion into the source chamber is estimated as 3.5 ~ 4.0 % of beam power. Beam acceleration up to 60 s which is the facility limit, has achieved at 683 keV, 100 A/m{sup 2} of negative ion beam, whose energy density increases two orders of magnitude since 2011.

  13. Iterative List Decoding

    DEFF Research Database (Denmark)

    Justesen, Jørn; Høholdt, Tom; Hjaltason, Johan

    2005-01-01

    We analyze the relation between iterative decoding and the extended parity check matrix. By considering a modified version of bit flipping, which produces a list of decoded words, we derive several relations between decodable error patterns and the parameters of the code. By developing a tree...... of codewords at minimal distance from the received vector, we also obtain new information about the code....

  14. Iterative software kernels

    Energy Technology Data Exchange (ETDEWEB)

    Duff, I.

    1994-12-31

    This workshop focuses on kernels for iterative software packages. Specifically, the three speakers discuss various aspects of sparse BLAS kernels. Their topics are: `Current status of user lever sparse BLAS`; Current status of the sparse BLAS toolkit`; and `Adding matrix-matrix and matrix-matrix-matrix multiply to the sparse BLAS toolkit`.

  15. Current ramps in tokamaks: from present experiments to ITER scenarios

    NARCIS (Netherlands)

    Imbeaux, F.; Citrin, J.; Hobirk, J.; Hogeweij, G. M. D.; Kochl, F.; Leonov, V. M.; Miyamoto, S.; Nakamura, Y.; Parail, V.; Pereverzev, G.; Polevoi, A.; Voitsekhovitch, I.; Basiuk, V.; Budny, R.; Casper, T.; Fereira, J.; Fukuyama, A.; Garcia, J.; Gribov, Y. V.; Hayashi, N.; Honda, M.; Hutchinson, I. H.; Jackson, G.; Kavin, A. A.; Kessel, C. E.; Khayrutdinov, R. R.; Labate, C.; Litaudon, X.; Lomas, P. J.; Lonnroth, J.; Luce, T.; Lukash, V. E.; Mattei, M.; Mikkelsen, D.; Nunes, I.; Peysson, Y.; Politzer, P.; Schneider, M.; Sips, G.; Tardini, G.; Wolfe, S. M.; Zhogolev, V. E.

    2011-01-01

    In order to prepare adequate current ramp-up and ramp-down scenarios for ITER, present experiments from various tokamaks have been analysed by means of integrated modelling in view of determining relevant heat transport models for these operation phases. A set of empirical heat transport models for

  16. Study of the choice of the decoupling layout for the ITER ICRH system

    Energy Technology Data Exchange (ETDEWEB)

    Vervier, M., E-mail: michel.vervier@rma.ac.be; Messiaen, A.; Ongena, J.; Durodié, F.

    2015-12-10

    10 decouplers are used to neutralize the mutual coupling effects and to control the current amplitude of the 24 straps array of the ITER ICRH antenna in the case of current drive phasing. In the case of heating phasing only 4 decouplers are active and the array current control needs to act on the ratio between the power delivered by the 4 generators. This ratio is very sensitive to the precise adjustment of the antenna array phasing. The maximum total radiated power capability is then limited when the power of one generator reaches its maximum value. With the addition of four switches all 10 installed decouplers are made active and can act on all mutual coupling effects with equal source power from the 4 generators. With four more switches the current drive phasing could work with a reduced poloidal phasing resulting in a 35% increase of its coupling to the plasma.

  17. Iterative Algorithms for Nonexpansive Mappings

    Directory of Open Access Journals (Sweden)

    Yao Yonghong

    2008-01-01

    Full Text Available Abstract We suggest and analyze two new iterative algorithms for a nonexpansive mapping in Banach spaces. We prove that the proposed iterative algorithms converge strongly to some fixed point of .

  18. Thermal behaviour analysis on ITER component cooling water system loop 2B

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Bin, E-mail: guobin@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Fu, Peng [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Dell’Orco, Giovanni; Liliana, Teodoros; Tao, Jun [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Yang, Lei [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2015-11-15

    Highlights: • Thermal hydraulic analysis model has been developed to perform thermal analysis on the component cooling water system loop 2B. • The cooling water temperature profile at client inlet and outlet during one cycle of the most demanding plasma operation scenario was obtained. • Operation behaviour of the main heat exchanger for CCWS-2B has been depicted. - Abstract: ITER cooling water system is composed by several cooling loops, the primary heat transfer loops that form the Tokamak Cooling Water System (TCWS), the secondary heat transfer loops that form the Component Cooling Water System (CCWS) and the Chilled Water System (CHWS) and a tertiary heat transfer loop which is the Heat Rejection System (HRS). The CCWS is further divided into CCWS-1, CCWS-2A, CCWS-2B, CCWS-2C, CCWS-2D depending on the water chemistry needs of clients and wetted area material. The component cooling water system loop 2B (CCWS-2B) has the function to remove heat load from coil power supply component, Neutral Beam Injectors (NBIs) system component and diagnostic system which are located in different buildings. As the total number of the client connections for the loop is a few hundreds, simplified thermal hydraulic analysis model has been developed to perform thermal analysis on the component cooling water system loop 2B. The curve of the cooling water temperature at client inlet and outlet during one cycle of the most demanding plasma operation scenario was obtained and the cooling water flow rate can meet the thermal removal requirement of client was also confirmed from this analysis. In addition, operation behaviour of the main heat exchanger for CCWS-2B in this thermal analysis was depicted for main heat exchanger selection purposes. This study has been carried out with the AFT Fathom code.

  19. Plasma Heating and Current Drive for Fusion Reactors

    Science.gov (United States)

    Holtkamp, Norbert

    2010-02-01

    ITER (in Latin ``the way'') is designed to demonstrate the scientific and technological feasibility of fusion energy. Fusion is the process by which two light atomic nuclei combine to form a heavier one and thus release energy. In the fusion process two isotopes of hydrogen - deuterium and tritium - fuse together to form a helium atom and a neutron. Thus fusion could provide large scale energy production without greenhouse effects; essentially limitless fuel would be available all over the world. The principal goals of ITER are to generate 500 megawatts of fusion power for periods of 300 to 500 seconds with a fusion power multiplication factor, Q, of at least 10. Q >= 10 (input power 50 MW / output power 500 MW). In a Tokamak the definition of the functionalities and requirements for the Plasma Heating and Current Drive are relevant in the determination of the overall plant efficiency, the operation cost of the plant and the plant availability. This paper summarise these functionalities and requirements in perspective of the systems under construction in ITER. It discusses the further steps necessary to meet those requirements. Approximately one half of the total heating will be provided by two Neutral Beam injection systems at with energy of 1 MeV and a beam power of 16 MW into the plasma. For ITER specific test facility is being build in order to develop and test the Neutral Beam injectors. Remote handling maintenance scheme for the NB systems, critical during the nuclear phase of the project, will be developed. In addition the paper will give an overview over the general status of ITER. )

  20. Iterative supervirtual refraction interferometry

    KAUST Repository

    Al-Hagan, Ola

    2014-05-02

    In refraction tomography, the low signal-to-noise ratio (S/N) can be a major obstacle in picking the first-break arrivals at the far-offset receivers. To increase the S/N, we evaluated iterative supervirtual refraction interferometry (ISVI), which is an extension of the supervirtual refraction interferometry method. In this method, supervirtual traces are computed and then iteratively reused to generate supervirtual traces with a higher S/N. Our empirical results with both synthetic and field data revealed that ISVI can significantly boost up the S/N of far-offset traces. The drawback is that using refraction events from more than one refractor can introduce unacceptable artifacts into the final traveltime versus offset curve. This problem can be avoided by careful windowing of refraction events.

  1. Iterative participatory design

    DEFF Research Database (Denmark)

    Simonsen, Jesper; Hertzum, Morten

    2010-01-01

    iterative process of mutual learning by designers and domain experts (users), who aim to change the users’ work practices through the introduction of information systems. We provide an illustrative case example with an ethnographic study of clinicians experimenting with a new electronic patient record......The theoretical background in this chapter is information systems development in an organizational context. This includes theories from participatory design, human-computer interaction, and ethnographically inspired studies of work practices. The concept of design is defined as an experimental...... system, focussing on emergent and opportunity-based change enabled by appropriating the system into real work. The contribution to a general core of design research is a reconstruction of the iterative prototyping approach into a general model for sustained participatory design....

  2. The potential role of Neutral Beam Injection in EU DEMO

    Science.gov (United States)

    Vincenzi, Pietro; Artaud, Jean-Francois; Bolzonella, Tommaso; Giruzzi, Gerardo

    2016-10-01

    EU DEMO studies for pulsed (DEMO1) and steady-state (DEMO2) concepts are currently in the pre-conceptual phase. Present DEMO1 design is based on ITER baseline H-mode scenario, while DEMO2 is based on advanced scenarios with moderate reversed q profile sustained by non-inductive currents. One of the possible flattop heating power systems currently considered is Neutral Beam Injection (NBI). In this work the role of NBI in DEMO1 and DEMO2 is investigated by means of integrated simulations of DEMO scenarios using METIS fast tokamak modelling tool. Limitations, requirements and benefits of the use of a NBI system are discussed. For DEMO1 pulsed concept, the role of NBI is mainly central plasma heating for scenario stability (high fusion power H-mode). As a by-product of the tangential injection, NBI is capable of current drive, which is favorable in order to extend the discharge duration. Regarding a steady-state DEMO2 concept, in addition to plasma heating, NBI becomes a direct actuator for the advanced scenario by driving a considerable part of the plasma current. This requires more than 100MW with off-axis injection. The effect of an increase of the injection energy on the driven current density profile is also presented for DEMO2.

  3. The Iterate Manual

    Science.gov (United States)

    1990-10-01

    is probably a bad idea. A better versica would use a temporary: (defmacro sum-of-squares (expr) (let ((temp ( gensym ))) ’(lot (,temp ,expr)) (sum...val ( gensym )) (tempi ( gensym )) (temp2 ( gensym )) (winner (or var iterate::*result-var*))) ’(progn (with ,max-val - nil) (with ,winner = nil) (cond ((null...the elements of a vector (disregards fill-pointer)" (let ((vect ( gensym )) (end ( gensym )) (index ( gensym ))) ’(progn (with ,vect - v) (with ,end = (array

  4. Iterative initial condition reconstruction

    Science.gov (United States)

    Schmittfull, Marcel; Baldauf, Tobias; Zaldarriaga, Matias

    2017-07-01

    Motivated by recent developments in perturbative calculations of the nonlinear evolution of large-scale structure, we present an iterative algorithm to reconstruct the initial conditions in a given volume starting from the dark matter distribution in real space. In our algorithm, objects are first moved back iteratively along estimated potential gradients, with a progressively reduced smoothing scale, until a nearly uniform catalog is obtained. The linear initial density is then estimated as the divergence of the cumulative displacement, with an optional second-order correction. This algorithm should undo nonlinear effects up to one-loop order, including the higher-order infrared resummation piece. We test the method using dark matter simulations in real space. At redshift z =0 , we find that after eight iterations the reconstructed density is more than 95% correlated with the initial density at k ≤0.35 h Mpc-1 . The reconstruction also reduces the power in the difference between reconstructed and initial fields by more than 2 orders of magnitude at k ≤0.2 h Mpc-1 , and it extends the range of scales where the full broadband shape of the power spectrum matches linear theory by a factor of 2-3. As a specific application, we consider measurements of the baryonic acoustic oscillation (BAO) scale that can be improved by reducing the degradation effects of large-scale flows. In our idealized dark matter simulations, the method improves the BAO signal-to-noise ratio by a factor of 2.7 at z =0 and by a factor of 2.5 at z =0.6 , improving standard BAO reconstruction by 70% at z =0 and 30% at z =0.6 , and matching the optimal BAO signal and signal-to-noise ratio of the linear density in the same volume. For BAO, the iterative nature of the reconstruction is the most important aspect.

  5. Integration of IC/EC systems in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Gassmann, T., E-mail: Thibault.Gassmann@iter.org [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Beaumont, B. [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Baruah, U.K. [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Bonicelli, T. [Fusion For Energy, C/ Josep Pla 2, Torres Diagonal Litoral-B3, E-08019 Barcelona (Spain); Chiocchio, S.; Cox, D.; Darbos, C.; Decamps, H. [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Denisov, G. [Institute of Applied Physics, Russian Academy of Sciences, 46 Ulyanov Street, Nizhny Novgorod, 603950 (Russian Federation); Henderson, M.; Kazarian, F.; Lamalle, P.U. [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Mukherjee, A. [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Rasmussen, D. [Oak Ridge National Laboratory, 055 Commerce Park, PO Box 2008, Oak Ridge, TN 37831-6483 (United States); Saibene, G.; Sartori, R. [Fusion For Energy, C/ Josep Pla 2, Torres Diagonal Litoral-B3, E-08019 Barcelona (Spain); Sakamoto, K. [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka-shi, Ibaraki 311-0193 (Japan); Tanga, A. [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

    2010-12-15

    The RF heating and current drive (H and CD) systems that are to be installed in ITER during the construction phase, are the electron cyclotron (EC) and ion cyclotron (IC) systems. They are complex assemblies of high voltage power supplies (HVPS), RF generators, transmission lines and antennas. Their design and integration are constrained by many interfaces, both internal, between the subsystems, and external, with the other ITER systems. In addition, some components must be compatible with a nuclear environment and are classified as Safety Important Component. This paper describes the processes implemented in ITER to ensure proper integration.

  6. Preliminary Design and Analysis of ITER In-Wall Shielding

    Institute of Scientific and Technical Information of China (English)

    LIU Changle; YU Jie; WU Songtao; CAI Yingxiang; PAN Wanjiang

    2007-01-01

    ITER in-wall shielding (IIS) is situated between the doubled shells of the ITER Vacuum Vessel (IVV). Its main functions are applied in shielding neutron, gamma-ray and toroidal field ripple reduction. The structure of IIS has been modelled according to the IVV design criteria which has been updated by the ITER team (IT). Static analysis and thermal expansion analysis were performed for the structure. Thermal-hydraulic analysis verified the heat removal capability and resulting temperature, pressure, and velocity changes in the coolant flow. Consequently, our design work is possibly suitable as a reference for IT's updated or final design in its next step.

  7. PTRANSP Tests Of TGLF And Predictions For ITER

    Energy Technology Data Exchange (ETDEWEB)

    Robert V. Budny, Xingqiu Yuan, S. Jardin, G. Hammett, G. Staebler, members of the ITPA Transport and Confinement Topical Group, and JET EFDA Contributions

    2012-02-28

    One of the physics goals for ITER is to achieve high fusion power PDT at a high gain QDT. This goal is important for studying the physics of reactor-relevant burning plasmas. Simulations of plasma performance in ITER can help achieve this goal by aiding in the design of systems such as diagnostics and in planning ITER plasma regimes. Simulations can indicate areas where further research in theory and experiments is needed. To have credible simulations integrated modeling is necessary since plasma profiles and applied heating, torque, and current drive are strongly coupled.

  8. Characterization of hierarchical α-MoO3 plates toward resistive heating synthesis: electrochemical activity of α-MoO3/Pt modified electrode toward methanol oxidation at neutral pH

    Science.gov (United States)

    Filippo, Emanuela; Baldassarre, Francesca; Tepore, Marco; Guascito, Maria Rachele; Chirizzi, Daniela; Tepore, Antonio

    2017-05-01

    The growth of MoO3 hierarchical plates was obtained by direct resistive heating of molybdenum foils at ambient pressure in the absence of any catalysts and templates. Plates synthesized after 60 min resistive heating typically grow in an single-crystalline orthorhombic structure that develop preferentially in the [001] direction, and are characterized by high resolution transmission electron microscopy, selected area diffraction pattern and Raman-scattering measurements. They are about 100-200 nm in thickness and a few tens of micrometers in length. As heating time proceeds to 80 min, plates of α-MoO3 form a branched structure. A more attentive look shows that primary plates formed at until 60 min could serve as substrates for the subsequent growth of secondary belts. Moreover, a full electrochemical characterization of α-MoO3 plates on platinum electrodes was done by cyclic voltammetric experiments, at pH 7 in phosphate buffer, to probe the activity of the proposed composite material as anode to methanol electro-oxidation. Reported results indicate that Pt MoO3 modified electrodes are appropriate to develop new an amperometric non-enzymatic sensor for methanol as well as to make anodes suitable to be used in direct methanol fuel cells working at neutral pH.

  9. Characterization of Hierarchical α-MoOsub>3sub> Plates Toward Resistive Heating Synthesis: Electrochemical Activity of α-MoOsub>3sub>/Pt Modified Electrode Toward Methanol Oxidation in Neutral pH.

    Science.gov (United States)

    Filippo, Emanuela; Baldassarre, Francesca; Tepore, Marco; Guascito, Maria Rachele; Chirizzi, Daniela; Tepore, Antonio

    2017-03-20

    The growth of MoOsub>3sub> hierarchical plates was obtained by direct resistive heating of molybdenum foil at ambient pressure in absence of any catalysts and templates. Plates synthesized after 60 min resistive heating typically growth in an single-crystalline orthorhombic structure that develop preferentially in [001] direction, as characterized by HRTEM, SAD and Raman-scattering measurements. They are about 100-200nm in thickness and a few tens micrometers in length. As heating time proceeds to 80 min, plates of α-MoOsub>3sub> form a branched structure. A more attentive look shows that a primary plates formed at until 60 min could serve as substrates for the subsequent growth of secondary belts. Moreover, a full electrochemical characterization of α-MoOsub>3sub> plates on platinum electrodes was done by Cyclic Voltammetric experiments, at pH 7 in phosphate buffer, to probe the activity of the proposed composite material as anode to methanol electrooxidation. Reported results indicate that Pt MoOsub>3sub> modified electrodes are appropriate to develop new amperometric non-enzymatic sensor for methanol measurements and as anode in Direct Methanol Fuel Cells (DMFCs) making at neutral pH.

  10. Binding of Gallic Acid and Epigallocatechin Gallate to Heat-Unfolded Whey Proteins at Neutral pH Alters Radical Scavenging Activity of in Vitro Protein Digests.

    Science.gov (United States)

    Cao, Yanyun; Xiong, Youling L

    2017-09-27

    Preheated (80 °C for 9 min) whey protein isolate (HWPI) was reacted with 20, 120, and 240 μmol/g (protein basis) gallic acid (GA) or epigallocatechin gallate (EGCG) at neutral pH and 25 °C. Isothermal titration calorimetry and fluorometry showed a similar trend that GA binding to HWPI was moderate but weaker than EGCG binding. However, the shift of maximal fluorescence emission wavelength in opposite directions in response to GA (blue) and EGCG (red) suggests discrepant binding patterns. Electrophoresis results showed that EGCG induced formation of HWPI complexes while GA only had a marginal effect. Both free and phenolic-bound HWPI exhibited mild antiradical activity. However, when subjected to in vitro digestion, synergistic radical-scavenging activity was produced between the phenolics and peptides with the highest synergism being observed on 120 μmol/g phenolics.

  11. 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.

  12. ERGODIC THEOREM FOR INFINITE ITERATED FUNCTION SYSTEMS

    Institute of Scientific and Technical Information of China (English)

    O Hyong-chol; Ro Yong-hwa; Kil Won-gun

    2005-01-01

    A set of contraction maps of a metric space is called an iterated function systems.Iterated function systems with condensation can be considered infinite iterated function systems. Infinite iterated function systems on compact metric spaces were studied. Using the properties of Banach limit and uniform contractiveness, it was proved that the random iterating algorithms for infinite iterated function systems on compact metric spaces satisfy ergodicity. So the random iterating algorithms for iterated function systems with condensation satisfy ergodicity, too.

  13. Runaway electrons and ITER

    Science.gov (United States)

    Boozer, Allen H.

    2017-05-01

    The potential for damage, the magnitude of the extrapolation, and the importance of the atypical—incidents that occur once in a thousand shots—make theory and simulation essential for ensuring that relativistic runaway electrons will not prevent ITER from achieving its mission. Most of the theoretical literature on electron runaway assumes magnetic surfaces exist. ITER planning for the avoidance of halo and runaway currents is focused on massive-gas or shattered-pellet injection of impurities. In simulations of experiments, such injections lead to a rapid large-scale magnetic-surface breakup. Surface breakup, which is a magnetic reconnection, can occur on a quasi-ideal Alfvénic time scale when the resistance is sufficiently small. Nevertheless, the removal of the bulk of the poloidal flux, as in halo-current mitigation, is on a resistive time scale. The acceleration of electrons to relativistic energies requires the confinement of some tubes of magnetic flux within the plasma and a resistive time scale. The interpretation of experiments on existing tokamaks and their extrapolation to ITER should carefully distinguish confined versus unconfined magnetic field lines and quasi-ideal versus resistive evolution. The separation of quasi-ideal from resistive evolution is extremely challenging numerically, but is greatly simplified by constraints of Maxwell’s equations, and in particular those associated with magnetic helicity. The physics of electron runaway along confined magnetic field lines is clarified by relations among the poloidal flux change required for an e-fold in the number of electrons, the energy distribution of the relativistic electrons, and the number of relativistic electron strikes that can be expected in a single disruption event.

  14. Iterative participatory design

    DEFF Research Database (Denmark)

    2010-01-01

    The theoretical background in this chapter is information systems development in an organizational context. This includes theories from participatory design, human-computer interaction, and ethnographically inspired studies of work practices. The concept of design is defined as an experimental...... iterative process of mutual learning by designers and domain experts (users), who aim to change the users’ work practices through the introduction of information systems. We provide an illustrative case example with an ethnographic study of clinicians experimenting with a new electronic patient record...

  15. Quantum iterated function systems.

    Science.gov (United States)

    Łoziński, Artur; Zyczkowski, Karol; Słomczyński, Wojciech

    2003-10-01

    An iterated function system (IFS) is defined by specifying a set of functions in a classical phase space, which act randomly on an initial point. In an analogous way, we define a quantum IFS (QIFS), where functions act randomly with prescribed probabilities in the Hilbert space. In a more general setting, a QIFS consists of completely positive maps acting in the space of density operators. This formalism is designed to describe certain problems of nonunitary quantum dynamics. We present exemplary classical IFSs, the invariant measure of which exhibits fractal structure, and study properties of the corresponding QIFSs and their invariant states.

  16. Iterative Magnetometer Calibration

    Science.gov (United States)

    Sedlak, Joseph

    2006-01-01

    This paper presents an iterative method for three-axis magnetometer (TAM) calibration that makes use of three existing utilities recently incorporated into the attitude ground support system used at NASA's Goddard Space Flight Center. The method combines attitude-independent and attitude-dependent calibration algorithms with a new spinning spacecraft Kalman filter to solve for biases, scale factors, nonorthogonal corrections to the alignment, and the orthogonal sensor alignment. The method is particularly well-suited to spin-stabilized spacecraft, but may also be useful for three-axis stabilized missions given sufficient data to provide observability.

  17. ITER LIDAR performance analysis.

    Science.gov (United States)

    Beurskens, M N A; Giudicotti, L; Kempenaars, M; Scannell, R; Walsh, M J

    2008-10-01

    The core LIDAR Thomson scattering for ITER is specified for core profile measurements with a spatial resolution of 7 cm (a/30) for the range of 500 eV3x10(19) m(-3) at an accuracy of system can meet its spatial and accuracy specifications for higher temperatures of T(e)>5 keV with a combination of a neodymium-doped yttrium aluminum garnet (Nd:YAG) laser (lambda(0)=1064 nm, Delta lambdanear infrared detectors.

  18. Experiments and simulations for the dynamics of cesium in negative hydrogen ion sources for ITER N-NBI

    Energy Technology Data Exchange (ETDEWEB)

    Gutser, Raphael

    2010-07-21

    The injection of fast neutral particles (NBI) into a fusion plasma is an important method for plasma heating and current drive. A source for negative deuterium ions delivering an 1 MeV beam that is accelerated to a specific energy and neutralized by a gas target is required for the ITER-NBI. Cesium seeding is required to extract high negative ion current densities from these sources. The optimization of the cesium homogeneity and control are major objectives to achieve the source requirements imposed by ITER. Within the scope of this thesis, the Monte Carlo based numerical transport simulation CsFlow3D was developed, which is the first computer model that is capable of simulating the flux and the accumulation of cesium on the surfaces of negative-ion sources. Basic studies that support the code development were performed at a dedicated experiment at the University of Augsburg. Input parameters of the ad- and desorption of cesium at ion source relevant conditions were taken from systematic measurements with a quartz micro balance, while the injection rate of the cesium oven at the ion source was determined by surface ionization detection. This experimental setup was used for further investigations of the work function of cesium-coated samples during plasma exposure. (orig.)

  19. Development of structural design criteria for ITER.

    Energy Technology Data Exchange (ETDEWEB)

    Majumdar, S.

    1998-06-22

    The irradiation environment experienced by the in-vessel components of fusion reactors such as HER presents structural design challenges not envisioned in the development of existing structural design criteria such as the ASME Code or RCC-MR. From the standpoint of design criteria, the most significant issues stem from the irradiation-induced changes in material properties, specifically the reduction of ductility, strain hardening capability, and fracture toughness with neutron irradiation. Recently, Draft 7 of the interim ITER structural design criteria (ISDC), which provide new rules for guarding against such problems, was released for trial use by the ITER designers. The new rules, which were derived from a simple model based on the concept of elastic follow up factor, provide primary and secondary stress limits as functions of uniform elongation and ductility. The implication of these rules on the allowable surface heat flux on typical first walls made of type 316 stainless steel and vanadium alloys are discussed.

  20. The Cryostat and Subsystems Development at ITER

    Science.gov (United States)

    Sekachev, Igor; Meekins, Michael; Sborchia, Carlo; Vitupier, Guillaume; Xie, Han; Zhou, Caipin

    ITER is a large experimental tokamak being built to research fusion power. The ITER cryostat is a multifunctional system which provides vacuum insulation for the superconducting magnets operating at 4.5 K and for the thermal shield operating at 80 K. It also serves as a structural support for the tokamak and provides access ways and corridors to the vacuum vessel for diagnostic lines of sight, additional heating beams and the deployment of remote handling equipment. The cryostat has feed-through penetrations for all the equipment connecting elements of systems outside the cryostat to the corresponding elements inside the cryostat. The cryostat is a vacuum containment vessel having a very large volume of ∼16000 m3 designed to be evacuated to a base pressure of 10-4 Pa. Design details of the cryostat and associated systems, including Torus Cryopump Housing (TCPH), are discussed. Status report of the cryostat developments is presented.

  1. ITER diagnostics ex-vessel engineering services

    Energy Technology Data Exchange (ETDEWEB)

    Arumugam, A.P., E-mail: arun.prakash@iter.org; Walker, C.I.; Andrew, P.; Barnsley, R.; Beltran, D.; Bertalot, L.; Dammann, A.; Direz, M.F.; Drevon, J.M.; Encheva, A.; Giacomin, T.; Hourtoule, J.; Kuehn, I.; Lanza, R.; Levesy, B.; Maquet, P.; Patel, K.M.; Patisson, L.; Pitcher, C.S.; Portales, M.; and others

    2013-10-15

    Highlights: • This paper describes about the ITER diagnostics ex-vessel engineering services. • It describes various diagnostics systems, its location and its environment. • Diagnostics interfaces with other services such as the buildings, HVAC, electrical services, cooling water, vacuum, liquid and gas distribution. • All the interfaces with these services are identified and defined. • Buildings services for diagnostics, such as penetrations, local shielding, embedment and temperature control are discussed. -- Abstract: Extensive diagnostics systems will be installed on the ITER machine to provide the measurements necessary to control, evaluate and optimize plasma performance in ITER and to further the understanding of plasma physics. These include measurements of temperature, density, impurity concentration, and particle and energy confinement times. ITER diagnostic systems extend from the center of the Tokamak to the various diagnostic areas, where they are controlled and acquired data is processed. This mainly includes the areas such as ports, port cells, gallery, diagnostics enclosures and cubicle areas. The diagnostics port plugs encloses the front end of the diagnostic systems and the diagnostics building houses the diagnostics equipment, instrumentation and control cubicles. There are several systems providing services to diagnostics. These mainly include ITER buildings, electrical power services, cooling water services, Heating Ventilation and Air Conditioning (HVAC), vacuum services, liquid and gas distribution services, cable engineering, de-tritiation systems, control cubicles, etc. Requirements of these service systems have to be defined, even though many of the diagnostics are at an early stage of development. It is a real challenge to define and to design diagnostics systems considering the constraints imposed by these service systems. This paper summarizes the provision of these services to the individual diagnostics and diagnostics areas

  2. Iterative guided image fusion

    Directory of Open Access Journals (Sweden)

    Alexander Toet

    2016-08-01

    Full Text Available We propose a multi-scale image fusion scheme based on guided filtering. Guided filtering can effectively reduce noise while preserving detail boundaries. When applied in an iterative mode, guided filtering selectively eliminates small scale details while restoring larger scale edges. The proposed multi-scale image fusion scheme achieves spatial consistency by using guided filtering both at the decomposition and at the recombination stage of the multi-scale fusion process. First, size-selective iterative guided filtering is applied to decompose the source images into approximation and residual layers at multiple spatial scales. Then, frequency-tuned filtering is used to compute saliency maps at successive spatial scales. Next, at each spatial scale binary weighting maps are obtained as the pixelwise maximum of corresponding source saliency maps. Guided filtering of the binary weighting maps with their corresponding source images as guidance images serves to reduce noise and to restore spatial consistency. The final fused image is obtained as the weighted recombination of the individual residual layers and the mean of the approximation layers at the coarsest spatial scale. Application to multiband visual (intensified and thermal infrared imagery demonstrates that the proposed method obtains state-of-the-art performance for the fusion of multispectral nightvision images. The method has a simple implementation and is computationally efficient.

  3. Runaway electrons and ITER

    Science.gov (United States)

    Boozer, Allen

    2016-10-01

    ITER planning for avoiding runaway damage depends on magnetic surface breakup in fast relaxations. These arise in thermal quenches and in the spreading of impurities from massive gas injection or shattered pellets. Surface breakup would prevent a runaway to relativistic energies were it not for non-intercepting flux tubes, which contain magnetic field lines that do not intercept the walls. Such tubes persist near the magnetic axis and in the cores of islands but must dissipate before any confining surfaces re-form. Otherwise, a highly dangerous situation arises. Electrons that were trapped and accelerated in these flux tubes can fill a large volume of stochastic field lines and serve as a seed for the transfer of the full plasma current to runaways. If the outer confining surfaces are punctured, as by a drift into the wall, then the full runaway inventory will be lost in a short pulse along a narrow flux tube. Although not part of ITER planning, currents induced in the walls by the fast magnetic relaxation could be used to passively prevent outer surfaces re-forming. If magnetic surface breakup can be avoided during impurity injection, the plasma current could be terminated in tens of milliseconds by plasma cooling with no danger of runaway. Support by DoE Office of Fusion Energy Science Grant De-FG02-03ER54696.

  4. Application of ECH to the Study of Transport in ITER Baseline Scenario-like Discharges in DIII-D

    Directory of Open Access Journals (Sweden)

    Pinsker R.I.

    2015-01-01

    Full Text Available Recent DIII-D experiments in the ITER Baseline Scenario (IBS have shown strong increases in fluctuations and correlated reduction of confinement associated with entering the electron-heating-dominated regime with strong electron cyclotron heating (ECH. The addition of 3.2 MW of 110 GHz EC power deposited at ρ∼0.42 to IBS discharges with ∼3 MW of neutral beam injection causes large increases in low-k and medium-k turbulent density fluctuations observed with Doppler backscatter (DBS, beam emission spectroscopy (BES and phase-contrast imaging (PCI diagnostics, correlated with decreases in the energy, particle, and momentum confinement times. Power balance calculations show the electron heat diffusivity χe increases significantly in the mid-radius region 0.4<ρ<0.8, which is roughly the same region where the DBS and BES diagnostics show the increases in turbulent density fluctuations. Confinement of angular momentum is also reduced during ECH. Studies with the TGYRO transport solver show that the model of turbulent transport embodied in the TGLF code quantitatively reproduces the measured transport in both the neutral beam (NB-only and in the NB plus EC cases. A simple model of the decrease in toroidal rotation with EC power is set forth, which exhibits a bifurcation in the rotational state of the discharge.

  5. Application of ECH to the Study of Transport in ITER Baseline Scenario-like Discharges in DIII-D

    Science.gov (United States)

    Pinsker, R. I.; Austin, M. E.; Ernst, D. R.; Garofalo, A. M.; Grierson, B. A.; Hosea, J. C.; Luce, T. C.; Marinoni, A.; McKee, G. R.; Perkins, R. J.; Petty, C. C.; Porkolab, M.; Rost, J. C.; Schmitz, L.; Solomon, W. M.; Taylor, G.; Turco, F.

    2015-03-01

    Recent DIII-D experiments in the ITER Baseline Scenario (IBS) have shown strong increases in fluctuations and correlated reduction of confinement associated with entering the electron-heating-dominated regime with strong electron cyclotron heating (ECH). The addition of 3.2 MW of 110 GHz EC power deposited at ρ˜0.42 to IBS discharges with ˜3 MW of neutral beam injection causes large increases in low-k and medium-k turbulent density fluctuations observed with Doppler backscatter (DBS), beam emission spectroscopy (BES) and phase-contrast imaging (PCI) diagnostics, correlated with decreases in the energy, particle, and momentum confinement times. Power balance calculations show the electron heat diffusivity χe increases significantly in the mid-radius region 0.4<ρ<0.8, which is roughly the same region where the DBS and BES diagnostics show the increases in turbulent density fluctuations. Confinement of angular momentum is also reduced during ECH. Studies with the TGYRO transport solver show that the model of turbulent transport embodied in the TGLF code quantitatively reproduces the measured transport in both the neutral beam (NB)-only and in the NB plus EC cases. A simple model of the decrease in toroidal rotation with EC power is set forth, which exhibits a bifurcation in the rotational state of the discharge.

  6. What is Neutrality?

    NARCIS (Netherlands)

    Pierik, R.; van der Burg, W.

    2014-01-01

    This paper reinvestigates the question of liberal neutrality. We contend that current liberal discussions have been dominated - if not hijacked - by one particular interpretation of what neutrality could imply, namely, exclusive neutrality, that aims to exclude religious and cultural expressions

  7. Burn control of an ITER-like fusion reactor using fuzzy logic

    Science.gov (United States)

    Garcia-Amador, A. Sair; Martinell, Julio J.

    2016-10-01

    The fuel burn in a fusion reactor has to be kept at a nearly constant rate in order to have a steady power exhaust. Here, we develop a control system based on a fuzzy logic controller in order that adjusts external parameters to keep the plasma temperature and density at the design values of a reactor of the characteristics of ITER. The control parameters chosen are the D-T refueling rate, the auxiliary heating power and a neutral helium beam. We use a fuzzy controller of the Mamdani type that uses a number of membership functions appropriate to produce a response to parameter deviations that minimizes the response time. The inference rules are determined in a way to provide stabilization to all perturbations of the temperature, density and alpha particle fraction. The dynamical response of the reactor is simulated with a 0D model that uses confinement times provided by the ITER scaling. We show that the system is feedback stabilized for a large range of parameters around the nominal values. The recovery time after a departure from the steady values is of the order of one second. We compare the results with another control system based on neural networks that was developed previously. Funded by projects PAPIIT IN109115 and Conacyt 152905.

  8. Design issues of the High Voltage platform and feedthrough for the ITER NBI Ion Source

    Energy Technology Data Exchange (ETDEWEB)

    Boldrin, M. [Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, I-35127 Padova (Italy)], E-mail: marco.boldrin@igi.cnr.it; Palma, M. Dalla; Milani, F. [Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, I-35127 Padova (Italy)

    2009-06-15

    In the ITER heating Neutral Beam Injector (NBI), a High Voltage air-insulated platform (named High Voltage Deck, HVD) will be installed to host the Ion Source and Extractor Power supply system and associated diagnostics referred to -1 MV DC potential. All power and control cables are routed from the HVD via a feedthrough (HV bushing) to the gas insulated transmission line which feeds the Injector. The paper focuses on insulation and mechanical issues for both HVD and HV bushing which are very special components, far from the present industrial standards as far as voltage (-1 MV DC) and dimensions are concerned. For this purpose, a preliminary design of the HVD has been carried out as concerns the mechanical structure and external shield. Then, the structure has been verified with a seismic analysis applying the seismic load excitation specified for the ITER construction site (Cadarache) and carrying out verifications according to relevant international standards. As regards the HV bushing design, proposals for the complex inner conductor structure and for interfaces to the HVD and transmission line are outlined; alternative installation layouts (aside or underneath the HVD) are compared from both mechanical and electrical points of view.

  9. Active beam spectroscopy for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Von Hellermann, M.; Giroud, C.; Jaspers, R. [Association Euratom-Fom, FOM Institute for Plasma Physics Rijnhuizen, Trilateral Euregio Cluster (Netherlands); Hawkes, N.C.; Mullane, M.O.; Zastrow, K.D. [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon (United Kingdom); Krasilnikov, A.; Tugarinov, S. [SRC RF TRINITI, Troitsk, Moscow region (Russian Federation); Lotte, P. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; McKee, G. [Wisconsin Univ., Madison, WI (United States); Malaquias, A. [Associacao EURATOM/IST, Instituto Superior Tecnico, Lisboa (Portugal); Rachlew, E. [Kungliga Tekniska Hoegskolan (KTH), Stockholm(Sweden)

    2003-07-01

    The latest status of 'Active Beam' related spectroscopy aspects as part of the ITER diagnostic scenario is presented. A key issue of the proposed scheme is based on the concept that in order to achieve the ultimate goal of global data consistency, all particles involved, that is, intrinsic and seeded impurity ions as well as helium ash ions and bulk plasma ions and also the plasma background data (e.g. magnetic and electric fields, electron density and temperature profiles) need to be addressed. A further sensible step in this direction is the decision of exploiting both a dedicated low-energy, low-power diagnostic beam (DNB, 2.2 MW 100 keV/amu) as well as the high-power, high-energy heating beams (HNB, 17 MW 500 keV/amu) for maximum diagnostic information. The authors report some new aspects referring to the use of DNB for motional Stark effect (MSE) where the main idea is to treat both beams (HNB and DNB) as potential diagnostic tools with complementary roles. The equatorial ports for the DNB promise excellent spatial resolution, however, the angles are less favourable for a polarimetric MSE exploitation. HNB can be used as probe beam for diagnosing slowing-down fusion alpha with a birth energy of 3,5 MeV.

  10. Shock acceleration in partially neutral plasmas

    CERN Document Server

    Morlino, G; Blasi, P; Caprioli, D

    2010-01-01

    We present the non-linear theory of shock acceleration applied to SNRs expanding into partially neutral plasma. Using this theory we show how the Balmer lines detected from young SNRs can be used to test the efficiency of shocks in the production of cosmic rays. In particular we investigate the effect of charge-exchange between protons and neutral hydrogen occurring in the precursor formed ahead of the shock. In this precursor the CR pressure accelerate the ionized component of the plasma and a relative velocity between protons and neutral hydrogen is established. On the other hand the charge-exchange process tends to equilibrate ions and neutrals resulting in the heating of both components. We show that even when the shock converts only a few per cent of the total bulk kinetic energy into CRs, the heating is efficient enough to produce a detectable broadening of the narrow Balmer lines emitted by the neutral hydrogen.

  11. Ultracold neutral plasmas

    Science.gov (United States)

    Lyon, M.; Rolston, S. L.

    2017-01-01

    By photoionizing samples of laser-cooled atoms with laser light tuned just above the ionization limit, plasmas can be created with electron and ion temperatures below 10 K. These ultracold neutral plasmas have extended the temperature bounds of plasma physics by two orders of magnitude. Table-top experiments, using many of the tools from atomic physics, allow for the study of plasma phenomena in this new regime with independent control over the density and temperature of the plasma through the excitation process. Characteristic of these systems is an inhomogeneous density profile, inherited from the density distribution of the laser-cooled neutral atom sample. Most work has dealt with unconfined plasmas in vacuum, which expand outward at velocities of order 100 m/s, governed by electron pressure, and with lifetimes of order 100 μs, limited by stray electric fields. Using detection of charged particles and optical detection techniques, a wide variety of properties and phenomena have been observed, including expansion dynamics, collective excitations in both the electrons and ions, and collisional properties. Through three-body recombination collisions, the plasmas rapidly form Rydberg atoms, and clouds of cold Rydberg atoms have been observed to spontaneously avalanche ionize to form plasmas. Of particular interest is the possibility of the formation of strongly coupled plasmas, where Coulomb forces dominate thermal motion and correlations become important. The strongest impediment to strong coupling is disorder-induced heating, a process in which Coulomb energy from an initially disordered sample is converted into thermal energy. This restricts electrons to a weakly coupled regime and leaves the ions barely within the strongly coupled regime. This review will give an overview of the field of ultracold neutral plasmas, from its inception in 1999 to current work, including efforts to increase strong coupling and effects on plasma properties due to strong coupling.

  12. Physics of thermo-nuclear fusion and the ITER project; La physique de la fusion thermonucleaire et le projet ITER

    Energy Technology Data Exchange (ETDEWEB)

    Garin, P. [CEA Cadarache, Dept. de Recherches sur la Fusion Controlee - DRFC, 13 - Saint-Paul-lez-Durance (France)

    2003-01-01

    This document gathers the slides of the 6 contributions to the workshop 'the physics of thermo-nuclear fusion and the ITER project': 1) the feasibility of magnetic confinement and the issue of heat recovery, 2) heating and current generation in tokamaks, 3) the physics of wall-plasma interaction, 4) recent results at JET, 5) inertial confinement and fast ignition, and 6) the technology of fusion machines based on magnetic confinement. This document presents the principles of thermo-nuclear fusion machines and gives a lot of technical information about JET, Tore-Supra and ITER.

  13. Wide-angle ITER-prototype tangential infrared and visible viewing system for DIII-D.

    Science.gov (United States)

    Lasnier, C J; Allen, S L; Ellis, R E; Fenstermacher, M E; McLean, A G; Meyer, W H; Morris, K; Seppala, L G; Crabtree, K; Van Zeeland, M A

    2014-11-01

    An imaging system with a wide-angle tangential view of the full poloidal cross-section of the tokamak in simultaneous infrared and visible light has been installed on DIII-D. The optical train includes three polished stainless steel mirrors in vacuum, which view the tokamak through an aperture in the first mirror, similar to the design concept proposed for ITER. A dichroic beam splitter outside the vacuum separates visible and infrared (IR) light. Spatial calibration is accomplished by warping a CAD-rendered image to align with landmarks in a data image. The IR camera provides scrape-off layer heat flux profile deposition features in diverted and inner-wall-limited plasmas, such as heat flux reduction in pumped radiative divertor shots. Demonstration of the system to date includes observation of fast-ion losses to the outer wall during neutral beam injection, and shows reduced peak wall heat loading with disruption mitigation by injection of a massive gas puff.

  14. Iterated crowdsourcing dilemma game

    Science.gov (United States)

    Oishi, Koji; Cebrian, Manuel; Abeliuk, Andres; Masuda, Naoki

    2014-02-01

    The Internet has enabled the emergence of collective problem solving, also known as crowdsourcing, as a viable option for solving complex tasks. However, the openness of crowdsourcing presents a challenge because solutions obtained by it can be sabotaged, stolen, and manipulated at a low cost for the attacker. We extend a previously proposed crowdsourcing dilemma game to an iterated game to address this question. We enumerate pure evolutionarily stable strategies within the class of so-called reactive strategies, i.e., those depending on the last action of the opponent. Among the 4096 possible reactive strategies, we find 16 strategies each of which is stable in some parameter regions. Repeated encounters of the players can improve social welfare when the damage inflicted by an attack and the cost of attack are both small. Under the current framework, repeated interactions do not really ameliorate the crowdsourcing dilemma in a majority of the parameter space.

  15. Comparisons of Predicted Plasma Performance in ITER H-mode Plasmas with Various Mixes of External He

    Energy Technology Data Exchange (ETDEWEB)

    R.V. Budny

    2009-03-20

    Performance in H-mode DT plasmas in ITER with various choices of heating systems are predicted and compared. Combinations of external heating by Negative Ion Neutral Beam Injection (NNBI), Ion Cyclotron Range of Frequencies (ICRF), and Electron Cyclotron Heating (ECH) are assumed. Scans with a range of physics assumptions about boundary temperatures in the edge pedestal, alpha ash transport, and toroidal momentum transport are used to indicate effects of uncertainties. Time-dependent integrated modeling with the PTRANSP code is used to predict profiles of heating, beam torque, and plasma profiles. The GLF23 model is used to predict temperature profiles. Either GLF23 or the assumption of a constant ratio for χø/χi is used to predict toroidal rotation profiles driven by the beam torques. Large differences for the core temperatures are predicted with different mixes of the external heating during the density and current ramp-up phase, but the profiles are similar during the flattop phase. With χø/χi = 0.5, the predicted toroidal rotation is relatively slow and the flow shear implied by the pressure, toroidal rotation, and neoclassical poloidal rotation are not sufficient to cause significant changes in the energy transport or steady state temperature profiles. The GLF23-predicted toroidal rotation is faster by a factor of six, and significant flow shear effects are predicted.

  16. Rocket Experiment For Neutral Upwelling

    Science.gov (United States)

    Kenward, D. R.; Lessard, M.

    2015-12-01

    Observations from the CHAMP satellite from 2004 show relatively small scale heating in the thermosphere. Several different mechanisms have been proposed to explain this phenomenon. The RENU 2 rocket mission includes a suite of 14 instruments which will acquire data to help understand processes involved in neutral upwelling in the cusp. Neutral, ion, and electron measurements will be made to provide an assessment of the upwelling process. SUPERDarn measurements of large- scale Joule heating in the cusp during overflight will also be acquired. Small-scale data which could possibly be associated with Alfvén waves, will be acquired using onboard electric field measurements. In-situ measurement of precipitating electrons and all other measurements will be used in thermodynamic and electrodynamic models for comparison to the observed upwelling.

  17. Vacuum Tight Threaded Junctions (VTTJ): A new solution for reliable heterogeneous junctions in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Agostinetti, P., E-mail: piero.agostinetti@igi.cnr.it; Palma, M. Dalla; Agostini, F. Degli; Marcuzzi, D.; Rizzolo, A.; Rossetto, F.; Sonato, P.; Zaccaria, P.

    2015-10-15

    Highlights: • Heterogeneous junctions represent a critical issue in Nuclear Fusion experiments. • We have developed a new technique for heterogeneous junctions, called VTTJ, whose main advantages are low cost, high reliability and easiness of construction. • The VTTJ junctions have passed all the tests required by ITER for the heterogeneous junctions of the divertor. • Further tests have demonstrated wide margins for operation (up to 700 °C and 500 bar). - Abstract: A new technique, called Vacuum Tight Threaded Junction (VTTJ), has been developed and patented by Consorzio RFX, permitting to obtain low-cost and reliable non-welded junctions, able to maintain vacuum tightness also in heavy loading conditions (high temperature and high mechanical loads). The technique can be applied also if the materials to be joint are not weldable and for heterogeneous junctions (for example, between steel and copper) and has been tested up to 500 bar internal pressure and up to 700 °C, showing excellent leak tightness in vacuum conditions and high mechanical resistance. The main advantages with respect to existing technologies (for example, friction welding and electron beam welding) are an easy construction, a low cost, a precise positioning of the junction and a high repeatability of the process. Due to these advantages, the new technique has been adopted for several components of the SPIDER experiment and it is proposed for ITER, in particular for the ITER Heat and Current Drive Neutral Beam Injector and for its prototype, the MITICA experiment, to be tested at Consorzio RFX. This paper gives a detailed description of the VTTJ technique, of the samples manufactured and of the qualification tests that have been carried out so far.

  18. LT-IIb(T13I, a non-toxic type II heat-labile enterotoxin, augments the capacity of a ricin toxin subunit vaccine to evoke neutralizing antibodies and protective immunity.

    Directory of Open Access Journals (Sweden)

    Christopher J Greene

    Full Text Available Currently, there is a shortage of adjuvants that can be employed with protein subunit vaccines to enhance protection against biological threats. LT-IIb(T13I is an engineered nontoxic derivative of LT-IIb, a member of the type II subfamily of heat labile enterotoxins expressed by Escherichia coli, that possesses potent mucosal adjuvant properties. In this study we evaluated the capacity of LT-IIb(T13I to augment the potency of RiVax, a recombinant ricin toxin A subunit vaccine, when co-administered to mice via the intradermal (i.d. and intranasal (i.n. routes. We report that co-administration of RiVax with LT-IIb(T13I by the i.d. route enhanced the levels of RiVax-specific serum IgG antibodies (Ab and elevated the ratio of ricin-neutralizing to non-neutralizing Ab, as compared to RiVax alone. Protection against a lethal ricin challenge was also augmented by LT-IIb(T13I. While local inflammatory responses elicited by LT-IIb(T13I were comparable to those elicited by aluminum salts (Imject®, LT-IIb(T13I was more effective than aluminum salts at augmenting production of RiVax-specific serum IgG. Finally, i.n. administration of RiVax with LT-IIb(T13I also increased levels of RiVax-specific serum and mucosal Ab and enhanced protection against ricin challenge. Collectively, these data highlight the potential of LT-IIb(T13I as an effective next-generation i.d., or possibly i.n. adjuvant for enhancing the immunogenicity of subunit vaccines for biodefense.

  19. EXISTENCE OF SOLUTION TO NONLINEAR SECOND ORDER NEUTRAL STOCHASTIC DIFFERENTIAL EQUATIONS WITH DELAY

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    This paper is concerned with nonlinear second order neutral stochastic differential equations with delay in a Hilbert space. Sufficient conditions for the existence of solution to the system are obtained by Picard iterations.

  20. Carbon fiber composites application in ITER plasma facing components

    Science.gov (United States)

    Barabash, V.; Akiba, M.; Bonal, J. P.; Federici, G.; Matera, R.; Nakamura, K.; Pacher, H. D.; Rödig, M.; Vieider, G.; Wu, C. H.

    1998-10-01

    Carbon Fiber Composites (CFCs) are one of the candidate armour materials for the plasma facing components of the International Thermonuclear Experimental Reactor (ITER). For the present reference design, CFC has been selected as armour for the divertor target near the plasma strike point mainly because of unique resistance to high normal and off-normal heat loads. It does not melt under disruptions and might have higher erosion lifetime in comparison with other possible armour materials. Issues related to CFC application in ITER are described in this paper. They include erosion lifetime, tritium codeposition with eroded material and possible methods for the removal of the codeposited layers, neutron irradiation effect, development of joining technologies with heat sink materials, and thermomechanical performance. The status of the development of new advanced CFCs for ITER application is also described. Finally, the remaining R&D needs are critically discussed.

  1. Analysis of active and passive magnetic field reduction systems (MFRS) of the ITER NBI

    Energy Technology Data Exchange (ETDEWEB)

    Roccella, M. [L.T. Calcoli S.a.S., Piazza Prinetti 26/B, Merate (Lecco) (Italy)], E-mail: roccella@ltcalcoli.it; Lucca, F.; Roccella, R. [L.T. Calcoli S.a.S., Piazza Prinetti 26/B, Merate (Lecco) (Italy); Pizzuto, A.; Ramogida, G. [Associazione EURATOM sulla Fusione - ENEA Frascati (Italy); Portone, A.; Tanga, A. [ITER EFDA (Italy); Formisano, A.; Martone, R. [CREATE Napoli (Italy)

    2007-10-15

    In ITER two heating (HNBI) and one diagnostic neutral beam injectors (DNBI) are foreseen. Inside these components there are very stringent limits on the magnetic field (the flux density must be below some G along the ion path and below 20 G in the neutralizing regions). To achieve these performances in an environment with high stray field due to the plasma and the poloidal field coils (PFC), both passive and active shielding systems have been foreseen. The present design of the magnetic field reduction systems (MFRS) is made of seven active coils and of a box surrounding the NBI region, consisting of ferromagnetic plates. The electromagnetic analyses of the effectiveness of these shields have been performed by a 3D FEM model using ANSYS code for the HNBI. The ANSYS models of the ferromagnetic box and of the active coils are fully parametric, thus any size change of the ferromagnetic box and coils (linear dimension or thickness) preserving the overall box shape could be easily reproduced by simply changing some parameter in the model.

  2. Neutral recirculation—the key to control of divertor operation

    Science.gov (United States)

    Kukushkin, A. S.; Pacher, H. D.

    2016-12-01

    Interaction of the plasma with neutral gas in the divertor affects virtually all aspects of divertor functionality (power loading of the targets, pumping and fuelling, sustaining the operational conditions of the core plasma). In the course of ITER design development, this interaction has been the subject of intense modelling analysis, supported by experiments on various tokamaks. Neutral gas puffing is found to be the most effective means of divertor control. The results of those studies are summarized and assessed in the paper.

  3. Preparation to manufacturing of ITER plasma facing components in Russia

    Energy Technology Data Exchange (ETDEWEB)

    Mazul, I.V., E-mail: mazuliv@niiefa.spb.su [Efremov Institute, St. Petersburg, 196641 (Russian Federation); Belyakov, V.A.; Giniatulin, R.N.; Gervash, A.A.; Kuznetsov, V.E.; Makhankov, A.N. [Efremov Institute, St. Petersburg, 196641 (Russian Federation); Sizenev, V.S. [Corporation ' Kompozit' , Korolev, 141070 (Russian Federation)

    2011-10-15

    The preparation of the procurement activities for the ITER plasma-facing-components (PFC) is currently well underway. Three ITER procurement packages associated with PFCs are currently allocated to the Russian Federation (RF): delivery of the central assembly of the divertor (dome and reflector plates assemblies), delivery of 40% of the first-wall (FW) panels and high heat flux testing of divertor components during the qualification and subsequent manufacturing phases. The results of the qualification process for these tasks undertaken by RF industry are presented. Qualification mockups of the dome divertor structure were successfully manufactured in accordance with the ITER specifications and tested at heat fluxes exceeding operational ones. The maturity and reliability of the proposed design and manufacturing technologies, proposed by RF industry, was therefore demonstrated. To confirm the manufacturing readiness of technologies proposed for the fabrication of the ITER first wall, three qualification mockups were fabricated. Two were heat flux tested in two facilities abroad. In addition to launching the qualification process, the PFC team at Efremov Institute is preparing the industrial facilities for serial production of above mentioned components. A brief description of such facilities is presented in this paper, together with the manufacturing technologies to be used. Two electron beam facilities (Tsefey and IDTF) for various high heat flux testing of PFC components are also described.

  4. Prospects for measuring the fuel ion ratio in burning ITER plasmas using a DT neutron emission spectrometer

    Science.gov (United States)

    Hellesen, C.; Skiba, M.; Dzysiuk, N.; Weiszflog, M.; Hjalmarsson, A.; Ericsson, G.; Conroy, S.; Andersson-Sundén, E.; Eriksson, J.; Binda, F.

    2014-11-01

    The fuel ion ratio nt/nd is an essential parameter for plasma control in fusion reactor relevant applications, since maximum fusion power is attained when equal amounts of tritium (T) and deuterium (D) are present in the plasma, i.e., nt/nd = 1.0. For neutral beam heated plasmas, this parameter can be measured using a single neutron spectrometer, as has been shown for tritium concentrations up to 90%, using data obtained with the MPR (Magnetic Proton Recoil) spectrometer during a DT experimental campaign at the Joint European Torus in 1997. In this paper, we evaluate the demands that a DT spectrometer has to fulfill to be able to determine nt/nd with a relative error below 20%, as is required for such measurements at ITER. The assessment shows that a back-scattering time-of-flight design is a promising concept for spectroscopy of 14 MeV DT emission neutrons.

  5. Fusion Neutron Flux Monitor for ITER

    Institute of Scientific and Technical Information of China (English)

    YANG Jinwei; YANG Qingwei; XIAO Gongshan; ZHANG Wei; SONG Xianying; LI Xu

    2008-01-01

    Neutron flux monitor (NFM) as an important diagnostic sub-system in ITER (international thermonuclear experimental reactor) provides a global neutron source intensity, fusion power and neutron flux in real time. Three types of neutron flux monitor assemblies with different sensitivities and shielding materials have been designed. Through MCNP (Mante-Carlo neutral particle transport code) calculations, this extended system of NFM can detect the neutron flux in a range of 104 n/(cm2·s) to 1014 n/(cm2·s). It is capable of providing accurate neutron yield measurements for all operational modes encountered in the ITER experiments including the in-situ calibration. Combining both the counting mode and Campbelling (MSV; Mean Square Voltage) mode in the signal processing units, the requirement of the dynamic range (107) for these NFMs and time resolution (1 ms) can be met. Based on a uncertainty analysis, the estimated absolute measurement accuracies of the total fusion neutron yield can reach the required 10% level in both the early stage of the DD-phase and the full power DT operation mode. In the advanced DD-phase, the absolute measurement accuracy would be better than 20%.

  6. Modelling the neutralisation process in neutral beam injectors

    OpenAIRE

    Fitzgerald, Niall J.

    2009-01-01

    High power neutral beams currently play an important role in heating, fuelling and diagnosing magnetically confined thermonuclear fusion plasmas. At the Joint European Torus (JET) in Oxfordshire, England, the formation of such a beam involves passing a positive ion beam through a neutral gas target wherein beam electron-capture collisions result in a neutral beam component. The subsequent beam injection into the fusion plasma requires the sole use of this neutral component, since the charged ...

  7. Predictions of H-mode performance in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Budny, R. V.; Andre, R.; Bateman, G.; Halpern, F.; Kessel, C. E.; Kritz, A.; McCune, D.

    2008-03-03

    Time-dependent integrated predictive modeling is carried out using the PTRANSP code to predict fusion power and parameters such as alpha particle density and pressure in ITER H-mode plasmas. Auxiliary heating by negative ion neutral beam injection and ion cyclotron heating of He3 minority ions are modeled, and the GLF23 transport model is used in the prediction of the evolution of plasma temperature profiles. Effects of beam steering, beam torque, plasma rotation, beam current drive, pedestal temperatures, sawtooth oscillations, magnetic diffusion, and accumulation of He ash are treated self-consistently. Variations in assumptions associated with physics uncertainties for standard base-line DT H-mode plasmas (with Ip=15 MA, BTF=5.3 T, and Greenwald fraction=0.86) lead to a range of predictions for DT fusion power PDT and quasi-steady state fusion QDT (≡ PDT/Paux). Typical predictions assuming Paux = 50-53 MW yield PDT = 250- 720 MW and QDT = 5 - 14. In some cases where Paux is ramped down or shut off after initial flat-top conditions, quasi-steady QDT can be considerably higher, even infinite. Adverse physics assumptions such as existence of an inward pinch of the helium ash and an ash recycling coefficient approaching unity lead to very low values for PDT. Alternative scenarios with different heating and reduced performance regimes are also considered including plasmas with only H or D isotopes, DT plasmas with toroidal field reduced 10 or 20%, and discharges with reduced beam voltage. In full-performance D-only discharges, tritium burn-up is predicted to generate central tritium densities up to 1016/m3 and DT neutron rates up to 5×1016/s, compared with the DD neutron rates of 6×1017/s. Predictions with the toroidal field reduced 10 or 20% below the planned 5.3 T and keeping the same q98, Greenwald fraction, and Βη indicate that the fusion yield PDT and QDT will be lower by about a factor of two (scaling as B3.5).

  8. Experimental studies of ITER demonstration discharges

    Science.gov (United States)

    Sips, A. C. C.; Casper, T. A.; Doyle, E. J.; Giruzzi, G.; Gribov, Y.; Hobirk, J.; Hogeweij, G. M. D.; Horton, L. D.; Hubbard, A. E.; Hutchinson, I.; Ide, S.; Isayama, A.; Imbeaux, F.; Jackson, G. L.; Kamada, Y.; Kessel, C.; Kochl, F.; Lomas, P.; Litaudon, X.; Luce, T. C.; Marmar, E.; Mattei, M.; Nunes, I.; Oyama, N.; Parail, V.; Portone, A.; Saibene, G.; Sartori, R.; Stober, J. K.; Suzuki, T.; Wolfe, S. M.; C-Mod Team; ASDEX Upgrade Team; DIII-D Team; JET EFDA Contributors

    2009-08-01

    Key parts of the ITER scenarios are determined by the capability of the proposed poloidal field (PF) coil set. They include the plasma breakdown at low loop voltage, the current rise phase, the performance during the flat top (FT) phase and a ramp down of the plasma. The ITER discharge evolution has been verified in dedicated experiments. New data are obtained from C-Mod, ASDEX Upgrade, DIII-D, JT-60U and JET. Results show that breakdown for Eaxis unassisted (ohmic) for large devices like JET and attainable in devices with a capability of using ECRH assist. For the current ramp up, good control of the plasma inductance is obtained using a full bore plasma shape with early X-point formation. This allows optimization of the flux usage from the PF set. Additional heating keeps li(3) < 0.85 during the ramp up to q95 = 3. A rise phase with an H-mode transition is capable of achieving li(3) < 0.7 at the start of the FT. Operation of the H-mode reference scenario at q95 ~ 3 and the hybrid scenario at q95 = 4-4.5 during the FT phase is documented, providing data for the li (3) evolution after the H-mode transition and the li (3) evolution after a back-transition to L-mode. During the ITER ramp down it is important to remain diverted and to reduce the elongation. The inductance could be kept <=1.2 during the first half of the current decay, using a slow Ip ramp down, but still consuming flux from the transformer. Alternatively, the discharges can be kept in H-mode during most of the ramp down, requiring significant amounts of additional heating.

  9. Some geometrical iteration methods for nonlinear equations

    Institute of Scientific and Technical Information of China (English)

    LU Xing-jiang; QIAN Chun

    2008-01-01

    This paper describes geometrical essentials of some iteration methods (e.g. Newton iteration,secant line method,etc.) for solving nonlinear equations and advances some geomet-rical methods of iteration that are flexible and efficient.

  10. Neutral Operator and Neutral Differential Equation

    Directory of Open Access Journals (Sweden)

    Jingli Ren

    2011-01-01

    Full Text Available In this paper, we discuss the properties of the neutral operator (Ax(t=x(t−cx(t−δ(t, and by applying coincidence degree theory and fixed point index theory, we obtain sufficient conditions for the existence, multiplicity, and nonexistence of (positive periodic solutions to two kinds of second-order differential equations with the prescribed neutral operator.

  11. PICARD ITERATION FOR NONSMOOTH EQUATIONS

    Institute of Scientific and Technical Information of China (English)

    Song-bai Sheng; Hui-fu Xu

    2001-01-01

    This paper presents an analysis of the generalized Newton method, approximate Newton methods, and splitting methods for solving nonsmooth equations from Picard iteration viewpoint. It is proved that the radius of the weak Jacobian (RGJ) of Picard iteration function is equal to its least Lipschitz constant. Linear convergence or superlinear convergence results can be obtained provided that RGJ of the Picard iteration function at a solution point is less than one or equal to zero. As for applications, it is pointed out that the approximate Newton methods, the generalized Newton method for piecewise C1problems and splitting methods can be explained uniformly with the same viewpoint.

  12. Remote maintenance development for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Tada, Eisuke [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Shibanuma, Kiyoshi

    1998-04-01

    This paper describes the overall ITER remote maintenance design concept developed mainly for in-vessel components such as diverters and blankets, and outlines the ITER R and D program to develop remote handling equipment and radiation hard components. Reactor structures inside the ITER cryostat must be maintained remotely due to DT operation, making remote handling technology basic to reactor design. The overall maintenance scenario and design concepts have been developed, and maintenance design feasibility, including fabrication and testing of full-scale in-vessel remote maintenance handling equipment and tool, is being verified. (author)

  13. The first fusion reactor: ITER

    Science.gov (United States)

    Campbell, D. J.

    2016-11-01

    Established by the signature of the ITER Agreement in November 2006 and currently under construction at St Paul-lez-Durance in southern France, the ITER project [1,2] involves the European Union (including Switzerland), China, India, Japan, the Russian Federation, South Korea and the United States. ITER (`the way' in Latin) is a critical step in the development of fusion energy. Its role is to provide an integrated demonstration of the physics and technology required for a fusion power plant based on magnetic confinement.

  14. Iterative optimization in inverse problems

    CERN Document Server

    Byrne, Charles L

    2014-01-01

    Iterative Optimization in Inverse Problems brings together a number of important iterative algorithms for medical imaging, optimization, and statistical estimation. It incorporates recent work that has not appeared in other books and draws on the author's considerable research in the field, including his recently developed class of SUMMA algorithms. Related to sequential unconstrained minimization methods, the SUMMA class includes a wide range of iterative algorithms well known to researchers in various areas, such as statistics and image processing. Organizing the topics from general to more

  15. Experimental Simulation of the Behaviour of Diagnostic First Mirrors Fabricated of Different Metals for ITER Conditions

    OpenAIRE

    Voitsenya, V.; Bardamid, A. F.; Donne, A. J. H.

    2016-01-01

    In the experimental fusion reactor ITER, the plasma-facing component of each optical and/or laser diagnostic needs to be based on reflective optics with at least one mirror (first mirror) facing the thermonuclear plasma. The different kinds of radiation emanating from the burning plasma (neutrons, neutral atoms, electromagnetic radiation) create hostile operating conditions for the first mirrors. Therefore, a special program has been set up under the ITER framework aimed at solving the first ...

  16. Nuclear Analyses of Indian LLCB Test Blanket System in ITER

    Science.gov (United States)

    Swami, H. L.; Shaw, A. K.; Danani, C.; Chaudhuri, Paritosh

    2017-04-01

    Heading towards the Nuclear Fusion Reactor Program, India is developing Lead Lithium Ceramic Breeder (LLCB) tritium breeding blanket for its future fusion Reactor. A mock-up of the LLCB blanket is proposed to be tested in ITER equatorial port no.2, to ensure the overall performance of blanket in reactor relevant nuclear fusion environment. Nuclear analyses play an important role in LLCB Test Blanket System design & development. It is required for tritium breeding estimation, thermal-hydraulic design, coolants process design, radioactive waste management, equipment maintenance & replacement strategies and nuclear safety. The nuclear behaviour of LLCB test blanket module in ITER is predicated in terms of nuclear responses such as tritium production, nuclear heating, neutron fluxes and radiation damages. Radiation shielding capability of LLCB TBS inside and outside bio-shield was also assessed to fulfill ITER shielding requirements. In order to supports the rad-waste and safety assessment, nuclear activation analyses were carried out and radioactivity data were generated for LLCB TBS components. Nuclear analyses of LLCB TBS are performed using ITER recommended nuclear analyses codes (i.e. MCNP, EASY), nuclear cross section data libraries (i.e. FENDL 2.1, EAF) and neutronic model (ITER C-lite v.l). The paper describes a comprehensive nuclear performance of LLCB TBS in ITER.

  17. Progress of ITER full tungsten divertor technology qualification in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Ezato, K., E-mail: ezato.koichiro@jaea.go.jp [Japan Atomic Energy Agency, 801-1, Mukoyma, Naka-shi, Ibaraki (Japan); Suzuki, S.; Seki, Y.; Mohri, K.; Yokoyama, K. [Japan Atomic Energy Agency, 801-1, Mukoyma, Naka-shi, Ibaraki (Japan); Escourbiac, F.; Hirai, T. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Kuznetcov, V. [NIIEFA, 3 doroga na Metallostroy, Metallostroy, St. Petersburg 196641 (Russian Federation)

    2015-10-15

    Highlights: • JAEA has demonstrated tungsten monoblock technology for ITER divertor that needs to withstand the repetitive heat load as high as 20 MW/m{sup 2}. This includes as follows; • Bonding technologies between W and Cu interlayer, and between Cu interlayer and CuCrZr tube. • Non-destructive examination techniques, especially, ultrasonic testing method, and. • Load carrying capability of W monoblock attachment to support structure of ITER divertor. - Abstract: Japan Atomic Energy Agency (JAEA) is in progress for technology qualification toward full-tungsten (W) ITER divertor outer vertical target (OVT), especially, tungsten monoblock technology that needs to withstand the repetitive heat load as high as 20 MW/m{sup 2}. To demonstrate the armor heat sink bonding technology and heat removal capability, 6 small-scale W monoblock mock-ups manufactured by different bonding technologies using different W materials in addition to 4 full-scale prototype plasma-facing units (PFUs). After non-destructive test, the W components were tested under high heat flux (HHF) in ITER Divertor Test Facility (IDTF) at NIIEFA. Consequently, all of the W monoblocks endured the repetitive heat load at 20 MW/m{sup 2} for 1000 cycles (requirements 20 MW/m{sup 2} for 300 cycles) without any failure. In addition to the armor to heat sink joints, the load carrying capability test on the W monoblock with a leg attachment was carried out. In uniaxial tensile test, all of the W monoblock attachments with different bonding technologies such as brazing and HIPping withstand the tensile load exceeding 20 kN that is the value more than twice the design value. The failures occurred at the leg attachments or the W monoblocks, rather than the bonding interface of the W monoblocks to the leg attachment.

  18. Rollout sampling approximate policy iteration

    NARCIS (Netherlands)

    Dimitrakakis, C.; Lagoudakis, M.G.

    2008-01-01

    Several researchers have recently investigated the connection between reinforcement learning and classification. We are motivated by proposals of approximate policy iteration schemes without value functions, which focus on policy representation using classifiers and address policy learning as a

  19. Iterative solution of linear systems

    Science.gov (United States)

    Freund, Roland W.; Golub, Gene H.; Nachtigal, Noel M.

    1992-01-01

    Recent advances in the field of iterative methods for solving large linear systems are reviewed. The main focus is on developments in the area of conjugate gradient-type algorithms and Krylov subspace methods for nonHermitian matrices.

  20. Cooperation between CERN and ITER

    CERN Multimedia

    2008-01-01

    CERN and the International Fusion Organisation ITER have just signed a first cooperation agreeement. Kaname Ikeda, the Director-General of the International Fusion Energy Organisation (ITER) (on the right) and Robert Aymar, Director-General of CERN, signing the agreement.The Director-General of the International Fusion Energy Organization, Mr Kaname Ikeda, and CERN Director-General, Robert Aymar, signed a cooperation agreement at a meeting on the Meyrin site on Thursday 6 March. One of the main purposes of this agreement is for CERN to give ITER the benefit of its experience in the field of technology as well as in administrative domains such as finance, procurement, human resources and informatics through the provision of consultancy services. Currently in its start-up phase at its Cadarache site, 70 km from Marseilles (France), ITER will focus its research on the scientific and technical feasibility of using fusion energy as a fu...

  1. Neutral Buoyancy Laboratory (NBL)

    Data.gov (United States)

    Federal Laboratory Consortium — The Neutral Buoyancy Laboratory (NBL) is an astronaut training facility and neutral buoyancy pool operated by NASA and located at the Sonny Carter Training Facility,...

  2. ITER leader to head CERN

    CERN Multimedia

    Feder, Toni

    2003-01-01

    After successfully chairing an external review committee for CERN last year, Robert Aymar will leave ITER to become director general of the European particle physics laboratory rom 2004. Before ITER he also successfully managed the startup or Tore Supra. He will attempt to ensure that the LHC begins operating in 2007 - two years late - and is paid for by 2010 and will also start the planning for life after the LHC (1 page)

  3. ITER diagnostic system: Vacuum interface

    Energy Technology Data Exchange (ETDEWEB)

    Patel, K.M., E-mail: Kaushal.Patel@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Udintsev, V.S.; Hughes, S.; Walker, C.I.; Andrew, P.; Barnsley, R.; Bertalot, L. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Drevon, J.M. [Bertin Technologies, BP 22, 13762 Aix-en Provence cedex 3 (France); Encheva, A. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Kashchuk, Y. [Institution “PROJECT CENTER ITER”, 1, Akademika Kurchatova pl., Moscow (Russian Federation); Maquet, Ph. [Bertin Technologies, BP 22, 13762 Aix-en Provence cedex 3 (France); Pearce, R.; Taylor, N.; Vayakis, G.; Walsh, M.J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France)

    2013-10-15

    Diagnostics play an essential role for the successful operation of the ITER tokamak. They provide the means to observe control and to measure plasma during the operation of ITER tokamak. The components of the diagnostic system in the ITER tokamak will be installed in the vacuum vessel, in the cryostat, in the upper, equatorial and divertor ports, in the divertor cassettes and racks, as well as in various buildings. Diagnostic components that are placed in a high radiation environment are expected to operate for the life of ITER. There are approx. 45 diagnostic systems located on ITER. Some diagnostics incorporate direct or independently pumped extensions to maintain their necessary vacuum conditions. They require a base pressure less than 10{sup −7} Pa, irrespective of plasma operation, and a leak rate of less than 10{sup −10} Pa m{sup 3} s{sup −1}. In all the cases it is essential to maintain the ITER closed fuel cycle. These directly coupled diagnostic systems are an integral part of the ITER vacuum containment and are therefore subject to the same design requirements for tritium and active gas confinement, for all normal and accidental conditions. All the diagnostics, whether or not pumped, incorporate penetration of the vacuum boundary (i.e. window assembly, vacuum feedthrough etc.) and demountable joints. Monitored guard volumes are provided for all elements of the vacuum boundary that are judged to be vulnerable by virtue of their construction, material, load specification etc. Standard arrangements are made for their construction and for the monitoring, evacuating and leak testing of these volumes. Diagnostic systems are incorporated at more than 20 ports on ITER. This paper will describe typical and particular arrangements of pumped diagnostic and monitored guard volume. The status of the diagnostic vacuum systems, which are at the start of their detailed design, will be outlined and the specific features of the vacuum systems in ports and extensions

  4. Modelling of transitions between L- and H-mode in JET high plasma current plasmas and application to ITER scenarios including tungsten behaviour

    Science.gov (United States)

    Koechl, F.; Loarte, A.; Parail, V.; Belo, P.; Brix, M.; Corrigan, G.; Harting, D.; Koskela, T.; Kukushkin, A. S.; Polevoi, A. R.; Romanelli, M.; Saibene, G.; Sartori, R.; Eich, T.; Contributors, JET

    2017-08-01

    The dynamics for the transition from L-mode to a stationary high Q DT H-mode regime in ITER is expected to be qualitatively different to present experiments. Differences may be caused by a low fuelling efficiency of recycling neutrals, that influence the post transition plasma density evolution on the one hand. On the other hand, the effect of the plasma density evolution itself both on the alpha heating power and the edge power flow required to sustain the H-mode confinement itself needs to be considered. This paper presents results of modelling studies of the transition to stationary high Q DT H-mode regime in ITER with the JINTRAC suite of codes, which include optimisation of the plasma density evolution to ensure a robust achievement of high Q DT regimes in ITER on the one hand and the avoidance of tungsten accumulation in this transient phase on the other hand. As a first step, the JINTRAC integrated models have been validated in fully predictive simulations (excluding core momentum transport which is prescribed) against core, pedestal and divertor plasma measurements in JET C-wall experiments for the transition from L-mode to stationary H-mode in partially ITER relevant conditions (highest achievable current and power, H 98,y ~ 1.0, low collisionality, comparable evolution in P net/P L-H, but different ρ *, T i/T e, Mach number and plasma composition compared to ITER expectations). The selection of transport models (core: NCLASS  +  Bohm/gyroBohm in L-mode/GLF23 in H-mode) was determined by a trade-off between model complexity and efficiency. Good agreement between code predictions and measured plasma parameters is obtained if anomalous heat and particle transport in the edge transport barrier are assumed to be reduced at different rates with increasing edge power flow normalised to the H-mode threshold; in particular the increase in edge plasma density is dominated by this edge transport reduction as the calculated neutral influx across the

  5. 1.5 MW RF Load for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Ives, Robert Lawrence [Calabazas Creek Research, Inc., San Mateo, CA (United States); Marsden, David [Calabazas Creek Research, Inc., San Mateo, CA (United States); Collins, George [Calabazas Creek Research, Inc., San Mateo, CA (United States); Karimov, Rasul [Calabazas Creek Research, Inc., San Mateo, CA (United States); Mizuhara, Max [Calabazas Creek Research, Inc., San Mateo, CA (United States); Neilson, Jeffrey [Lexam Research, Redwood City, CA (United States)

    2016-09-01

    Calabazas Creek Research, Inc. developed a 1.5 MW RF load for the ITER fusion research facility currently under construction in France. This program leveraged technology developed in two previous SBIR programs that successfully developed high power RF loads for fusion research applications. This program specifically focused on modifications required by revised technical performance, materials, and assembly specification for ITER. This program implemented an innovative approach to actively distribute the RF power inside the load to avoid excessive heating or arcing associated with constructive interference. The new design implemented materials and assembly changes required to meet specifications. Critical components were built and successfully tested during the program.

  6. RF sources for ITER Ion Cyclotron H and CD system

    Energy Technology Data Exchange (ETDEWEB)

    Kazarian, F., E-mail: fabienne.kazarian@iter.org [ITER Organization, CS 90 046, 13067 Sain-Paul-Les-Durance (France); Beaumont, B.; Arambhadiya, B.; Gassmann, T.; Lamalle, Ph.; Rathi, D. [ITER Organization, CS 90 046, 13067 Sain-Paul-Les-Durance (France); Mukherjee, A.; Ajesh, P.; Machchhar, H.; Patadia, D.; Patel, M.; Rajnish, K.; Singh, R.; Suthar, G.; Trivedi, R. [ITER India, IPR, Bhat, Gandhinagar 382428, Gujarat (India); Kumazawa, R.; Seki, T.; Saito, K.; Kasahara, H.; Mutoh, T. [National Institute for Fusion Science, Toki 509-5292 (Japan)

    2011-10-15

    The Ion Cyclotron Heating and Current Drive (IC H and CD) system for ITER will provide 20 MW to the plasma. The associated Radio Frequency (RF) source system has to be compliant with all operation modes foreseen in that frame. Their specifications are fully described in this paper and constraints on IC RF source components are detailed, in particular concerning the final stage tube of the amplifier. Results of tests performed under a collaborative work at the National Institute for Fusion Science (NIFS) facility are presented. Consequences on the procurement process by ITER India (II) are deduced.

  7. Safety analysis of the US dual coolant liquid lead lithium ITER test blanket module

    Science.gov (United States)

    Merrill, Brad; Reyes, Susana; Sawan, Mohamed; Wong, Clement

    2007-07-01

    The US is proposing a prototype of a dual coolant liquid lead-lithium DEMO blanket concept for testing in the International Thermonuclear Experimental Reactor (ITER) as an ITER test blanket module (TBM). Because safety considerations are an integral part of the design process to ensure that this TBM does not adversely impact the safety of ITER, a safety assessment has been conducted for this TBM and its ancillary systems as requested by the ITER project. Four events were selected by the ITER international team (IT) to address specific reactor safety concerns, such as vaccum vessel (VV) pressurization, confinement building pressure build-up, TBM decay heat removal capability, tritium and activation products release from the TBM system and hydrogen and heat production from chemical reactions. This paper summarizes the results of this safety assessment conducted with the MELCOR computer code.

  8. ITER Central Solenoid Module Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Smith, John [General Atomics, San Diego, CA (United States)

    2016-09-23

    The fabrication of the modules for the ITER Central Solenoid (CS) has started in a dedicated production facility located in Poway, California, USA. The necessary tools have been designed, built, installed, and tested in the facility to enable the start of production. The current schedule has first module fabrication completed in 2017, followed by testing and subsequent shipment to ITER. The Central Solenoid is a key component of the ITER tokamak providing the inductive voltage to initiate and sustain the plasma current and to position and shape the plasma. The design of the CS has been a collaborative effort between the US ITER Project Office (US ITER), the international ITER Organization (IO) and General Atomics (GA). GA’s responsibility includes: completing the fabrication design, developing and qualifying the fabrication processes and tools, and then completing the fabrication of the seven 110 tonne CS modules. The modules will be shipped separately to the ITER site, and then stacked and aligned in the Assembly Hall prior to insertion in the core of the ITER tokamak. A dedicated facility in Poway, California, USA has been established by GA to complete the fabrication of the seven modules. Infrastructure improvements included thick reinforced concrete floors, a diesel generator for backup power, along with, cranes for moving the tooling within the facility. The fabrication process for a single module requires approximately 22 months followed by five months of testing, which includes preliminary electrical testing followed by high current (48.5 kA) tests at 4.7K. The production of the seven modules is completed in a parallel fashion through ten process stations. The process stations have been designed and built with most stations having completed testing and qualification for carrying out the required fabrication processes. The final qualification step for each process station is achieved by the successful production of a prototype coil. Fabrication of the first

  9. Design progress on ITER port plug test facility

    Energy Technology Data Exchange (ETDEWEB)

    Levesy, B., E-mail: bruno.levesy@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Beaumont, B. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Bruno, L.; Cerisier, T. [CNIM, Z.I de Bregaillon, 83507 La Seyne sur Mer (France); Cordier, J.J.; Dammann, A.; Giancarli, L.; Henderson, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Romannikov, A. [ITER Agency, Russian Research Center ' Kurchatov Institute' , pl. Kurchatova I., 1, Moscow 123182 (Russian Federation); Rumyantsev, Y. [JSC ' Cryogenmash' , 143907 Moscow reg., Balashikha (Russian Federation); Udintsev, V.S. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2012-08-15

    To achieve the overall ITER machine availability target, the availability of diagnostics and heating port plugs shall be as high as 99.5%. To fulfill these requirements, it is mandatory to test the port plugs at operating temperature before installation on the machine and after refurbishment. The ITER port plug test facility (PPTF) provides the possibility to test upper and equatorial port plugs before installation on the machine. The port plug test facility is composed of several test stands. These test stands are first used in the domestic agencies and on the ITER Organization site to test the port plugs at the end of manufacturing. Two of these stands are installed later in the ITER hot cell facility to test the port plugs after refurbishment. The port plugs to be tested are the Ion Cyclotron (IC) heating and current drive antennas, Electron Cyclotron (EC) heating and current drive launchers, diagnostics and test blanket modules port plugs. Test stands shall be capable to perform environmental and functional tests. The test stands are composed of one vacuum tank (3.3 m in diameter, 5.6 m long) and the associated heating, vacuum and control systems. The vacuum tank shall achieve an ultimate pressure of 1 Multiplication-Sign 10{sup -5} Pa at 100 Degree-Sign C containing a port plug. The heating system shall provide water at 240 Degree-Sign C and 4.4 MPa to heat up the port plugs. Openings are provided on the back of the vacuum tank to insert probes for the functional tests. This paper describes the tests to be performed on the port plugs and the conceptual design of the port plug test facility. The configuration of the standalone test stands and the integration in the hot cell facility are presented.

  10. Compatibility of ITER scenarios with full tungsten wall in ASDEX Upgrade

    Science.gov (United States)

    Gruber, O.; Sips, A. C. C.; Dux, R.; Eich, T.; Fuchs, J. C.; Herrmann, A.; Kallenbach, A.; Maggi, C. F.; Neu, R.; Pütterich, T.; Schweinzer, J.; Stober, J.; ASDEX Upgrade Team

    2009-11-01

    The transition of ASDEX Upgrade (AUG) from a graphite device to a full tungsten device is demonstrated with a reduction by an order of magnitude in both the carbon deposition and deuterium retention. The tungsten source is dominated by sputtering from intrinsic light impurities, and the tungsten influxes from the outboard limiters are the main source for the plasma. In H-mode discharges, central heating (neutral beams, ECRH) is used to increase turbulent outward transport avoiding tungsten accumulation. ICRH can only be used after boronization as its application otherwise results in large W influxes due to light impurities accelerated by electrical fields at the ICRH antennas. ELMs are important in reducing the inward transport of tungsten in the H-mode edge barrier and are controlled by gas puffing. Even without boronization, stationary, ITER baseline H-modes (confinement enhancement factor from ITER 98(y, 2) scaling H98 ~ 1, normalized beta βN ~ 2), with W concentrations below 3 × 10-5 were routinely achieved up to 1.2 MA plasma current. The compatibility of high performance improved H-modes with unboronized W wall was demonstrated, achieving H98 = 1.1 and βN up to 2.6 at modest triangularities δ cooled by N2 seeding. N2 seeding does not only protect the divertor tiles but also considerably improves the performance of improved H-mode discharges. The energy confinement increased to H98-factors of 1.25 (βN ~ 2.7) and thereby exceeded the best values in a carbon-dominated AUG machine under similar conditions. Recent investigations show that this improvement is due to higher temperatures rather than to peaking of the electron density profile. Further ITER discharge scenario tests include the demonstration of ECRF assisted low voltage plasma start-up and current rise to q95 = 3 at toroidal electric fields below 0.3 V m-1, to achieve a ITER compatible range of plasma internal inductance of 0.71-0.97. The results reported here strongly support tungsten as a first

  11. A recursion identity for formal iterated logarithms and iterated exponentials

    CERN Document Server

    Robinson, Thomas J

    2010-01-01

    We prove a recursive identity involving formal iterated logarithms and formal iterated exponentials. These iterated logarithms and exponentials appear in a natural extension of the logarithmic formal calculus used in the study of logarithmic intertwining operators and logarithmic tensor category theory for modules for a vertex operator algebra. This extension has a variety of interesting arithmetic properties. We develop one such result here, the aforementioned recursive identity. We have applied this identity elsewhere to certain formal series expansions related to a general formal Taylor theorem and these series expansions in turn yield a sequence of combinatorial identities which have as special cases certain classical combinatorial identities involving (separately) the Stirling numbers of the first and second kinds.

  12. Fluctuation BES measurements with the ITER core CXRS prototype spectrometer

    Energy Technology Data Exchange (ETDEWEB)

    Pokol, G.I., E-mail: pokol@reak.bme.hu [Institute of Nuclear Techniques, Budapest University of Technology and Economics, EURATOM Association, PO Box 91, H-1521, Budapest (Hungary); Zoletnik, S.; Dunai, D. [WIGNER RCP, RMKI, EURATOM Association, PO Box 91, H-1521, Budapest (Hungary); Marchuk, O. [Institut für Energieforschung – Plasmaphysik, Forschungszentrum Jülich Gmbh, Association EURATOM-FZJ, member of Trilateral Euregio Cluster, 52425 Jülich (Germany); Baross, T. [WIGNER RCP, RMKI, EURATOM Association, PO Box 91, H-1521, Budapest (Hungary); Erdei, G. [Department of Atomic Physics, Budapest University of Technology and Economics, EURATOM Association, PO Box 91, H-1521, Budapest (Hungary); Grunda, G.; Kiss, I.G. [WIGNER RCP, RMKI, EURATOM Association, PO Box 91, H-1521, Budapest (Hungary); Kovacsik, A. [Institute of Nuclear Techniques, Budapest University of Technology and Economics, EURATOM Association, PO Box 91, H-1521, Budapest (Hungary); Hellermann, M. von; Lischtschenko, O. [Dutch-Institute for Fundamental Energy Research, Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster and ITER-NL, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Biel, W. [Institut für Energieforschung – Plasmaphysik, Forschungszentrum Jülich Gmbh, Association EURATOM-FZJ, member of Trilateral Euregio Cluster, 52425 Jülich (Germany); Jaspers, R.J.E. [Science and Technology of Nuclear Fusion, Eindhoven University of Technology (Netherlands); Durkut, M. [TNO Science and Industry, Partner in ITER-NL, PO Box 155, 2600 AD Delft (Netherlands)

    2013-10-15

    Highlights: • We integrated a fluctuation beam emission measurement into the ITER CXRS prototype spectrometer. • The fluctuation BES measurement provided data at TEXTOR that agree well with the simulation based on the Simulation Of Spectra package. • The same simulation method has been used to evaluate the feasibility of a fluctuation BES measurement on the ITER DNB using the CXRS periscopes. -- Abstract: The ITER core CXRS diagnostic system collects the light emitted from the interaction of the diagnostic neutral beam with the core plasma and guides it via a mirror labyrinth through the upper port plug no. 3 towards a fiber bundle, which then transmits the light into a set of spectrometers for spectral analysis. In order to test the accessibility of the special parameter range required for the ITER measurement, a prototype spectrometer was built and operated successfully at the TEXTOR tokamak. In addition to the He/Be, C/Ne and H/D/T regular spectral channels, a fluctuation beam emission spectroscopy (BES) system has been integrated to measure core MHD activity, and validate corresponding ITER simulation results. The fluctuation system can be operated as an alternative to the spectral BES measurement, and has 8 spatial channels sampled at 2 MHz. In this paper, we present details of the fluctuation BES system and its interface to the ITER prototype spectrometer along with simulation and measurement results at TEXTOR. We show that the measurement fully confirms the simulation results on achievable photon current at the detector and on the signal to noise ratio.

  13. High power 1 MeV neutral beam system and its application plan for the international tokamak experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hemsworth, R.S. [ITER Joint Central Team, Naka, Ibaraki (Japan)

    1997-03-01

    This paper describes the Neutral Beam Injection system which is presently being designed for the International Tokamak Experimental Reactor, ITER, in Europe Japan and Russia, with co-ordination by the Joint Central Team of ITER at Naka, Japan. The proposed system consists of three negative ion based neutral injectors, delivering a total of 50 MW of 1 MeV D{sup 0} to the ITER plasma for a pulse length of >1000 s. Each injectors uses a single caesiated volume arc discharge negative ion source, and a multi-grid, multi-aperture accelerator, to produce about 40 A of 1 MeV D{sup -}. This will be neutralized by collisions with D{sub 2} in a sub-divided gas neutralizer, which has a conversion efficiency of about 60%. The charged fraction of the beam emerging from the neutralizer is dumped in an electrostatic residual ion dump. A water cooled calorimeter can be moved into the beam path to intercept the neutral beam, allowing commissioning of the injector independent of ITER. ITER is scheduled to produce its first plasma at the beginning of 2008, and the planning of the R and D, construction and installation foresees the neutral injection system being available from the start of ITER operations. (author)

  14. Relaxation Criteria for Iterated Traffic Simulations

    Science.gov (United States)

    Kelly, Terence; Nagel, Kai

    Iterative transportation microsimulations adjust traveler route plans by iterating between a microsimulation and a route planner. At each iteration, the route planner adjusts individuals' route choices based on the preceding microsimulations. Empirically, this process yields good results, but it is usually unclear when to stop the iterative process when modeling real-world traffic. This paper investigates several criteria to judge relaxation of the iterative process, emphasizing criteria related to traveler decision-making.

  15. Thermo-mechanical analysis of ITER first mirrors and its use for the ITER equatorial visible∕infrared wide angle viewing system optical design.

    Science.gov (United States)

    Joanny, M; Salasca, S; Dapena, M; Cantone, B; Travère, J M; Thellier, C; Fermé, J J; Marot, L; Buravand, O; Perrollaz, G; Zeile, C

    2012-10-01

    ITER first mirrors (FMs), as the first components of most ITER optical diagnostics, will be exposed to high plasma radiation flux and neutron load. To reduce the FMs heating and optical surface deformation induced during ITER operation, the use of relevant materials and cooling system are foreseen. The calculations led on different materials and FMs designs and geometries (100 mm and 200 mm) show that the use of CuCrZr and TZM, and a complex integrated cooling system can limit efficiently the FMs heating and reduce their optical surface deformation under plasma radiation flux and neutron load. These investigations were used to evaluate, for the ITER equatorial port visible∕infrared wide angle viewing system, the impact of the FMs properties change during operation on the instrument main optical performances. The results obtained are presented and discussed.

  16. ITER transient consequences for material damage: modelling versus experiments

    Energy Technology Data Exchange (ETDEWEB)

    Bazylev, B [Forschungszentrum Karlsruhe, IHM, P O Box 3640, 76021 Karlsruhe (Germany); Janeschitz, G [Forschungszentrum Karlsruhe, Fusion, P O Box 3640, 76021 Karlsruhe (Germany); Landman, I [Forschungszentrum Karlsruhe, IHM, P O Box 3640, 76021 Karlsruhe (Germany); Pestchanyi, S [Forschungszentrum Karlsruhe, IHM, P O Box 3640, 76021 Karlsruhe (Germany); Loarte, A [EFDA Close Support Unit Garching, Boltmannstr 2, D-85748 Garching (Germany); Federici, G [ITER International Team, Garching Working Site, Boltmannstr 2, D-85748 Garching (Germany); Merola, M [ITER International Team, Garching Working Site, Boltmannstr 2, D-85748 Garching (Germany); Linke, J [Forschungszentrum Juelich, EURATOM-Association, D-52425 Juelich (Germany); Zhitlukhin, A [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Podkovyrov, V [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Klimov, N [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Safronov, V [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation)

    2007-03-15

    Carbon-fibre composite (CFC) and tungsten macrobrush armours are foreseen as PFC for the ITER divertor. In ITER the main mechanisms of metallic armour damage remain surface melting and melt motion erosion. In the case of CFC armour, due to rather different heat conductivities of CFC fibres a noticeable erosion of the PAN bundles may occur at rather small heat loads. Experiments carried out in the plasma gun facilities QSPA-T for the ITER like edge localized mode (ELM) heat load also demonstrated significant erosion of the frontal and lateral brush edges. Numerical simulations of the CFC and tungsten (W) macrobrush target damage accounting for the heat loads at the face and lateral brush edges were carried out for QSPA-T conditions using the three-dimensional (3D) code PHEMOBRID. The modelling results of CFC damage are in a good qualitative and quantitative agreement with the experiments. Estimation of the droplet splashing caused by the Kelvin-Helmholtz (KH) instability was performed.

  17. A Linear Iterative Unfolding Method

    CERN Document Server

    Laszlo, Andras

    2011-01-01

    A frequently faced task in experimental physics is to measure the probability distribution of some quantity. Often this quantity to be measured is smeared by a non-ideal detector response or by some physical process. The procedure of removing this smearing effect from the measured distribution is called unfolding, and is a delicate problem in signal processing. Due to the numerical ill-posedness of this task, various methods were invented which, given some assumptions on the initial probability distribution, try to regularize the problem. Most of these methods definitely introduce bias on the estimate of the initial probability distribution. We propose a linear iterative method (motivated by the Neumann series / Landweber iteration known in functional analysis), which has the advantage that no assumptions on the initial probability distribution is needed, and the only regularization parameter is the stopping order of the iteration. Convergence is proved under certain quite general conditions, which hold for p...

  18. Construction Safety Forecast for ITER

    Energy Technology Data Exchange (ETDEWEB)

    cadwallader, lee charles

    2006-11-01

    The International Thermonuclear Experimental Reactor (ITER) project is poised to begin its construction activity. This paper gives an estimate of construction safety as if the experiment was being built in the United States. This estimate of construction injuries and potential fatalities serves as a useful forecast of what can be expected for construction of such a major facility in any country. These data should be considered by the ITER International Team as it plans for safety during the construction phase. Based on average U.S. construction rates, ITER may expect a lost workday case rate of < 4.0 and a fatality count of 0.5 to 0.9 persons per year.

  19. Development of design Criteria for ITER In-vessel Components

    Energy Technology Data Exchange (ETDEWEB)

    Sannazzaro, G., E-mail: Giulio.Sannazzaro@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Barabash, V.; Kang, S.C. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Fernandez, E. [Fusion For Energy, Josep Pla, 2, Torres Diagonal Mar B3, 08019 Barcelona (Spain); Kalinin, G.; Obushev, A. [The Research and Development Institute of Power Engineering, P.O. Box 788, Moscow (Russian Federation); Martínez, V.J.; Vázquez, I. [IDESA, Parque Científico Tecnológico, C/Profesor Potter 105, 33203 Gijón (Spain); Fernández, F.; Guirao, J. [NATEC, C/Marqués de San Esteban 52 Entlo, 33209 Gijón (Spain)

    2013-10-15

    Absrtract: The components located inside the ITER vacuum chamber (in-vessel components – IC), due to their specific nature and the environments they are exposed to (neutron radiation, high heat fluxes, electromagnetic forces, etc.), have specific design criteria which are, in this paper, referred as Structural Design Criteria for In-vessel Components (SDC-IC). The development of these criteria started in the very early phase of the ITER design and followed closely the criteria of the RCC-MR code. Specific rules to include the effect of neutron irradiation were implemented. In 2008 the need of an update of the SDC-IC was identified to add missing specifications, to implement improvements, to modernise rules including recent evolutions in international codes and regulations (i.e. PED). Collaboration was set up between ITER Organization (IO), European (EUDA) and Russian Federation (RFDA) Domestic Agencies to generate a new version of SDC-IC. A Peer Review Group (PRG) composed by members of the ITER Organization and all ITER Domestic Agencies and code experts was set-up to review the proposed modifications, to provide comments, contributions and recommendations.

  20. Simulation of divertor targets shielding during transients in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Pestchanyi, Sergey, E-mail: serguei.pestchanyi@kit.edu [KIT, Hermann-von-Helmholtz-Platz 1, Eggenstein-Leopoldshafen (Germany); Pitts, Richard; Lehnen, Michael [ITER Organization,Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2016-11-01

    Highlights: • We simulated plasma shielding effect during disruption in ITER using the TOKES code. • It has been found that vaporization is unavoidable under action of ITER transients, but plasma shielding drastically reduces the divertor target damage: the melt pool and the vaporization region widths reduced 10–15 times. • A simplified 1D model describing the melt pool depth and the shielded heat flux to the divertor targets have been developed. • The results of the TOKES simulations have been compared with the analytic model when the model is valid. - Abstract: Direct extrapolation of the disruptive heat flux on ITER conditions predicts severe melting and vaporization of the divertor targets causing their intolerable damage. However, tungsten vaporized from the target at initial stage of the disruption can create plasma shield in front of the target, which effectively protects the target surface from the rest of the heat flux. Estimation of this shielding efficiency has been performed using the TOKES code. The shielding effect under ITER conditions is found to be very strong: the maximal depth of the melt layer reduced 4 times, the melt layer width—more than 10 times and vaporization region shrinks 10–15 times due to shielding for unmitigated disruption of 350 MJ discharge. The simulation results show complex, 2D plasma dynamics of the shield under ITER conditions. However, a simplified analytic model, valid for rough estimation of the maximum value for the shielded flux to the target and for the melt depth at the target surface has been developed.

  1. Design study of ITER-like divertor target for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Crescenzi, Fabio, E-mail: fabio.crescenzi@enea.it [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Bachmann, C. [EFDA, Power Plant Physics and Technology, Boltzmannstraße 2, 85748 Garching (Germany); Richou, M. [CEA, IRFM, F-13108 Saint Paul Lez Durance (France); Roccella, S.; Visca, E. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); You, J.-H. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany)

    2015-10-15

    Highlights: • ‘DEMO’ is a near-term Power Plant Conceptual Study (PPCS). • The ITER-like design concept represents a promising solution also for DEMO plasma facing units. • The optimization of PFUs aims to enhance the thermo-mechanical behaviour of the component. • The optimized geometry was evaluated by ITER SDC-IC criteria and in terms of low cycle fatigue (LCF). - Abstract: A near-term water-cooled target solution has to be evaluated together with the required technologies and its power exhaust limit under ‘DEMO’ conditions. The ITER-like design concept based on the mono-block technology using W as armour material and the CuCrZr-IG as structural material with an interlayer of pure copper represents a promising solution also for DEMO. This work reports the design study of an “optimized” ITER-like Water Cooled Divertor able to withstand a heat flux of 10 MW m{sup −2}, as requested for DEMO operating conditions. The optimization of plasma facing unit (PFU) aims to enhance the thermo-mechanical behaviour of the component by varying some geometrical parameters (monoblock size, interlayer thickness and, tube diameter and thickness). The optimization was performed by means of the multi-variable optimization algorithms using the FEM code ANSYS. The coolant hydraulic conditions (inlet pressure, temperature and velocity) were fixed for simplicity. This study is based on elastic analysis and 3 dimensional modelling. The resulting optimized geometry was evaluated on the basis of the ITER SDC-IC criteria and in terms of low cycle fatigue (LCF). The margin to the critical heat flux (CHF) was also estimated. Further design study (taking into account the effect of neutron radiation on the material properties) together with mock-up fabrication and high-heat-flux (HHF) tests are foreseen in next work programmes.

  2. Rollout Sampling Approximate Policy Iteration

    CERN Document Server

    Dimitrakakis, Christos

    2008-01-01

    Several researchers have recently investigated the connection between reinforcement learning and classification. We are motivated by proposals of approximate policy iteration schemes without value functions which focus on policy representation using classifiers and address policy learning as a supervised learning problem. This paper proposes variants of an improved policy iteration scheme which addresses the core sampling problem in evaluating a policy through simulation as a multi-armed bandit machine. The resulting algorithm offers comparable performance to the previous algorithm achieved, however, with significantly less computational effort. An order of magnitude improvement is demonstrated experimentally in two standard reinforcement learning domains: inverted pendulum and mountain-car.

  3. Kinetic modeling of divertor heat load fluxes in the Alcator C-Mod and DIII-D tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Pankin, A. Y. [Tech-X Corporation, Boulder, Colorado 80303 (United States); Rafiq, T.; Kritz, A. H. [Department of Physics, Lehigh University, Bethlehem, Pennsylvania 18015 (United States); Park, G. Y. [National Fusion Research Institute, Daejeon, 305-333 (Korea, Republic of); Chang, C. S.; Ku, S. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Brunner, D.; Hughes, J. W.; LaBombard, B.; Terry, J. L. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Groebner, R. J. [General Atomics, San Diego, California 92121 (United States)

    2015-09-15

    The guiding-center kinetic neoclassical transport code, XGC0 [Chang et al., Phys. Plasmas 11, 2649 (2004)], is used to compute the heat fluxes and the heat-load width in the outer divertor plates of Alcator C-Mod and DIII-D tokamaks. The dependence of the width of heat-load fluxes on neoclassical effects, neutral collisions, and anomalous transport is investigated using the XGC0 code. The XGC0 code includes realistic X-point geometry, a neutral source model, the effects of collisions, and a diffusion model for anomalous transport. It is observed that the width of the XGC0 neoclassical heat-load is approximately inversely proportional to the total plasma current I{sub p.} The scaling of the width of the divertor heat-load with plasma current is examined for an Alcator C-Mod discharge and four DIII-D discharges. The scaling of the divertor heat-load width with plasma current is found to be weaker in the Alcator C-Mod discharge compared to scaling found in the DIII-D discharges. The effect of neutral collisions on the 1/I{sub p} scaling of heat-load width is shown not to be significant. Although inclusion of poloidally uniform anomalous transport results in a deviation from the 1/I{sub p} scaling, the inclusion of the anomalous transport that is driven by ballooning-type instabilities results in recovering the neoclassical 1/I{sub p} scaling. The Bohm or gyro-Bohm scalings of anomalous transport do not strongly affect the dependence of the heat-load width on plasma current. The inclusion of anomalous transport, in general, results in widening the width of neoclassical divertor heat-load and enhances the neoclassical heat-load fluxes on the divertor plates. Understanding heat transport in the tokamak scrape-off layer plasmas is important for strengthening the basis for predicting divertor conditions in ITER.

  4. NCSX Plasma Heating Methods

    Energy Technology Data Exchange (ETDEWEB)

    H.W. Kugel; D. Spong; R. Majeski; M. Zarnstorff

    2003-02-28

    The NCSX (National Compact Stellarator Experiment) has been designed to accommodate a variety of heating systems, including ohmic heating, neutral-beam injection, and radio-frequency. Neutral beams will provide one of the primary heating methods for NCSX. In addition to plasma heating, beams are also expected to provide a means for external control over the level of toroidal plasma rotation velocity and its profile. The plan is to provide 3 MW of 50 keV balanced neutral-beam tangential injection with pulse lengths of 500 msec for initial experiments, and to be upgradeable to pulse lengths of 1.5 sec. Subsequent upgrades will add 3 MW of neutral-beam injection. This Chapter discusses the NCSX neutral-beam injection requirements and design issues, and shows how these are provided by the candidate PBX-M (Princeton Beta Experiment-Modification) neutral-beam injection system. In addition, estimations are given for beam-heating efficiencies, scaling of heating efficiency with machine size an d magnetic field level, parameter studies of the optimum beam-injection tangency radius and toroidal injection location, and loss patterns of beam ions on the vacuum chamber wall to assist placement of wall armor and for minimizing the generation of impurities by the energetic beam ions. Finally, subsequent upgrades could add an additional 6 MW of radio-frequency heating by mode-conversion ion-Bernstein wave (MCIBW) heating, and if desired as possible future upgrades, the design also will accommodate high-harmonic fast-wave and electron-cyclotron heating. The initial MCIBW heating technique and the design of the radio-frequency system lend themselves to current drive, so that if current drive became desirable for any reason only minor modifications to the heating system described here would be needed. The radio-frequency system will also be capable of localized ion heating (bulk or tail), and possibly ion-Bernstein-wave-generated sheared flows.

  5. Conceptual design of the ITER upper port plug for charge exchange diagnostic

    NARCIS (Netherlands)

    Sadakov, S.; Baross, T.; Biel, W.; Borsuk, V.; Hawkes, N.; Hellermann, M. von; Gille, P.; Kiss, G.; Koning, J.; Knaup, M.; Klinkhamer, J.F.F.; Krasikov, Yu.; Litnovsky, A.; Neubauer, O.; Panin, A.

    2009-01-01

    A plug for the ITER core charge exchange recombination spectroscopy (core CXRS) is located in the upper port 3. It transfers the light emitted by interaction of plasma ions with the diagnostic neutral beam (DNB). The plug consists of a main shell, a shielding cassette and a retractable tube. The tub

  6. The Linear Stability Properties of Medium- to High- n TAEs in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Gorelenkov, N N; Budny, R V; Kessel, C E; Kramer, G J; McCune, D; Manickam, J; Nazikian, R

    2008-02-14

    This document provides a detailed report on the successful completion of the DOE OFES Theory Milestone for FY2007: Improve the simulation resolution of linear stability properties of Toroidal Alfvén Eigenmodes (TAE) driven by energetic particles and neutral beams in ITER by increasing the numbers of toroidal modes used to 15.

  7. Negative ion source development for a photoneutralization based neutral beam system for future fusion reactors

    Science.gov (United States)

    Simonin, A.; Agnello, R.; Bechu, S.; Bernard, J. M.; Blondel, C.; Boeuf, J. P.; Bresteau, D.; Cartry, G.; Chaibi, W.; Drag, C.; Duval, B. P.; de Esch, H. P. L.; Fubiani, G.; Furno, I.; Grand, C.; Guittienne, Ph; Howling, A.; Jacquier, R.; Marini, C.; Morgal, I.

    2016-12-01

    In parallel to the developments dedicated to the ITER neutral beam (NB) system, CEA-IRFM with laboratories in France and Switzerland are studying the feasibility of a new generation of NB system able to provide heating and current drive for the future DEMOnstration fusion reactor. For the steady-state scenario, the NB system will have to provide a high NB power level with a high wall-plug efficiency (η ˜ 60%). Neutralization of the energetic negative ions by photodetachment (so called photoneutralization), if feasible, appears to be the ideal solution to meet these performances, in the sense that it could offer a high beam neutralization rate (>80%) and a wall-plug efficiency higher than 60%. The main challenge of this new injector concept is the achievement of a very high power photon flux which could be provided by 3 MW Fabry-Perot optical cavities implanted along the 1 MeV D- beam in the neutralizer stage. The beamline topology is tall and narrow to provide laminar ion beam sheets, which will be entirely illuminated by the intra-cavity photon beams propagating along the vertical axis. The paper describes the present R&D (experiments and modelling) addressing the development of a new ion source concept (Cybele source) which is based on a magnetized plasma column. Parametric studies of the source are performed using Langmuir probes in order to characterize and compare the plasma parameters in the source column with different plasma generators, such as filamented cathodes, radio-frequency driver and a helicon antenna specifically developed at SPC-EPFL satisfying the requirements for the Cybele (axial magnetic field of 10 mT, source operating pressure: 0.3 Pa in hydrogen or deuterium). The paper compares the performances of the three plasma generators. It is shown that the helicon plasma generator is a very promising candidate to provide an intense and uniform negative ion beam sheet.

  8. Development of the Butt Joint for the ITER Central Solenoid

    Energy Technology Data Exchange (ETDEWEB)

    Martovetsky, N N

    2006-08-23

    The ITER Central Solenoid (CS) requires compact and reliable joints for its Cable-in-Conduit Conductor (CICC). The baseline design is a diffusion bonded butt joint. In such a joint the mating cables are compacted to a very low void fraction in a copper sleeve and then heat treated. After the heat treatment the ends are cut, polished and aligned against each other and then diffusion bonded under high compression in a vacuum chamber at 750 C. The jacket is then welded on the conductor to complete the joint, which remarkably does not require more room than a regular conductor. This joint design is based on a proven concept developed for the ITER CS Model Coil that was successfully tested in the previous R&D phase.

  9. Bounded Fixed-Point Iteration

    DEFF Research Database (Denmark)

    Nielson, Hanne Riis; Nielson, Flemming

    1992-01-01

    they obtain a quadratic bound. These bounds are shown to be tight. Specializing the case of strict and additive functions to functionals of a form that would correspond to iterative programs they show that a linear bound is tight. This is related to several analyses studied in the literature (including...

  10. Iterative method for interferogram processing

    Science.gov (United States)

    Kotlyar, Victor V.; Seraphimovich, P. G.; Zalyalov, Oleg K.

    1994-12-01

    We have developed and numerically evaluated an iterative algorithm for interferogram processing including the Fourier-transform method, the Gerchberg-Papoulis algorithm and Wiener's filter-based regularization used in combination. Using a signal-to-noise ratio not less than 1, it has been possible to reconstruct the phase of an object field with accuracy better than 5%.

  11. Iterative Specialisation of Horn Clauses

    DEFF Research Database (Denmark)

    Nielsen, Christoffer Rosenkilde; Nielson, Flemming; Nielson, Hanne Riis

    2008-01-01

    We present a generic algorithm for solving Horn clauses through iterative specialisation. The algorithm is generic in the sense that it can be instantiated with any decidable fragment of Horn clauses, resulting in a solution scheme for general Horn clauses that guarantees soundness and terminatio...

  12. Cooperation between CERN and ITER

    CERN Multimedia

    CERN Audiovisual Service

    2008-01-01

    CERN and the International Fusion Organisation ITER have just signed a first cooperation agreeement. The Director-General of the International Fusion Energy Organization, Mr Kaname Ikeda, and CERN Director-General, Robert Aymar, signed a cooperation agreement at a meeting on the Meyrin site on Thursday 6 March.

  13. Influence of the heat-treatment conditions, microchemistry, and microstructure on the irreversible strain limit of a selection of Ti-doped internal-tin Nb3Sn ITER wires

    Science.gov (United States)

    Cheggour, N.; Lee, P. J.; Goodrich, L. F.; Sung, Z.-H.; Stauffer, T. C.; Splett, J. D.; Jewell, M. C.

    2014-10-01

    Systematic studies of the intrinsic irreversible strain limit ɛirr,0, microstructure, and microchemistry were made on several internal-tin Nb3Sn pre-production wires, fabricated for the domestic agencies of the USA and China participating in the International Thermonuclear Experimental Reactor. These wires were produced by Luvata, Oxford Superconducting Technology (OST), and Western Superconducting Technologies (WST), and were intended for the tokamak’s toroidal-field coils. The results of this study show that, for a final heat-treatment at 650 °C to form the A15 phase, both ɛirr,0 and the de-pinning field Bc2* improved by increasing heat-treatment duration beyond 100 h for the Luvata wires. On the other hand, we saw no improvement in these two parameters as a function of heat-treatment duration in the OST wires. Furthermore, micro-chemical analysis of OST wires revealed that some Nb3Sn filaments have a Sn- and Ti-rich phase at the interface between Cu(Sn) matrix and Nb3Sn in the form of a shell around individual filaments. This phase is far less prominent in the Luvata and WST conductors, and could inhibit diffusion of Sn and Ti into Nb3Sn filaments during the reaction and may potentially be the reason for the lack of noticeable change in Bc2* with heat-treatment duration in the OST wires. The increase of ɛirr,0 and Bc2* with heat-treatment duration in the Luvata wires and the lack of increase in the OST wires may suggest a possible correlation between ɛirr,0 and the stoichiometry of the A15 composition. Investigation of the samples’ microstructure revealed only a small number of cracked Nb3Sn filaments despite the significant and permanent degradation of their critical current Ic when subjected to longitudinal tensile strain ɛ beyond ɛirr,0. The scarcity of cracks indicate that Ic(ɛ) measurements are highly sensitive to crack formation in Nb3Sn filaments, especially at low electric-field criteria ≦̸0.1 μV cm-1, even when the sizes of the

  14. Design and manufacturing of the ITER ECRH Upper launcher mirrors

    OpenAIRE

    Sanchez, Francisco; Bertizzolo, Robert; Chavan, Rene; Collazos, Andres; Landis, Jean Daniel

    2008-01-01

    Four of the 16 ITER upper port plugs will be devoted to electron cyclotron resonance heating (ECRH) in order to control the magneto-hydrodynamic (MHD) instabilities. In order to achieve the stabilisation of the neoclassical tearing modes (NTM) and sawtooth oscillation, a deposition of a very localized and peaked current density profile over a broad poloidal steering range is required. In the present optical configuration eight 2MW mm-wave beams enter each of the four upper launchers (UL) thro...

  15. Neutralization of English Consonants

    Institute of Scientific and Technical Information of China (English)

    庞彬彬

    2011-01-01

    This paper gives a brief account of English consonant cluster's structure and phonetic features from the perspective of the definition and cause of neutralization of English consonants as well as their distinctive features and oppositions.It comes up with the final conclusion that neutralization exists in only thirteen English consonant clusters,among a large number of consonant clusters.

  16. Three-dimensional modeling of plasma edge transport and divertor fluxes during application of resonant magnetic perturbations on ITER

    Science.gov (United States)

    Schmitz, O.; Becoulet, M.; Cahyna, P.; Evans, T. E.; Feng, Y.; Frerichs, H.; Loarte, A.; Pitts, R. A.; Reiser, D.; Fenstermacher, M. E.; Harting, D.; Kirschner, A.; Kukushkin, A.; Lunt, T.; Saibene, G.; Reiter, D.; Samm, U.; Wiesen, S.

    2016-06-01

    Results from three-dimensional modeling of plasma edge transport and plasma-wall interactions during application of resonant magnetic perturbation (RMP) fields for control of edge-localized modes in the ITER standard 15 MA Q  =  10 H-mode are presented. The full 3D plasma fluid and kinetic neutral transport code EMC3-EIRENE is used for the modeling. Four characteristic perturbed magnetic topologies are considered and discussed with reference to the axisymmetric case without RMP fields. Two perturbation field amplitudes at full and half of the ITER ELM control coil current capability using the vacuum approximation are compared to a case including a strongly screening plasma response. In addition, a vacuum field case at high q 95  =  4.2 featuring increased magnetic shear has been modeled. Formation of a three-dimensional plasma boundary is seen for all four perturbed magnetic topologies. The resonant field amplitudes and the effective radial magnetic field at the separatrix define the shape and extension of the 3D plasma boundary. Opening of the magnetic field lines from inside the separatrix establishes scrape-off layer-like channels of direct parallel particle and heat flux towards the divertor yielding a reduction of the main plasma thermal and particle confinement. This impact on confinement is most accentuated at full RMP current and is strongly reduced when screened RMP fields are considered, as well as for the reduced coil current cases. The divertor fluxes are redirected into a three-dimensional pattern of helical magnetic footprints on the divertor target tiles. At maximum perturbation strength, these fingers stretch out as far as 60 cm across the divertor targets, yielding heat flux spreading and the reduction of peak heat fluxes by 30%. However, at the same time substantial and highly localized heat fluxes reach divertor areas well outside of the axisymmetric heat flux decay profile. Reduced RMP amplitudes due to screening or reduced RMP

  17. Safety Analysis Results for Cryostat Ingress Accidents in ITER

    Science.gov (United States)

    Merrill, B. J.; Cadwallader, L. C.; Petti, D. A.

    1997-06-01

    Accidents involving the ingress of air, helium, or water into the cryostat of the International Thermonuclear Experimental Reactor (ITER) tokamak design have been analyzed with a modified version of the MELCOR code for the ITER Non-site Specific Safety Report (NSSR-1). The air ingress accident is the result of a postulated breach of the cryostat boundary into an adjoining room. MELCOR results for this accident demonstrate that the condensed air mass and increased heat loads are not a magnet safety concern, but that the partial vacuum in the adjoining room must be accommodated in the building design. The water ingress accident is the result of a postulated magnet arc that results in melting of a Primary Heat Transport System (PHTS) coolant pipe, discharging PHTS water and PHTS water activated corrosion products and HTO into the cryostat. MELCOR results for this accident demonstrate that the condensed water mass and increased heat loads are not a magnet safety concern, that the cryostat pressure remains below design limits, and that the corrosion product and HTO releases are well within the ITER release limits.

  18. Net neutrality towards a co-regulatory solution

    CERN Document Server

    Marsden, Christopher T

    2010-01-01

    In considering market developments and policy responses to some of the most heated net-neutrality debates in Europe and the United States, Net Neutrality is the first, fully comprehensive overview of the subject. This book is also unique in providing readers with a supplementary outline of recommended policy prescriptives.

  19. Design of ITER NBI power supply system

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Kazuhiro; Ohara, Yoshihiro; Okumura, Yoshikazu [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Higa, Osamu; Kawashima, Syuichi; Ono, Youichi; Tanaka, Masanobu; Yasutomi, Sei

    1997-07-01

    Power supply system for the ITER neutral beam injector (NBI) whose total injection power is 1 MeV, 50 MW from three modules, has been designed. The power supply system consists of a source power supply for negative ion production/extraction and a DC 1 MV, 45 A power supply for negative ion acceleration. An inverter controlled multi-transformer/rectifier system has been adopted to the acceleration power supply. An inverter frequency of 150 Hz was selected to satisfy required specifications which are rise time of <100 ms, voltage ripple of <10% peak to peak and cut off speed of <200{mu}s. It was confirmed that the rise time, the ripple and the cut off speed is about 50 ms, 7% and <200{mu}s respectively by computation. It was also confirmed that a surge current and an energy input to the ion source at the breakdown can be suppressed lower than 3 kA and 10 J, which are considered to be lower than allowable values. A 1 MV transmission line has been designed from a view point of electric field on the inner conductors and grounded conductor. The results from the design study indicate that all the required specification to the power supply system can be satisfied and that R and D on the transmission line is one of the most important subjects. (author)

  20. ITERATIVE ALGORITHMS FOR DATA ASSIMILATION PROBLEMS

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    Iterative algorithms for solving the data assimilation problems are considered, based on the main and adjoint equations. Spectral properties of the control operators of the problem are studied, the iterative algorithms are justified.

  1. Existence test for asynchronous interval iterations

    DEFF Research Database (Denmark)

    Madsen, Kaj; Caprani, O.; Stauning, Ole

    1997-01-01

    In the search for regions that contain fixed points ofa real function of several variables, tests based on interval calculationscan be used to establish existence ornon-existence of fixed points in regions that are examined in the course ofthe search. The search can e.g. be performed...... as a synchronous (sequential) interval iteration:In each iteration step all components of the iterate are calculatedbased on the previous iterate. In this case it is straight forward to base simple interval existence and non-existencetests on the calculations done in each step of the iteration. The search can also...... be performed as an asynchronous (parallel) iteration: Only a few components are changed in each stepand this calculation is in general based on components from differentprevious iterates. For the asynchronous iteration it turns out thatsimple tests of existence and non-existence can be based...

  2. Influence of impurity seeding on plasma burning scenarios for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Ivanova-Stanik, I., E-mail: irena.ivanova-stanik@ifpilm.pl [Institute of Plasma Physics and Laser Microfusion, Hery 23, 01-497 Warsaw (Poland); Zagórski, R. [Institute of Plasma Physics and Laser Microfusion, Hery 23, 01-497 Warsaw (Poland); Voitsekhovitch, I. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Brezinsek, S. [Forschungszentrum Jülich GmbH Institut für Energie-und Klimaforschung—Plasmaphysik, Jülich 52425 (Germany)

    2016-11-01

    Highlights: • The self-consistent (core-edge) COREDIV code has been used to analyze ITER standard inductive scenarios with neon and argon seeding. • In order to achieve wide operational window with the power crossing separatrix above the H-L threshold and simultaneously with tolerable heat load to target plates (<40 MW) relatively strong impurity transport in the core and SOL regions is necessary. • For argon seeding, the operational window is much smaller than for neon case due to enhanced core radiation (in comparison to Ne). - Abstract: ITER expects to produce fusion power of about 0.5GW when operating with tungsten (W) divertor and beryllium (Be) wall. The influx of W from divertor can have significant influence on the discharge performance. This work describes predictive integrated numerical modeling of ITER discharges using the COREDIV code, which self-consistently solves the 1D radial energy and particle transport in the core region and 2D multi-fluid transport in the SOL. Calculations are performed for inductive ITER scenarios with intrinsic (W, Be and He) impurities and with seeded impurities (Ne and Ar) for different particle and heat transport in the core and different radial transport in the SOL. Simulations show, that only for sufficiently high radial diffusion (both in the core and in the SOL regions), it is possible to achieve H-mode mode plasma operation (power to SOL > L-H threshold power) with acceptable low level of power reaching the divertor plates. For argon seeding, the operational window is much smaller than for neon case due to enhanced core radiation (in comparison to Ne). Particle transport in the core characterized by the ratio of particle diffusion to thermal conductivity) has strong influence on the predicted ITER performance.

  3. Overview of recent developments in pellet injection for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Combs, Stephen Kirk, E-mail: combssk@ornl.gov [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6169 (United States); Baylor, L.R.; Meitner, S.J.; Caughman, J.B.O.; Rasmussen, D.A. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6169 (United States); Maruyama, S. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer Status of the ITER pellet injection system. Black-Right-Pointing-Pointer Fueling requirements for ITER. Black-Right-Pointing-Pointer Summarizes the design/operating parameters and highlights recent developments. Black-Right-Pointing-Pointer Benefits of plasma fueling by the injection of pellets, composed of frozen hydrogen isotopes and millimeters in size, into magnetically confined plasmas (core fueling). Black-Right-Pointing-Pointer ELM mitigation with pellets (ELM pacing). - Abstract: Pellet injection is the primary fueling technique planned for core fueling of ITER burning plasmas. Also, the injection of relatively small pellets to purposely trigger rapid small edge localized modes (ELMs) has been proposed as a possible solution to the heat flux damage from larger natural ELMs likely to be an issue on the ITER divertor surfaces. The ITER pellet injection system is designed to inject pellets into the plasma through both inner and outer wall guide tubes. The inner wall guide tubes will provide high throughput pellet fueling while the outer wall guide tubes will be used primarily to trigger ELMs at a high frequency (>15 Hz). The pellet fueling rate of each injector is to be up to 120 Pa m{sup 3}/s, which will require the formation of solid D-T at a volumetric rate of {approx}1500 mm{sup 3}/s. Two injectors are to be provided for ITER at the startup with a provision for up to six injectors during the D-T phase. The required throughput of each injector is greater than that of any injector built to date, and a novel twin-screw continuous extrusion system is being developed to meet the challenging design parameters. Status of the development activities is presented, highlighting recent progress.

  4. Design and R&D for manufacturing the beamline components of MITICA and ITER HNBs

    Energy Technology Data Exchange (ETDEWEB)

    Dalla Palma, M., E-mail: mauro.dallapalma@igi.cnr.it [Consorzio RFX, Padova (Italy); Sartori, E. [Consorzio RFX, Padova (Italy); Blatchford, P.; Chuilon, B. [CCFE, Culham Science Centre, Oxfordshire (United Kingdom); Graceffa, J. [ITER Organization, St Paul Lez Durance (France); Hanke, S. [KIT, Institute for Technical Physics, Eggenstein-Leopoldshafen (Germany); Hardie, C. [CCFE, Culham Science Centre, Oxfordshire (United Kingdom); Masiello, A. [F4E, Barcelona (Spain); Muraro, A. [Consorzio RFX, Padova (Italy); Ochoa, S. [KIT, Institute for Technical Physics, Eggenstein-Leopoldshafen (Germany); Shah, D. [ITER Organization, St Paul Lez Durance (France); Veltri, P.; Zaccaria, P.; Zaupa, M. [Consorzio RFX, Padova (Italy)

    2015-10-15

    Highlights: • Particle beam-component interaction was analysed developing and applying numerical codes. • Gas density distribution was calculated with AVOCADO code and applied for electrical analyses. • High heat flux components were designed, analysed with subcooled boiling, verified for fatigue. • Fracture behaviour of ceramics was analysed by finite element modelling and was verified. • R&D supports the design of the beamline components, especially for water-vacuum barriers. - Abstract: The design of the beamline components of MITICA, the full prototype of the ITER heating neutral beam injectors, is almost finalised and technical specifications for the procurement are under preparation. These components are the gas neutraliser, the electrostatic residual ion dump, and the calorimeter. Electron dump panels are foreseen each side of the upstream end of the neutraliser to protect the cryo-panels from electrons, created by stripping and other processes, that exit the 1 MeV accelerator. As the design of the components must fulfil requirements on the beam physics, insight on physical processes is required to identify performance trade-offs and constraints. The spatial gas distribution was simulated to verify the pumping requirements with electron dump panels and local conditions for breakdown voltage. Electrostatic analyses were carried out for the insulating elements of the RID to verify the limits of the electric field intensity. Different criteria were approached to investigate the fracture behaviour of ceramics considering the manufacturing implications and extrapolating the conditions for proof testing. Severe heating conditions will be applied steadily, as the maximum pulse duration is 1 h, and cyclically so requiring to fulfil fatigue and ratcheting verifications. High heat fluxes, up to 13 MW/m{sup 2} on the calorimeter, with enhanced heat transfer in subcooled boiling conditions will occur in the actively cooled CuCr1Zr panel elements provided with

  5. BEAMS3D Neutral Beam Injection Model

    Science.gov (United States)

    McMillan, Matthew; Lazerson, Samuel A.

    2014-09-01

    With the advent of applied 3D fields in Tokamaks and modern high performance stellarators, a need has arisen to address non-axisymmetric effects on neutral beam heating and fueling. We report on the development of a fully 3D neutral beam injection (NBI) model, BEAMS3D, which addresses this need by coupling 3D equilibria to a guiding center code capable of modeling neutral and charged particle trajectories across the separatrix and into the plasma core. Ionization, neutralization, charge-exchange, viscous slowing down, and pitch angle scattering are modeled with the ADAS atomic physics database. Elementary benchmark calculations are presented to verify the collisionless particle orbits, NBI model, frictional drag, and pitch angle scattering effects. A calculation of neutral beam heating in the NCSX device is performed, highlighting the capability of the code to handle 3D magnetic fields. Notice: this manuscript has been authored by Princeton University under Contract Number DE-AC02-09CH11466 with the US Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes.

  6. REMARK ON STABILITY OF ISHIKAWA ITERATIVE PROCEDURES

    Institute of Scientific and Technical Information of China (English)

    薛志群; 田虹

    2002-01-01

    The stability of the Ishikawa iteration procedures was studied for one class ofcontinuity strong pseudocontraction and continuity strongly accretive operators with boundedrange in real uniformly smooth Banach space. Under parameters satisfying certainconditions, the convergence of iterative sequences was proved. The results improve andextend the recent corresponding results, and supply the basis of theory for further discussingconvergence of iteration procedures with errors.

  7. On One-Point Iterations and DIIS

    DEFF Research Database (Denmark)

    Østerby, Ole; Sørensen, Hans Henrik Brandenborg

    2009-01-01

    We analyze various iteration procedures in many dimensions inspired by the SCF iteration used in first principles electronic structure calculations. We show that the simple mixing of densities can turn a divergent (or slowly convergent) iteration into a (faster) convergent process provided all th...

  8. NCSX Plasma Heating Methods

    Energy Technology Data Exchange (ETDEWEB)

    Kugel, H. W.; Spong, D.; Majeski, R.; Zarnstorff, M.

    2008-01-18

    The National Compact Stellarator Experiment (NCSX) has been designed to accommodate a variety of heating systems, including ohmic heating, neutral beam injection, and radio-frequency (rf). Neutral beams will provide one of the primary heating methods for NCSX. In addition to plasma heating, neutral beams are also expected to provide a means for external control over the level of toroidal plasma rotation velocity and its profile. The experimental plan requires 3 MW of 50-keV balanced neutral beam tangential injection with pulse lengths of 500 ms for initial experiments, to be upgradeable to pulse lengths of 1.5 s. Subsequent upgrades will add 3MW of neutral beam injection (NBI). This paper discusses the NCSX NBI requirements and design issues and shows how these are provided by the candidate PBX-M NBI system. In addition, estimations are given for beam heating efficiencies, scaling of heating efficiency with machine size and magnetic field level, parameter studies of the optimum beam injection tangency radius and toroidal injection location, and loss patterns of beam ions on the vacuum chamber wall to assist placement of wall armor and for minimizing the generation of impurities by the energetic beam ions. Finally, subsequent upgrades could add an additional 6 MW of rf heating by mode conversion ion Bernstein wave (MCIBW) heating, and if desired as possible future upgrades, the design also will accommodate high-harmonic fast-wave and electron cyclotron heating. The initial MCIBW heating technique and the design of the rf system lend themselves to current drive, so if current drive became desirable for any reason, only minor modifications to the heating system described here would be needed. The rf system will also be capable of localized ion heating (bulk or tail), and possiblyIBW-generated sheared flows.

  9. An Iterative Rejection Sampling Method

    CERN Document Server

    Sherstnev, A

    2008-01-01

    In the note we consider an iterative generalisation of the rejection sampling method. In high energy physics, this sampling is frequently used for event generation, i.e. preparation of phase space points distributed according to a matrix element squared $|M|^2$ for a scattering process. In many realistic cases $|M|^2$ is a complicated multi-dimensional function, so, the standard von Neumann procedure has quite low efficiency, even if an error reducing technique, like VEGAS, is applied. As a result of that, many of the $|M|^2$ calculations go to ``waste''. The considered iterative modification of the procedure can extract more ``unweighted'' events, i.e. distributed according to $|M|^2$. In several simple examples we show practical benefits of the technique and obtain more events than the standard von Neumann method, without any extra calculations of $|M|^2$.

  10. ITER LHCD plans and design

    Energy Technology Data Exchange (ETDEWEB)

    Bibet, Ph.; Beaumont, B.; Delpech, L.; Ekedahl, A.; Kazarian, F.; Litaudon, X.; Prou, M. [CEA Cadarache, Dept. de Recherches sur la Fusion Controlee (DRFC), 13 - Saint-Paul-lez-Durance (France); Belo, J.H.; Bizarro, J.P.S. [Centro de Fusao Nuclear, Associacao Euratom-IST, Instituto Superior Tecnico, Lisboa (Portugal); Granucci, G. [Associazione EURATOM-ENEA sulla Fusione, Milano (Italy); Kuzikov, S. [Institute of Applied Physics, Nizhny Novgorod (Russian Federation); Mailloux, J. [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon (United Kingdom); Mirizzi, F.; Pericoli, V.; Tuccillo, A.A. [Association Euratom-ENEA sulla Fusione, Centro Ricerche Energia Frascati (Italy); Rantamaki, K. [Association Euratom-Tekes, VTT (Finland)

    2005-07-01

    LH waves experimentally exhibit the highest Current Drive efficiency at low plasma temperature, therefore they are the most suitable candidates for controlling the current profile in the off axis part of ITER Steady State plasmas. For this purpose, a 5 GHz, 20 MW CW LH system has been designed, that relies on a generator made of 24 klystrons, 1 MW each, 60 metres long circular oversized transmission lines, one antenna, based on the Passive Active Multi-function (PAM) concept. High reliability of the launcher is achieved, by limiting the power density to 33 MW/m{sup 2}. Together with the overall system description, the present results achieved toward ITER are presented. The different ongoing project are listed. The remaining outstanding problems are depicted. (authors)

  11. Matlab modeling of ITER CODAC

    Energy Technology Data Exchange (ETDEWEB)

    Pangione, L. [Associazione Euratom/ENEA Ssulla Fusione, Centro Ricerche Frascati, CP 65, 00044 Frascati, Roma (Italy)], E-mail: pangione@frascati.enea.it; Lister, J.B. [CRPP-EPFL, Association EURATOM-Suisse, Station 13, 1015 Lausanne (Switzerland)

    2008-04-15

    The ITER CODAC (COntrol, Data Access and Communication) conceptual design resulted from 2 years of activity. One result was a proposed functional partitioning of CODAC into different CODAC Systems, each of them partitioned into other CODAC Systems. Considering the large size of this project, simple use of human language assisted by figures would certainly be ineffective in creating an unambiguous description of all interactions and all relations between these Systems. Moreover, the underlying design is resident in the mind of the designers, who must consider all possible situations that could happen to each system. There is therefore a need to model the whole of CODAC with a clear and preferably graphical method, which allows the designers to verify the correctness and the consistency of their project. The aim of this paper is to describe the work started on ITER CODAC modeling using Matlab/Simulink. The main feature of this tool is the possibility of having a simple, graphical, intuitive representation of a complex system and ultimately to run a numerical simulation of it. Using Matlab/Simulink, each CODAC System was represented in a graphical and intuitive form with its relations and interactions through the definition of a small number of simple rules. In a Simulink diagram, each system was represented as a 'black box', both containing, and connected to, a number of other systems. In this way it is possible to move vertically between systems on different levels, to show the relation of membership, or horizontally to analyse the information exchange between systems at the same level. This process can be iterated, starting from a global diagram, in which only CODAC appears with the Plant Systems and the external sites, and going deeper down to the mathematical model of each CODAC system. The Matlab/Simulink features for simulating the whole top diagram encourage us to develop the idea of completing the functionalities of all systems in order to finally

  12. Iterative Goal Refinement for Robotics

    Science.gov (United States)

    2014-06-01

    Iterative Goal Refinement for Robotics Mark Roberts1, Swaroop Vattam1, Ronald Alford2, Bryan Auslander3, Justin Karneeb3, Matthew Molineaux3... robotics researchers and practitioners. We present a goal lifecycle and define a formal model for GR that (1) relates distinct disciplines concerning...researchers to collaborate in exploring this exciting frontier. 1. Introduction Robotic systems often act using incomplete models in environments

  13. Truncated States Obtained by Iteration

    Institute of Scientific and Technical Information of China (English)

    W.B.Cardoso; N.G.de Almeida

    2008-01-01

    We introduce the concept of truncated states obtained via iterative processes(TSI)and study its statistical features,making an analogy with dynamical systems theory(DST).As a specific example,we have studied TSI for the doubring and the logistic functions,which are standard functions in studying chaos.TSI for both the doubling and logistic functions exhibit certain similar patterns when their statistical features are compared from the point of view of DST.

  14. Calculation of Ion Equilibrium Temperature in Ultracold Neutral Plasmas

    Institute of Scientific and Technical Information of China (English)

    李金星; 曹明涛; 韩亮; 齐越蓉; 张首刚; 高宏; 李福利; T.C.Killian

    2011-01-01

    We provide a fast iteration method to calculate the ion equilibrium temperature in an ultracold neutral plasma (UNP). The temperature as functions of electron initial temperature and ion density is obtained and compared with the recent UNP experimental data. The theoretical predictions agree with the experimental results very well. The calculated ion equilibrium temperature by this method can be applied to study the UNP expansion process more effectively.%We provide a fast iteration method to calculate the ion equilibrium temperature in an ultracold neutral plasma (UNP).The temperature as functions of electron initial temperature and ion density is obtained and compared with the recent UNP experimental data.The theoretical predictions agree with the experimental results very well.The calculated ion equilibrium temperature by this method can be applied to study the UNP expansion process more effectively.

  15. Manufacturing preparations for the European Vacuum Vessel Sector for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Jones, Lawrence, E-mail: lawrence.jones@f4e.europa.eu [F4E, c/Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019, Barcelona (Spain); Arbogast, Jean Francois; Bayon, Angel [F4E, c/Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019, Barcelona (Spain); Bianchi, Aldo [Ansaldo Nucleare, Corso F.M. Perrone, 25, I-16152, Genoa (Italy); Caixas, Joan [F4E, c/Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019, Barcelona (Spain); Facca, Aldo; Fachin, Gianbattista [Mangiarotti, Pannellia di Sedegliano, I-33039, Sedegliano (UD) (Italy); Fernandez, Jose [F4E, c/Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019, Barcelona (Spain); Giraud, Benoit [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Losasso, Marcello [F4E, c/Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019, Barcelona (Spain); Loewer, Thorsten [Pro-Beam, Behringstrasse 6, D-82152 Planegg (Germany); Mico, Gonzalo; Pacheco, Jose Miguel [F4E, c/Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019, Barcelona (Spain); Paoletti, Roberto [Walter Tosto, Via Erasmo Piaggio, 72, I-66100 Chieti Scalo (Italy); Sanguinetti, Gian Paolo [Ansaldo Nucleare, Corso F.M. Perrone, 25, I-16152, Genoa (Italy); Stamos, Vassilis [F4E, c/Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019, Barcelona (Spain); Tacconelli, Massimiliano [Walter Tosto, Via Erasmo Piaggio, 72, I-66100 Chieti Scalo (Italy); Trentea, Alexandru [F4E, c/Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019, Barcelona (Spain); Utin, Yuri [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2012-08-15

    The contract for the seven European Sectors of the ITER Vacuum Vessel, which has very tight tolerances and high density of welding, was placed at the end of 2010 with AMW, a consortium of three companies. The start-up of the engineering, including R and D, design and analysis activities of this large and complex contract, one of the largest placed by F4E, the European Domestic Agency for ITER, is described. The statutory and regulatory requirements of ITER Organization and the French Nuclear Safety regulations have made the design development subject to rigorous controls. AMW was able to make use of the previous extensive R and D and prototype work carried out during the past 9 years, especially in relation to advanced welding and inspection techniques. The paper describes the manufacturing methodology with the focus on controlling distortion with predictions by analysis, avoiding use of welded-on jigs, and making use of low heat input narrow-gap welding with electron beam welding as far as possible and narrow-gap TIG when not. Further R and D and more than ten significant mock-ups are described. All these preparations will help to assure the successful manufacture of this critical path item of ITER.

  16. Iterative methods for mixed finite element equations

    Science.gov (United States)

    Nakazawa, S.; Nagtegaal, J. C.; Zienkiewicz, O. C.

    1985-01-01

    Iterative strategies for the solution of indefinite system of equations arising from the mixed finite element method are investigated in this paper with application to linear and nonlinear problems in solid and structural mechanics. The augmented Hu-Washizu form is derived, which is then utilized to construct a family of iterative algorithms using the displacement method as the preconditioner. Two types of iterative algorithms are implemented. Those are: constant metric iterations which does not involve the update of preconditioner; variable metric iterations, in which the inverse of the preconditioning matrix is updated. A series of numerical experiments is conducted to evaluate the numerical performance with application to linear and nonlinear model problems.

  17. Fundamental ion cyclotron resonance heating of JET deuterium plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Krasilnikov, A. V. [Troitsk Institute of Nuclear Physics (TRINITI), Russia; Van Eester, D. [Laboratory for Plasma Physics-ERM/KMS (LPP-ERM/KMS), Brussels, Belgium; Lerche, E. [Laboratory for Plasma Physics-ERM/KMS (LPP-ERM/KMS), Brussels, Belgium; Ongena, J. [Laboratory for Plasma Physics-ERM/KMS (LPP-ERM/KMS), Brussels, Belgium; Amosov, V. N. [Troitsk Institute of Nuclear Physics (TRINITI), Russia; Biewer, Theodore M [ORNL; Bonheure, G. [Laboratory for Plasma Physics-ERM/KMS (LPP-ERM/KMS), Brussels, Belgium; Crombe, K. [Ghent University, Belgium; Ericsson, G. [Uppsala University, Uppsala, Sweden; Esposito, Basilio [ENEA, Frascati; Giacomelli, L. [Uppsala University, Uppsala, Sweden; Hellesen, C. [Uppsala University, Uppsala, Sweden; Hjalmarsson, A. [Uppsala University, Uppsala, Sweden; Jachmich, S. [EURATOM / UKAEA, UK; Kallne, J. [Uppsala University, Uppsala, Sweden; Kaschuck, Yu A [Troitsk Institute of Nuclear Physics (TRINITI), Russia; Kiptily, V. [EURATOM / UKAEA, UK; Leggate, H. [EURATOM / UKAEA, UK; Mailloux, J. [EURATOM / UKAEA, UK; Marocco, D. [ENEA, Frascati; Mayoral, M.-L. [EURATOM / UKAEA, UK; Popovichev, S. [EURATOM / UKAEA, UK; Riva, M. [ENEA, Frascati; Santala, M. [EURATOM / UKAEA, UK; Stamp, M. F. [EURATOM / UKAEA, UK; Vdovin, V. [Russian Research Center, Kurchatov Institute, Moscow, Russia; Walden, A. [EURATOM / UKAEA, UK

    2009-03-01

    Radio frequency heating of majority ions is of prime importance for understanding the basic role of auxiliary heating in the activated D T phase of ITER. Majority deuterium ion cyclotron resonance heating (ICRH) experiments at the fundamental cyclotron frequency were performed in JET. In spite of the poor antenna coupling at 25 MHz, this heating scheme proved promising when adopted in combination with D neutral beam injection (NBI). The effect of fundamental ICRH of a D population was clearly demonstrated in these experiments: by adding ~25% of heating power the fusion power was increased up to 30 50%, depending on the type of NBI adopted. At this power level, the ion and electron temperatures increased from Ti ~ 4.0 keV and Te ~ 4.5 keV (NBI-only phase) to Ti ~ 5.5 keV and Te ~ 5.2 keV (ICRH + NBI phase), respectively. The increase in the neutron yield was stronger when 80 keV rather than 130 keV deuterons were injected in the plasma. It is shown that the neutron rate, the diamagnetic energy and the electron as well as the ion temperature scale roughly linearly with the applied RF power. A synergistic effect of the combined use of ICRF and NBI heating was observed: (i) the number of neutron counts measured by the neutron camera during the combined ICRF + NBI phases of the discharges exceeded the sum of the individual counts of the NBI-only and ICRF-only phases; (ii) a substantial increase in the number of slowing-down beam ions was detected by the time of flight neutron spectrometer when ICRF power was switched on; (iii) a small D subpopulation with energies slightly above the NBI launch energy was detected by the neutral particle analyzer and -ray spectroscopy.

  18. Engineering estimates of impurity fluxes on the ITER port plugs

    Science.gov (United States)

    Kotov, Vladislav

    2016-10-01

    Predictions of impurity fluxes are required for design analysis of the ITER optical diagnostics. In the present paper a simplified model is proposed for calculation of the neutral impurity fluxes on the recessed surfaces which are not in direct contact with plasma. The method is based on the Monte-Carlo simulation of the neutral particles transport in prescribed and fixed plasma background. The plasma parameters are projected from experimental observations, scalings and ITER modelling results. Blobs are approximated as stationary hot species. Results of 2D simulations with toroidally uniform wall and of the ‘2.5D model’ are presented. In this latter the 3D geometry of ports is implemented, but details of the incident ion flux distribution on the first wall panels are neglected. The calculated worst case gross deposition rate of Be in the middle of the port plug faces reaches almost 0.1 nm s-1. At the same time, the obtained Be erosion to deposition ratio at those locations is always larger than 5, indicating high probability of net erosion conditions there.

  19. Meromorphic iterative roots of linear fractional functions

    Institute of Scientific and Technical Information of China (English)

    SHI YongGuo; CHEN Li

    2009-01-01

    Iterative root problem can be regarded as a weak version of the problem of embedding a homeomorphism into a flow. There are many results on iterative roots of monotone functions. However, this problem gets more difficult in non-monotone cases. Therefore, it is interesting to find iterative roots of linear fractional functions (abbreviated as LFFs), a class of non-monotone functions on R. In this paper, iterative roots of LFFs are studied on C. An equivalence between the iterative functional equation for non-constant LFFs and the matrix equation is given. By means of a method of finding matrix roots, general formulae of all meromorphic iterative roots of LFFs are obtained and the precise number of roots is also determined in various cases. As applications, we present all meromorphic iterative roots for functions z and 1/z.

  20. Performance bounds for Lambda Policy Iteration

    CERN Document Server

    Scherrer, Bruno

    2007-01-01

    We consider the discrete-time infinite-horizon discounted stationary optimal control problem formalized by Markov Decision Processes. We study Lambda Policy Iteration, a family of algorithms parameterized by lambda, originally introduced by Ioffe and Bertsekas. Lambda Policy Iteration generalizes the standard algorithms Value Iteration and Policy Iteration, and has some connections with TD(Lambda) introduced by Sutton & Barto. We deepen the original theory developped by Ioffe and Bertsekas by providing convergence rate bounds which generalize standard bounds for Value Iteration described for instance by Puterman. We also develop the theory of this algorithm when it is used in an approximate form. Doing so, we extend and unify the separate analyses developped by Munos for Approximate Value Iteration and Approximate Policy Iteration.

  1. Thermal-hydraulics of LLCB TBM under different ITER operational conditions

    Energy Technology Data Exchange (ETDEWEB)

    Chaudhuri, Paritosh, E-mail: paritc@gmail.com; Ranjithkumar, S.; Sharma, Deepak; Danani, Chandan; Kumar, E. Rajendra

    2016-11-01

    Highlights: • Thermal analysis of LLCB TBM has been performed for different ITER operations during normal and power excursion scenarios. • It is observed that in all ITER operations, the temperatures in all components of LLCB TBM are well within the limits. • The maximum temperature of CB-1 in steady state is 920 °C, which is 72% of the maximum temperature obtained in ITER pulse. - Abstract: Lead–lithium cooled ceramic breeder (LLCB) TBM is the Indian DEMO relevant blanket module, which will be tested in one-half of a designated ITER port from day 1 operation of ITER. LLCB TBM will be tested in various ITER operational phases. Different TBM may be tested corresponding to the four different phases of plasma operation. The paper present a review of thermal analyses based on the various heat loads on the LLCB TBM obtained from neutronics calculations to confirm heat removal and structural integrity under different ITER operational conditions. This includes the normal plasma operation, fusion power excursions, surface heat flux due to MARFES, disruption heat loads during thermal and current quench etc., in inductive and non-inductive plasma operations. Transient events, like plasma disruption, vertical displacement events (VDEs) and edge localized modes (ELMs) deliver considerable heat flux (high power densities) for short durations onto the first wall of TBM. Detailed thermal-hydraulic simulation studies have been performed using ANSYS. These analyses, therefore, will provide important information about the temperature distribution in different materials used in TBM and their temperature-time histories.

  2. Nuclear analysis of ITER Test Blanket Module Port Plug

    Energy Technology Data Exchange (ETDEWEB)

    Villari, Rosaria, E-mail: rosaria.villari@enea.it [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Kim, Byoung Yoon; Barabash, Vladimir; Giancarli, Luciano; Levesy, Bruno; Loughlin, Michael; Merola, Mario [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 Saint Paul-lez-Durance Cedex (France); Moro, Fabio [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Pascal, Romain [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 Saint Paul-lez-Durance Cedex (France); Petrizzi, Luigino [European Commission, DG Research & Innovation G5, CDMA 00/030, B-1049 Brussels (Belgium); Polunovsky, Eduard; Van Der Laan, Jaap G. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 Saint Paul-lez-Durance Cedex (France)

    2015-10-15

    Highlights: • 3D nuclear analysis of the ITER TBM Port Plug (PP). • Calculations of neutron fluxes, nuclear heating, damage and He-production in TBM PP components. • Shutdown dose rate assessment with Advanced D1S method considering different configurations. • Potential design improvements to reduce the shutdown dose rate in the port interspace. - Abstract: Nuclear analyses have been performed for the ITER Test Blanket Module Port Plug (TBM PP) using the MCNP-5 Monte Carlo Code. A detailed 3D model of the TBM Port Plug with dummy TBM has been integrated into the ITER MCNP model (B-lite v.3). Neutron fluxes, nuclear heating, helium production and neutron damage have been calculated in all the TBM PP components. Global shutdown dose rate calculations have also been performed with Advanced D1S method for different configurations of the TBM PP system. This paper presents the results of these analyses and discusses potential design improvements aiming to further reduce the shutdown dose rate in the port interspace.

  3. EC power management and NTM control in ITER

    Science.gov (United States)

    Poli, Francesca; Fredrickson, E.; Henderson, M.; Bertelli, N.; Farina, D.; Figini, L.; Nowak, S.; Poli, E.; Sauter, O.

    2016-10-01

    The suppression of Neoclassical Tearing Modes (NTMs) is an essential requirement for the achievement of the demonstration baseline in ITER. The Electron Cyclotron upper launcher is specifically designed to provide highly localized heating and current drive for NTM stabilization. In order to assess the power management for shared applications, we have performed time-dependent simulations for ITER scenarios covering operation from half to full field. The free-boundary TRANSP simulations evolve the magnetic equilibrium and the pressure profiles in response to the heating and current drive sources and are interfaced with a GRE for the evolution of size and frequency of the magnetic islands. Combined with a feedback control of the EC power and the steering angle, these simulations are used to model the plasma response to NTM control, accounting for the misalignment of the EC deposition with the resonant surfaces, uncertainties in the magnetic equilibrium reconstruction and in the magnetic island detection threshold. Simulations indicate that the threshold for detection of the island should not exceed 2-3cm, that pre-emptive control is a preferable option, and that for safe operation the power needed for NTM control should be reserved, rather than shared with other applications. Work supported by ITER under IO/RFQ/13/9550/JTR and by DOE under DE-AC02-09CH11466.

  4. Cryogenic instrumentation for ITER magnets

    Science.gov (United States)

    Poncet, J.-M.; Manzagol, J.; Attard, A.; André, J.; Bizel-Bizellot, L.; Bonnay, P.; Ercolani, E.; Luchier, N.; Girard, A.; Clayton, N.; Devred, A.; Huygen, S.; Journeaux, J.-Y.

    2017-02-01

    Accurate measurements of the helium flowrate and of the temperature of the ITER magnets is of fundamental importance to make sure that the magnets operate under well controlled and reliable conditions, and to allow suitable helium flow distribution in the magnets through the helium piping. Therefore, the temperature and flow rate measurements shall be reliable and accurate. In this paper, we present the thermometric chains as well as the venturi flow meters installed in the ITER magnets and their helium piping. The presented thermometric block design is based on the design developed by CERN for the LHC, which has been further optimized via thermal simulations carried out by CEA. The electronic part of the thermometric chain was entirely developed by the CEA and will be presented in detail: it is based on a lock-in measurement and small signal amplification, and also provides a web interface and software to an industrial PLC. This measuring device provides a reliable, accurate, electromagnetically immune, and fast (up to 100 Hz bandwidth) system for resistive temperature sensors between a few ohms to 100 kΩ. The flowmeters (venturi type) which make up part of the helium mass flow measurement chain have been completely designed, and manufacturing is on-going. The behaviour of the helium gas has been studied in detailed thanks to ANSYS CFX software in order to obtain the same differential pressure for all types of flowmeters. Measurement uncertainties have been estimated and the influence of input parameters has been studied. Mechanical calculations have been performed to guarantee the mechanical strength of the venturis required for pressure equipment operating in nuclear environment. In order to complete the helium mass flow measurement chain, different technologies of absolute and differential pressure sensors have been tested in an applied magnetic field to identify equipment compatible with the ITER environment.

  5. Spatial calibration of a tokamak neutral beam diagnostic using in situ neutral beam emission

    Energy Technology Data Exchange (ETDEWEB)

    Chrystal, C. [Oak Ridge Associated Universities, Oak Ridge, Tennessee 37831 (United States); Burrell, K. H.; Pace, D. C. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Grierson, B. A. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States)

    2015-10-15

    Neutral beam injection is used in tokamaks to heat, apply torque, drive non-inductive current, and diagnose plasmas. Neutral beam diagnostics need accurate spatial calibrations to benefit from the measurement localization provided by the neutral beam. A new technique has been developed that uses in situ measurements of neutral beam emission to determine the spatial location of the beam and the associated diagnostic views. This technique was developed to improve the charge exchange recombination (CER) diagnostic at the DIII-D tokamak and uses measurements of the Doppler shift and Stark splitting of neutral beam emission made by that diagnostic. These measurements contain information about the geometric relation between the diagnostic views and the neutral beams when they are injecting power. This information is combined with standard spatial calibration measurements to create an integrated spatial calibration that provides a more complete description of the neutral beam-CER system. The integrated spatial calibration results are very similar to the standard calibration results and derived quantities from CER measurements are unchanged within their measurement errors. The methods developed to perform the integrated spatial calibration could be useful for tokamaks with limited physical access.

  6. Beryllium in the ITER blanket

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.C.

    1995-01-01

    This paper consists of viewgraphs used in a presentation on the application of beryllium in breeding blankets for ITER and JET. The paper brings together data on the physical, thermal, mechanical, and chemical properties of beryllium and beryllium oxide for this type of application, as well as issues of compatibility with construction materials, and irradiation experience. It includes the results from testing programs carried out to arrive at some of the information, including fabrication work, irradiation experiments, and sample tests performed both in and out of the irradiation piles.

  7. Measurement and control system for ITER remote maintenance equipment

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Kiyoshi; Kakudate, Satoshi; Takeda, Nobukazu; Takiguchi, Yuji; Akou, Kentaro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    ITER in-vessel components such as blankets and divertors are categorized as scheduled maintenance components because they are subjected to severe plasma heat and particle loads. Blanket maintenance requires remote handling equipment and tools able to handle Heavy payloads of about 4 tons within a 2 mm precision tolerance. Divertor maintenance requires remote replacement of 60 cassettes with a dead weight of about 25 tons each. In the ITER R and D program, full-scale remote handling equipment for blanket and divertor maintenance has been designed and assembled for demonstration tests. This paper reviews the measurement and control system developed for full-scale remote handling equipment, the Japan Home Team contribution. (author)

  8. Conceptual design of the ITER fast-ion loss detector

    Science.gov (United States)

    Garcia-Munoz, M.; Kocan, M.; Ayllon-Guerola, J.; Bertalot, L.; Bonnet, Y.; Casal, N.; Galdon, J.; Garcia Lopez, J.; Giacomin, T.; Gonzalez-Martin, J.; Gunn, J. P.; Jimenez-Ramos, M. C.; Kiptily, V.; Pinches, S. D.; Rodriguez-Ramos, M.; Reichle, R.; Rivero-Rodriguez, J. F.; Sanchis-Sanchez, L.; Snicker, A.; Vayakis, G.; Veshchev, E.; Vorpahl, Ch.; Walsh, M.; Walton, R.

    2016-11-01

    A conceptual design of a reciprocating fast-ion loss detector for ITER has been developed and is presented here. Fast-ion orbit simulations in a 3D magnetic equilibrium and up-to-date first wall have been carried out to revise the measurement requirements for the lost alpha monitor in ITER. In agreement with recent observations, the simulations presented here suggest that a pitch-angle resolution of ˜5° might be necessary to identify the loss mechanisms. Synthetic measurements including realistic lost alpha-particle as well as neutron and gamma fluxes predict scintillator signal-to-noise levels measurable with standard light acquisition systems with the detector aperture at ˜11 cm outside of the diagnostic first wall. At measurement position, heat load on detector head is comparable to that in present devices.

  9. A new technique to measure the neutralizer cell gas line density applied to a DIII-D neutral beamline

    Energy Technology Data Exchange (ETDEWEB)

    Kessler, D.N.; Hong, R.M.; Riggs, S.P.

    1995-10-01

    The DIII-D tokamak employs eight ion sources for plasma heating. In order to obtain the maximum neutralization of energetic ions (providing maximum neutral beam power) and reduce the heat load on beamline internal components caused by residual energetic ions, sufficient neutral gas must be injected into the beamline neutralizer cell. The neutral gas flow rate must be optimized, however, since excessive gas will increase power losses due to neutral beam scattering and reionization. It is important, therefore, to be able to determine the neutralizer cell gas line density. A new technique which uses the ion source suppressor grid current to obtain the neutralizer cell gas line density has been developed. The technique uses the fact that slow ions produced by beam-gas interactions in the neutralizer cell during beam extraction are attracted to the negative potential applied to the suppressor grid, inducing current flow in the grid. By removing the dependence on beam energy and beam current a normalized suppressor grid current function can be formed which is dependent only on the gas line density. With this technique it is possible to infer the gas line density on a shot by shot basis.

  10. IHadoop: Asynchronous iterations for MapReduce

    KAUST Repository

    Elnikety, Eslam Mohamed Ibrahim

    2011-11-01

    MapReduce is a distributed programming frame-work designed to ease the development of scalable data-intensive applications for large clusters of commodity machines. Most machine learning and data mining applications involve iterative computations over large datasets, such as the Web hyperlink structures and social network graphs. Yet, the MapReduce model does not efficiently support this important class of applications. The architecture of MapReduce, most critically its dataflow techniques and task scheduling, is completely unaware of the nature of iterative applications; tasks are scheduled according to a policy that optimizes the execution for a single iteration which wastes bandwidth, I/O, and CPU cycles when compared with an optimal execution for a consecutive set of iterations. This work presents iHadoop, a modified MapReduce model, and an associated implementation, optimized for iterative computations. The iHadoop model schedules iterations asynchronously. It connects the output of one iteration to the next, allowing both to process their data concurrently. iHadoop\\'s task scheduler exploits inter-iteration data locality by scheduling tasks that exhibit a producer/consumer relation on the same physical machine allowing a fast local data transfer. For those iterative applications that require satisfying certain criteria before termination, iHadoop runs the check concurrently during the execution of the subsequent iteration to further reduce the application\\'s latency. This paper also describes our implementation of the iHadoop model, and evaluates its performance against Hadoop, the widely used open source implementation of MapReduce. Experiments using different data analysis applications over real-world and synthetic datasets show that iHadoop performs better than Hadoop for iterative algorithms, reducing execution time of iterative applications by 25% on average. Furthermore, integrating iHadoop with HaLoop, a variant Hadoop implementation that caches

  11. ETR/ITER systems code

    Energy Technology Data Exchange (ETDEWEB)

    Barr, W.L.; Bathke, C.G.; Brooks, J.N.; Bulmer, R.H.; Busigin, A.; DuBois, P.F.; Fenstermacher, M.E.; Fink, J.; Finn, P.A.; Galambos, J.D.; Gohar, Y.; Gorker, G.E.; Haines, J.R.; Hassanein, A.M.; Hicks, D.R.; Ho, S.K.; Kalsi, S.S.; Kalyanam, K.M.; Kerns, J.A.; Lee, J.D.; Miller, J.R.; Miller, R.L.; Myall, J.O.; Peng, Y-K.M.; Perkins, L.J.; Spampinato, P.T.; Strickler, D.J.; Thomson, S.L.; Wagner, C.E.; Willms, R.S.; Reid, R.L. (ed.)

    1988-04-01

    A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs.

  12. ITER Port Interspace Pressure Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, Juan J [ORNL; Van Hove, Walter A [ORNL

    2016-01-01

    The ITER Vacuum Vessel (VV) is equipped with 54 access ports. Each of these ports has an opening in the bioshield that communicates with a dedicated port cell. During Tokamak operation, the bioshield opening must be closed with a concrete plug to shield the radiation coming from the plasma. This port plug separates the port cell into a Port Interspace (between VV closure lid and Port Plug) on the inner side and the Port Cell on the outer side. This paper presents calculations of pressures and temperatures in the ITER (Ref. 1) Port Interspace after a double-ended guillotine break (DEGB) of a pipe of the Tokamak Cooling Water System (TCWS) with high temperature water. It is assumed that this DEGB occurs during the worst possible conditions, which are during water baking operation, with water at a temperature of 523 K (250 C) and at a pressure of 4.4 MPa. These conditions are more severe than during normal Tokamak operation, with the water at 398 K (125 C) and 2 MPa. Two computer codes are employed in these calculations: RELAP5-3D Version 4.2.1 (Ref. 2) to calculate the blowdown releases from the pipe break, and MELCOR, Version 1.8.6 (Ref. 3) to calculate the pressures and temperatures in the Port Interspace. A sensitivity study has been performed to optimize some flow areas.

  13. Iterated Stretching of Viscoelastic Jets

    Science.gov (United States)

    Chang, Hsueh-Chia; Demekhin, Evgeny A.; Kalaidin, Evgeny

    1999-01-01

    We examine, with asymptotic analysis and numerical simulation, the iterated stretching dynamics of FENE and Oldroyd-B jets of initial radius r(sub 0), shear viscosity nu, Weissenberg number We, retardation number S, and capillary number Ca. The usual Rayleigh instability stretches the local uniaxial extensional flow region near a minimum in jet radius into a primary filament of radius [Ca(1 - S)/ We](sup 1/2)r(sub 0) between two beads. The strain-rate within the filament remains constant while its radius (elastic stress) decreases (increases) exponentially in time with a long elastic relaxation time 3We(r(sup 2, sub 0)/nu). Instabilities convected from the bead relieve the tension at the necks during this slow elastic drainage and trigger a filament recoil. Secondary filaments then form at the necks from the resulting stretching. This iterated stretching is predicted to occur successively to generate high-generation filaments of radius r(sub n), (r(sub n)/r(sub 0)) = square root of 2[r(sub n-1)/r(sub 0)](sup 3/2) until finite-extensibility effects set in.

  14. Bulk Ion Heating with ICRF Waves in Tokamaks

    DEFF Research Database (Denmark)

    Mantsinen, M. J.; Bilato, R.; Bobkov, V. V.

    2015-01-01

    Heating with ICRF waves is a well-established method on present-day tokamaks and one of the heating systems foreseen for ITER. However, further work is still needed to test and optimize its performance in fusion devices with metallic high-Z plasma facing components (PFCs) in preparation of ITER a...

  15. Neutral particle lithography

    Science.gov (United States)

    Craver, Barry Paul

    Neutral particle lithography (NPL) is a high resolution, proximity exposure technique where a broad beam of energetic neutral particles floods a stencil mask and transmitted beamlets transfer the mask pattern to resist on a substrate, such that each feature is printed in parallel, rather than in the serial manner of electron beam lithography. It preserves the advantages of ion beam lithography (IBL), including extremely large depth-of-field, sub-5 nm resist scattering, and the near absence of diffraction, yet is intrinsically immune to charge-related artifacts including line-edge roughness and pattern placement errors due to charge accumulation on the mask and substrate. In our experiments, a neutral particle beam is formed by passing an ion beam (e.g., 30 keV He+) through a high pressure helium gas cell (e.g., 100 mTorr) to convert the ions to energetic neutrals through charge transfer scattering. The resolution of NPL is generally superior to that of IBL for applications involving insulating substrates, large proximity gaps, and ultra-small features. High accuracy stepped exposures with energetic neutral particles, where magnetic or electrostatic deflection is impossible, have been obtained by clamping the mask to the wafer, setting the proximity gap with a suitable spacer, and mechanically inclining the mask/wafer stack relative to the beam. This approach is remarkably insensitive to vibration and thermal drift; nanometer scale image offsets have been obtained with +/-2 nm placement accuracy for experiments lasting over one hour. Using this nanostepping technique, linewidth versus dose curves were obtained, from which the NPL lithographic blur was determined as 4.4+/-1.4 nm (1sigma), which is 2-3 times smaller than the blur of electron beam lithography. Neutral particle lithography has the potential to form high density, periodic patterns with sub-10 nm resolution.

  16. Characterization of plasma sprayed beryllium ITER first wall mockups

    Energy Technology Data Exchange (ETDEWEB)

    Castro, R.G.; Vaidya, R.U.; Hollis, K.J. [Los Alamos National Lab., NM (United States). Material Science and Technology Div.

    1998-01-01

    ITER first wall beryllium mockups, which were fabricated by vacuum plasma spraying the beryllium armor, have survived 3000 thermal fatigue cycles at 1 MW/m{sup 2} without damage during high heat flux testing at the Plasma Materials Test Facility at Sandia National Laboratory in New Mexico. The thermal and mechanical properties of the plasma sprayed beryllium armor have been characterized. Results are reported on the chemical composition of the beryllium armor in the as-deposited condition, the through thickness and normal to the through thickness thermal conductivity and thermal expansion, the four-point bend flexure strength and edge-notch fracture toughness of the beryllium armor, the bond strength between the beryllium armor and the underlying heat sink material, and ultrasonic C-scans of the Be/heat sink interface. (author)

  17. Benchmarking ICRF Full-wave Solvers for ITER

    Energy Technology Data Exchange (ETDEWEB)

    R. V. Budny, L. Berry, R. Bilato, P. Bonoli, M. Brambilla, R. J. Dumont, A. Fukuyama, R. Harvey, E. F. Jaeger, K. Indireshkumar, E. Lerche, D. McCune, C. K. Phillips, V. Vdovin, J. Wright, and members of the ITPA-IOS

    2011-01-06

    Abstract Benchmarking of full-wave solvers for ICRF simulations is performed using plasma profiles and equilibria obtained from integrated self-consistent modeling predictions of four ITER plasmas. One is for a high performance baseline (5.3 T, 15 MA) DT H-mode. The others are for half-field, half-current plasmas of interest for the pre-activation phase with bulk plasma ion species being either hydrogen or He4. The predicted profiles are used by six full-wave solver groups to simulate the ICRF electromagnetic fields and heating, and by three of these groups to simulate the current-drive. Approximate agreement is achieved for the predicted heating power for the DT and He4 cases. Factor of two disagreements are found for the cases with second harmonic He3 heating in bulk H cases. Approximate agreement is achieved simulating the ICRF current drive.

  18. Nuclear analysis of the ITER Cryopump Ports

    Energy Technology Data Exchange (ETDEWEB)

    Moro, Fabio, E-mail: fabio.moro@enea.it [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Villari, Rosaria; Flammini, Davide [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Antipenkov, Alexander; Dremel, Matthias; Levesy, Bruno; Loughlin, Michael [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-lez-Durance (France); Juarez, Rafael; Perez, Lucia [UNED, Energetic Engineering Department, C/Juan del Rosal 12, Madrid (Spain); Petrizzi, Luigino [European Commission, DG Research & Innovation G5, CDMA 00/030, B-1049 Brussels (Belgium)

    2015-10-15

    Highlights: • Evaluation the shielding effectiveness of the TCPHs by means of 3-D neutrons and gamma maps. • Assessment of the nuclear heating induced by neutron and photons on the TCP and TCPHs. • Calculation of the dose rate at 12 days after shutdown in the maintenance area of the Lower Ports with the Advanced D1S method, in order to verify the design target (100 μSv/h). • Potential improvements of the shielding configuration aimed at the reduction of the dose level in the Port Cell have been proposed and discussed. - Abstract: The ITER machine will be equipped with 6 torus Cryopumps (TCP) that are positioned in their housings (TCPH) and integrated into the cryostat walls at B1 level in the port cells. A comprehensive nuclear analysis of the Cryopump Ports #4 and #12 has been carried out by means of the MCNP-5 Monte Carlo code in a full 3-D geometry, providing guidelines for the design of the embedded components. Radiation transport calculations have been performed in order to determine the radiation field inside the Lower Ports, up the Port Cell: 3-D neutrons and gamma maps have been provided in order to evaluate the shielding effectiveness of the TCPHs. Nuclear heating induced by neutron and photons have been estimated on the TCP and TCPH to assess the nuclear loads during plasma operations. The shutdown dose rate in the maintenance area of the Lower Ports has been assessed with the Advanced D1S method to verify the design limits.

  19. Electron cyclotron emission diagnostic for ITER.

    Science.gov (United States)

    Rowan, W; Austin, M; Beno, J; Ellis, R; Feder, R; Ouroua, A; Patel, A; Phillips, P

    2010-10-01

    Electron temperature measurements and electron thermal transport inferences will be critical to the nonactive and deuterium phases of ITER operation and will take on added importance during the alpha heating phase. The diagnostic must meet stringent criteria on spatial coverage and spatial resolution during full field operation. During the early phases of operation, it must operate equally well at half field. The key to the diagnostic is the front end design. It consists of a quasioptical antenna and a pair of calibration sources. The radial resolution of the diagnostic is less than 0.06 m. The spatial coverage extends at least from the core to the separatrix with first harmonic O-mode being used for the core and second harmonic X-mode being used for the pedestal. The instrumentation used for the core measurement at full field can be used for detection at half field by changing the detected polarization. Intermediate fields are accessible. The electron cyclotron emission systems require in situ calibration, which is provided by a novel hot calibration source. The critical component for the hot calibration source, the emissive surface, has been successfully tested. A prototype hot calibration source has been designed, making use of extensive thermal and mechanical modeling.

  20. A FAST CONVERGENT METHOD OF ITERATED REGULARIZATION

    Institute of Scientific and Technical Information of China (English)

    Huang Xiaowei; Wu Chuansheng; Wu Di

    2009-01-01

    This article presents a fast convergent method of iterated regularization based on the idea of Landweber iterated regularization, and a method for a-posteriori choice by the Morozov discrepancy principle and the optimum asymptotic convergence order of the regularized solution is obtained. Numerical test shows that the method of iterated regu-larization can quicken the convergence speed and reduce the calculation burden efficiently.

  1. Preconditioned iterations to calculate extreme eigenvalues

    Energy Technology Data Exchange (ETDEWEB)

    Brand, C.W.; Petrova, S. [Institut fuer Angewandte Mathematik, Leoben (Austria)

    1994-12-31

    Common iterative algorithms to calculate a few extreme eigenvalues of a large, sparse matrix are Lanczos methods or power iterations. They converge at a rate proportional to the separation of the extreme eigenvalues from the rest of the spectrum. Appropriate preconditioning improves the separation of the eigenvalues. Davidson`s method and its generalizations exploit this fact. The authors examine a preconditioned iteration that resembles a truncated version of Davidson`s method with a different preconditioning strategy.

  2. Iterative solution of the reduced eigenvalue problem

    Energy Technology Data Exchange (ETDEWEB)

    Sauer, G. (Technischer Ueberwachungs-Verein Bayern e.V., Muenchen (Germany, F.R.))

    1991-04-01

    The Guyan method of reducing the stiffness and mass matrices of large linear structures introduces errors in the reduced mass matrix. These errors cannot be completely avoided even if the analysis coordinates are chosen optimally. However, they can be elimiated by iterating on the eigenvectors found from the Guyan reduced matrices. The necessary iteration steps follow directly from the eigenvalue problem. The resulting iteration procedures are presented and applied to two test problems showing that the iterations enable the exact eigensolutions to be extracted. All errors from the Guyan reduced matrices are removed or substantially decreased. (orig.).

  3. Interaction of a neutral cloud moving through a magnetized plasma

    Science.gov (United States)

    Goertz, C. K.; Lu, G.

    1990-01-01

    Current collection by outgassing probes in motion relative to a magnetized plasma may be significantly affected by plasma processes that cause electron heating and cross field transport. Simulations of a neutral gas cloud moving across a static magnetic field are discussed. The authors treat a low-Beta plasma and use a 2-1/2 D electrostatic code linked with the authors' Plasma and Neutral Interaction Code (PANIC). This study emphasizes the understanding of the interface between the neutral gas cloud and the surrounding plasma where electrons are heated and can diffuse across field lines. When ionization or charge exchange collisions occur a sheath-like structure is formed at the surface of the neutral gas. In that region the crossfield component of the electric field causes the electron to E times B drift with a velocity of the order of the neutral gas velocity times the square root of the ion to electron mass ratio. In addition a diamagnetic drift of the electron occurs due to the number density and temperature inhomogeneity in the front. These drift currents excite the lower-hybrid waves with the wave k-vectors almost perpendicular to the neutral flow and magnetic field again resulting in electron heating. The thermal electron current is significantly enhanced due to this heating.

  4. Bleach Neutralizes Mold Allergens

    Science.gov (United States)

    Science Teacher, 2005

    2005-01-01

    Researchers at National Jewish Medical and Research Center have demonstrated that dilute bleach not only kills common household mold, but may also neutralize the mold allergens that cause most mold-related health complaints. The study, published in the Journal of Allergy and Clinical Immunology, is the first to test the effect on allergic…

  5. CO2-Neutral Fuels

    NARCIS (Netherlands)

    Goede, A.; van de Sanden, M. C. M.

    2016-01-01

    Mimicking the biogeochemical cycle of System Earth, synthetic hydrocarbon fuels are produced from recycled CO2 and H2O powered by renewable energy. Recapturing CO2 after use closes the carbon cycle, rendering the fuel cycle CO2 neutral. Non-equilibrium molecular CO2 vibrations are key to high energy

  6. CO2-Neutral Fuels

    Science.gov (United States)

    Goede, Adelbert; van de Sanden, Richard

    2016-06-01

    Mimicking the biogeochemical cycle of System Earth, synthetic hydrocarbon fuels are produced from recycled CO2 and H2O powered by renewable energy. Recapturing CO2 after use closes the carbon cycle, rendering the fuel cycle CO2 neutral. Non-equilibrium molecular CO2 vibrations are key to high energy efficiency.

  7. CO2-Neutral Fuels

    NARCIS (Netherlands)

    Goede, A.; van de Sanden, M. C. M.

    2016-01-01

    Mimicking the biogeochemical cycle of System Earth, synthetic hydrocarbon fuels are produced from recycled CO2 and H2O powered by renewable energy. Recapturing CO2 after use closes the carbon cycle, rendering the fuel cycle CO2 neutral. Non-equilibrium molecular CO2 vibrations are key to high energy

  8. Truncated states obtained by iteration

    CERN Document Server

    Cardoso, W B

    2007-01-01

    Quantum states of the electromagnetic field are of considerable importance, finding potential application in various areas of physics, as diverse as solid state physics, quantum communication and cosmology. In this paper we introduce the concept of truncated states obtained via iterative processes (TSI) and study its statistical features, making an analogy with dynamical systems theory (DST). As a specific example, we have studied TSI for the doubling and the logistic functions, which are standard functions in studying chaos. TSI for both the doubling and logistic functions exhibit certain similar patterns when their statistical features are compared from the point of view of DST. A general method to engineer TSI in the running-wave domain is employed, which includes the errors due to the nonidealities of detectors and photocounts.

  9. Planning as an Iterative Process

    Science.gov (United States)

    Smith, David E.

    2012-01-01

    Activity planning for missions such as the Mars Exploration Rover mission presents many technical challenges, including oversubscription, consideration of time, concurrency, resources, preferences, and uncertainty. These challenges have all been addressed by the research community to varying degrees, but significant technical hurdles still remain. In addition, the integration of these capabilities into a single planning engine remains largely unaddressed. However, I argue that there is a deeper set of issues that needs to be considered namely the integration of planning into an iterative process that begins before the goals, objectives, and preferences are fully defined. This introduces a number of technical challenges for planning, including the ability to more naturally specify and utilize constraints on the planning process, the ability to generate multiple qualitatively different plans, and the ability to provide deep explanation of plans.

  10. ITER Safety Analyses with ISAS

    Science.gov (United States)

    Gulden, W.; Nisan, S.; Porfiri, M.-T.; Toumi, I.; de Gramont, T. Boubée

    1997-06-01

    Detailed analyses of accident sequences for the International Thermonuclear Experimental Reactor (ITER), from an initiating event to the environmental release of activity, have involved in the past the use of different types of computer codes in a sequential manner. Since these codes were developed at different time scales in different countries, there is no common computing structure to enable automatic data transfer from one code to the other, and no possibility exists to model or to quantify the effect of coupled physical phenomena. To solve this problem, the Integrated Safety Analysis System of codes (ISAS) is being developed, which allows users to integrate existing computer codes in a coherent manner. This approach is based on the utilization of a command language (GIBIANE) acting as a “glue” to integrate the various codes as modules of a common environment. The present version of ISAS allows comprehensive (coupled) calculations of a chain of codes such as ATHENA (thermal-hydraulic analysis of transients and accidents), INTRA (analysis of in-vessel chemical reactions, pressure built-up, and distribution of reaction products inside the vacuum vessel and adjacent rooms), and NAUA (transport of radiological species within buildings and to the environment). In the near future, the integration of S AFALY (simultaneous analysis of plasma dynamics and thermal behavior of in-vessel components) is also foreseen. The paper briefly describes the essential features of ISAS development and the associated software architecture. It gives first results of a typical ITER accident sequence, a loss of coolant accident (LOCA) in the divertor cooling loop inside the vacuum vessel, amply demonstrating ISAS capabilities.

  11. Simulation of dust production in ITER transient events

    Energy Technology Data Exchange (ETDEWEB)

    Pestchanyi, S. [Forschungszentrum Karlsruhe (Germany)

    2007-07-01

    The tritium retention problem is a critical issue for the tokamak ITER performance. Tritium is trapped in redeposited T-C layers and at the surface of carbon dust, where it is retained in form of various hydrocarbons. The area of dust surface and hence, the amount of tritium deposited on the surface depends on the dust amount and of the dust sizes. The carbon dust appears as a result of brittle destruction at the surface of the carbon fibre composite (CFC) which is now the reference armour material for the most loaded part of tokamak divertor. Stationary heat flux on the ITER divertor armour does not cause its brittle destruction and does not produce dust. However, according to the modern understanding of tokamak fusion devices performance, the most attractive regime of ITER operation is the ELMy H mode. This regime is associated with a repetitive short time increase of heat flux at the CFC divertor armour of 2-3 orders of magnitude over its stationary value during edge localized modes (ELMs). Under influence of these severe heat shocks CFC armour can crack due to the thermostress, producing a dust of carbon. Besides, a carbon dust produced during disruptions due to brittle destruction of the armour under influence of thermoshock. Most of the modern tokamaks do not produce the ELMs powerful enough to cause CFC brittle destruction at the divertor surface, except of very special regimes in JET. This is why the CFC erosion and dust production could be investigated now only theoretically and experimentally in plasma guns and electron beam facilities. Simulation of the CFC brittle destruction has been done using the code PEGASUS already developed and tested in FZK for simulation of erosion for ITER candidate materials under the heat shocks. After upgrades the code was used for simulation of the amount of carbon dust particles and of the distribution of their sizes. The code has been tested against available experimental data from the plasma gun MK-200UG and from the

  12. Estimation of carbon fibre composites as ITER divertor armour

    Science.gov (United States)

    Pestchanyi, S.; Safronov, V.; Landman, I.

    2004-08-01

    Exposure of the carbon fibre composites (CFC) NB31 and NS31 by multiple plasma pulses has been performed at the plasma guns MK-200UG and QSPA. Numerical simulation for the same CFCs under ITER type I ELM typical heat load has been carried out using the code PEGASUS-3D. Comparative analysis of the numerical and experimental results allowed understanding the erosion mechanism of CFC based on the simulation results. A modification of CFC structure has been proposed in order to decrease the armour erosion rate.

  13. Aggregation-iterative analogues and generalizations of projection-iterative methods

    Directory of Open Access Journals (Sweden)

    Shuvar B.F.

    2013-06-01

    Full Text Available Aggregation-iterative algorithms for linear operator equations are constructed and investigated. These algorithms cover methods of iterative aggregation and projection-iterative methods. In convergence conditions there is neither requirement for the corresponding operator of fixed sign no restriction to the spectral radius to be less than one.

  14. Installation of the ITER committee industry. Participants guide; Installation du Comite industrie ITER. Dossier des participants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-07-01

    ITER is an international project to design and build an experimental fusion reactor based on the tokamak concept. This guide presents the ITER project and objectives and the associated organizations in France, the recommendations and actions for ITER, the industrial mobilization, the industrial committee and its members, technological sheets for the enterprises and the statistical document of the SESSI. (A.L.B.)

  15. Design and manufacturing of the ITER ECRH upper launcher mirrors

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, Francisco [Ecole Polytechnique Federale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, Association EURATOMConfederation, Suisse, CH-1015 Lausanne (Switzerland)], E-mail: francisco.sanchez@epfl.ch; Bertizzolo, R.; Chavan, R.; Collazos, A. [Ecole Polytechnique Federale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, Association EURATOMConfederation, Suisse, CH-1015 Lausanne (Switzerland); Henderson, M. [ITER Organization, Cadarache Centre, Saint Paul Lez Durance (France); Landis, J.D. [Ecole Polytechnique Federale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, Association EURATOMConfederation, Suisse, CH-1015 Lausanne (Switzerland)

    2009-06-15

    Four of the 16 ITER upper port plugs will be devoted to electron cyclotron resonance heating (ECRH) in order to control magnetohydrodynamic (MHD) instabilities . In order to achieve the stabilisation of the neoclassical tearing modes (NTM) and sawtooth oscillation, a deposition of a very localized and peaked current density profile over a broad poloidal steering range is required. In the present optical configuration eight 2 MW mm-wave beams enter each of the four upper launchers (UL) through waveguides into the vacuum vessel. Each beam line comprises consecutive corrugated waveguide sections with two mitre bends, orientating the poloidal and toroidal directions and three sections of quasi-optical transmission . The beam waist locations and beam shaping properties in free space propagation are defined by two additional mirrors, the first being a static focusing mirror and the second a plane poloidally steerable mirror. Each mirror reflects a group of 4 mm-wave beams. The three types of UL mirrors (mitre bend, focusing and steering) absorb heat generated essentially by three sources: the ohmic loss of the RF beam reflected at the mirror surfaces and the nuclear and thermal radiation coming from the plasma. While the average heat load is within reasonable engineering limits, three elements condition the actual mirror design, the peak ohmic heat load (Gaussian or Bessel type heat deposition profiles), the electromagnetic forces generated in vertical disruption events (VDE), and the ITER cooling water requirements. This paper provides an overview of the different upper port-plug mirror designs and cooling schemes and an outlook on the prototype manufacturing activities and the future test program. The optimized mm-wave layout within the ECH port plugs is also presented.

  16. Rater Variables Associated with ITER Ratings

    Science.gov (United States)

    Paget, Michael; Wu, Caren; McIlwrick, Joann; Woloschuk, Wayne; Wright, Bruce; McLaughlin, Kevin

    2013-01-01

    Advocates of holistic assessment consider the ITER a more authentic way to assess performance. But this assessment format is subjective and, therefore, susceptible to rater bias. Here our objective was to study the association between rater variables and ITER ratings. In this observational study our participants were clerks at the University of…

  17. Dense Iterative Contextual Pixel Classification using Kriging

    DEFF Research Database (Denmark)

    Ganz, Melanie; Loog, Marco; Brandt, Sami

    2009-01-01

    have been proposed to this end, e.g., iterative contextual pixel classification, iterated conditional modes, and other approaches related to Markov random fields. A problem of these methods, however, is their computational complexity, especially when dealing with high-resolution images in which...

  18. Iterative Brinkman penalization for remeshed vortex methods

    DEFF Research Database (Denmark)

    Hejlesen, Mads Mølholm; Koumoutsakos, Petros; Leonard, Anthony;

    2015-01-01

    We introduce an iterative Brinkman penalization method for the enforcement of the no-slip boundary condition in remeshed vortex methods. In the proposed method, the Brinkman penalization is applied iteratively only in the neighborhood of the body. This allows for using significantly larger time s...

  19. ITER Fast Ion Collective Thomson Scattering

    DEFF Research Database (Denmark)

    Bindslev, Henrik; Meo, Fernando; Korsholm, Søren Bang

    for measurements of the confined fusion alpha particles in ITER set by the ITER team. Then we outline the considerations, which enter into the selection and evaluation of CTS systems. System definition includes choice of probe frequency, geometry of probe and receiver beam patterns and probe power, but ultimately...

  20. An iterative method for spherical bounces

    CERN Document Server

    Buniy, Roman V

    2016-01-01

    We develop a new iterative method for finding approximate solutions for spherical bounces associated with the decay of the false vacuum in scalar field theories. The method works for any generic potential in any number of dimensions, contains Coleman's thin-wall approximation as its first iteration, and greatly improves its accuracy by including higher order terms.

  1. Iterative methods for weighted least-squares

    Energy Technology Data Exchange (ETDEWEB)

    Bobrovnikova, E.Y.; Vavasis, S.A. [Cornell Univ., Ithaca, NY (United States)

    1996-12-31

    A weighted least-squares problem with a very ill-conditioned weight matrix arises in many applications. Because of round-off errors, the standard conjugate gradient method for solving this system does not give the correct answer even after n iterations. In this paper we propose an iterative algorithm based on a new type of reorthogonalization that converges to the solution.

  2. New concurrent iterative methods with monotonic convergence

    Energy Technology Data Exchange (ETDEWEB)

    Yao, Qingchuan [Michigan State Univ., East Lansing, MI (United States)

    1996-12-31

    This paper proposes the new concurrent iterative methods without using any derivatives for finding all zeros of polynomials simultaneously. The new methods are of monotonic convergence for both simple and multiple real-zeros of polynomials and are quadratically convergent. The corresponding accelerated concurrent iterative methods are obtained too. The new methods are good candidates for the application in solving symmetric eigenproblems.

  3. Experimental studies of ITER demonstration discharges

    NARCIS (Netherlands)

    Sips, A.C.C.; Casper, T. A.; Doyle, E. J.; Giruzzi, G.; Gribov, Y.; Hobirk, J.; Hogeweij, G. M. D.; Horton, L. D.; Hubbard, A. E.; Hutchinson, I.; Ide, S.; Isayama, A.; Imbeaux, F.; Jackson, G. L.; Kamada, Y.; Kessel, C.; Kochl, F.; Lomas, P.; Litaudon, X.; Luce, T. C.; Marmar, E.; Mattei, M.; Nunes, I.; Oyama, N.; Parail, V.; Portone, A.; Saibene, G.; Sartori, R.; Stober, J. K.; Suzuki, T.; Wolfe, S. M.

    2009-01-01

    Key parts of the ITER scenarios are determined by the capability of the proposed poloidal field (PF) coil set. They include the plasma breakdown at low loop voltage, the current rise phase, the performance during the flat top (FT) phase and a ramp down of the plasma. The ITER discharge evolution has

  4. Materials development for ITER shielding and test blanket in China

    Energy Technology Data Exchange (ETDEWEB)

    Chen, J.M., E-mail: Chenjm@swip.ac.cn [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Wu, J.H.; Liu, X.; Wang, P.H. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Wang, Z.H.; Li, Z.N. [Ningxia Orient Non-ferrous Metals Group Co. Ltd., P.O. Box 105, Shizuishan (China); Wang, X.S.; Zhang, P.C. [China Academy of Engineering Physics, P.O. Box 919-71, Mianyang 621900 (China); Zhang, N.M.; Fu, H.Y.; Liu, D.H. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China)

    2011-10-01

    China is a member of the ITER program and is developing her own materials for its shielding and test blanket modules. The materials include vacuum-hot-pressing (VHP) Be, CuCrZr alloy, 316L(N) and China low activation ferritic/martensitic (CLF-1) steels. Joining technologies including Be/Cu hot isostatic pressing (HIP) and electron beam (EB) weldability of 316L(N) were investigated. Chinese VHP-Be showed good properties, with BeO content and ductility that satisfy the ITER requirements. Be/Cu mock-ups were fabricated for Be qualification tests at simulated ITER vertical displacement event (VDE) and heat flux cycling conditions. Fine microstructure and good mechanical strength of the CuCrZr alloy were achieved by a pre-forging treatment, while the weldability of 316L(N) by EB was demonstrated for welding depths varying from 5 to 80 mm. Fine microstructure, high strength, and good ductility were achieved in CLF-1 steel by an optimized normalizing, tempering and aging procedure.

  5. Preliminary design of the ITER ECH Upper Launcher

    Energy Technology Data Exchange (ETDEWEB)

    Strauss, D., E-mail: dirk.strauss@kit.edu [Karlsruhe Institute of Technology, Assoc. KIT-EURATOM, D-76021 Karlsruhe (Germany); Aiello, G. [Karlsruhe Institute of Technology, Assoc. KIT-EURATOM, D-76021 Karlsruhe (Germany); Chavan, R. [Centre de Recherches en Physique des Plasmas, CRPP–EPFL, CH-1015 Lausanne (Switzerland); Cirant, S. [Istituto di Fisica del Plasma CNR, Euratom Association, 20125 Milano (Italy); Baar, M. de [FOM, Van Vollenhovenlaan 659, 3527 JP, Utrecht (Netherlands); Farina, D. [Istituto di Fisica del Plasma CNR, Euratom Association, 20125 Milano (Italy); Gantenbein, G. [Karlsruhe Institute of Technology, Assoc. KIT-EURATOM, D-76021 Karlsruhe (Germany); Goodman, T.P. [Centre de Recherches en Physique des Plasmas, CRPP–EPFL, CH-1015 Lausanne (Switzerland); Henderson, M.A. [ITER Organization, 13108 Saint-Paul-lez-Durance (France); Kasparek, W. [Institut für Plasmaforschung, IPF, D-70569 Stuttgart (Germany); Kleefeldt, K. [Karlsruhe Institute of Technology, Assoc. KIT-EURATOM, D-76021 Karlsruhe (Germany); Landis, J.-D. [Centre de Recherches en Physique des Plasmas, CRPP–EPFL, CH-1015 Lausanne (Switzerland); Meier, A. [Karlsruhe Institute of Technology, Assoc. KIT-EURATOM, D-76021 Karlsruhe (Germany); Moro, A.; Platania, P. [Istituto di Fisica del Plasma CNR, Euratom Association, 20125 Milano (Italy); Plaum, B. [Institut für Plasmaforschung, IPF, D-70569 Stuttgart (Germany); Poli, E. [Max-Planck-IPP, Euratom Association, D-85748 Garching (Germany); Ramponi, G. [Istituto di Fisica del Plasma CNR, Euratom Association, 20125 Milano (Italy); Ronden, D. [FOM, Van Vollenhovenlaan 659, 3527 JP, Utrecht (Netherlands); Saibene, G. [Fusion for Energy, Barcelona (Spain); and others

    2013-11-15

    Highlights: • Front steering mirror design. • Plasma facing blanket shield module/first wall panel design. • Fixed frequency torus CVD diamond window serving as first tritium barrier. • Prototypes and tests of the above key components in the Launcher Handling and Testing Facility. -- Abstract: The design of the ITER electron cyclotron launchers recently reached the preliminary design level - the last major milestone before design finalization. The ITER ECH system contains 24 installed gyrotrons providing a maximum ECH injected power of 20 MW through transmission lines towards the tokamak. There are two EC launcher types both using a front steering mirror; one equatorial launcher (EL) for plasma heating and four upper launchers (UL) for plasma mode stabilization (neoclassical tearing modes and the sawtooth instability). A wide steering angle range of the ULs allows focusing of the beam on magnetic islands which are expected on the rational magnetic flux surfaces q = 1 (sawtooth instability), q = 3/2 and q = 2 (NTMs). In this paper the preliminary design of the ITER ECH UL is presented, including the optical system and the structural components. Highlights of the design include the torus CVD-diamond windows, the frictionless, front steering mechanism and the plasma facing blanket shield module (BSM). Numerical simulations as well as prototype tests are used to verify the design.

  6. Fusion Plasma Physics and ITER - An Introduction (1/4)

    CERN Document Server

    CERN. Geneva

    2011-01-01

    In November 2006, ministers representing the world’s major fusion research communities signed the agreement formally establishing the international project ITER. Sited at Cadarache in France, the project involves China, the European Union (including Switzerland), India, Japan, the Russian Federation, South Korea and the United States. ITER is a critical step in the development of fusion energy: its role is to confirm the feasibility of exploiting magnetic confinement fusion for the production of energy for peaceful purposes by providing an integrated demonstration of the physics and technology required for a fusion power plant. The ITER tokamak is designed to study the “burning plasma” regime in deuterium-tritium (D-T) plasmas by achieving a fusion amplification factor, Q (the ratio of fusion output power to plasma heating input power), of 10 for several hundreds of seconds with a nominal fusion power output of 500MW. It is also intended to allow the study of steady-state plasma operation at Q≥5 by me...

  7. Piping structural design for the ITER thermal shield manifold

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Chang Hyun, E-mail: chnoh@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Chung, Wooho, E-mail: whchung@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Nam, Kwanwoo; Kang, Kyoung-O. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Bae, Jing Do; Cha, Jong Kook [Korea Marine Equipment Research Institute, Busan 606-806 (Korea, Republic of); Kim, Kyoung-Kyu [Mecha T& S, Jinju-si 660-843 (Korea, Republic of); Hamlyn-Harris, Craig; Hicks, Robby; Her, Namil; Jun, Chang-Hoon [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2015-10-15

    Highlights: • We finalized piping design of ITER thermal shield manifold for procurement. • Support span is determined by stress and deflection limitation. • SQP, which is design optimization method, is used for the pipe design. • Benchmark analysis is performed to verify the analysis software. • Pipe design is verified by structural analyses. - Abstract: The thermal shield (TS) provides the thermal barrier in the ITER tokamak to minimize heat load transferred by thermal radiation from the hot components to the superconducting magnets operating at 4.2 K. The TS is actively cooled by 80 K pressurized helium gas which flows from the cold valve box to the cooling tubes on the TS panels via manifold piping. This paper describes the manifold piping design and analysis for the ITER thermal shield. First, maximum allowable span for the manifold support is calculated based on the simple beam theory. In order to accommodate the thermal contraction in the manifold feeder, a contraction loop is designed and applied. Sequential Quadratic Programming (SQP) method is used to determine the optimized dimensions of the contraction loop to ensure adequate flexibility of manifold pipe. Global structural behavior of the manifold is investigated when the thermal movement of the redundant (un-cooled) pipe is large.

  8. Ultracold Neutral Plasmas

    CERN Document Server

    Killian, T C; Gupta, P; Laha, S; Martinez, Y N; Mickelson, P G; Nagel, S B; Saenz, A D; Simien, C E; Killian, Thomas C.

    2005-01-01

    Ultracold neutral plasmas are formed by photoionizing laser-cooled atoms near the ionization threshold. Through the application of atomic physics techniques and diagnostics, these experiments stretch the boundaries of traditional neutral plasma physics. The electron temperature in these plasmas ranges from 1-1000 K and the ion temperature is around 1 K. The density can approach $10^{11}$ cm$^{-3}$. Fundamental interest stems from the possibility of creating strongly-coupled plasmas, but recombination, collective modes, and thermalization in these systems have also been studied. Optical absorption images of a strontium plasma, using the Sr$^+$ ${^2S_{1/2}} -> {^2P_{1/2}}$ transition at 422 nm, depict the density profile of the plasma, and probe kinetics on a 50 ns time-scale. The Doppler-broadened ion absorption spectrum measures the ion velocity distribution, which gives an accurate measure of the ion dynamics in the first microsecond after photoionization.

  9. Between detection and neutralization.

    Energy Technology Data Exchange (ETDEWEB)

    Snell, Mark Kamerer; Green, Mary Wilson; Adams, Douglas Glenn; Pritchard, Daniel Allison

    2005-08-01

    Security system analytical performance analysis is generally based on the probability of system effectiveness. The probability of effectiveness is a function of the probabilities of interruption and neutralization. Interruption occurs if the response forces are notified in sufficient time to engage the adversary. Neutralization occurs if the adversary attack is defeated after the security forces have actively engaged the adversary. Both depend upon communications of data. This paper explores details of embedded communications functions that are often assumed to be inconsequential. It is the intent of the authors to bring focus to an issue in security system modeling that, if not well understood, has the potential to be a deciding factor in the overall system failure or effectiveness.

  10. Landweber iterative regularization for nearfield acoustic holography

    Institute of Scientific and Technical Information of China (English)

    BI Chuanxing; CHEN Xinzhao; ZHOU Rong; CHEN Jian

    2006-01-01

    On the basis of the distributed source boundary point method (DSBPM)-based nearfield acoustic holography (NAH), Landweber iterative regularization method is proposed to stabilize the NAH reconstruction process, control the influence of measurement errors on the reconstructed results and ensure the validity of the reconstructed results. And a new method, the auxiliary surface method, is proposed to determine the optimal iterative number for optimizing the regularization effect. Here, the optimal number is determined by minimizing the relative error between the calculated pressure on the auxiliary surface corresponding to each iterative number and the measured pressure. An experiment on a speaker is investigated to demonstrate the high sensitivity of the reconstructed results to measurement errors and to validate the chosen method of the optimal iterative number and the Landweber iterative regularization method for controlling the influence of measurement errors on the reconstructed results.

  11. Turbo iterative equalization for HSDPA systems

    Institute of Scientific and Technical Information of China (English)

    WU QiHui; ZHAO ChunMing; WANG JinLong

    2007-01-01

    In this paper, a turbo iterative receiver structure with chip equalization is proposed for the 3G high-speed downlink packet access (HSDPA) systems. The receiver equalizes the channel prior to the dispreading and then performs two successive soft-output decisions, achieved by a soft-input soft-output (SISO) multi-code detector and a SISO turbo decoder through an iterative process. At each iteration, extrinsic information is extracted from detection and decoding stages and is then used as a priori information in the next iteration, just as in turbo decoding. Computer simulations show that the turbo iterative receiver structure with chip equalization offers significant performance gain over the traditional receiver structure.

  12. Techniques in Iterative Proton CT Image Reconstruction

    CERN Document Server

    Penfold, Scott

    2015-01-01

    This is a review paper on some of the physics, modeling, and iterative algorithms in proton computed tomography (pCT) image reconstruction. The primary challenge in pCT image reconstruction lies in the degraded spatial resolution resulting from multiple Coulomb scattering within the imaged object. Analytical models such as the most likely path (MLP) have been proposed to predict the scattered trajectory from measurements of individual proton location and direction before and after the object. Iterative algorithms provide a flexible tool with which to incorporate these models into image reconstruction. The modeling leads to a large and sparse linear system of equations that can efficiently be solved by projection methods-based iterative algorithms. Such algorithms perform projections of the iterates onto the hyperlanes that are represented by the linear equations of the system. They perform these projections in possibly various algorithmic structures, such as block-iterative projections (BIP), string-averaging...

  13. Gargamelle: neutral current event

    CERN Multimedia

    1973-01-01

    This event shows real tracks of particles from the 1200 litre Gargamelle bubble chamber that ran on the PS from 1970 to 1976 and on the SPS from 1976 to 1979. In this image a neutrino passes close to a nucleon and reemerges as a neutrino. Such events are called neutral curent, as they are mediated by the Z0 boson which has no electric charge.

  14. Neutral atom traps.

    Energy Technology Data Exchange (ETDEWEB)

    Pack, Michael Vern

    2008-12-01

    This report describes progress in designing a neutral atom trap capable of trapping sub millikelvin atom in a magnetic trap and shuttling the atoms across the atom chip from a collection area to an optical cavity. The numerical simulation and atom chip design are discussed. Also, discussed are preliminary calculations of quantum noise sources in Kerr nonlinear optics measurements based on electromagnetically induced transparency. These types of measurements may be important for quantum nondemolition measurements at the few photon limit.

  15. Magnetization study of ITER-type internal-Sn Nb3Sn superconducting wire

    Institute of Scientific and Technical Information of China (English)

    Zhang Chao-Wu; Zhou Lian; Andre Sulpice; Jean-Louis Soubeyroux; Christophe Verwaerde; Gia Ky Hoang; Zhang Ping-Xiang; Lu Ya-Feng; Tang Xian-De

    2007-01-01

    Through magnetization measurement with a SQUID magnetometer the heat treatment optimization of an international thermonuclear experimental reactor (ITER)-type internal-Sn Nb3Sn superconducting wire has been investigated. The irreversibility temperature T*(H), which is mainly dependent on A15 phase composition, was obtained by a warming and cooling cycle at a fixed field. The hysteresis width △M(H) which reflects the flux pinning situation of the A15 phase is determined by the sweeping of magnetic field at a constant temperature. The results obtained from differently heat-treated samples show that the combination of T*(H) with △M(H) measurement is very effective for optimizing the heat reaction process. The heat treatment condition of the ITER-type wire is optimized at 675 ℃/128 h, which results in a composition closer to stoichiometric Nb3Sn and a state with best flux pinning.

  16. Erosion of tungsten armor after multiple intense transient events in ITER

    Science.gov (United States)

    Bazylev, B. N.; Janeschitz, G.; Landman, I. S.; Pestchanyi, S. E.

    2005-03-01

    Macroscopic erosion by melt motion is the dominating damage mechanism for tungsten armour under high-heat loads with energy deposition W > 1 MJ/m 2 and τ > 0.1 ms. For ITER divertor armour the results of a fluid dynamics simulation of the melt motion erosion after repetitive stochastically varying plasma heat loads of consecutive disruptions interspaced by ELMs are presented. The heat loads for particular single transient events are numerically simulated using the two-dimensional MHD code FOREV-2D. The whole melt motion is calculated by the fluid dynamics code MEMOS-1.5D. In addition for the ITER dome melt motion erosion of tungsten armour caused by the lateral radiation impact from the plasma shield at the disruption and ELM heat loads is estimated.

  17. Design analysis of the hinge support for the ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B.Y. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Ahn, H.J., E-mail: hjahn@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Ha, M.S. [DNDE, Busan 612-021 (Korea, Republic of); Kim, Y.K. [Hyundai Heavy Industries Co. Ltd., Ulsan 682-792 (Korea, Republic of); Kim, H.S.; Sa, J.W.; Kim, B.C.; Bak, J.S. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Choi, C.H.; Ioki, K.; Wang, X.; Bachmann, C. [ITER Organization, CS 90 046, 13067 St Paul-lez-Durance CEDEX (France)

    2011-10-15

    In order to verify design feasibility and structural integrity of a hinge type support for the ITER VV support system, the design analysis has been performed in detail, which includes heat transfer, elastic stress and limit analyses. The structural analyses were performed to confirm the transfer of forces through the supporting structure and to determine the maximum allowable loads according to the RCC-MR. From the heat transfer analysis for VV baking stage, total heat flow into the support was obtained to confirm the thermal heat flux into the cryostat under baking condition. In addition, the design modification was also discussed to enhance the structural performance of the supporting system.

  18. Validation of the ITER CXRS design by tests on TEXTORa)

    Science.gov (United States)

    Jaspers, R. J. E.; von Hellermann, M. G.; Delabie, E.; Biel, W.; Marchuk, O.; Yao, L.

    2008-10-01

    The charge exchange recombination spectroscopy system (CXRS) for ITER is designed to measure the core helium concentration, and in addition, profiles of ion temperature and rotation. This highly demanding task, due to the huge background radiation (bremsstrahlung) and the high attenuation of the dedicated diagnostic neutral beam, requires high throughput spectrometers with high resolution. On TEXTOR, a CXRS system has been developed with the aim to test the physics implications of these specifications. (i) A relevant spectrometer has been tested. (ii) A method to determine the helium concentrations from the CXRS intensity, using the beam emission has been evaluated. A 20% discrepancy in beam emission was revealed. (iii) The determination of the magnetic pitch angle by the ratio of Balmer lines showed qualitatively the right behavior, although the accuracy was limited by the polarization sensitivity of the first mirror. (iv) The simulation code developed for the prediction of the CXRS spectra was quantitatively confronted with experimental data.

  19. Neutralization principles for the Extraction and Transport of Ion Beams

    CERN Document Server

    Riege, H

    2000-01-01

    The strict application of conventional extraction techniques of ion beams from a plasma source is characterized by a natural intensity limit determined by space charge.The extracted current may be enhanced far beyond this limit by neutralizing the space charge of the extracted ions in the first extraction gap of the source with electrons injected from the opposite side. The transverse and longitudinal emittances of a neutralized ion beam, hence its brightness, are preserved. Results of beam compensation experiments, which have been carried out with a laser ion source, are resumed for proposing a general scheme of neutralizing ion sources and their adjacent low-energy beam transport channels with electron beams. Many technical applications of high-mass ion beam neutralization technology may be identified: the enhancement of ion source output for injection into high-intensity, low-and high-energy accelerators, or ion thrusters in space technology, for the neutral beams needed for plasma heating of magnetic conf...

  20. Optimization of tungsten castellated structures for the ITER divertor

    Science.gov (United States)

    Litnovsky, A.; Hellwig, M.; Matveev, D.; Komm, M.; van den Berg, M.; De Temmerman, G.; Rudakov, D.; Ding, F.; Luo, G.-N.; Krieger, K.; Sugiyama, K.; Pitts, R. A.; Petersson, P.

    2015-08-01

    In ITER, the plasma-facing components (PFCs) of the first wall and the divertor armor will be castellated to improve their thermo-mechanical stability and to limit forces due to induced currents. The fuel accumulation in the gaps may significantly contribute to the in-vessel fuel inventory. Castellation shaping may be the most straightforward way to minimize the fuel inventory and to alleviate the thermal loads onto castellations. A new castellation shape was proposed and comparative modeling of conventional (rectangular) and shaped castellation was performed for ITER conditions. Shaped castellation was predicted to be capable to operate under stationary heat load of 20 MW/m2. An 11-fold decrease of beryllium (Be) content in the gaps of the shaped cells alone with a 7-fold decrease of carbon content was predicted. In order to validate the predictive capabilities of modeling tools used for ITER conditions, the dedicated modeling with the same codes was made for existing tokamaks and benchmarked with the results of multi-machine experiments. For the castellations exposed in TEXTOR and DIII-D, the carbon amount in the gaps of shaped cells was 1.9-2.3 times smaller than that of rectangular ones. Modeling for TEXTOR conditions yielded to 1.5-fold decrease of carbon content in the gaps of shaped castellation outlining fair agreement with the experiment. At the same time, a number of processes, like enhanced erosion of molten layer yet need to be implemented in the codes in order to increase the accuracy of predictions for ITER.

  1. New developments in the diagnostics for the fusion products on JET in preparation for ITER (invited)

    Energy Technology Data Exchange (ETDEWEB)

    Murari, A.; Angelone, M.; Pillon, M. [JET-EFDA, Culham Science Centre, OX14 3DB, Abingdon (United Kingdom); Consorzio RFX, Assoc. EURATOM ENEA sulla Fusione, Corso Stati Uniti, 4, I-35127, Padova (Italy); Bonheure, G. [JET-EFDA, Culham Science Centre, OX14 3DB, Abingdon (United Kingdom); Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, C.P. 65, I-00044 Frascati, Roma (Italy); Cecil, E. [JET-EFDA, Culham Science Centre, OX14 3DB, Abingdon (United Kingdom); Laboratory for Plasma Physics, Association ' ' Euratom-Belgian State' ' , Royal Military Academy, Avenue de la Renaissance, 30, Kunstherlevinglaan, B-1000 Brussels (Belgium); Craciunescu, T.; Zoita, V. L. [JET-EFDA, Culham Science Centre, OX14 3DB, Abingdon (United Kingdom); Colorado School of Mines, 1500 Illinois St., Golden, Colorado (United States); Darrow, D. [JET-EFDA, Culham Science Centre, OX14 3DB, Abingdon (United Kingdom); EURATOM-MEdC Association, National Institute for Laser, Plasma, and Radiation Physics, Bucharest (Romania); Edlington, T.; Kiptily, V.; Popovichev, S.; Syme, B. [JET-EFDA, Culham Science Centre, OX14 3DB, Abingdon (United Kingdom); Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, New Jersey (United States); Ericsson, G.; Gatu-Johnson, M.; Hellesen, C. [JET-EFDA, Culham Science Centre, OX14 3DB, Abingdon (United Kingdom); EURATOM/CCFE Fusion Association, Culham Science Centre, Oxon. OX14 3DB, Abingdon, OXON (United Kingdom); Gorini, G.; Tardocchi, M. [JET-EFDA, Culham Science Centre, OX14 3DB, Abingdon (United Kingdom); Department of Physics and Astronomy, Uppsala University, EURATOM-VR Association, Uppsala (Sweden); Mlynar, J. [JET-EFDA, Culham Science Centre, OX14 3DB, Abingdon (United Kingdom); Associazione EURATOM-ENEA sulla Fusione, IFP Milano (Italy); Perez von Thun, C. [JET-EFDA, Culham Science Centre, OX14 3DB, Abingdon (United Kingdom); Association Euratom-IPP.CR, Institute of Plasma Physics AS CR, v.v.i., Za Slovankou 3, CZ-182 00 Praha 8 (Czech Republic); Collaboration: EFDA-JET Contributors

    2010-10-15

    Notwithstanding the advances of the past decades, significant developments are still needed to satisfactorily diagnose ''burning plasmas.'' D-T plasmas indeed require a series of additional measurements for the optimization and control of the configuration: the 14 MeV neutrons, the isotopic composition of the main plasma, the helium ash, and the redistribution and losses of the alpha particles. Moreover a burning plasma environment is in general much more hostile for diagnostics than purely deuterium plasmas. Therefore, in addition to the development and refinement of new measuring techniques, technological advances are also indispensable for the proper characterization of the next generation of devices. On JET an integrated program of diagnostic developments, for JET future and in preparation for ITER, has been pursued and many new results are now available. In the field of neutron detection, the neutron spectra are now routinely measured in the energy range of 1-18 MeV by a time of flight spectrometer and they have allowed studying the effects of rf heating on the fast ions.A new analysis method for the interpretation of the neutron cameras measurements has been refined and applied to the data of the last trace tritium campaign (TTE). With regard to technological upgrades, chemical vapor deposition diamond detectors have been qualified both as neutron counters and as neutron spectrometers, with a potential energy resolution of about one percent. The in situ calibration of the neutron diagnostics, in preparation for the operation with the ITER-like wall, is also promoting important technological developments. With regard to the fast particles, for the first time the temperature of the fast particle tails has been obtained with a new high purity Germanium detector measuring the gamma emission spectrum from the plasma. The effects of toroidal Alfven eigenmodes modes and various MHD instabilities on the confinement of the fast particles have been

  2. New developments in the diagnostics for the fusion products on JET in preparation for ITER (invited).

    Science.gov (United States)

    Murari, A; Angelone, M; Bonheure, G; Cecil, E; Craciunescu, T; Darrow, D; Edlington, T; Ericsson, G; Gatu-Johnson, M; Gorini, G; Hellesen, C; Kiptily, V; Mlynar, J; Perez von Thun, C; Pillon, M; Popovichev, S; Syme, B; Tardocchi, M; Zoita, V L

    2010-10-01

    Notwithstanding the advances of the past decades, significant developments are still needed to satisfactorily diagnose “burning plasmas.” D–T plasmas indeed require a series of additional measurements for the optimization and control of the configuration: the 14 MeV neutrons, the isotopic composition of the main plasma, the helium ash, and the redistribution and losses of the alpha particles. Moreover a burning plasma environment is in general much more hostile for diagnostics than purely deuterium plasmas. Therefore, in addition to the development and refinement of new measuring techniques, technological advances are also indispensable for the proper characterization of the next generation of devices. On JET an integrated program of diagnostic developments, for JET future and in preparation for ITER, has been pursued and many new results are now available. In the field of neutron detection, the neutron spectra are now routinely measured in the energy range of 1–18 MeV by a time of flight spectrometer and they have allowed studying the effects of rf heating on the fast ions. A new analysis method for the interpretation of the neutron cameras measurements has been refined and applied to the data of the last trace tritium campaign (TTE). With regard to technological upgrades, chemical vapor deposition diamond detectors have been qualified both as neutron counters and as neutron spectrometers, with a potential energy resolution of about one percent. The in situ calibration of the neutron diagnostics, in preparation for the operation with the ITER-like wall, is also promoting important technological developments. With regard to the fast particles, for the first time the temperature of the fast particle tails has been obtained with a new high purity Germanium detector measuring the gamma emission spectrum from the plasma. The effects of toroidal Alfven eigenmodes modes and various MHD instabilities on the confinement of the fast particles have been determined

  3. Highly Nonlinear Temperature-Dependent Fin Analysis by Variational Iteration Method

    DEFF Research Database (Denmark)

    Fouladi, F.; Hosseinzadeh, E.; Barari, Amin

    2010-01-01

    In this research, the variational iteration method as an approximate analytical method is utilized to overcome some inherent limitations arising as uncontrollability to the nonzero endpoint boundary conditions and is used to solve some examples in the field of heat transfer. The available exact s...

  4. Analysis of Diffusion Problems using Homotopy Perturbation and Variational Iteration Methods

    DEFF Research Database (Denmark)

    Barari, Amin; Poor, A. Tahmasebi; Jorjani, A.

    2010-01-01

    In this paper, variational iteration method and homotopy perturbation method are applied to different forms of diffusion equation. The diffusion equations have found wide applications in heat transfer problems, theory of consolidation and many other problems in engineering. The methods proposed t...

  5. Improved Confinement in JET High {beta} Plasmas with an ITER-Like Wall

    CERN Document Server

    Challis, C D; Beurskens, M; Buratti, P; Delabie, E; Drewelow, P; Frassinetti, L; Giroud, C; Hawkes, N; Hobirk, J; Joffrin, E; Keeling, D; King, D B; Maggi, C F; Mailloux, J; Marchetto, C; McDonald, D; Nunes, I; Pucella, G; Saarelma, S; Simpson, J

    2015-01-01

    The replacement of the JET carbon wall (C-wall) by a Be/W ITER-like wall (ILW) has affected the plasma energy confinement. To investigate this, experiments have been performed with both the C-wall and ILW to vary the heating power over a wide range for plasmas with different shapes.

  6. Impact and mitigation of disruptions with the ITER-like wall in JET

    NARCIS (Netherlands)

    Lehnen, M.; Arnoux, G.; Brezinsek, S.; Flanagan, J.; Gerasimov, S. N.; Hartmann, N.; Hender, T. C.; Huber, A.; Jachmich, S.; Kiptily, V.; Kruezi, U.; Matthews, G. F.; Morris, J.; Plyusnin, V. V.; Reux, C.; Riccardo, V.; Sieglin, B.; de Vries, P. C.; JET-EFDA Contributors,

    2013-01-01

    Disruptions are a critical issue for ITER because of the high thermal and magnetic energies that are released on short timescales, which results in extreme forces and heat loads. The choice of material of the plasma-facing components (PFCs) can have significant impact on the loads that arise during

  7. Expanding the operating space of ICRF on JET with a view to ITER

    DEFF Research Database (Denmark)

    Lamalle, P.U.; Mantsinen, M.J.; Noterdaeme, J.M.

    2006-01-01

    This paper reports on ITER-relevant ion cyclotron resonance frequency (ICRF) physics investigated on JET in 2003 and early 2004. Minority heating of helium three in hydrogen plasmas-(He-3)H-was systematically explored by varying the 3 He concentration and the toroidal phasing of the antenna array...

  8. The ITER full size plasma source device design

    Energy Technology Data Exchange (ETDEWEB)

    Sonato, P. [Consorzio RFX, EURATOM-ENEA Association, Corso Stati Uniti 4, I-35127 Padova (Italy)], E-mail: piergiorgio.sonato@igi.cnr.it; Agostinetti, P.; Anaclerio, G.; Antoni, V.; Barana, O.; Bigi, M.; Boldrin, M. [Consorzio RFX, EURATOM-ENEA Association, Corso Stati Uniti 4, I-35127 Padova (Italy); Cavenago, M. [INFN, Legnaro, Padova (Italy); Dal Bello, S.; Palma, M. Dalla [Consorzio RFX, EURATOM-ENEA Association, Corso Stati Uniti 4, I-35127 Padova (Italy); Daniele, A. [ENEA, Frascati, Roma (Italy); D' Arienzo, M.; De Lorenzi, A.; Ferro, A.; Fiorentin, A.; Gaio, E.; Gazza, E.; Grando, L.; Fantini, F.; Fellin, F. [Consorzio RFX, EURATOM-ENEA Association, Corso Stati Uniti 4, I-35127 Padova (Italy)] (and others)

    2009-06-15

    In the framework of the strategy for the development and the procurement of the NB systems for ITER, it has been decided to build in Padova a test facility, including two experimental devices: a full size plasma source with low voltage extraction and a full size NB injector at full beam power (1 MV). These two different devices will separately address the main scientific and technological issues of the 17 MW NB injector for ITER. In particular the full size plasma source of negative ions will address the ITER performance requirements in terms of current density and uniformity, limitation of the electron/ion ratio and stationary operation at full current with high reliability and constant performances for the whole operating time up to 1 h. The required negative ion current density to be extracted from the plasma source ranges from 290 A/m{sup 2} in D{sub 2} (D{sup -}) and 350 A/m{sup 2} in H{sub 2} (H{sup -}) and these values should be obtained at the lowest admissible neutral pressure in the plasma source volume, nominally at 0.3 Pa. The electron to ion ratio should be limited to less than 1 and the admissible ion inhomogeneity extracted from the grids should be better than 10% on the whole plasma cross-section having a surface exposed to the extraction grid of the order of 1 m{sup 2}. The main design choices will be presented in the paper as well as an overview of the design of the main components and systems.

  9. Neutron measurements in ITER using the Radial Neutron Camera

    Science.gov (United States)

    Marocco, D.; Esposito, B.; Moro, F.

    2012-03-01

    The Radial Neutron Camera (RNC) is one of the key diagnostic systems of the ITER international fusion experiment. It is designed to measure the uncollided 14 MeV and 2.5 MeV neutrons from deuterium-tritium (DT) and deuterium-deuterium (DD) fusion reactions taking place in the ITER plasma through an array of 45 detectors positioned along collimated lines of sight. Scintillators and diamonds coupled to fast digital acquisition electronics are among the detectors presently considered for the RNC. The RNC will provide spatially resolved measurements of several plasma parameters needed for fusion power estimation, plasma control and plasma physics studies. The line-integrated RNC neutron fluxes are used to evaluate the local profile of the neutron emission (neutron emissivity, s-1m-3) and therefore the total neutron yield and the birth profile of the alpha particles. The temperature profile of the bulk ions can be derived from the Doppler broadened widths of the RNC line-integrated spectra, that also provide insight on the supra-thermal ions produced by the injection in the plasma of electromagnetic waves and neutral particles. The RNC emissivity and temperature measurements can be employed to estimate the composition of the ITER fuel, namely the ratio between the tritium and deuterium densities. Data processing techniques involving spatial inversion and spectra unfolding are necessary to deduce the profile quantities from the line-integrated RNC measurements. The expected performances of the RNC as a diagnostic for the neutron emissivity/ion temperature/fuel ratio profile (measurement range, time resolution, accuracy, precision) have been estimated by means of synthetic data simulating actual RNC measurements. The results of the simulations, together with an overall description of the diagnostic and of the measurement techniques, are presented.

  10. Iterants, Fermions and Majorana Operators

    Science.gov (United States)

    Kauffman, Louis H.

    Beginning with an elementary, oscillatory discrete dynamical system associated with the square root of minus one, we study both the foundations of mathematics and physics. Position and momentum do not commute in our discrete physics. Their commutator is related to the diffusion constant for a Brownian process and to the Heisenberg commutator in quantum mechanics. We take John Wheeler's idea of It from Bit as an essential clue and we rework the structure of that bit to a logical particle that is its own anti-particle, a logical Marjorana particle. This is our key example of the amphibian nature of mathematics and the external world. We show how the dynamical system for the square root of minus one is essentially the dynamics of a distinction whose self-reference leads to both the fusion algebra and the operator algebra for the Majorana Fermion. In the course of this, we develop an iterant algebra that supports all of matrix algebra and we end the essay with a discussion of the Dirac equation based on these principles.

  11. On the interplay between inner and outer iterations for a class of iterative methods

    Energy Technology Data Exchange (ETDEWEB)

    Giladi, E. [Stanford Univ., CA (United States)

    1994-12-31

    Iterative algorithms for solving linear systems of equations often involve the solution of a subproblem at each step. This subproblem is usually another linear system of equations. For example, a preconditioned iteration involves the solution of a preconditioner at each step. In this paper, the author considers algorithms for which the subproblem is also solved iteratively. The subproblem is then said to be solved by {open_quotes}inner iterations{close_quotes} while the term {open_quotes}outer iteration{close_quotes} refers to a step of the basic algorithm. The cost of performing an outer iteration is dominated by the solution of the subproblem, and can be measured by the number of inner iterations. A good measure of the total amount of work needed to solve the original problem to some accuracy c is then, the total number of inner iterations. To lower the amount of work, one can consider solving the subproblems {open_quotes}inexactly{close_quotes} i.e. not to full accuracy. Although this diminishes the cost of solving each subproblem, it usually slows down the convergence of the outer iteration. It is therefore interesting to study the effect of solving each subproblem inexactly on the total amount of work. Specifically, the author considers strategies in which the accuracy to which the inner problem is solved, changes from one outer iteration to the other. The author seeks the `optimal strategy`, that is, the one that yields the lowest possible cost. Here, the author develops a methodology to find the optimal strategy, from the set of slowly varying strategies, for some iterative algorithms. This methodology is applied to the Chebychev iteration and it is shown that for Chebychev iteration, a strategy in which the inner-tolerance remains constant is optimal. The author also estimates this optimal constant. Then generalizations to other iterative procedures are discussed.

  12. Design of the RF ion source for the ITER NBI

    Energy Technology Data Exchange (ETDEWEB)

    Marcuzzi, D. [Consorzio RFX, Euratom-ENEA Association, Corso Stati Uniti 4, I-35127 Padova (Italy)], E-mail: diego.marcuzzi@igi.cnr.it; Agostinetti, P.; Dalla Palma, M. [Consorzio RFX, Euratom-ENEA Association, Corso Stati Uniti 4, I-35127 Padova (Italy); Falter, H.D.; Heinemann, B.; Riedl, R. [Max-Planck-Institut fuer Plasmaphysik, D-85748 Garching (Germany)

    2007-10-15

    A radio frequency (RF) driven negative ion source has been designed for the ITER neutral beam injectors, as an alternative to the traditional arc driven solution. The main advantage of this technology is to avoid the presence of the filaments, that require periodic maintenance and consequently frequent shutdowns. The requirements for the ion source of the ITER NBI are to provide a uniform flux of D{sup -}/H{sup -} to the plasma grid of the accelerator that will result in a beam current of 40 A at 1 MeV. The present specification is for a filling pressure of 0.3 Pa. The ion source needs to provide 20/28 mA/cm{sup 2} D{sup -}/H{sup -} current density across the 0.58 m x 1.54 m aperture array for 3600 s. The source, consisting of a main chamber facing the plasma grid, of eight RF drivers and the auxiliary systems for power transfer, cooling and diagnostic purposes, is housed in the same quasi-cylindrical structure that supports the arc driven solution. Specific electric and hydraulic circuits have been designed and verified. In the paper the analyses performed for the design of the components are presented in detail.

  13. Stabilization of burn conditions in an ITER FEAT like tokamak with uncertainities in the helium ash confinement time

    CERN Document Server

    Vitela, J E

    2004-01-01

    In this work we demostrate using a two-temperature volume average 0D model that robust stabilization, with regard the hellium ash confinement time, of the burn conditions of a tokamak reactor with the ITER FEAT design parameters can be achieved using Radial Basis Neural Networks (RBNN). Alpha particle thermalization time delay is taken into account in this model. The control actions implemented by means of a RBNN, include the modulation of the DT refueling rate, a neutral He-4 injection beam and auxiliary heating powers to ions and to electrons; all of them constrained to lie within allowable range values. Here we assume that the tokamak follows the IPB98(y,2) scaling for the energy confinement time, while the helium ash confinement time is assumed to be independently estimated on-line. The DT and helium ash confinement times are assumed to keep a constant relationship at all times. An on-line noisy estimation of the helium ash confinement time is simulated by corrupting it with pseudo Gaussian noise.

  14. ITER lip seal welding and cutting developments

    Energy Technology Data Exchange (ETDEWEB)

    Levesy, B.; Cordier, J.J.; Jokinen, T. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Kujanpää, V.; Karhu, M. [VTT Technical Research Centre of Finland (Finland); Le Barbier, R. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Määttä, T. [VTT Technical Research Centre of Finland (Finland); Martins, J.P.; Utin, Y. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2015-10-15

    Highlights: • Different TIG and Laser welding techniques are tested. • Twin spot laser welding techniques is the best. • Limited heat input gives a stable weld pool in all positions. • Penetrations is achieved. • Lip seal welding and cutting with a robotic arm is successfully performed on a representative mock-up. - Abstract: The welded lip seals form part of the torus primary vacuum boundary in between the port plugs and the vacuum vessel, and are classified as Protection Important Component. In order to refurbish the port plugs or the in-vessel components, port plugs have to be removed from the machine. The lip seal design must enable up to ten opening of the vacuum vessel during the life time operation of the ITER machine. Therefore proven, remote reliable cutting and re-welding are essential, as these operations need to be performed in the port cells in a nuclear environment, where human presence will be restricted. Moreover, the combination of size of the components to be welded (∼10 m long vacuum compatible thin welds) and the congested environment close to the core of the machine constraint the type and size of tools to be used. This paper describes the lip seal cutting and welding development programme performed at the VTT Technical Research Centre, Finland. Potential cutting and welding techniques are analyzed and compared. The development of the cutting, TIG and laser welding techniques on samples are presented. Effects of lip seal misalignments and optimization of the 2 welding processes are discussed. Finally, the manufacturing and test of the two 1.2 m × 1 m representative mock-ups are presented. The set-up and use of a robotic arm for the mock-up cutting and welding operations are also described.

  15. ITER breeding blanket module design and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kuroda, Toshimasa; Enoeda, Mikio; Kikuchi, Shigeto [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    1998-11-01

    The ITER breeding blanket employs a ceramic breeder and Be neutron multiplier both in small spherical pebble form. Radial-poloidal cooling panels are arranged in the blanket box to remove the nuclear heating in these materials and to reinforce the blanket structure. At the first wall, Be armor is bonded onto the stainless steel (SS) structure to provide a low Z plasma-compatible surface and to protect the first wall/blanket structure from the direct contact with the plasma during off-normal events. Thermo-mechanical analyses and investigation of fabrication procedure have been performed for this breeding blanket. To evaluate thermo-mechanical behavior of the pebble beds including the dependency of the effective thermal conductivity on stress, analysis methods have been preliminary established by the use of special calculation option of ABAQUS code, which are briefly summarized in this report. The structural response of the breeding blanket module under internal pressure of 4 MPa (in case of in-blanket LOCA) resulted in rather high stress in the blanket side (toroidal end) wall, thus addition of a stiffening rib or increase of the wall thickness will be needed. Two-dimensional elasto-plastic analyses have been performed for the Be/SS bonded interface at the first wall taking a fabrication process based on HIP bonding and thermal cycle due to pulsed plasma operation into account. The stress-strain hysteresis during these process and operation was clarified, and a procedure to assess and/or confirm the bonding integrity was also proposed. Fabrication sequence of the breeding blanket module was preliminarily developed based on the procedure to fabricate part by part and to assemble them one by one. (author)

  16. Progress of the ECRH Upper Launcher design for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Strauss, D., E-mail: dirk.strauss@kit.edu [Karlsruhe Institute of Technology, Assoc. KIT-EURATOM, D-76021 Karlsruhe (Germany); Aiello, G. [Karlsruhe Institute of Technology, Assoc. KIT-EURATOM, D-76021 Karlsruhe (Germany); Bruschi, A. [Istituto di Fisica del Plasma CNR, Euratom Association, 20125 Milano (Italy); Chavan, R. [Centre de Recherches en Physique des Plasmas, CRPP–EPFL, CH-1015 Lausanne (Switzerland); Farina, D.; Figini, L. [Istituto di Fisica del Plasma CNR, Euratom Association, 20125 Milano (Italy); Gagliardi, M.; Garcia, V. [Fusion for Energy, Barcelona (Spain); Goodman, T.P. [Centre de Recherches en Physique des Plasmas, CRPP–EPFL, CH-1015 Lausanne (Switzerland); Grossetti, G. [Karlsruhe Institute of Technology, Assoc. KIT-EURATOM, D-76021 Karlsruhe (Germany); Heemskerk, C. [Heemskerk Innovative Technology, Merelhof 2, 2172 HZ Sassenheim (Netherlands); Henderson, M.A. [ITER Organization, 13108 Saint-Paul-lez-Durance (France); Kasparek, W. [Institut für Plasmaforschung, IPF, D-70569 Stuttgart (Germany); Krause, A.; Landis, J.-D. [Centre de Recherches en Physique des Plasmas, CRPP–EPFL, CH-1015 Lausanne (Switzerland); Meier, A. [Karlsruhe Institute of Technology, Assoc. KIT-EURATOM, D-76021 Karlsruhe (Germany); Moro, A.; Platania, P. [Istituto di Fisica del Plasma CNR, Euratom Association, 20125 Milano (Italy); Plaum, B. [Institut für Plasmaforschung, IPF, D-70569 Stuttgart (Germany); Poli, E. [Max-Planck-IPP, Euratom Association, D-85748 Garching (Germany); and others

    2014-10-15

    The design of the ITER ECRH system provides 20 MW millimeter wave power for central plasma heating and MHD stabilization. The system consists of an array of 24 gyrotrons with power supplies coupled to a set of transmission lines guiding the beams to the four upper and the equatorial launcher. The front steering upper launcher design described herein has passed successfully the preliminary design review, and it is presently in the final design stage. The launcher consists of a millimeter wave system and steering mechanism with neutron shielding integrated into an upper port plug with the plasma facing blanket shield module (in-vessel) and a set of ex-vessel waveguides connecting the launcher to the transmission lines. Part of the transmission lines are the ultra-low loss CVD torus diamond windows and a shutter valve, a miter bend section and the feedthroughs integrated in the plug closure plate. These components are connected by corrugated waveguides and form together the first confinement system (FCS). In-vessel, the millimeter-wave system includes a quasi-optical beam propagation system including four mirror sets and a front steering mirror. The millimeter wave system is integrated into a specifically optimized upper port plug providing structural stability to withstand plasma disruption forces and the high heat load from the plasma side with a dedicated blanket shield module. A recent update in the ITER interface definition has resulted in the recession of the upper port plug first wall panels, which is now integrated into the design. Apart from the millimeter wave system the upper port plug houses also a set of shield blocks which provide neutron shielding. An overview of the actual ITER ECRH Upper Launcher is given together with some highlights of the design.

  17. Newton—Like Iteration Method for Solving Algebraic Equations

    Institute of Scientific and Technical Information of China (English)

    JihuanHE

    1998-01-01

    In this paper,a Newton-like iteration method is proposed to solve an approximate solution of an algebraic equation.The iteration formula obtained by homotopy perturbation method contains the well-known Newton iteration formulain logic.

  18. Disruption scenarios, their mitigation and operation window in ITER

    Science.gov (United States)

    Sugihara, M.; Shimada, M.; Fujieda, H.; Gribov, Yu.; Ioki, K.; Kawano, Y.; Khayrutdinov, R.; Lukash, V.; Ohmori, J.

    2007-04-01

    The impacts of plasma disruptions on ITER have been investigated in detail to confirm the robustness of the design of the machine to the potential consequential loads. The loads include both electro-magnetic (EM) and heat loads on the in-vessel components and the vacuum vessel. Several representative disruption scenarios are specified based on newly derived physics guidelines for the shortest current quench time as well as the maximum product of halo current fraction and toroidal peaking factor arising from disruptions in ITER. Disruption simulations with the DINA code and EM load analyses with a 3D finite element method code are performed for these scenarios. Some margins are confirmed in the EM load on in-vessel components due to induced eddy and halo currents for these representative scenarios. However, the margins are not very large. The heat load on various parts of the first wall due to the vertical movement and the thermal quench (TQ) is calculated with a 2D heat conduction code based on the database of heat deposition during disruptions and simulation results with the DINA code. For vertical displacement event, it is found that the beryllium (Be) wall does not melt during the vertical movement, prior to the TQ. Significant melting is anticipated for the upper Be wall and the tungsten divertor baffle due to TQ after the vertical movement. However, its impact could be substantially mitigated by implementing a reliable detection system of the vertical movement and a mitigation system, e.g. massive noble gas injection. Some melting of the upper Be wall is anticipated at major disruptions. At least several tens of unmitigated disruptions must be considered even if an advanced prediction/mitigation system is implemented. With these unmitigated disruptions, the loss of the Be layer is expected to be within ap30-100 µm/event out of a 10 mm thick Be first wall.

  19. Validation of Helium Inlet Design for ITER Toroidal Field Coil

    CERN Document Server

    Boyer, C; Hamada, K; Foussat, A; Le Rest, M; Mitchell, N; Decool, P; Savary, F; Sgobba, S; Weiss, K-P

    2014-01-01

    The ITER organization has performed design and its validation tests on a helium inlet structure for the ITER Toroidal Field (TF) coil under collaboration with CERN, KIT, and CEA-Cadarache. Detailed structural analysis was performed in order to optimize the weld shape. A fatigue resistant design on the fillet weld between the shell covers and the jacket is an important point on the helium inlet structure. A weld filler material was selected based on tensile test at liquid helium temperature after Nb3Sn reaction heat treatment. To validate the design of the weld joint, fatigue tests at 7 K were performed using heat-treated butt weld samples. A pressure drop measurement of a helium inlet mock-up was performed by using nitrogen gas at room temperature in order to confirm uniform flow distribution and pressure drop characteristic. These tests have validated the helium inlet design. Based on the validation, Japanese and European Union domestic agencies, which have responsibilities of the TF coil procurement, are pr...

  20. Effects of ELMs on ITER divertor armour materials

    Energy Technology Data Exchange (ETDEWEB)

    Zhitlukhin, A. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation)]. E-mail: zhitlukh@triniti.ru; Klimov, N. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Landman, I. [Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe (Germany); Linke, J. [Forschungszentrum Juelich, EURATOM-Association, Juelich (Germany)]. E-mail: j.linke@fz-juelich.de; Loarte, A. [EFDA, Boltzmannstr. 2, 85748 Garching (Germany); Merola, M. [EFDA, Boltzmannstr. 2, 85748 Garching (Germany); Podkovyrov, V. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Federici, G. [ITER JWS Garching, Boltzmannstr. 2, 85748 Garching (Germany); Bazylev, B. [Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe (Germany); Pestchanyi, S. [Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe (Germany); Safronov, V. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Hirai, T. [Forschungszentrum Juelich, EURATOM-Association, Juelich (Germany); Maynashev, V. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Levashov, V. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Muzichenko, A. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation)

    2007-06-15

    This paper is concerned with investigation of an erosion of the ITER-like divertor plasma facing components under plasma heat loads expected during the Type I ELMs in ITER. These experiments were carried out on plasma accelerator QSPA at the SRC RF TRINITI under EU/RF collaboration. Targets were exposed by series repeated plasma pulses with heat loads in a range of 0.5-1.5 MJ/m{sup 2} and pulse duration 0.5 ms. Erosion of CFC macrobrushes was determined mainly by sublimation of PAN-fibres that was less than 2.5 {mu}m per pulse. The CFC erosion was negligible at the energy density less than 0.5 MJ/m{sup 2} and was increased to the average value 0.3 {mu}m per pulse at 1.5 MJ/m{sup 2}. The pure tungsten macrobrushes erosion was small in the energy range of 0.5-1.3 MJ/m{sup 2}. The sharp growth of tungsten erosion and the intense droplet ejection were observed at the energy density of 1.5 MJ/m{sup 2}.

  1. Effects of ELMs on ITER divertor armour materials

    Science.gov (United States)

    Zhitlukhin, A.; Klimov, N.; Landman, I.; Linke, J.; Loarte, A.; Merola, M.; Podkovyrov, V.; Federici, G.; Bazylev, B.; Pestchanyi, S.; Safronov, V.; Hirai, T.; Maynashev, V.; Levashov, V.; Muzichenko, A.

    2007-06-01

    This paper is concerned with investigation of an erosion of the ITER-like divertor plasma facing components under plasma heat loads expected during the Type I ELMs in ITER. These experiments were carried out on plasma accelerator QSPA at the SRC RF TRINITI under EU/RF collaboration. Targets were exposed by series repeated plasma pulses with heat loads in a range of 0.5-1.5 MJ/m2 and pulse duration 0.5 ms. Erosion of CFC macrobrushes was determined mainly by sublimation of PAN-fibres that was less than 2.5 μm per pulse. The CFC erosion was negligible at the energy density less than 0.5 MJ/m2 and was increased to the average value 0.3 μm per pulse at 1.5 MJ/m2. The pure tungsten macrobrushes erosion was small in the energy range of 0.5-1.3 MJ/m2. The sharp growth of tungsten erosion and the intense droplet ejection were observed at the energy density of 1.5 MJ/m2.

  2. Halpern Iteration in CAT(κ) Spaces

    Institute of Scientific and Technical Information of China (English)

    Bo(z)ena PI(A)TEK

    2011-01-01

    In this paper we show that an iterative sequence generated by the Halpern algorithm converges to a fixed point in the case of complete CAT(κ) spaces. Similar results for Hadamard manifolds were obtained in[Li,C.,López, G., Martín-Márquez, V.:Iterative algorithms for nonexpansive mappings on Hadamard manifolds. Taiwanese J. Math., 14, 541-559 (2010)], but we study a much more general case. Moreover, we discuss the Halpern iteration procedure for set-valued mappings.

  3. Iterative restoration algorithms for nonlinear constraint computing

    Science.gov (United States)

    Szu, Harold

    A general iterative-restoration principle is introduced to facilitate the implementation of nonlinear optical processors. The von Neumann convergence theorem is generalized to include nonorthogonal subspaces which can be reduced to a special orthogonal projection operator by applying an orthogonality condition. This principle is shown to permit derivation of the Jacobi algorithm, the recursive principle, the van Cittert (1931) deconvolution method, the iteration schemes of Gerchberg (1974) and Papoulis (1975), and iteration schemes using two Fourier conjugate domains (e.g., Fienup, 1981). Applications to restoring the image of a double star and division by hard and soft zeros are discussed, and sample results are presented graphically.

  4. Frozen Landweber Iteration for Nonlinear Ill-Posed Problems

    Institute of Scientific and Technical Information of China (English)

    J.Xu; B.Han; L.Li

    2007-01-01

    In this paper we propose a modification of the Landweber iteration termed frozen Landweber iteration for nonlinear ill-posed problems.A convergence analysis for this iteration is presented.The numerical performance of this frozen Landweber iteration for a nonlinear Hammerstein integral equation is compared with that of the Landweber iteration.We obtain a shorter running time of the frozen Landweber iteration based on the same convergence accuracy.

  5. The Weak Neutral Current

    CERN Document Server

    Erler, Jens

    2013-01-01

    This is a review of electroweak precision physics with particular emphasis on low-energy precision measurements in the neutral current sector of the electroweak theory and includes future experimental prospects and the theoretical challenges one faces to interpret these observables. Within the minimal Standard Model they serve as determinations of the weak mixing angle which are competitive with and complementary to those obtained near the Z-resonance. In the context of new physics beyond the Standard Model these measurements are crucial to discriminate between models and to reduce the allowed parameter space within a given model. We illustrate this for the minimal supersymmetric Standard Model with or without R-parity.

  6. Mod en neutral seksualitet!

    DEFF Research Database (Denmark)

    Leer, Jonatan

    2013-01-01

    the paradigm”. This notion was presented at a series of lectures at Collège de France in 1977. Through a reading of Barthes’s autobiography, Roland Barthes par Roland Barthes (1975), the article demonstrates how Barthes in this text tries to outplay the paradigms that rules over the hegemonic understanding...... of gender and sexuality; also the fragmented text presents a vision of a sexual utopia, a neutral sexuality, that tries – like the queer theory – to go and think beyond a binary conception of gender and sexuality. Finally, it is suggested that we should start to think about a movement of “French queer...

  7. Structural analysis of the ITER Vacuum Vessel regarding 2012 ITER Project-Level Loads

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, J.-M., E-mail: jean-marc.martinez@live.fr [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul lez Durance (France); Jun, C.H.; Portafaix, C.; Choi, C.-H.; Ioki, K.; Sannazzaro, G.; Sborchia, C. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul lez Durance (France); Cambazar, M.; Corti, Ph.; Pinori, K.; Sfarni, S.; Tailhardat, O. [Assystem EOS, 117 rue Jacquard, L' Atrium, 84120 Pertuis (France); Borrelly, S. [Sogeti High Tech, RE2, 180 rue René Descartes, Le Millenium – Bat C, 13857 Aix en Provence (France); Albin, V.; Pelletier, N. [SOM Calcul – Groupe ORTEC, 121 ancien Chemin de Cassis – Immeuble Grand Pré, 13009 Marseille (France)

    2014-10-15

    Highlights: • ITER Vacuum Vessel is a part of the first barrier to confine the plasma. • ITER Vacuum Vessel as Nuclear Pressure Equipment (NPE) necessitates a third party organization authorized by the French nuclear regulator to assure design, fabrication, conformance testing and quality assurance, i.e. Agreed Notified Body (ANB). • A revision of the ITER Project-Level Load Specification was implemented in April 2012. • ITER Vacuum Vessel Loads (seismic, pressure, thermal and electromagnetic loads) were summarized. • ITER Vacuum Vessel Structural Margins with regards to RCC-MR code were summarized. - Abstract: A revision of the ITER Project-Level Load Specification (to be used for all systems of the ITER machine) was implemented in April 2012. This revision supports ITER's licensing by accommodating requests from the French regulator to maintain consistency with the plasma physics database and our present understanding of plasma transients and electro-magnetic (EM) loads, to investigate the possibility of removing unnecessary conservatism in the load requirements and to review the list and definition of incidental cases. The purpose of this paper is to present the impact of this 2012 revision of the ITER Project-Level Load Specification (LS) on the ITER Vacuum Vessel (VV) loads and the main structural margins required by the applicable French code, RCC-MR.

  8. Overview of recent nuclear analyses for the Upper ECH launcher in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Serikov, A., E-mail: serikov@inr.fzk.d [Association FZK-EURATOM, Forschungszentrum Karlsruhe, P.O. Box 3640, D-76021 Karlsruhe (Germany); Fischer, U.; Grosse, D.; Heidinger, R.; Kleefeldt, K.; Spaeh, P.; Strauss, D.; Vaccaro, A. [Association FZK-EURATOM, Forschungszentrum Karlsruhe, P.O. Box 3640, D-76021 Karlsruhe (Germany)

    2010-12-15

    An overview is given for the analyses of the nuclear physics characteristics in support of the development of the Electron Cyclotron Heating (ECH) launcher installed in the ITER upper port. The launcher's Quasi-Optical (QO) system of millimeter-wave guides represents a pathway for the neutron streaming which results in radiation loads on the launcher internals, mainly on the front steering (FS) mirrors and neighboring ITER components, in particular the vacuum vessel (VV) and the superconducting magnets surrounding the launcher. The radiation transport calculations were performed with the Monte Carlo code MCNP5 employing the recent MCNP geometry model of ITER called Alite and the FENDL-2.1 nuclear data. A dedicated CAD model of the QO ECH launcher, generated with CATIA V5, was converted at FZK into the MCNP5 geometry representation using the McCad conversion tool. The neutron and photon fluxes and critical nuclear responses, such as nuclear heating, neutron damage, and the helium production rate were calculated in the paper and compared with the ITER nuclear design limits. On the basis of the results obtained from the nuclear analyses, it is concluded that the recent design of the QO ECH launcher satisfies the ITER radiation requirements, and thus, from this point of view, can be operated safely.

  9. Clinical applications of iterative reconstruction

    Energy Technology Data Exchange (ETDEWEB)

    Eberl, S. [Royal Prince Alfred Hospital, Camperdown, NSW (Australia). Department of PET and Nuclear Medicine

    1998-03-01

    Expectation maximisation (EM) reconstruction largely eliminates the hot and cold streaking artifacts characteristic of filtered-back projection (FBP) reconstruction around localised hot areas, such as the bladder. It also substantially reduces the problem of decreased inferior wall counts in MIBI myocardial perfusion studies due to ``streaking`` from high liver uptake. Non-uniform attenuation and scatter correction, resolution recovery, anatomical information, e.g. from MRI or CT tracer kinetic modelling, can all be built into the EM reconstruction imaging model. The properties of ordered subset EM (OSEM) have also been used to correct for known patient motion as part of the reconstruction process. These uses of EM are elaborated more fully in some of the other abstracts of this meeting. Currently we use OSEM routinely for: (i) studies where streaking is a problem, including all MIBI myocardial perfusion studies, to avoid hot liver inferior wall artifact, (ii) all whole body FDG PET, all lung V/Q SPECT (which have a short acquisition time) and all gated {sup 201}TI myocardial perfusion studies due to improved noise characteristics of OSEM in these studies; (iii) studies with measured, non-uniform attenuation correction. With the accelerated OSEM algorithm, iterative reconstruction is practical for routine clinical applications and we have found OSEM to provide clearly superior reconstructions for the areas listed above and are investigating its application to other studies. In clinical use, we have not found OSEM to introduce artifacts which would not also occur with FBP, e.g. uncorrected patient motion will cause artifacts with both OSEM and FBP.

  10. Numerical Stability Test of Neutral Delay Differential Equations

    Directory of Open Access Journals (Sweden)

    Z. H. Wang

    2008-01-01

    Full Text Available The stability of a delay differential equation can be investigated on the basis of the root location of the characteristic function. Though a number of stability criteria are available, they usually do not provide any information about the characteristic root with maximal real part, which is useful in justifying the stability and in understanding the system performances. Because the characteristic function is a transcendental function that has an infinite number of roots with no closed form, the roots can be found out numerically only. While some iterative methods work effectively in finding a root of a nonlinear equation for a properly chosen initial guess, they do not work in finding the rightmost root directly from the characteristic function. On the basis of Lambert W function, this paper presents an effective iterative algorithm for the calculation of the rightmost roots of neutral delay differential equations so that the stability of the delay equations can be determined directly, illustrated with two examples.

  11. Several atomic-physics issues connected with the use of neutral beams in fusion experiments

    Energy Technology Data Exchange (ETDEWEB)

    Post, D.E.; Grisham, L.R.; Fonck, R.J.

    1982-08-01

    Energetic neutral beams are used for heating and diagnostics in present magnetic fusion experiments. They are also being considered for use in future large experiments. Atomic physics issues are important for both the production of the neutral beams and the interaction of the beams and the plasma. Interest in neutral beams based on negative hydrogen ions is growing, largely based on advances in producing high current ion sources. An extension of the negative ion approach has been the suggestion to use negative ions of Z > 1 elements, such as carbon and oxygen, to form high power neutral beams for plasma heating.

  12. Is /h/ phonetically neutral?

    Science.gov (United States)

    Robb, Michael P; Chen, Yang

    2009-11-01

    Use of /h/ in the phrase, 'Say /hVC/ again' has been tacitly assumed to provide a neutral phonetic context in which to study the articulatory characteristics of speech either preceding or following /h/ articulation. Yet, assessment of the stability or neutrality of /h/ has gone untested. The current study sought to determine whether articulation of /h/ differs according to sex and language accent, as well as to examine its influence on subsequent vowel articulation. Selected acoustic features of /hVC/ were measured in 40 speakers of American English (AE) and 40 speakers of Mandarin-accented English (MAE). Results of an analysis of /h/ duration revealed no sex differences within each language group, however considerable variation was found according to accented vs unaccented English. Clear sex differences were found for the production of /h/, occurring more often among male speakers regardless of language variety. Considerable variation in production of /h/ was found between language groups. Analysis of vowel formant frequencies immediately following /h/ articulation indicated minimal coarticulatory effects for both AE and MAE speakers. The present results appear to support the suggestion that /h/ is not exclusively sex-linked and may indeed vary according to non-biological factors. In spite of these variations, /h/ articulation appears to have a negligible influence on neighbouring vowel articulation.

  13. Conceptual design of the beam source for the DEMO Neutral Beam Injectors

    Science.gov (United States)

    Sonato, P.; Agostinetti, P.; Fantz, U.; Franke, T.; Furno, I.; Simonin, A.; Tran, M. Q.

    2016-12-01

    DEMO (DEMOnstration Fusion Power Plant) is a proposed nuclear fusion power plant that is intended to follow the ITER experimental reactor. The main goal of DEMO will be to demonstrate the possibility to produce electric energy from the fusion reaction. The injection of high energy neutral beams is one of the main tools to heat the plasma up to fusion conditions. A conceptual design of the Neutral Beam Injector (NBI) for the DEMO fusion reactor, is currently being developed by Consorzio RFX in collaboration with other European research institutes. High efficiency and low recirculating power, which are fundamental requirements for the success of DEMO, have been taken into special consideration for the DEMO NBI. Moreover, particular attention has been paid to the issues related to reliability, availability, maintainability and inspectability. A conceptual design of the beam source for the DEMO NBI is here presented featuring 20 sub-sources (two adjacent columns of 10 sub-sources each), following a modular design concept, with each sub-source featuring its radio frequency driver, capable of increasing the reliability and availability of the DEMO NBI. Copper grids with increasing size of the apertures have been adopted in the accelerator, with three main layouts of the apertures (circular apertures, slotted apertures and frame-like apertures for each sub-source). This design, permitting to significantly decrease the stripping losses in the accelerator without spoiling the beam optics, has been investigated with a self-consistent model able to study at the same time the magnetic field, the electrostatic field and the trajectory of the negative ions. Moreover, the status on the R&D carried out in Europe on the ion sources is presented.

  14. Applications of the ergodic iteration theorem

    OpenAIRE

    Zapletal, J.

    2010-01-01

    I prove several natural preservation theorems for the countable support iteration. This solves a question of Roslanowski regarding the preservation of localization properties and greatly simplifies the proofs in the area.

  15. Archimedes' Pi--An Introduction to Iteration.

    Science.gov (United States)

    Lotspeich, Richard

    1988-01-01

    One method (attributed to Archimedes) of approximating pi offers a simple yet interesting introduction to one of the basic ideas of numerical analysis, an iteration sequence. The method is described and elaborated. (PK)

  16. Stability of Jungck-type iterative procedures

    Directory of Open Access Journals (Sweden)

    S. L. Singh

    2005-01-01

    Full Text Available We introduce and discuss the stability of Jungck and Jungck-Mann iterative procedures for a pair of Jungck-Osilike-type maps on an arbitrary set with values in a metric or linear metric space.

  17. Anderson Acceleration for Fixed-Point Iterations

    Energy Technology Data Exchange (ETDEWEB)

    Walker, Homer F. [Worcester Polytechnic Institute, MA (United States)

    2015-08-31

    The purpose of this grant was to support research on acceleration methods for fixed-point iterations, with applications to computational frameworks and simulation problems that are of interest to DOE.

  18. Accelerating Iterative Big Data Computing Through MPI

    Institute of Scientific and Technical Information of China (English)

    梁帆; 鲁小亿

    2015-01-01

    Current popular systems, Hadoop and Spark, cannot achieve satisfied performance because of the inefficient overlapping of computation and communication when running iterative big data applications. The pipeline of computing, data movement, and data management plays a key role for current distributed data computing systems. In this paper, we first analyze the overhead of shuffle operation in Hadoop and Spark when running PageRank workload, and then propose an event-driven pipeline and in-memory shuffle design with better overlapping of computation and communication as DataMPI-Iteration, an MPI-based library, for iterative big data computing. Our performance evaluation shows DataMPI-Iteration can achieve 9X∼21X speedup over Apache Hadoop, and 2X∼3X speedup over Apache Spark for PageRank and K-means.

  19. Overview and status of ITER internal components

    Energy Technology Data Exchange (ETDEWEB)

    Merola, Mario, E-mail: mario.merola@iter.org; Escourbiac, Frederic; Raffray, René; Chappuis, Philippe; Hirai, Takeshi; Martin, Alex

    2014-10-15

    Highlights: • Manufacturing technologies for the ITER internal components have been developed. • The Blanket System successfully went through its Final Design Review in April 2013. • The decision to start operation with a Divertor with a full-W armour has been taken. - Abstract: The internal components of ITER are one of the most design and technically challenging components of the ITER machine, and include the Blanket System and the Divertor. The Blanket System successfully went through its Final Design Review in April 2013 and now it is entering into the procurement phase. The design and qualification of the Divertor with a full-tungsten armour was successfully completed and this enabled the decision in November 2013 to start operation with this material option. This paper summarizes the engineering design, the R and D, the technology qualification and procurement status of the Blanket System and of the Divertor of the ITER machine.

  20. Erosion of beryllium under ITER – Relevant transient plasma loads

    Energy Technology Data Exchange (ETDEWEB)

    Kupriyanov, I.B., E-mail: igkupr@gmail.com [A.A. Bochvar High Technology Research Institute of Inorganic Materials, Rogova St. 5a, 123060 Moscow (Russian Federation); Nikolaev, G.N.; Kurbatova, L.A.; Porezanov, N.P. [A.A. Bochvar High Technology Research Institute of Inorganic Materials, Rogova St. 5a, 123060 Moscow (Russian Federation); Podkovyrov, V.L.; Muzichenko, A.D.; Zhitlukhin, A.M. [TRINITI, Troitsk, Moscow reg. (Russian Federation); Gervash, A.A. [Efremov Research Institute, S-Peterburg (Russian Federation); Safronov, V.M. [Project Center of ITER, Moscow (Russian Federation)

    2015-08-15

    Highlights: • We study the erosion, mass loss/gain and surface structure evolution of Be/CuCrZr mock-ups, armored with beryllium of TGP-56FW grade after irradiation by deuterium plasma heat load of 0.5 MJ/m{sup 2} at 250 °C and 500 °C. • Beryllium mass loss/erosion under plasma heat load at 250 °C is rather small (no more than 0.2 g/m{sup 2} shot and 0.11 μm/shot, correspondingly, after 40 shots) and tends to decrease with increasing number of shots. • Beryllium mass loss/erosion under plasma heat load at 500 °C is much higher (∼2.3 g/m{sup 2} shot and 1.2 μm/shot, correspondingly, after 10 shot) and tends to decrease with increasing the number of shots (∼0.26 g/m{sup 2} pulse and 0.14 μm/shot, correspondingly, after 100 shot). • Beryllium erosion value derived from the measurements of profile of irradiated surface is much higher than erosion value derived from mass loss data. - Abstract: Beryllium will be used as a armor material for the ITER first wall. It is expected that erosion of beryllium under transient plasma loads such as the edge-localized modes (ELMs) and disruptions will mainly determine a lifetime of the ITER first wall. This paper presents the results of recent experiments with the Russian beryllium of TGP-56FW ITER grade on QSPA-Be plasma gun facility. The Be/CuCrZr mock-ups were exposed to up to 100 shots by deuterium plasma streams (5 cm in diameter) with pulse duration of 0.5 ms and heat loads range of 0.2–0.5 MJ/m{sup 2} at different temperature of beryllium tiles. The temperature of Be tiles has been maintained about 250 and 500 °C during the experiments. After 10, 40 and 100 shots, the beryllium mass loss/gain under erosion process were investigated as well as evolution of surface microstructure and cracks morphology.

  1. Experiences with iterated traffic microsimulations in Dallas

    CERN Document Server

    Nagel, K

    1997-01-01

    This paper reports experiences with iterated traffic microsimulations in the context of a Dallas study. ``Iterated microsimulations'' here means that the information generated by a microsimulation is fed back into the route planner so that the simulated individuals can adjust their routes to circumvent congestion. This paper gives an overview over what has been done in the Dallas context to better understand the relaxation process, and how to judge the robustness of the results.

  2. Threshold power and energy confinement for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Takizuka, T.

    1996-12-31

    In order to predict the threshold power for L-H transition and the energy confinement performance in ITER, assembling of database and analyses of them have been progressed. The ITER Threshold Database includes data from 10 divertor tokamaks. Investigation of the database gives a scaling of the threshold power of the form P{sub thr} {proportional_to} B{sub t} n{sub e}{sup 0.75} R{sup 2} {times} (n{sub e} R{sup 2}){sup +-0.25}, which predicts P{sub thr} = 100 {times} 2{sup 0{+-}1} MW for ITER at n{sub e} = 5 {times} 10{sup 19} m{sup {minus}3}. The ITER L-mode Confinement Database has also been expanded by data from 14 tokamaks. A scaling of the thermal energy confinement time in L-mode and ohmic phases is obtained; {tau}{sub th} {approximately} I{sub p} R{sup 1.8} n{sub e}{sup 0.4{sub P{sup {minus}0.73}}}. At the ITER parameter, it becomes about 2.2 sec. For the ignition in ITER, more than 2.5 times of improvement will be required from the L-mode. The ITER H-mode Confinement Database is expanded from data of 6 tokamaks to data of 11 tokamaks. A {tau}{sub th} scaling for ELMy H-mode obtained by a standard regression analysis predicts the ITER confinement time of {tau}{sub th} = 6 {times} (1 {+-} 0.3) sec. The degradation of {tau}{sub th} with increasing n{sub e} R{sup 2} (or decreasing {rho}{sub *}) is not found for ELMy H-mode. An offset linear law scaling with a dimensionally correct form also predicts nearly the same {tau}{sub th} value.

  3. Overview and status of ITER Cryostat manufacturing

    Energy Technology Data Exchange (ETDEWEB)

    Bhardwaj, Anil K., E-mail: anil.bhardwaj@iter-india.org [ITER-India, Institute For Plasma Research, A-29, GIDC Electronics Estate, Sector-25, Gandhinagar 382016 (India); Gupta, Girish; Prajapati, Rajnikant; Joshi, Vaibhav; Patel, Mitul; Bhavsar, Jagrut; More, Vipul; Jindal, Mukesh; Bhattacharya, Avik; Jogi, Gourav; Palaliya, Amit; Jha, Saroj; Pandey, Manish; Shukla, Dileep [ITER-India, Institute For Plasma Research, A-29, GIDC Electronics Estate, Sector-25, Gandhinagar 382016 (India); Iyer, Ganesh; Jadhav, Pandurang; Goyal, Dipesh; Desai, Anish [Larsen & Toubro Limited, Heavy Engineering, Hazira Manufacturing Complex, Gujarat (India); Sekachev, I.; Vitupier, Guillaume [ITER Organization, Route de Vinon sur Verdon – CS 90046, 13067 Saint Paul Lez Durance Cedex (France); and others

    2016-11-01

    Highlights: • Manufacturing status of one of the largest and the heaviest fully welded stainless steel vacuum chambers in the world (ITER Cryostat). • Overview of manufacturing stages and its segmentation. • Overview of manufacturing procedures and assembly and installation. - Abstract: One of ITER-India's commitments to the ITER Organization is procurement of the ITER Cryostat. It is a large vacuum vessel (∼29 m dia. and ∼29 m height), which is made up of 304/304 L dual marked stainless steel and has a total mass over 3500 t. The thickness of the vessel wall varies from 50 mm to 190 mm. It is one of the largest and the heaviest fully welded stainless steel vacuum chambers in the world which provides vacuum thermal insulation for the superconducting magnets operating at 4.5 K and for the thermal shield operating at 80 K. It also mechanically supports the magnet system along with the vacuum vessel (VV). The cryostat is designed and constructed according to ASME Section-VIII Division-2 with additional ITER Vacuum Handbook requirements and it is classified as protection important component (PIC-2). Manufacturing of cryostat segments is ongoing in India; sub-assembly of four major sections of the cryostat from the segments will be done at the ITER site in a temporary workshop building and the final assembly will be done in the pit of the tokamak building, the final location. The cryostat manufacturing contract has been awarded to Larsen and Toubro Limited in August 2012 after completion of design [4] and signing of Procurement Arrangement [1] with ITER Organization. Manufacturing of the cryostat was started in January 2014 after approval of the manufacturing drawings and procedures. The temporary workshop of 44 m × 110 m × 26 m in height has been completed in November 2014 at the ITER site with a 200 t crane installed. This paper gives an overview and the status of the cryostat manufacturing.

  4. Efficient iterative adaptive pole placement algorithm

    Institute of Scientific and Technical Information of China (English)

    李俊民; 李靖; 杨磊

    2004-01-01

    An iterative adaptive pole placement algorithm is presented. The stability and the convergence of the algorithm are respectively established. Since one-step iterative formulation in computing controller's parameters is used, the on-line computation cost is greatly reduced with respected to the traditional algorithm. The algorithm with the feed-forward can follow arbitrarily bounded output. The algorithm is also extended to multivariate case. Simulation examples show the efficiency and robustness of the algorithm.

  5. Iterative consolidation of unorganized point clouds.

    Science.gov (United States)

    Liu, Shengjun; Chan, Kwan-Chung; Wang, Charlie C L

    2012-01-01

    Unorganized point clouds obtained from 3D shape acquisition devices usually present noise, outliers, and nonuniformities. The proposed framework consolidates unorganized points through an iterative procedure of interlaced downsampling and upsampling. Selection operations remove outliers while preserving geometric details. The framework improves the uniformity of points by moving the downsampled particles and refining point samples. Surface extrapolation fills missed regions. Moreover, an adaptive sampling strategy speeds up the iterations. Experimental results demonstrate the framework's effectiveness.

  6. Accelerated Schwarz iterations for Helmholtz equation

    Science.gov (United States)

    Nagid, Nabila; Belhadj, Hassan; Amattouch, Mohamed Ridouan

    2017-01-01

    In this paper, the Restricted additive Schwarz (RAS) method is applied to solve Helmholtz equation. To accelerate the RAS iterations, we propose to apply the vector ɛ-algorithm. Some convergence analysis of the proposed method is presented, and applied succeffully to Helmholtz problem. The obtained results show the efficiency of the proposed approach. Moreover, the algorithm yields much faster convergence than the classical Schwarz iterations.

  7. IMM Iterated Extended Particle Filter Algorithm

    OpenAIRE

    Yang Wan; Shouyong Wang; Xing Qin

    2013-01-01

    In order to solve the tracking problem of radar maneuvering target in nonlinear system model and non-Gaussian noise background, this paper puts forward one interacting multiple model (IMM) iterated extended particle filter algorithm (IMM-IEHPF). The algorithm makes use of multiple modes to model the target motion form to track any maneuvering target and each mode uses iterated extended particle filter (IEHPF) to deal with the state estimation problem of nonlinear non-Gaussian system. IEH...

  8. Algorithmic Optimisations for Iterative Deconvolution Methods

    OpenAIRE

    Welk, Martin; Erler, Martin

    2013-01-01

    We investigate possibilities to speed up iterative algorithms for non-blind image deconvolution. We focus on algorithms in which convolution with the point-spread function to be deconvolved is used in each iteration, and aim at accelerating these convolution operations as they are typically the most expensive part of the computation. We follow two approaches: First, for some practically important specific point-spread functions, algorithmically efficient sliding window or list processing tech...

  9. Simulation of the neutral inventory in the pilot-PSI beam

    Energy Technology Data Exchange (ETDEWEB)

    Wieggers, R.C.; Groen, P.W.C.; Blank, H.J. de; Goedheer, W.J. [FOM Institute DIFFER - Dutch Institute for Fundamental Energy Research, Association EURATOM-FOM, Nieuwegein (Netherlands)

    2012-06-15

    The Eunomia code is used to study the neutral species in and near a hydrogen plasma beam. Eunomia is a non-linear Monte Carlo transport code that solves the neutral equilibrium, given a fixed background plasma. The code is developed to study the neutral inventory of Pilot-PSI and Magnum-PSI, linear devices developed to study plasma surface interactions in similar conditions as expected in the ITER divertor. Results show the influence of elastic collisions and the outer vessel wall on the neutral species. In the center of the 2 cm diameter Pilot-PSI beam the results show a strong coupling to the plasma. Only millimeters away from the center, the neutral flow, temperature and density are strongly influenced by recombination processes at the vessel wall (copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  10. Experimental and numerical studies of neutral gas depletion in an inductively coupled plasma

    Science.gov (United States)

    Shimada, Masashi

    The central theme of this dissertation is to explore the impact of neutral depletion and coupling between plasma and neutral gas in weakly ionized unmagnetized plasma. Since there have been few systematic studies of the mechanism which leads to non-uniform neutral distribution in processing plasmas, this work investigated the spatial profiles of neutral temperature and pressure experimentally, and the mechanism of resulting neutral depletion by simulation. The experimental work is comprised of neutral temperature measurements using high resolution atomic spectroscopy and molecular spectroscopy, and neutral pressure measurements considering thermal transpiration. When thermal transpiration effects are used to correct the gas pressure measurements, the total pressure remains constant regardless of the plasma condition. Since the neutral gas follows the ideal gas law, the neutral gas density profile is also obtained from the meas