WorldWideScience

Sample records for iter heating neutral

  1. Status of the Negative Ion Based Heating and Diagnostic Neutral Beams for ITER

    Science.gov (United States)

    Schunke, B.; Bora, D.; Hemsworth, R.; Tanga, A.

    2009-03-01

    The current baseline of ITER foresees 2 Heating Neutral Beam (HNB's) systems based on negative ion technology, each accelerating to 1 MeV 40 A of D- and capable of delivering 16.5 MW of D0 to the ITER plasma, with a 3rd HNB injector foreseen as an upgrade option [1]. In addition a dedicated Diagnostic Neutral Beam (DNB) accelerating 60 A of H- to 100 keV will inject ≈15 A equivalent of H0 for charge exchange recombination spectroscopy and other diagnostics. Recently the RF driven negative ion source developed by IPP Garching has replaced the filamented ion source as the reference ITER design. The RF source developed at IPP, which is approximately a quarter scale of the source needed for ITER, is expected to have reduced caesium consumption compared to the filamented arc driven ion source. The RF driven source has demonstrated adequate accelerated D- and H- current densities as well as long-pulse operation [2, 3]. It is foreseen that the HNB's and the DNB will use the same negative ion source. Experiments with a half ITER-size ion source are on-going at IPP and the operation of a full-scale ion source will be demonstrated, at full power and pulse length, in the dedicated Ion Source Test Bed (ISTF), which will be part of the Neutral Beam Test Facility (NBTF), in Padua, Italy. This facility will carry out the necessary R&D for the HNB's for ITER and demonstrate operation of the full-scale HNB beamline. An overview of the current status of the neutral beam (NB) systems and the chosen configuration will be given and the ongoing integration effort into the ITER plant will be highlighted. It will be demonstrated how installation and maintenance logistics have influenced the design, notably the top access scheme facilitating access for maintenance and installation. The impact of the ITER Design Review and recent design change requests (DCRs) will be briefly discussed, including start-up and commissioning issues. The low current hydrogen phase now envisaged for start

  2. Status of the Negative Ion Based Heating and Diagnostic Neutral Beams for ITER

    International Nuclear Information System (INIS)

    Schunke, B.; Bora, D.; Hemsworth, R.; Tanga, A.

    2009-01-01

    The current baseline of ITER foresees 2 Heating Neutral Beam (HNB's) systems based on negative ion technology, each accelerating to 1 MeV 40 A of D - and capable of delivering 16.5 MW of D 0 to the ITER plasma, with a 3rd HNB injector foreseen as an upgrade option. In addition a dedicated Diagnostic Neutral Beam (DNB) accelerating 60 A of H - to 100 keV will inject ≅15 A equivalent of H 0 for charge exchange recombination spectroscopy and other diagnostics. Recently the RF driven negative ion source developed by IPP Garching has replaced the filamented ion source as the reference ITER design. The RF source developed at IPP, which is approximately a quarter scale of the source needed for ITER, is expected to have reduced caesium consumption compared to the filamented arc driven ion source. The RF driven source has demonstrated adequate accelerated D - and H - current densities as well as long-pulse operation. It is foreseen that the HNB's and the DNB will use the same negative ion source. Experiments with a half ITER-size ion source are on-going at IPP and the operation of a full-scale ion source will be demonstrated, at full power and pulse length, in the dedicated Ion Source Test Bed (ISTF), which will be part of the Neutral Beam Test Facility (NBTF), in Padua, Italy. This facility will carry out the necessary R and D for the HNB's for ITER and demonstrate operation of the full-scale HNB beamline. An overview of the current status of the neutral beam (NB) systems and the chosen configuration will be given and the ongoing integration effort into the ITER plant will be highlighted. It will be demonstrated how installation and maintenance logistics have influenced the design, notably the top access scheme facilitating access for maintenance and installation. The impact of the ITER Design Review and recent design change requests (DCRs) will be briefly discussed, including start-up and commissioning issues. The low current hydrogen phase now envisaged for start

  3. ITER neutral beam system

    International Nuclear Information System (INIS)

    Mondino, P.L.; Di Pietro, E.; Bayetti, P.

    1999-01-01

    The Neutral Beam (NB) system for the International Thermonuclear Experimental Reactor (ITER) has reached a high degree of integration with the tokamak and with the rest of the plant. Operational requirements and maintainability have been considered in the design. The paper considers the integration with the tokamak, discusses design improvements which appear necessary and finally notes R and D progress in key areas. (author)

  4. ITER Neutral Beam Injection System

    International Nuclear Information System (INIS)

    Ohara, Yoshihiro; Tanaka, Shigeru; Akiba, Masato

    1991-03-01

    A Japanese design proposal of the ITER Neutral Beam Injection System (NBS) which is consistent with the ITER common design requirements is described. The injection system is required to deliver a neutral deuterium beam of 75MW at 1.3MeV to the reactor plasma and utilized not only for plasma heating but also for current drive and current profile control. The injection system is composed of 9 modules, each of which is designed so as to inject a 1.3MeV, 10MW neutral beam. The most important point in the design is that the injection system is based on the utilization of a cesium-seeded volume negative ion source which can produce an intense negative ion beam with high current density at a low source operating pressure. The design value of the source is based on the experimental values achieved at JAERI. The utilization of the cesium-seeded volume source is essential to the design of an efficient and compact neutral beam injection system which satisfies the ITER common design requirements. The critical components to realize this design are the 1.3MeV, 17A electrostatic accelerator and the high voltage DC acceleration power supply, whose performances must be demonstrated prior to the construction of ITER NBI system. (author)

  5. Structural analysis of the Passive Magnetic Shield for the ITER Heating Neutral Beam Injector system

    Energy Technology Data Exchange (ETDEWEB)

    Cabrera, Santiago, E-mail: santiago.cabrera@ciemat.es [CIEMAT Laboratorio Nacional de Fusión, Avda. Complutense 40, 28040 Madrid (Spain); Rincón, Esther; Ahedo, Begoña; Alonso, Javier; Barrera, Germán; Ramos, Francisco; Ríos, Luis [CIEMAT Laboratorio Nacional de Fusión, Avda. Complutense 40, 28040 Madrid (Spain); El-Ouazzani, Anass; García, Pablo [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 ITER Passive Magnetic Shield (PMS) main function is to protect the Neutral Beam Injector (NBI) from the external magnetic field coming from the tokamak, and to shield the NB cell from the radiation coming from all activated components. The shielding from the external magnetic field is performed in association with the Active Compensation Cooled Correction Coils (ACCC). The Bushing and Transmission Line (TL) PMS also provides structural support for HV bushing, allowing its maintenance and providing air sealing function between NBI cell and High Voltage deck room. The paper summarizes the structural analyses performed in order to evaluate the mechanical behaviour of the HNB PMS under operation combined with seismic event. The RCC-MR Code is used to validate the design, assuming creep is negligible, since the structure is expected to be at room temperature. P-type damage is assessed.

  6. 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

  7. Neutral particle beam alternative concept for ITER

    International Nuclear Information System (INIS)

    Sedgley, D.; Brook, J.; Luzzi, T.; Deutsch, L.

    1989-01-01

    An analysis of an ITER neutral particle beam system is presented. The analysis covers the neutralizer, ion dumps, pumping, and geometric aspects. The US beam concept for ITER consists of three or four clusters of beamlines delivering approximately 80 MW total of 1.6-MeV deuterium to three or four reactor ports. Each cluster has three self-contained beamlines featuring plasma neutralizers and electrostatic ion dumps. In this study, each of the beamlines has two source assemblies with separate gas neutralizers and magnetic ion dumps. Deuterium is injected into the gas neutralizers by a separate system. Saddle-shaped copper coils augment the tokamak poloidal field to turn the charged particles into the ion dumps. The gas flow from the source, neutralizer, and ion dump is pumped by regenerable cryopanels. The effect of the port between the TF coils and the beam injection angle on the plasma footprint was studied

  8. Design of the ITER Neutral Beam injectors

    International Nuclear Information System (INIS)

    Hemsworth, R.S.; Feist, J.; Hanada, M.; Heinemann, B.; Inoue, T.; Kuessel, E.; Kulygin, V.; Krylov, A.; Lotte, P.; Miyamoto, K.; Miyamoto, N.; Murdoch, D.; Nagase, A.; Ohara, Y.; Okumura, Y.; Pamela, J.; Panasenkov, A.; Shibata, K.; Tanii, M.

    1996-01-01

    This paper describes the Neutral Beam Injection system which is presently being designed 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 0 to the ITER plasma for pulse length of ≥1000 s. The injectors each use 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 - . This will be neutralized 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. copyright 1996 American Institute of Physics

  9. 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.

  10. Guidelines for Remote Handling Maintenance of ITER Neutral Beam Components

    International Nuclear Information System (INIS)

    Cordier, J.-J.; Hemsworth, R.; Bayetti, P.

    2006-01-01

    Remote handling maintenance of ITER components is one of the main challenges of the ITER project. This type of maintenance shall be operational for the nuclear phase of exploitation of ITER, and be considered at a very early stage since it significantly impacts on the components design, interfaces management and integration business. A large part of the R/H equipment will be procured by the EU partner, in particular the whole Neutral Beam Remote Handling (RH) equipment package. A great deal of work has already been done in this field during the EDA phase of ITER project, but improvements and alternative option that are now proposed by ITER lead to added RH and maintenance engineering studies. The Neutral Beam Heating -and- Current Drive system 1 is being revisited by the ITER project. The vertical maintenance scheme that is presently considered by ITER, may significantly impact on the reference design of the Neutral Beam (NB) system and associated components and lead to new design of the NB box itself. In addition, revision of both NB cell radiation level zoning and remote handling classification of the beam line injector will also significantly impact on components design and maintenance. Based on the experience gained on the vertical maintenance scheme, developed in detail for the ITER Neutral Beam Test Facility 2 to be built in Europe in a near future, guidelines for the revision of the design and preliminary feasibility study of the remote handling vertical maintenance scheme of beam line components are described in the paper. A maintenance option for the SINGAP3 accelerator is also presented. (author)

  11. EPICS - MDSplus integration in the ITER Neutral Beam Test Facility

    International Nuclear Information System (INIS)

    Luchetta, Adriano; Manduchi, Gabriele; Barbalace, Antonio; Soppelsa, Anton; Taliercio, Cesare

    2011-01-01

    SPIDER, the ITER-size ion-source test bed in the ITER Neutral Beam Test Facility, is a fusion device requiring a complex central system to provide control and data acquisition, referred to as CODAS. The CODAS software architecture will rely on EPICS and MDSplus, two open-source, collaborative software frameworks, targeted at control and data acquisition, respectively. EPICS has been selected as ITER CODAC middleware and, as the final deliverable of the Neutral Beam Test Facility is the procurement of the ITER Heating Neutral Beam Injector, we decided to adopt this ITER technology. MDSplus is a software package for data management, supporting advanced concepts, such as platform and underlying hardware independence, self description data, and data driven model. The combined use of EPICS and MDSplus is not new in fusion, but their level of integration will be new in SPIDER, achieved by a more refined data access layer. The paper presents the integration software to use effectively EPICS and MDSplus, including the definition of appropriate EPICS records to interact with MDSplus. The MDSplus and EPICS archive concepts are also compared on the basis of performance tests and data streaming is investigated by ad-hoc measurements.

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

    International Nuclear Information System (INIS)

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

    2008-01-01

    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

  13. ITER neutral beam system US conceptual design

    International Nuclear Information System (INIS)

    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 - 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

  14. Intense diagnostic neutral beam development for ITER

    International Nuclear Information System (INIS)

    Rej, D.J.; Henins, I.; Fonck, R.J.; Kim, Y.J.

    1992-01-01

    For the next-generation, burning tokamak plasmas such as ITER, diagnostic neutral beams and beam spectroscopy will continue to be used to determine a variety of plasma parameters such as ion temperature, rotation, fluctuations, impurity content, current density profile, and confined alpha particle density and energy distribution. Present-day low-current, long-pulse beam technology will be unable to provide the required signal intensities because of higher beam attenuation and background bremsstrahlung radiation in these larger, higher-density plasmas. To address this problem, we are developing a short-pulse, intense diagnostic neutral beam. Protons or deuterons are accelerated using magnetic-insulated ion-diode technology, and neutralized in a transient gas cell. A prototype 25-kA, 100-kV, 1-μs accelerator is under construction at Los Alamos. Initial experiments will focus on ITER-related issues of beam energy distribution, current density, pulse length, divergence, propagation, impurity content, reproducibility, and maintenance

  15. Modeling of the lithium based neutralizer for ITER neutral beam injector

    Energy Technology Data Exchange (ETDEWEB)

    Dure, F., E-mail: franck.dure@u-psud.fr [LPGP, Laboratoire de Physique des Gaz et Plasmas, CNRS-Universite Paris Sud, Orsay (France); Lifschitz, A.; Bretagne, J.; Maynard, G. [LPGP, Laboratoire de Physique des Gaz et Plasmas, CNRS-Universite Paris Sud, Orsay (France); Simonin, A. [IRFM, Institut de Recherche sur la Fusion Magnetique, CEA Cadarache, 13108 Saint-Paul lez Durance (France); Minea, T. [LPGP, Laboratoire de Physique des Gaz et Plasmas, CNRS-Universite Paris Sud, Orsay (France)

    2012-04-04

    Highlights: Black-Right-Pointing-Pointer We compare different lithium based neutraliser configurations to the deuterium one. Black-Right-Pointing-Pointer We study characteristics of the secondary plasma and the propagation of the 1 MeV beam. Black-Right-Pointing-Pointer Using lithium increases the neutralisation effiency keeping correct beam focusing. Black-Right-Pointing-Pointer Using lithium also reduces the backstreaming effect in direction of the ion source. - Abstract: To achieve thermonuclear temperatures necessary to produce fusion reactions in the ITER Tokamak, additional heating systems are required. One of the main method to heat the plasma ions in ITER will be the injection of energetic neutrals (NBI). In the neutral beam injector, negative ions (D{sup -}) are electrostatically accelerated to 1 MeV, and then stripped of their extra electron via collisions with a target gas, in a structure known as neutralizer. In the current ITER specification, the target gas is deuterium. It has been recently proposed to use lithium vapor instead of deuterium as target gas in the neutralizer. This would allow to reduce the gas load in the NBI vessel and to improve the neutralization efficiency. A Particle-in-Cell Monte Carlo code has been developed to study the transport of the beams and the plasma formation in the neutralizer. A comparison between Li and D{sub 2} based neutralizers made with this code is presented here, as well as a parametric study on the geometry of the Li based neutralizer. Results demonstrate the feasibility of a Li based neutralizer, and its advantages with respect to the deuterium based one.

  16. Neutral transport and helium pumping of ITER

    International Nuclear Information System (INIS)

    Ruzic, D.N.

    1990-08-01

    A 2-D Monte-Carlo simulation of the neutral atom densities in the divertor, divertor throat and pump duct of ITER was made using the DEGAS code. Plasma conditions in the scrape-off layer and region near the separatrix were modeled using the B2 plasma transport code. Wall reflection coefficients including the effect of realistic surface roughness were determined by using the fractal TRIM code. The DEGAS and B2 coupling was iterated until a consistent recycling was predicted. Results were obtained for a helium and a deuterium/tritium mixture on 7 different ITER divertor throat geometries for both the physics phase reference base case and a technology phase case. The geometry with a larger structure on the midplane-side of the throat opening closing the divertor throat and a divertor plate which maintains a steep slope well into the throat removed helium 1.5 times better than the reference geometry for the physics phase case and 2.2 times better for the technology phase case. At the same time the helium to hydrogen pumping ratio shows a factor of 2.34 ± .41 enhancement over the ratio of helium to hydrogen incident on the divertor plate in the physics phase and an improvement of 1.61 ± .31 in the technology phase. If the helium flux profile on the divertor plate is moved outward by 20 cm with respect to the D/T flux profile for this particular geometry, the enhancement increases to 4.36 ± .90 in the physics phase and 5.10 ± .92 in the technology phase

  17. 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

  18. International Thermonuclear Experimental Reactor (ITER) neutral beam design

    International Nuclear Information System (INIS)

    Myers, T.J.; Brook, J.W.; Spampinato, P.T.; Mueller, J.P.; Luzzi, T.E.; Sedgley, D.W.

    1990-10-01

    This report discusses the following topics on ITER neutral beam design: ion dump; neutralizer and module gas flow analysis; vacuum system; cryogenic system; maintainability; power distribution; and system cost

  19. An Absolute Valve for the ITER Neutral Beam Injector

    International Nuclear Information System (INIS)

    Jones, Ch.; Chuilon, B.; Michael, W.

    2006-01-01

    In the ITER reference design a fast shutter was included to limit tritium migration into the beamline vacuum enclosures. The need was recently identified to extend the functionality of the fast shutter to that of an absolute valve in order to facilitate injector maintenance procedures and to satisfy safety requirements in case of an in-vessel loss of coolant event. Three concepts have been examined satisfying the ITER requirements for speed of actuation, sealing performance over the required lifetime, and pressure differential in fault scenarios, namely: a rectangular closure section; a circular cross section; and a rotary JET-type valve. The rectangular section represents the most efficient usage of the available space envelope and leads to a minimum-mass system, although it requires greater total force for a given load per unit length of seal. However, a metallic seal of the '' hard/hard '' type, where the seal relies on the elastic properties of the material and does not utilise any type of spring device, can provide the required seal performance with typical loading of 200 kg/cm. The conceptual design of the proposed absolute valve will be presented. The aperture dimensions are 1.45 m high by 0.6 m wide, with minimum achievable leak rate of 1 · 10 -9 mbarl/s and maximum pressure differential of 3 bar across the valve. Sealing force is provided using two seal plates, linked by a 3 mm thick ' omega ' diaphragm, by pressurisation of the interspace to 8 bar; this allows for a relative movement of the plates of 2 mm. Movement of the device perpendicular to the beam direction is carried out using a novel magnetic drive in order to transmit the motive force across the vacuum boundary, similar to that demonstrated on a test-rig in an earlier study. The conceptual design includes provision of all the services such as pneumatics and water cooling to cope with the heat loads from neutral beams in quasi steady-state operation and from the ITER plasma. A future programme

  20. 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.

  1. Maintenance schemes for the ITER neutral beam test facility

    International Nuclear Information System (INIS)

    Zaccaria, P.; Dal Bello, S.; Marcuzzi, D.; Masiello, A.; Coniglio, A.; Antoni, V.; Cordier, J.J.; Hemsworth, R.; Jones, T.; Di Pietro, E.; Mondino, P.L.

    2004-01-01

    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 - 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)

  2. 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.

  3. Neutral-beam-heating applications and development

    International Nuclear Information System (INIS)

    Menon, M.M.

    1981-01-01

    The technique of heating the plasma in magnetically confined fusion devices by the injection of intense beams of neutral atoms is described. The basic principles governing the physics of neutral beam heating and considerations involved in determining the injection energy, power, and pulse length required for a fusion reactor are discussed. The pertinent experimental results from various fusion devices are surveyed to illustrate the efficacy of this technique. The second part of the paper is devoted to the technology of producing the neutral beams. A state-of-the-art account o the development of neutral injectors is presented, and the prospects for utilizing neutral injection to heat the plasma in a fusion reactor are examined

  4. Optimal neutral beam heating scenario for FED

    International Nuclear Information System (INIS)

    Hively, L.M.; Houlberg, W.A.; Attenberger, S.E.

    1981-01-01

    Optimal neutral beam heating scenarios are determined for FED based on a 1/one-half/-D transport analysis. Tradeoffs are examined between neutral beam energy, power, and species mix for positive ion systems. A ramped density startup is found to provide the most economical heating. The resulting plasma power requirements are reduced by 10-30% from a constant density startup. For beam energies between 100 and 200 keV, the power needed to heat the plasma does not decrease significantly as beam energy is increased. This is due to reduced ion heating, more power in the fractional energy components, and rising power supply requirements as beam energy increases

  5. Integrating supervision, control and data acquisition—The ITER Neutral Beam Test Facility experience

    Energy Technology Data Exchange (ETDEWEB)

    Luchetta, A., E-mail: adriano.luchetta@igi.cnr.it; Manduchi, G.; Taliercio, C.; Breda, M.; Capobianco, R.; Molon, F.; Moressa, M.; Simionato, P.; Zampiva, E.

    2016-11-15

    Highlights: • The paper describes the experience gained in the integration of different systems for the control and data acquisition system of the ITER Neutral Beam Test Facility. • It describes the way the different frameworks have been integrated. • It reports some lessons learnt during system integration. • It reports some authors’ considerations about the development the ITER CODAC. - Abstract: The ITER Neutral Beam (NBI) Test Facility, under construction in Padova, Italy consists in the ITER full scale ion source for the heating neutral beam injector, referred to as SPIDER, and the full size prototype injector, referred to as MITICA. The Control and Data Acquisition System (CODAS) for SPIDER has been developed and is going to be in operation in 2016. The system is composed of four main components: Supervision, Slow Control, Fast Control and Data Acquisition. These components interact with each other to carry out the system operation and, since they represent a common pattern in fusion experiments, software frameworks have been used for each (set of) component. In order to reuse as far as possible the architecture developed for SPIDER, it is important to clearly define the boundaries and the interfaces among the system components so that the implementation of any component can be replaced without affecting the overall architecture. This work reports the experience gained in the development of SPIDER components, highlighting the importance in the definition of generic interfaces among component, showing how the specific solutions have been adapted to such interfaces and suggesting possible approaches for the development of other ITER subsystems.

  6. Integrating supervision, control and data acquisition—The ITER Neutral Beam Test Facility experience

    International Nuclear Information System (INIS)

    Luchetta, A.; Manduchi, G.; Taliercio, C.; Breda, M.; Capobianco, R.; Molon, F.; Moressa, M.; Simionato, P.; Zampiva, E.

    2016-01-01

    Highlights: • The paper describes the experience gained in the integration of different systems for the control and data acquisition system of the ITER Neutral Beam Test Facility. • It describes the way the different frameworks have been integrated. • It reports some lessons learnt during system integration. • It reports some authors’ considerations about the development the ITER CODAC. - Abstract: The ITER Neutral Beam (NBI) Test Facility, under construction in Padova, Italy consists in the ITER full scale ion source for the heating neutral beam injector, referred to as SPIDER, and the full size prototype injector, referred to as MITICA. The Control and Data Acquisition System (CODAS) for SPIDER has been developed and is going to be in operation in 2016. The system is composed of four main components: Supervision, Slow Control, Fast Control and Data Acquisition. These components interact with each other to carry out the system operation and, since they represent a common pattern in fusion experiments, software frameworks have been used for each (set of) component. In order to reuse as far as possible the architecture developed for SPIDER, it is important to clearly define the boundaries and the interfaces among the system components so that the implementation of any component can be replaced without affecting the overall architecture. This work reports the experience gained in the development of SPIDER components, highlighting the importance in the definition of generic interfaces among component, showing how the specific solutions have been adapted to such interfaces and suggesting possible approaches for the development of other ITER subsystems.

  7. Design of neutral beam injection power supplies for ITER

    International Nuclear Information System (INIS)

    Watanabe, Kazuhiro; Okumura, Yoshikazu; Ono, Youichi; Tanaka, Masanobu

    2000-03-01

    Design study on a power supply system for the ITER neutral beam injector(NBI) has been performed. Circuits of converter/inverter system and other components of the acceleration power supply whose capacity is 1 MV, 45 A have been designed in detail. Performance of the negative ion production power supplies such as an arc and an extraction power supplies was investigated using the EMTDC code. It was confirmed that ripples of 0.34%p-p for the extraction power supply and 1.7%p-p for the arc power supply are small enough. It was also confirmed that an energy input to a negative ion generator from the arc power supply at an arcing can be suppressed smaller than 8 J. The extraction power supply was designed to suppress the energy input lower than 13 J at the breakdown in the extractor. These performances satisfy the required specification of the power supply system. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Sasao, M.; Tanaka, N.; Terai, K.; Kaneko, O. [Graduate school of Engineering, Tohoku University, Sendai 980-8579 (Japan); Kisaki, M.; Kobuchi, T.; Tsumori, K.; Okamoto, A.; Kitajima, S. [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Shinto, K. [IFMIF R and D Center, Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan); Wada, M. [Graduate School of Engineering, Doshisha University, Kyotanabe, Kyoto 610-0321 (Japan)

    2012-02-15

    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{sup +} ion source (2 A, the beam radius of 11.3 mm, the beam energy less than 20 keV). Present strong-focusing He{sup +} 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.

  9. Control systems for ITER diagnostics, heating and current drive

    Energy Technology Data Exchange (ETDEWEB)

    Simrock, Stefan, E-mail: stefan.simrock@iter.org

    2016-11-15

    The ITER Diagnostic, Heating and Current Drive systems might appear, on the face of it, to have very different control requirements. There are approximately 45 diagnostic systems, including magnetic sensors for plasma position and shape determination, imaging systems in the IR and visible, Thompson scattering for electron temperature and density, neutron detectors and collective scattering for alpha particle density and energy distribution. The H&CD systems encompass Electron Cyclotron Heating, using 24 1MW, 170 GHz gyrotrons and 5 steerable launchers to deliver 20 MW to the plasma, Ion Cyclotron Heating, using 8 3MW, 40–55 MHz sources and two multi-element launchers to deliver 20 MW to the plasma, and 2 Negative Ion Neutral Beam Injectors, each of which can deliver up to 16.5 MW of 1 MeV beams to the plasma. Although there are substantial differences in the needs for protection, when handling multi-MW heating systems, and in data throughput for many diagnostics, the formal processes needed to translate system requirements into Instrumentation and Control are identical. Due to the distributed procurement of ITER sub-systems and the need to integrate as painlessly as possible to CODAC, the formal processes, together with a substantial degree of standardization, are even more than usually essential. Starting from the technical, safety and protection, integration and operation requirements, a loop of functional analysis and signal listing is used to generate the controller configuration and the conceptual architecture. These elements in their turn lead to the physical and software design. The paper will describe the formal processes of control system design and the methods used by the ITER project to achieve the standardization of systems engineering practices. These have been applied to several use-cases covering all operation relevant phases such as plasma operation, maintenance, testing and conditioning. There are a number of running contracts that are developing

  10. Guidelines for remote handling maintenance of ITER neutral beam line components: Proposal of an alternate supporting system

    International Nuclear Information System (INIS)

    Cordier, J.J.; Bayetti, P.; Hemsworth, R.; David, O.; Friconneau, J.P.

    2007-01-01

    Remote handling (R/H) maintenance of ITER components is one of the main challenges of the ITER project. This type of maintenance shall be operational for the assembly and nuclear phase of exploitation of ITER. It must be considered at a very early stage since it significantly impacts on the components design, interfaces management, assembly, maintenance and integration aspects. A large part of the R/H equipment will be procured by the EU Participating Team, including the whole Neutral Beam R/H Equipment. The Neutral Beam Heating and Current Drive system (NB and CD) design is being revisited by the ITER project. A vertical maintenance scheme is presently considered which may significantly impact on the reference design and associated components and lead to a new design of the NB and CD vacuum tank. In addition, NB line components remote handling solutions are being studied. The neutral beam test facility ITER to be built in Europe in the near future is also based on the vertical NB maintenance scheme of beam line components. New design guidelines compliant for both the ITER NB and CD system and the NB test facility proposed by the CEA association are described in the paper

  11. Solutions to mitigate heat loads due to electrons on sensitive components of ITER HNB beamlines

    Energy Technology Data Exchange (ETDEWEB)

    Sartori, Emanuele, E-mail: emanuele.sartori@gmail.com [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), C.so Stati Uniti 4, 35127 Padova (Italy); Veltri, Pierluigi; Dalla Palma, Mauro; Agostinetti, Piero [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), C.so Stati Uniti 4, 35127 Padova (Italy); Hemsworth, Ronald; Singh, Mahendrajit [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Serianni, Gianluigi [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), C.so Stati Uniti 4, 35127 Padova (Italy)

    2016-11-01

    Highlights: • Energetic electrons leaking out of the ITER HNB accelerator are simulated. • Electrons generated along the ITER HNB beamline are simulated. • Heat loads and heat load maps on cryopumps are calculated for ITER HNB and test facility. • Protection solutions that will be installed are presented and their effect discussed. - Abstract: The operation of neutral beam injectors for plasma heating and current drive in a fusion device provides challenges in the thermal management of beamline components. Sensitive components such as the cryogenic pumps at beamline periphery shall be protected from the heat flux due to stray electrons. These are emitted by the negative ion accelerator or generated along the beamline by interaction of fast electrons, ions or atoms with background gas and surfaces. In this article the case of the ITER Heating Neutral Beam (HNB) and its test facility MITICA is discussed, for which the beam parameters and the required pulse length of one hour is a major leap forward with respect to the present experience with neutral beam systems. The engineering solutions adopted for effective cryopump protection against the heat load from electrons are described. The use of three-dimensional numerical simulations of particle trajectories in the complex geometry of the beamline was needed for the quantitative estimations of the heat loads. The presented solutions were optimized to minimize the impact on gas pumping and on the functionality of other components.

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

    International Nuclear Information System (INIS)

    Cordier, J.J.; Hemsworth, R.; Chantant, M.; Gravil, B.; Henry, D.; Sabathier, F.; Doceul, L.; Thomas, E.; Houtte, D. van; Zaccaria, P.; Antoni, V.; Bello, S. Dal; Marcuzzi, D.; Antipenkov, A.; Day, C.; Dremel, M.; Mondino, P.L.

    2005-01-01

    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

  13. A suite of diagnostics to validate and optimize the prototype ITER neutral beam injector

    Science.gov (United States)

    Pasqualotto, R.; Agostini, M.; Barbisan, M.; Brombin, M.; Cavazzana, R.; Croci, G.; Dalla Palma, M.; Delogu, R. S.; De Muri, M.; Muraro, A.; Peruzzo, S.; Pimazzoni, A.; Pomaro, N.; Rebai, M.; Rizzolo, A.; Sartori, E.; Serianni, G.; Spagnolo, S.; Spolaore, M.; Tardocchi, M.; Zaniol, B.; Zaupa, M.

    2017-10-01

    The ITER project requires additional heating provided by two neutral beam injectors using 40 A negative deuterium ions accelerated at 1 MV. As the beam requirements have never been experimentally met, a test facility is under construction at Consorzio RFX, which hosts two experiments: SPIDER, full-size 100 kV ion source prototype, and MITICA, 1 MeV full-size ITER injector prototype. Since diagnostics in ITER injectors will be mainly limited to thermocouples, due to neutron and gamma radiation and to limited access, it is crucial to thoroughly investigate and characterize in more accessible experiments the key parameters of source plasma and beam, using several complementary diagnostics assisted by modelling. In SPIDER and MITICA the ion source parameters will be measured by optical emission spectroscopy, electrostatic probes, cavity ring down spectroscopy for H^- density and laser absorption spectroscopy for cesium density. Measurements over multiple lines-of-sight will provide the spatial distribution of the parameters over the source extension. The beam profile uniformity and its divergence are studied with beam emission spectroscopy, complemented by visible tomography and neutron imaging, which are novel techniques, while an instrumented calorimeter based on custom unidirectional carbon fiber composite tiles observed by infrared cameras will measure the beam footprint on short pulses with the highest spatial resolution. All heated components will be monitored with thermocouples: as these will likely be the only measurements available in ITER injectors, their capabilities will be investigated by comparison with other techniques. SPIDER and MITICA diagnostics are described in the present paper with a focus on their rationale, key solutions and most original and effective implementations.

  14. Behaviour of the ASDEX pressure gauge at high neutral gas pressure and applications for ITER

    International Nuclear Information System (INIS)

    Scarabosio, A.; Haas, G.

    2008-01-01

    The ASDEX Pressure Gauge is, at present, the main candidate for in-vessel neutral pressure measurement in ITER. Although the APG output is found to saturate at around 15 Pa, below the ITER requirement of 20 Pa. We show, here, that with small modifications of the gauge geometry and potentials settings we can achieve satisfactory behaviour up to 30 Pa at 6 T

  15. Control and data acquisition of the ITER full-scale ion source for the neutral beam test facility

    International Nuclear Information System (INIS)

    Luchetta, Adriano; Manduchi, Gabriele; Taliercio, Cesare; Paolucci, Francesco; Sartori, Filippo; Svensson, Lennart; Labate, Carmelo Vincenzo; Breda, Mauro; Capobianco, Roberto; Molon, Federico; Moressa, Modesto; Simionato, Paola; Zampiva, Enrico; Barbato, Paolo; Polato, Sandro

    2015-01-01

    Highlights: • This paper describes the requirements and architecture of the control and data acquisition system of the ITER full-ion source experiment in the neutral beam test facility. • The system architecture integrates various popular software frameworks. • Slow control is based on the EPICS (Experimental Physics and Industrial Control System) framework. • Fast control is based on the MARTe (Multi-threaded Application Real-Time executor) framework. • Data acquisition is based on the MDSplus framework. - Abstract: The neutral beam test facility, which is under construction in Padova, Italy, is developing the ITER full-scale ion source for the ITER heating neutral beam injectors, referred to as the SPIDER experiment, and the full-size prototype injector, referred to as MITICA. The SPIDER control and data acquisition system (CODAS) has been developed and its construction will start in 2014. Slow control and data acquisition will be based on the ITER CODAC core system software suite that has been designed to facilitate the integration of ITER plant systems with CODAC. Fast control and data acquisition will use solutions specific to the test facility, as the corresponding concepts are not ready-to-use in the ITER design. The ITER hardware catalog for fast control has been taken into consideration. The software development will be based on the integration of MDSplus and MARTe, two framework software packages that are well known in the fusion community, targeting data organization and fast real-time control, respectively. The paper revises the system requirements and the system design and shows the results already achieved in terms of system integration. In addition, the paper will report the experience in the usage of different cooperating software frameworks and in the integration of industrial procured plant systems.

  16. Control and data acquisition of the ITER full-scale ion source for the neutral beam test facility

    Energy Technology Data Exchange (ETDEWEB)

    Luchetta, Adriano, E-mail: adriano.luchetta@igi.cnr.it [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), Padova (Italy); Manduchi, Gabriele; Taliercio, Cesare [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), Padova (Italy); Paolucci, Francesco; Sartori, Filippo [Fusion for Energy, Barcelona (Spain); Svensson, Lennart [ITER Organization, Route de Vinon-sur-Verdon, CS 90046 St. Paul Lez Durance (France); Labate, Carmelo Vincenzo [Association ENEA-CREATE, Department of Engineering, University of Naples “Parthenope” (Italy); Breda, Mauro; Capobianco, Roberto; Molon, Federico; Moressa, Modesto; Simionato, Paola; Zampiva, Enrico; Barbato, Paolo; Polato, Sandro [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), Padova (Italy)

    2015-10-15

    Highlights: • This paper describes the requirements and architecture of the control and data acquisition system of the ITER full-ion source experiment in the neutral beam test facility. • The system architecture integrates various popular software frameworks. • Slow control is based on the EPICS (Experimental Physics and Industrial Control System) framework. • Fast control is based on the MARTe (Multi-threaded Application Real-Time executor) framework. • Data acquisition is based on the MDSplus framework. - Abstract: The neutral beam test facility, which is under construction in Padova, Italy, is developing the ITER full-scale ion source for the ITER heating neutral beam injectors, referred to as the SPIDER experiment, and the full-size prototype injector, referred to as MITICA. The SPIDER control and data acquisition system (CODAS) has been developed and its construction will start in 2014. Slow control and data acquisition will be based on the ITER CODAC core system software suite that has been designed to facilitate the integration of ITER plant systems with CODAC. Fast control and data acquisition will use solutions specific to the test facility, as the corresponding concepts are not ready-to-use in the ITER design. The ITER hardware catalog for fast control has been taken into consideration. The software development will be based on the integration of MDSplus and MARTe, two framework software packages that are well known in the fusion community, targeting data organization and fast real-time control, respectively. The paper revises the system requirements and the system design and shows the results already achieved in terms of system integration. In addition, the paper will report the experience in the usage of different cooperating software frameworks and in the integration of industrial procured plant systems.

  17. 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.

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

    International Nuclear Information System (INIS)

    Veltri, Pierluigi; Agostinetti, Piero; Marcuzzi, Diego; Sartori, Emanuele; Serianni, Gianluigi

    2016-01-01

    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

  19. Simulation of an ITER-like dissipative divertor plasma with a combined edge plasma Navier-Stokes neutral model

    International Nuclear Information System (INIS)

    Knoll, D.A.; McHugh, P.R.; Krasheninnikov, S.I.; Sigmar, D.J.

    1996-01-01

    A combined edge plasma/Navier-Stokes neutral transport model is used to simulate dissipative divertor plasmas in the collisional limit for neutrals on a simplified two-dimensional slab geometry with ITER-like plasma conditions and scale lengths. The neutral model contains three momentum equations which are coupled to the plasma through ionization, recombination, and ion-neutral elastic collisions. The neutral transport coefficients are evaluated including both ion-neutral and neutral-neutral collisions. (orig.)

  20. Feasibility of a fast optical pressure interlock for the ITER neutral beam injectors

    International Nuclear Information System (INIS)

    Ash, Andrew; Surrey, Elizabeth

    2009-01-01

    The feasibility of using Balmer-α emission for a high-speed pressure diagnostic and beam interlock for the ITER neutral beam heating system is investigated. An interlock is needed to prevent excessive re-ionisation of the neutral beam when rapid excursions of pressure occur in either the electrostatic residual ion dump (ERID), or the neutral beam duct (NBD). The re-ionised fraction of the beam, will be deflected by stray tokamak fields, potentially causing excessive thermal loads on beam line components. Experience from JET indicates that a response time of order 100 μs is required in order to prevent fast pressure excursions. Fast penning gauges have a time response of around 30-50 ms, however, a faster response (around 1 μs) is possible by monitoring the H α (656.3 nm)/D α (656.1 nm) emission from collisional excitation of the background gas and neutral beam. Published total cross-sections are used to calculate a signal of 3.5x10 13 -3.0x10 17 photons s -1 m -2 sr -1 for normal conditions. This signal must be distinguished from the background light of the tokamak plasma (line emission and bremsstrahlung). The beam emission is Doppler shifted by up to 21 nm (D operation) and up to 27 nm (H operation) depending on angle of observation and this can be used to help distinguish against background line emission. The distribution of background light along the beam line is calculated with a two-dimensional radiosity code, solving the equilibrium energy balance within the beam line enclosure. The Balmer-α signal and background signal due to bremsstrahlung are compared for a 500-MW reference plasma.

  1. High heat flux (HHF) elements for negative ion systems on ITER

    International Nuclear Information System (INIS)

    Milnes, J.; Chuilon, B.; Xue, Y.; Martin, D.; Waldon, C.

    2007-01-01

    Negative Ion Neutral Beam systems on ITER will require actively cooled scrapers and dumps to process and shape the beam before injection into the tokamak. The scale of the systems is much larger than any presently operating, bringing challenges for designers in terms of available sub cooling, total pressure drop, deflection and mandatory remote maintenance. High heat fluxes (∼15-20 MW/m 2 ), pulse lengths in excess of 3000 s and high number of cycles pose new challenges in terms of stress and fatigue life. The designs outlined in the Design Description Document for the ITER Neutral Beam System [N53 DDD 29 01-07-03 R 0.1. ITER Design Description Document, DDD 5.3, Neutral Beam H and CD system (including Appendices).], based on swirl tubes, have been reviewed as part of the design process and recommendations made. Additionally, alternative designs have been proposed based on the Hypervapotron high heat flux elements with modified geometry and drawing upon a vast background knowledge of large scale equipment procurement and integration. A full thermo-mechanical analysis of all HHF components has also been undertaken based on ITER design criteria and the limited material data available. The advantages and disadvantages of all designs are presented and recommendations for improvements discussed

  2. Seventh meeting of the ITER physics expert group on energetic particles, heating and steady state operations

    International Nuclear Information System (INIS)

    Gormezano, C.

    1999-01-01

    The seventh meeting of the ITER Physics Group on energetic particles, heating and steady state operation was held at CEN/Cadarache from 14 to 18 September 1999. This was the first meeting following the redefinition of the Expert Group structure and it was also the first meeting without participation of US physicists. The main topics covered were: 1. Energetic Particles, 2. Ion Cyclotron Resonance Heating, 3. Lower Hybrid Current Drive, 4. Electron Cyclotron Resonance Heating and Current Drive, 5. Neutral Beam Injection, 6. Steady-State Aspects

  3. ITER radio frequency systems

    International Nuclear Information System (INIS)

    Bosia, G.

    1998-01-01

    Neutral Beam Injection and RF heating are two of the methods for heating and current drive in ITER. The three ITER RF systems, which have been developed during the EDA, offer several complementary services and are able to fulfil ITER operational requirements

  4. 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

    Auxiliary heating such as neutral beam injection (NBI) and ion cyclotron resonance heating (ICRH) will accelerate ions in ITER up to energies in the MeV range, i.e. energies which are also typical for alpha particles. Fast ions of any of these populations will elevate the collective Thomson...... functions of fast ions generated by NBI and ICRH are calculated for a steady-state ITER burning plasma equilibrium with the ASCOT and PION codes, respectively. The parameters for the auxiliary heating systems correspond to the design currently foreseen for ITER. The geometry of the CTS system for ITER...... is chosen such that near perpendicular and near parallel velocity components are resolved. In the investigated ICRH scenario, waves at 50MHz resonate with tritium at the second harmonic off-axis on the low field side. Effects of a minority heating scheme with He-3 are also considered. CTS scattering...

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

    International Nuclear Information System (INIS)

    Heinemann, B.; Falter, H.; Fantz, U.; Franzen, P.; Froeschle, M.; Gutser, R.; Kraus, W.; Nocentini, R.; Riedl, R.; Speth, E.; Staebler, A.; Wuenderlich, D.; Agostinetti, P.; Jiang, T.

    2009-01-01

    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].

  6. Progress in the design of the ITER Neutral Beam cell Remote Handling System

    Energy Technology Data Exchange (ETDEWEB)

    Shuff, R., E-mail: robin.shuff@f4e.europa.eu [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Van Uffelen, M.; Damiani, C. [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Tesini, A.; Choi, C.-H. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-lez-Durance (France); Meek, R. [Oxford Technologies Limited, 7 Nuffield Way, Abingdon OX14 1RL (United Kingdom)

    2014-10-15

    The ITER Neutral Beam cell will include a suite of Remote Handling equipment for maintenance tasks. This paper summarises the current status and recent developments in the design of the ITER Neutral Beam Remote Handling System. Its concept design was successfully completed in July 2012 by CCFE in the frame of a grant agreement with F4E, in collaboration with the ITER Organisation, including major systems like monorail crane, Beam Line Transporter, beam source equipment, upper port and neutron shield equipment and associated tooling. Research and development activities are now underway on the monorail crane radiation hardened on-board control system and first of a kind remote pipe and lip seal maintenance tooling for the beam line vessel, reported in this paper.

  7. Progress in the design of the ITER Neutral Beam cell Remote Handling System

    International Nuclear Information System (INIS)

    Shuff, R.; Van Uffelen, M.; Damiani, C.; Tesini, A.; Choi, C.-H.; Meek, R.

    2014-01-01

    The ITER Neutral Beam cell will include a suite of Remote Handling equipment for maintenance tasks. This paper summarises the current status and recent developments in the design of the ITER Neutral Beam Remote Handling System. Its concept design was successfully completed in July 2012 by CCFE in the frame of a grant agreement with F4E, in collaboration with the ITER Organisation, including major systems like monorail crane, Beam Line Transporter, beam source equipment, upper port and neutron shield equipment and associated tooling. Research and development activities are now underway on the monorail crane radiation hardened on-board control system and first of a kind remote pipe and lip seal maintenance tooling for the beam line vessel, reported in this paper

  8. Multi-scale transport in the DIII-D ITER baseline scenario with direct electron heating and projection to ITER

    Science.gov (United States)

    Grierson, B. A.; Staebler, G. M.; Solomon, W. M.; McKee, G. R.; Holland, C.; Austin, M.; Marinoni, A.; Schmitz, L.; Pinsker, R. I.; DIII-D Team

    2018-02-01

    Multi-scale fluctuations measured by turbulence diagnostics spanning long and short wavelength spatial scales impact energy confinement and the scale-lengths of plasma kinetic profiles in the DIII-D ITER baseline scenario with direct electron heating. Contrasting discharge phases with ECH + neutral beam injection (NBI) and NBI only at similar rotation reveal higher energy confinement and lower fluctuations when only NBI heating is used. Modeling of the core transport with TGYRO using the TGLF turbulent transport model and NEO neoclassical transport reproduces the experimental profile changes upon application of direct electron heating and indicates that multi-scale transport mechanisms are responsible for changes in the temperature and density profiles. Intermediate and high-k fluctuations appear responsible for the enhanced electron thermal flux, and intermediate-k electron modes produce an inward particle pinch that increases the inverse density scale length. Projection to ITER is performed with TGLF and indicates a density profile that has a finite scale length due to intermediate-k electron modes at low collisionality and increases the fusion gain. For a range of E × B shear, the dominant mechanism that increases fusion performance is suppression of outward low-k particle flux and increased density peaking.

  9. Design Features of the Neutral Particle Diagnostic System for the ITER Tokamak

    Science.gov (United States)

    Petrov, S. Ya.; Afanasyev, V. I.; Melnik, A. D.; Mironov, M. I.; Navolotsky, A. S.; Nesenevich, V. G.; Petrov, M. P.; Chernyshev, F. V.; Kedrov, I. V.; Kuzmin, E. G.; Lyublin, B. V.; Kozlovski, S. S.; Mokeev, A. N.

    2017-12-01

    The control of the deuterium-tritium (DT) fuel isotopic ratio has to ensure the best performance of the ITER thermonuclear fusion reactor. The diagnostic system described in this paper allows the measurement of this ratio analyzing the hydrogen isotope fluxes (performing neutral particle analysis (NPA)). The development and supply of the NPA diagnostics for ITER was delegated to the Russian Federation. The diagnostics is being developed at the Ioffe Institute. The system consists of two analyzers, viz., LENPA (Low Energy Neutral Particle Analyzer) with 10-200 keV energy range and HENPA (High Energy Neutral Particle Analyzer) with 0.1-4.0MeV energy range. Simultaneous operation of both analyzers in different energy ranges enables researchers to measure the DT fuel ratio both in the central burning plasma (thermonuclear burn zone) and at the edge as well. When developing the diagnostic complex, it was necessary to account for the impact of several factors: high levels of neutron and gamma radiation, the direct vacuum connection to the ITER vessel, implying high tritium containment, strict requirements on reliability of all units and mechanisms, and the limited space available for accommodation of the diagnostic hardware at the ITER tokamak. The paper describes the design of the diagnostic complex and the engineering solutions that make it possible to conduct measurements under tokamak reactor conditions. The proposed engineering solutions provide a safe—with respect to thermal and mechanical loads—common vacuum channel for hydrogen isotope atoms to pass to the analyzers; ensure efficient shielding of the analyzers from the ITER stray magnetic field (up to 1 kG); provide the remote control of the NPA diagnostic complex, in particular, connection/disconnection of the NPA vacuum beamline from the ITER vessel; meet the ITER radiation safety requirements; and ensure measurements of the fuel isotopic ratio under high levels of neutron and gamma radiation.

  10. 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.

  11. Active ion temperature measurement with heating neutral beam

    International Nuclear Information System (INIS)

    Miura, Yukitoshi; Matsuda, Toshiaki; Yamamoto, Shin

    1987-03-01

    When the heating neutral-beam (hydrogen beam) is injected into a deuterium plasma, the density of neutral particles is increased locally. By using this increased neutral particles, the local ion temperature is measured by the active charge-exchange method. The analyzer is the E//B type mass-separated neutral particle energy analyzer and the measured position is about one third outside of the plasma radius. The deuterium energy spectrum is Maxwellian, and the temperature is increased from 350 eV to 900 eV during heating. Since the local hydrogen to deuterium density concentration and the density of the heating neutral-beam as well as the ion temperature can be obtained good S/N ratio, the usefulness of this method during neutral-beam heating is confirmed by this experiment. (author)

  12. Protections Against Grid Breakdowns in the ITER Neutral Beam Injector Power Supplies

    International Nuclear Information System (INIS)

    Bigi, M.; Toigo, V.; Zanotto, L.

    2006-01-01

    The ITER Neutral Beam Injector (NBI) is designed to deliver 16.5 MW of additional heating power to the plasma, accelerating negative ions up to -1 MV with a current up to 40 A. Two main power supplies are foreseen to feed the system: the Acceleration Grid Power Supply (AGPS), which provides power to the acceleration grids, and the Ion Source Power Supply (ISPS), devoted to supplying the ion source components. For the accelerator, two different concepts are under investigation: the MAMuG (Multiple Aperture, Multiple Gap) and the SINGAP (SINgle Aperture). During operation of the NBI, the breakdown of the acceleration grids will occur regularly; as a consequence the AGPS is expected to experience frequent load short-circuits during a pulse. For each grid breakdown, energy and current peaks are delivered from the power supply systems that could damage the grids, if not limited. In previous NBI, rated for a lower accelerating voltage, the protection system in case of grid breakdowns was based on dc circuit breakers able to quickly disconnect the power supply from the grids. In the ITER case, a similar solution is not feasible, as the voltage level is too high for present dc breaker technology. Therefore, the protection strategy has to rely on fast switch-off of the power supplies, on the optimisation of the filter elements and core snubbers placed downstream the AGPS and on the introduction of additional passive elements. However, achieving a satisfactory protection against grid breakdowns is a challenging task, as the optimisation of each single provision can result in drawbacks for other aspects of the design; for instance, the optimisation of the filter elements, obtained by reducing the filter capacitance, produces an increase of the output voltage ripple. Therefore, the design of the protections must be carried out considering all the relevant aspects of the specifications, also those that are not strictly related to the limitations of the current peaks and energy

  13. Surface heat loads on the ITER divertor vertical targets

    Czech Academy of Sciences Publication Activity Database

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

    2017-01-01

    Roč. 57, č. 4 (2017), č. článku 046025. ISSN 0029-5515 Institutional support: RVO:61389021 Keywords : ITER * divertor * ELM heat load * inter-ELM heat load * tungsten Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 3.307, year: 2016 http://iopscience.iop.org/article/10.1088/1741-4326/aa5e2a

  14. A kind of iteration algorithm for fast wave heating

    International Nuclear Information System (INIS)

    Zhu Xueguang; Kuang Guangli; Zhao Yanping; Li Youyi; Xie Jikang

    1998-03-01

    The standard normal distribution for particles in Tokamak geometry is usually assumed in fast wave heating. In fact, due to the quasi-linear diffusion effect, the parallel and vertical temperature of resonant particles is not equal, so, this will bring some error. For this case, the Fokker-Planck equation is introduced, and iteration algorithm is adopted to solve the problem well

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

    Energy Technology Data Exchange (ETDEWEB)

    Dalla Palma, M., E-mail: mauro.dallapalma@igi.cnr.it; Pasqualotto, R.; Spagnolo, S.; Spolaore, M. [Consorzio RFX, Padova 35127 (Italy); Sartori, E. [Consorzio RFX, Padova 35127 (Italy); Università degli Studi di Padova, Padova 35122 (Italy); Veltri, P. [Consorzio RFX, Padova 35127 (Italy); INFN-LNL, Legnaro (PD) 35020 (Italy)

    2016-11-15

    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.

  16. Design of the ITER (International Thermonuclear Experimental Reactor) neutral beam system beamline, United States concept

    International Nuclear Information System (INIS)

    Purgalis, P.; Anderson, O.A.; Cooper, W.S.; DeVries, G.E.; Lietzke, A.F.; Kunkel, W.B.; Kwan, J.W.; Matuk, C.A.; Nakai, T.; Stearns, J.W.; Soroka, L.; Wells, R.P.; Lindquist, W.B.; Neef, W.S.; Reginato, L.L.; Sedgley, D.W.; Brook, J.W.; Luzzi, T.E.; Myers, T.J.

    1989-01-01

    Design of a neutral beamline for ITER (International Thermonuclear Experimental Reactor) is described. The design incorporates a barium surface conversion D - 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 watercooled 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 that can be removed for remote maintenance. The neutral beam system delivers 75 MW of D degree into three ports with a total of nine modules arranged in stacks of three modules per port. To increase reliability each module is designed to deliver up to 10 MW at 1.3 MeV; this allows eight modules operating at partial capacity to deliver the required power in the event one module is removed from service. Radiation protection is provided by shielding and by locating critical components in the source and accelerator 35 m from the port into the torus. Neutron shielding in the drift duct provides the added feature of limiting conductance and thus reducing gas flow to and from the torus. Alternative component choices are also discussed for the evolving design. 8 refs., 4 figs., 1 tab

  17. An Indian test facility to characterise diagnostic neutral beam for ITER

    International Nuclear Information System (INIS)

    Singh, M.J.; Bandyopadhyay, M.; Rotti, C.; Singh, N.P.; Shah, Sejal; Bansal, G.; Gahlaut, A.; Soni, J.; Lakdawala, H.; Waghela, Harshad; Ahmed, I.; Roopesh, G.; Baruah, U.K.; Chakraborty, A.K.

    2011-01-01

    The diagnostic neutral beam (DNB) line shall be used to diagnose the He ash content in the D-T phase of the ITER machine using the charge exchange recombination spectroscopy (CXRS). Implementation of a successful DNB at ITER requires several challenges related to the production, neutralization and transport of the neutral beam over path lengths of 20.665 m, to be overcome. The delivery is aided if the above effects are tested prior to onsite commissioning. As DNB is a procurement package for INDIA, an ITER approved Indian test facility, INTF, is under construction at Institute for Plasma Research (IPR), India and is envisaged to be operational in 2015. The timeline for this facility is synchronized with the RADI, ELISE (IPP, Garching), SPIDER (RFX, Padova) in a manner that best utilization of configurational inputs available from them are incorporated in the design. This paper describes the facility in detail and discusses the experiments planned to optimise the beam transmission and testing of the beam line components using various diagnostics.

  18. The ion source development for neutral injection heating at JAERI

    International Nuclear Information System (INIS)

    Shirakata, H.; Itoh, T.; Kondoh, U.; Matsuda, S.; Ohara, Y.; Ohga, T.; Shibata, T.; Sugawara, T.; Tanaka, S.

    1976-01-01

    The neutral beam research and development effort at JAERI has been mainly concentrated on design, construction and testing of ion sources needed for present and planned heating experiments. Fundamental characteristics of the ion sources developed are described

  19. Progress on radio frequency auxiliary heating system designs in ITER

    International Nuclear Information System (INIS)

    Makowski, M.; Bosia, G.; Elio, F.

    1996-09-01

    ITER will require over 100 MW of auxiliary power for heating, on- and off-axis current drive, accessing the H-mode, and plasma shut-down. The Electron Cyclotron Range of Frequencies (ECRF) and Ion Cyclotron Range of Frequencies (ICRF) are two forms of Radio Frequency (RF) auxiliary power being developed for these applications. Design concepts for both the ECRF and ICRF systems are presented, key features and critical design issues are discussed, and projected performances outlined

  20. 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.

  1. Plasma heating with multi-MeV neutral atom beams

    International Nuclear Information System (INIS)

    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 0 formed from D - . 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

  2. Plasma Heating and Current Drive by Neutral Beam and Alpha Particles

    Energy Technology Data Exchange (ETDEWEB)

    Kikuchi, M; Okumura, Y [Fusion Research and Development Directorate, Japan Atomic Energy Agency (Japan)

    2012-09-15

    The purpose of plasma heating is to raise the plasma temperature enough to produce a deuterium and tritium reaction (D + T {yields} {sup 4}He + n). The required plasma temperature T is in the range of 10-30 keV. Since the high temperature plasma is confined by a strong magnetic field, injection of energetic ions from outside to heat the plasma is difficult due to the Lorenz force. The most efficient way to heat the plasma by energetic particles is to inject high energy 'neutrals' which get ionized in the plasma. Neutral beam injection (NBI) with a beam energy much above the average kinetic energy of the plasma electrons or ions is used (beam energy typically {approx}40 keV - 1 MeV). This heating scheme is similar to warming up cold water by pouring in hot water. There are two types of neutral beam, called P-NBI and N-NBI (P- and N- means 'positive' and 'negative', respectively). P-NBI uses the acceleration of positively charged ions and their neutralization, while N-NBI uses the acceleration of negative ions (electrons attached to neutral atoms) and their neutralization. Details are given in NBI technology Section. The first demonstration of plasma heating by P-NBI was made in ORMAK and ATC in 1974, while that by N-NBI was made in JT-60U for the first time in 1996. ITER has also adopted the N-NBI system as the heating and current drive system with a beam energy of 1 MeV. Figure A typical bird's eye view of a tokamak with N-NBI and N-NBI (JT-60U) is shown. (author)

  3. Study of Heating and Fusion Power Production in ITER Discharges

    International Nuclear Information System (INIS)

    Rafiq, T.; Kritz, A. H.; Bateman, G.; Kessel, C.; McCune, D. C.; Budny, R. V.; Pankin, A. Y.

    2011-01-01

    ITER simulations, in which the temperatures, toroidal angular frequency and currents are evolved, are carried out using the PTRANSP code starting with initial profiles and boundary conditions obtained from TSC code studies. The dependence of heat deposition and current drive on ICRF frequency, number of poloidal modes, beam orientation, number of Monte Carlo particles and ECRH launch angles is studied in order to examine various possibilities and contingencies for ITER steady state and hybrid discharges. For the hybrid discharges, the fusion power production and fusion Q, computed using the Multi-Mode MMM v7.1 anomalous transport model, are compared with those predicted using the GLF23 model. The simulations of the hybrid scenario indicate that the fusion power production at 1000 sec will be approximately 500 MW corresponding to a fusion Q = 10.0. The discharge scenarios simulated aid in understanding the conditions for optimizing fusion power production and in examining measures of plasma performance.

  4. Upgrade of the TCV tokamak, first phase: Neutral beam heating system

    Energy Technology Data Exchange (ETDEWEB)

    Karpushov, Alexander N., E-mail: alexander.karpushov@epfl.ch [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Alberti, Stefano; Chavan, René [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Davydenko, Vladimir I. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Duval, Basil P. [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Ivanov, Alexander A. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Fasel, Damien; Fasoli, Ambrogio [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Gorbovsky, Aleksander I. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Goodman, Timothy [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Kolmogorov, Vyacheslav V. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Martin, Yves; Sauter, Olivier [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Sorokin, Aleksey V. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); and others

    2015-10-15

    Highlights: • Widening the parameter range of reactor relevant regimes on the TCV tokamak. • Installation of 1 MW, 30 keV neutral beam, direct ion heating, access to T{sub i}/T{sub e} ≥ 1. • ASTRA simulation of plasma response to NB and EC heating in different regimes. • Specific low divergency neutral beam injector with tunable beam power and energy. - Abstract: Experiments on TCV are designed to complement the work at large integrated tokamak facilities (such as JET) to provide a stepwise approach to extrapolation to ITER and DEMO in areas where medium-size tokamaks can often exploit their experimental capabilities and flexibility. Improving the understanding and control requirements of burning plasmas is a major scientific challenge, requiring access to plasma regimes and configurations with high normalized plasma pressure and a wide range of ion to electron temperature ratios, including T{sub e}/T{sub i} ∼ 1. These conditions will be explored by adding a 1 MW neutral heating beam to TCV's auxiliary for direct ion heating (2015) and increasing the ECH power injected in X-mode at the third harmonic (2 MW in 2015–2016). The manufacturing of the neutral beam injector was launched in 2014.

  5. Progress in control and data acquisition for the ITER neutral beam test facility

    Energy Technology Data Exchange (ETDEWEB)

    Luchetta, Adriano, E-mail: adriano.luchetta@igi.cnr.it [Consorzio RFX, Euratom-ENEA Association, Padova (Italy); Manduchi, Gabriele; Taliercio, Cesare; Soppelsa, Anton [Consorzio RFX, Euratom-ENEA Association, Padova (Italy); Paolucci, Francesco; Sartori, Filippo [Fusion for Energy, Barcelona (Spain); Barbato, Paolo; Capobianco, Roberto; Breda, Mauro; Molon, Federico; Moressa, Modesto; Polato, Sandro; Simionato, Paola; Zampiva, Enrico [Consorzio RFX, Euratom-ENEA Association, Padova (Italy)

    2013-10-15

    Highlights: ► An ion source experiment, referred to as SPIDER, is under construction in the ITER neutral beam test facility. ► The progress in designing and testing the SPIDER control and data acquisition system is reported. ► An original approach is proposed in using ITER CODAC and non-ITER CODAC technology. -- Abstract: SPIDER, the ion source test bed in the ITER neutral beam test facility, is under construction and its operation is expected to start in 2014. Control and data acquisition for SPIDER are undergoing final design. SPIDER CODAS, as the control and data acquisition system is referred to, is requested to manage 25 plant units, to acquire 1000 analogue signals with sampling rates ranging from a few S/s to 10 MS/s, to acquire images with up to 100 frames per second, to operate with long pulses lasting up to 1 h, and to sustain 200 MB/s data throughput into the data archive with an annual data storage amount of up to 50 TB. SPIDER CODAS software architecture integrates three open-source software frameworks each addressing specific system requirements. Slow control exploits the synergy among EPICS and Siemens S7 programmable controllers. Data handling is by MDSplus a data-centric framework that is geared towards the collection and organization of scientific data. Diagnostics based on imaging drive the design of data throughput and archive size. Fast control is implemented by using MARTe, a data-driven, object-oriented, real-time environment. The paper will describe in detail the progress of the system hardware and software architecture and will show how the software frameworks interact to provide the functions requested by SPIDER CODAS. The paper will focus on how the performance requirements can be met with the described SPIDER CODAS architecture, describing the progress achieved by carrying out prototyping activities.

  6. Progress in control and data acquisition for the ITER neutral beam test facility

    International Nuclear Information System (INIS)

    Luchetta, Adriano; Manduchi, Gabriele; Taliercio, Cesare; Soppelsa, Anton; Paolucci, Francesco; Sartori, Filippo; Barbato, Paolo; Capobianco, Roberto; Breda, Mauro; Molon, Federico; Moressa, Modesto; Polato, Sandro; Simionato, Paola; Zampiva, Enrico

    2013-01-01

    Highlights: ► An ion source experiment, referred to as SPIDER, is under construction in the ITER neutral beam test facility. ► The progress in designing and testing the SPIDER control and data acquisition system is reported. ► An original approach is proposed in using ITER CODAC and non-ITER CODAC technology. -- Abstract: SPIDER, the ion source test bed in the ITER neutral beam test facility, is under construction and its operation is expected to start in 2014. Control and data acquisition for SPIDER are undergoing final design. SPIDER CODAS, as the control and data acquisition system is referred to, is requested to manage 25 plant units, to acquire 1000 analogue signals with sampling rates ranging from a few S/s to 10 MS/s, to acquire images with up to 100 frames per second, to operate with long pulses lasting up to 1 h, and to sustain 200 MB/s data throughput into the data archive with an annual data storage amount of up to 50 TB. SPIDER CODAS software architecture integrates three open-source software frameworks each addressing specific system requirements. Slow control exploits the synergy among EPICS and Siemens S7 programmable controllers. Data handling is by MDSplus a data-centric framework that is geared towards the collection and organization of scientific data. Diagnostics based on imaging drive the design of data throughput and archive size. Fast control is implemented by using MARTe, a data-driven, object-oriented, real-time environment. The paper will describe in detail the progress of the system hardware and software architecture and will show how the software frameworks interact to provide the functions requested by SPIDER CODAS. The paper will focus on how the performance requirements can be met with the described SPIDER CODAS architecture, describing the progress achieved by carrying out prototyping activities

  7. Applying remote handling attributes to the ITER neutral beam cell monorail crane

    Energy Technology Data Exchange (ETDEWEB)

    Crofts, O., E-mail: Oliver.Crofts@CCFE.ac.uk [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Allan, P.; Raimbach, J. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Tesini, A.; Choi, C.-H. [ITER Organisation, CS90 046, 13067 St. Paul les Durance Cedex (France); Damiani, C.; Van Uffelen, M. [Fusion for Energy, C/Josep Pla 2, Torres Diagonal Litoral-B3, E-08019 Barcelona (Spain)

    2013-10-15

    The maintenance requirements for the equipment in the ITER neutral beam cell require 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.

  8. Applying remote handling attributes to the ITER neutral beam cell monorail crane

    International Nuclear Information System (INIS)

    Crofts, O.; Allan, P.; 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 require 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

  9. Heating of toroidal plasmas by neutral injection

    International Nuclear Information System (INIS)

    Stix, T.H.

    1971-08-01

    This paper presents a brief review of the physics of ion acceleration, charge exchange and ionization, trajectories for fast ions in toroidal magnetic fields, and fast-ion thermalization. The injection of fast atoms is found to be a highly competitive method both for heating present-day experimental toroidal plasmas and for bringing full-scale toroidal CTR plasmas to low-density ignition. 13 refs., 9 figs

  10. Tungsten recrystallization and cracking under ITER-relevant heat loads

    Energy Technology Data Exchange (ETDEWEB)

    Budaev, V.P., E-mail: Budaev@mail.ru [NRC «Kurchatov Institute», Akademika Kurchatova pl., Moscow (Russian Federation); Martynenko, Yu.V. [NRC «Kurchatov Institute», Akademika Kurchatova pl., Moscow (Russian Federation); National Research Nuclear University MEPhI, Kashirskoe sh. 31, Moscow (Russian Federation); Karpov, A.V.; Belova, N.E. [NRC «Kurchatov Institute», Akademika Kurchatova pl., Moscow (Russian Federation); Zhitlukhin, A.M. [SRC RF TRINITI, Moscow Region (Russian Federation); Klimov, N.S., E-mail: klimov@triniti.ru [SRC RF TRINITI, Moscow Region (Russian Federation); National Research Nuclear University MEPhI, Kashirskoe sh. 31, Moscow (Russian Federation); Podkovyrov, V.L.; Barsuk, V.A.; Putrik, A.B.; Yaroshevskaya, A.D. [SRC RF TRINITI, Moscow Region (Russian Federation); Giniyatulin, R.N. [Efremov Institute, St. Petersburg (Russian Federation); Safronov, V.M. [Institution «Project Center ITER», Moscow (Russian Federation); SRC RF TRINITI, Moscow Region (Russian Federation); Khimchenko, L.N. [Institution «Project Center ITER», Moscow (Russian Federation)

    2015-08-15

    The tungsten surface structure was analyzed after the test in the QSPA-T under heat loads relevant to those expected in the ITER during disruptions. Repeated pulses lead to the melting and the resolidification of the tungsten surface layer of ∼50 μm thickness. There is ∼50 μm thickness intermediate layer between the original structure and the resolidified layer. The intermediate layer is recrystallized and has a random grains’ orientation whereas the resolidified layer and basic structure have texture with preferable orientation 〈1 0 0〉 normal to the surface. The cracks which were normal to the surface were observed in the resolidified layer as well as the cracks which were parallel to the surface at the depth up to 300 μm. Such cracks can result in the brittle destruction which is a hazard for the full tungsten divertor of the ITER. The theoretical analysis of the crack formation reasons and a possible consequence for the ITER are given.

  11. Tokamak heating by neutral beams and adiabatic compression

    International Nuclear Information System (INIS)

    Furth, H.P.

    1973-08-01

    ''Realistic'' models of tokamak energy confinement strongly favor reactor operation at the maximum MHD-stable β-value, in order to maximize plasma density. Ohmic heating is unsuitable for this purpose. Neutral-beam heating plus compression is well suited; however, very large requirements on device size and injection power seem likely for a DT ignition experiment using a Maxwellian plasma. Results of the ATC experiment are reviewed, including Ohmic heating, neutral-beam heating, and production of two-energy-component plasmas (energetic deuteron population in deuterium ''target plasma''). A modest extrapolation of present ATC parameters could give zero-power conditions in a DT experiment of the two-energy-component type. (U.S.)

  12. Theory of neutral injection heating of toroidal plasmas

    International Nuclear Information System (INIS)

    Cordey, J.G.

    1976-01-01

    The present state of injection theory is reviewed with particular emphasis on the consequences of high power injection. The subject is divided into the following six sections: fast ion deposition; the slowing down and scattering of the fast ions; energy and momentum transfer rates; heating of the thermal ions; other perturbations; microinstabilities. The theory is compared with the experimental results. The questions that remain to be answered to establish neutral injection as a useful heating technique in reactors, are listed (26 references)

  13. Simulation of cracks in tungsten under ITER specific heat loads

    International Nuclear Information System (INIS)

    Peschany, S.

    2006-01-01

    The problem of high tritium retention in co-deposited carbon layers on the walls of ITER vacuum chamber motivates investigation of materials for the divertor armour others than carbon fibre composite (CFC). Tungsten is most probable material for CFC replacement as the divertor armour because of high vaporisation temperature and heat conductivity. In the modern ITER design tungsten is a reference material for the divertor cover, except for the separatrix strike point armoured with CFC. As divertor armour, tungsten should withstand severe heat loads at off-normal ITER events like disruptions, ELMs and vertical displacement events. Experiments on tungsten heating with plasma streams and e-beams have shown an intense crack formation at the surface of irradiated sample [ V.I. Tereshin, A.N. Bandura, O.V. Byrka et al. Repetitive plasma loads typical for ITER type-I ELMs: Simulation at QSPA Kh-50.PLASMA 2005. ed. By Sadowski M.J., AIP Conference Proceedings, American Institute of Physics, 2006, V 812, p. 128-135., J. Linke. Private communications.]. The reason for tungsten cracking under severe heat loads is thermo stress. It appears as due to temperature gradient in solid tungsten as in resolidified layer after cooling down. Both thermo stresses are of the same value, but the gradiental stress is compressive and the stress in the resolidified layer is tensile. The last one is most dangerous for crack formation and it was investigated in this work. The thermo stress in tungsten that develops during cooling from the melting temperature down to room temperature is ∼ 8-16 GPa. Tensile strength of tungsten is much lower, < 1 GPa at room temperature, and at high temperatures it drops at least for one order of magnitude. As a consequence, various cracks of different characteristic scales appear at the heated surface of the resolidified layer. For simulation of the cracks in tungsten the numeric code PEGASUS-3D [Pestchanyi and I. Landman. Improvement of the CFC structure to

  14. Sawtooth stability in neutral beam heated plasmas in TEXTOR

    NARCIS (Netherlands)

    Chapman, I.T.; Pinches, S. D.; Koslowski, H. R.; Liang, Y.; Kramer-Flecken, A.; De Bock, M.

    2008-01-01

    The experimental sawtooth behaviour in neutral beam injection (NBI) heated plasmas in TEXTOR is described. It is found that the sawtooth period is minimized with a low NBI power oriented in the same direction as the plasma current. As the beam power is increased in the opposite direction to the

  15. Analysis of secondary particle behavior in multiaperture, multigrid accelerator for the ITER neutral beam injector.

    Science.gov (United States)

    Mizuno, T; Taniguchi, M; Kashiwagi, M; Umeda, N; Tobari, H; Watanabe, K; Dairaku, M; Sakamoto, K; Inoue, T

    2010-02-01

    Heat load on acceleration grids by secondary particles such as electrons, neutrals, and positive ions, is a key issue for long pulse acceleration of negative ion beams. Complicated behaviors of the secondary particles in multiaperture, multigrid (MAMuG) accelerator have been analyzed using electrostatic accelerator Monte Carlo code. The analytical result is compared to experimental one obtained in a long pulse operation of a MeV accelerator, of which second acceleration grid (A2G) was removed for simplification of structure. The analytical results show that relatively high heat load on the third acceleration grid (A3G) since stripped electrons were deposited mainly on A3G. This heat load on the A3G can be suppressed by installing the A2G. Thus, capability of MAMuG accelerator is demonstrated for suppression of heat load due to secondary particles by the intermediate grids.

  16. Heat treatment trials for ITER toroidal field coils

    International Nuclear Information System (INIS)

    Matsui, Kunihiro; Hemmi, Tsutomu; Koizumi, Norikiyo; Nakajima, Hideo; Kimura, Satoshi; Nakamoto, Kazunari

    2012-01-01

    Cable-in-conduit (CIC) conductors using Nb 3 Sn strands are used in ITER toroidal fields (TF) coils. Heat treatment generates thermal strain in CIC conductors because of the difference in thermal expansion between the Nb 3 Sn strands and the stainless-steel jacket. The elongation/shrinkage of the TF conductor may make it impossible to insert a wound TF conductor into the groove of a radial plate. In addition, it is expected that the deformation of the winding due to heat treatment-based release of the residual force in the jacket may also make it impossible to insert the winding in the groove, and that correcting the winding geometry to allow insertion of the winding may influence the superconducting performance of the TF conductor. The authors performed several trials using heat treatment as the part of activities in Phase II of TF coil procurement aiming to resolve the above-mentioned technical issues, and evaluated the elongations of 0.064, 0.074 and 0.072% for the straight and curved conductors and 1/3-scale double-pancake (DP) winding, respectively. It was confirmed that correction if the deformed winding did not influence the superconducting performance of the conductor. (author)

  17. Laplace transform overcoming principle drawbacks in application of the variational iteration method to fractional heat equations

    Directory of Open Access Journals (Sweden)

    Wu Guo-Cheng

    2012-01-01

    Full Text Available This note presents a Laplace transform approach in the determination of the Lagrange multiplier when the variational iteration method is applied to time fractional heat diffusion equation. The presented approach is more straightforward and allows some simplification in application of the variational iteration method to fractional differential equations, thus improving the convergence of the successive iterations.

  18. HVPTF-The high voltage laboratory for the ITER Neutral Beam test facility

    Energy Technology Data Exchange (ETDEWEB)

    De Lorenzi, A., E-mail: antonio.delorenzi@igi.cnr.it [Consorzio RFX-Associazione EURATOM-ENEA per la Fusione Corso Stati Uniti 4, 35127 Padova (Italy); Pilan, N.; Lotto, L.; Fincato, M. [Consorzio RFX-Associazione EURATOM-ENEA per la Fusione Corso Stati Uniti 4, 35127 Padova (Italy); Pesavento, G.; Gobbo, R. [DIE, Universita di Padova, Via Gradenigo 6A, I-35100 Padova (Italy)

    2011-10-15

    In the MITICA research program for the construction of the ITER Neutral Beam Injector prototype, a Laboratory for the investigation on high voltage holding in vacuum has been set up. This Laboratory - HVPTF: High Voltage Padova Test Facility - is presently capable of experiments up to 300 kV dc, and planned for the upgrade to 800 kV. The specific mission for this ancillary lab is the support to the electrostatic design and construction of the MITICA accelerator and the development and testing of HV components to be installed inside the MITICA accelerator during its operation. The paper describes the structure of the lab, characterized by a high degree of automation and reports the results of the commissioning at 300 kV and the first results of voltage holding between test electrodes.

  19. HVPTF-The high voltage laboratory for the ITER Neutral Beam test facility

    International Nuclear Information System (INIS)

    De Lorenzi, A.; Pilan, N.; Lotto, L.; Fincato, M.; Pesavento, G.; Gobbo, R.

    2011-01-01

    In the MITICA research program for the construction of the ITER Neutral Beam Injector prototype, a Laboratory for the investigation on high voltage holding in vacuum has been set up. This Laboratory - HVPTF: High Voltage Padova Test Facility - is presently capable of experiments up to 300 kV dc, and planned for the upgrade to 800 kV. The specific mission for this ancillary lab is the support to the electrostatic design and construction of the MITICA accelerator and the development and testing of HV components to be installed inside the MITICA accelerator during its operation. The paper describes the structure of the lab, characterized by a high degree of automation and reports the results of the commissioning at 300 kV and the first results of voltage holding between test electrodes.

  20. 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.

  1. ANL ITER high-heat-flux blanket-module heat transfer experiments

    International Nuclear Information System (INIS)

    Kasza, K.E.

    1992-02-01

    An Argonne National Laboratory facility for conducting tests on multilayered slab models of fusion blanket designs is being developed; some of its features are described. This facility will allow testing under prototypic high heat fluxes, high temperatures, thermal gradients, and variable mechanical loadings in a helium gas environment. Steady and transient heat flux tests are possible. Electrical heating by a two-sided, thin stainless steel (SS) plate electrical resistance heater and SS water-cooled cold panels placed symmetrically on both sides of the heater allow achievement of global one-dimensional heat transfer across blanket specimen layers sandwiched between the hot and cold plates. The heat transfer characteristics at interfaces, as well as macroscale and microscale thermomechanical interactions between layers, can be studied in support of the ITER engineering design effort. The engineering design of the test apparatus has shown that it is important to use multidimensional thermomechanical analysis of sandwich-type composites to adequately analyze heat transfer. This fact will also be true for the engineering design of ITER

  2. Neutral beam heating in stellarators: a numerical approach

    International Nuclear Information System (INIS)

    Hokin, S.A.; Rome, J.A.; Hender, T.C.; Fowler, R.H.

    1983-03-01

    Calculation of neutral beam deposition and heating in stellarators is complicated by the twisty stellarator geometry and by the usual beam focusing, divergence, and cross-sectional shape considerations. A new deposition code has been written that takes all of this geometry into account. A unique feature of this code is that it gives particle deposition in field-line coordinates, enabling the thermalization problem to be solved more efficiently

  3. Measurements of low energy neutral hydrogen efflux during ICRF heating

    International Nuclear Information System (INIS)

    Cohen, S.A.; Ruzic, D.; Voss, D.E.

    1984-09-01

    Using the Low Energy Neutral Atom Spectrometer, measurements were made of the H 0 and D 0 efflux from PLT during ion cyclotron heating experiments. The application of rf power at frequencies appropriate to fundamental and 2nd-harmonic heating results in a rapid, toroidally uniform rise in the charge-exchange efflux at a rate of about 10 15 cm -2 s -1 MW -1 . This flux increase is larger at lower plasma currents. The cause of this flux and its impact on plasma behavior are discussed

  4. Novel manufacturing method by using stainless steel pipes expanded into aluminium profiles for the ITER Neutral Beam cryopumps

    Energy Technology Data Exchange (ETDEWEB)

    Dremel, Matthias, E-mail: matthias.dremel@iter.org; Boissin, Jean-Claude; Déléage, Vincent; Quinn, Eamonn; Pearce, Robert

    2015-10-15

    This paper describes the novel engineering and manufacturing solution of stainless steel pipe expansion into aluminium extrusion profiles for use at cryogenic temperatures up to 400 K. This fabrication method will be used for the thermal radiation shields and the cryopanels of the ITER Neutral Beam cryopumps. The use of stainless steel pipes expanded into aluminium extrusion profiles is a solution that combines standard stainless steel welding procedures for the manifolds of the cooling circuits with extended aluminium structures taking advantage of the high thermal conductivity of aluminium. The cryogenic cooling circuits of the pump are a first confinement barrier in the ITER vacuum vessel and the risk of a leakage needs to be minimized as far as possible. The expansion method avoids the need of joints of dissimilar materials in the primary confinement barrier. The fabrication method and results of the prototyping of full scaled components for the ITER Neutral Beam cryopumps are outlined in this paper.

  5. 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

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

    International Nuclear Information System (INIS)

    Loewenhoff, Th.; Linke, J.; Pintsuk, G.; Pitts, R.A.; Riccardi, B.

    2015-01-01

    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 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

  7. Biological shield around the neutral beam injector ducts in the ITER conceptual design

    International Nuclear Information System (INIS)

    Maki, Koichi; Takatsu, Hideyuki; Satoh, Satoshi; Seki, Yasushi

    1994-01-01

    There are gaps between the toroidal field coils and neutral beam injector (NBI) duct wall for the thermal insulator in tokamak reactors such as ITER (International Thermonuclear Experimental Reactor). Neutrons stream through the duct, and some of them penetrate the wall and stream through the gaps. These neutrons activate the materials composing the duct wall, toroidal field coil (TFC) case and cryostat wall surfaces. The dose rate is enhanced just outside the cryostat around the ducts in the reactor room after reactor operation by activation. We investigated the gamma-ray dose rate just outside the cryostat after shutdown due to gamma-rays from activity induced by the neutrons streaming through the gaps. By evaluating the difference between the dose rate in models with and without gaps, we decided whether the thickness of the cryostat as biological shielding is sufficient or not. From these investigations, we recommend a cryostat design suitable for radiation shielding. Dose rates after shutdown at a point just outside the cryostat around the NBI ducts in the model with gaps are two orders larger than those without gaps. The value at this point is approximately 400 mrem h -1 (4 mSv h -1 ), which is two orders larger than the design value for workers to enter the reactor room. In order to reduce the dose rate after shutdown, a method of providing the shielding function of the cryostat is suggested. ((orig.))

  8. Plasma heating by injection of neutral beams into TFR 600

    International Nuclear Information System (INIS)

    1981-01-01

    Experimental results from quasi-perpendicular high power (up to 1.2 MW) neutral beam injection in the TFR 600 tokamak are reported. The trapped fast ions show all the characteristics of a classical feature. This allows us to study the behaviour of a dense plasma (n approximately equal to 10 14 cm -3 ) whose electron and ion temperatures are significantly changed by fast neutrals injection (ΔTsub(e,i)>300 eV). No increase of the global energy confinement time has been observed, but at low q value a large increase of internal disruptions appears. This fact permits to partly enlighten the internal disruptions mechanism and to emphasize their importance. 1-D simulation calculations are also reported; changes in the electron and ion heat conduction, necessary to explain most of the experimental results observed during the internal disruptions will be discussed

  9. Electron and ion heat transport with lower hybrid current drive and neutral beam injection heating in ASDEX

    International Nuclear Information System (INIS)

    Soeldner, F.X.; Pereverzev, G.V.; Bartiromo, R.; Fahrbach, H.U.; Leuterer, F.; Murmann, H.D.; Staebler, A.; Steuer, K.H.

    1993-01-01

    Transport code calculations were made for experiments with the combined operation of lower hybrid current drive and heating and of neutral beam injection heating on ASDEX. Peaking or flattening of the electron temperature profile are mainly explained by modifications of the MHD induced electron heat transport. They originate from current profile changes due to lower hybrid and neutral beam current drive and to contributions from the bootstrap current. Ion heat transport cannot be described by one single model for all heating scenarios. The ion heat conductivity is reduced during lower hybrid heated phases with respect to Ohmic and neutral beam heating. (author). 13 refs, 5 figs

  10. Qualification test for ITER HCCR-TBS mockups with high heat flux test facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Suk-Kwon, E-mail: skkim93@kaeri.re.kr [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Seong 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)

    2016-11-01

    Highlights: • The test mockups for ITER HCCR (Helium Cooled Ceramic Reflector) TBS (Test Blanket System) in Korea were designed and fabricated. • A thermo-hydraulic analysis was performed using a high heat flux test facility by using electron beam. • The plan for qualification tests was developed to evaluate the thermo-hydraulic efficiency in accordance with the requirements of the ITER Organization. - Abstract: The test mockups for ITER HCCR (Helium Cooled Ceramic Reflector) TBS (Test Blanket System) in Korea were designed and fabricated, and an integrity and thermo-hydraulic performance test should be completed under the same or similar operation conditions of ITER. The test plan for a thermo-hydraulic analysis was developed by using a high heat flux test facility, called the Korean heat load test facility by using electron beam (KoHLT-EB). This facility is utilized for a qualification test of the plasma facing component (PFC) for the ITER first wall and DEMO divertor, and for the thermo-hydraulic experiments. In this work, KoHLT-EB will be used for the plan of the performance qualification test of the ITER HCCR-TBS mockups. This qualification tests should be performed to evaluate the thermo-hydraulic efficiency in accordance with the requirements of the ITER Organization (IO), which describe the specifications and qualifications of the heat flux test facility and test procedure for ITER PFC.

  11. Development of neutral beams for fusion plasma heating

    International Nuclear Information System (INIS)

    Haselton, H.H.; Pyle, R.V.

    1980-01-01

    A state-of-the-art account of neutral beam technology at the LBL/LLNL and ORNL facilities is given with emphasis on positive-ion-based systems. The advances made in the last few years are elaborated and problem areas are identified. The ORNL program has successfully completed the neutral injection systems for PLT, ISX-B, and most recently, PDX and the ISX-B upgrade. All of these are high current (60 to 100 A), medium energy (40 to 50 keV) systems. This program is also engaged in the development of a reactor-grade advanced positive ion system (150 to 200 kV/100 A/5 to 10 s) and a multimegawatt, long pulse (30 s) heating system for ISX-C. In a joint program, LBL and LLNL are developing and testing neutral beam injection systems based on the acceleration of positive ions for application in the 80- to 160-keV range on MFTF-B, D-III, TFTR/TFM, ETF, MNS, etc. A conceptual design of a 160-keV injection system for the German ZEPHYR project is in progress at LBL/LLNL and independently at ORNL. The laboratories are also engaged in the development of negative-ion-based systems for future applications at higher energies

  12. Development of neutral beams for fusion plasma heating

    Energy Technology Data Exchange (ETDEWEB)

    Haselton, H.H.; Pyle, R.V.

    1980-01-01

    A state-of-the-art account of neutral beam technology at the LBL/LLNL and ORNL facilities is given with emphasis on positive-ion-based systems. The advances made in the last few years are elaborated and problem areas are identified. The ORNL program has successfully completed the neutral injection systems for PLT, ISX-B, and most recently, PDX and the ISX-B upgrade. All of these are high current (60 to 100 A), medium energy (40 to 50 keV) systems. This program is also engaged in the development of a reactor-grade advanced positive ion system (150 to 200 kV/100 A/5 to 10 s) and a multimegawatt, long pulse (30 s) heating system for ISX-C. In a joint program, LBL and LLNL are developing and testing neutral beam injection systems based on the acceleration of positive ions for application in the 80- to 160-keV range on MFTF-B, D-III, TFTR/TFM, ETF, MNS, etc. A conceptual design of a 160-keV injection system for the German ZEPHYR project is in progress at LBL/LLNL and independently at ORNL. The laboratories are also engaged in the development of negative-ion-based systems for future applications at higher energies.

  13. Heat-exchanger concepts for neutral-beam calorimeters

    International Nuclear Information System (INIS)

    Thompson, C.C.; Polk, D.H.; McFarlin, D.J.; Stone, R.

    1981-01-01

    Advanced cooling concepts that permit the design of water cooled heat exchangers for use as calorimeters and beam dumps for advanced neutral beam injection systems were evaluated. Water cooling techniques ranging from pool boiling to high pressure, high velocity swirl flow were considered. Preliminary performance tests were carried out with copper, inconel and molybdenum tubes ranging in size from 0.19 to 0.50 in. diameter. Coolant flow configurations included (1) smooth tube/straight flow, (2) smooth tube with swirl flow created by tangential injection of the coolant, and (3) axial flow in internally finned tubes. Additionally, the effect of tube L/D was evaluated. A CO 2 laser was employed to irradiate a sector of the tube exterior wall; the laser power was incrementally increased until burnout (as evidenced by a coolant leak) occurred. Absorbed heat fluxes were calculated by dividing the measured coolant heat load by the area of the burn spot on the tube surface. Two six element thermopiles were used to accurately determine the coolant temperature rise. A maximum burnout heat flux near 14 kW/cm 2 was obtained for the molybdenum tube swirl flow configuration

  14. 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

  15. 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

  16. 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.

  17. On choosing a nonlinear initial iterate for solving the 2-D 3-T heat conduction equations

    International Nuclear Information System (INIS)

    An Hengbin; Mo Zeyao; Xu Xiaowen; Liu Xu

    2009-01-01

    The 2-D 3-T heat conduction equations can be used to approximately describe the energy broadcast in materials and the energy swapping between electron and photon or ion. To solve the equations, a fully implicit finite volume scheme is often used as the discretization method. Because the energy diffusion and swapping coefficients have a strongly nonlinear dependence on the temperature, and some physical parameters are discontinuous across the interfaces between the materials, it is a challenge to solve the discretized nonlinear algebraic equations. Particularly, as time advances, the temperature varies so greatly in the front of energy that it is difficult to choose an effective initial iterate when the nonlinear algebraic equations are solved by an iterative method. In this paper, a method of choosing a nonlinear initial iterate is proposed for iterative solving this kind of nonlinear algebraic equations. Numerical results show the proposed initial iterate can improve the computational efficiency, and also the convergence behavior of the nonlinear iteration.

  18. Negative ion based neutral beams for plasma heating

    International Nuclear Information System (INIS)

    Prelec, K.

    1978-01-01

    Neutral beam systems based on negative ions have been considered because of a high expected power efficiency. Methods for the production, acceleration and neutralization of negative ions will be reviewed and possibilities for an application in neutral beam lines explored

  19. 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.

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

    International Nuclear Information System (INIS)

    Bhatnagar, V.P.; Eriksson, L.; Gormezano, C.; Jacquinot, J.; Kaye, A.; Start, D.F.H.

    1994-01-01

    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

  1. High Heat flux (HHF) elements for Negative Ion Systems on ITER

    International Nuclear Information System (INIS)

    Milnes, J.; Chuilon, B.; Martin, D.; Waldon, Ch.; Yong Xue

    2006-01-01

    Negative Ion Neutral Beam systems on ITER will require actively cooled scrapers and dumps to process and shape the beam before injection into the tokamak. The scale of the systems is much larger than any presently operating, bringing challenges for designers in terms of available sub cooling, total pressure drop, deflection and mandatory remote maintenance. In common with Positive Ion systems, flux densities in the order of 15-20 MW/m 2 are commonplace but with much longer pulses. A pulse length in excess of 3000 seconds and the anticipated beam breakdown rate pose new challenges in terms of stress and fatigue life. The cooling system specification (up to 26 bar, 80 o C) adds further constraints impacting the material choice and operating temperature. The DDD designs, based on swirl tubes, have been reviewed as part of the design process and recommendations made. Additionally, alternative designs have been proposed based on the Hypervapotron high heat flux elements with modified geometry and drawing upon a vast background knowledge of large scale equipment procurement and integration. Existing operational and design experience has been applied to give a simple, robust and low maintenance alternative. A full thermomechanical analysis of all HHF components has been undertaken based on ITER design criteria and the limited material data available. The results of this analysis will be presented, highlighting areas where further R(and)D is necessary to reach the operating limits set out in the functional specification. Extensive comparison of these analyses is made with the large operational database of existing JET beamline components for benchmarking purposes. A particular feature of the thermo-mechanical analyses is a fully self-consistent description in which ageing characteristics are related to the local temperature, and the components' power loading takes into account the thermal distortion. The advantages and disadvantages of all designs will be presented and

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

    Energy Technology Data Exchange (ETDEWEB)

    Bazylev, B.; Janeschitz, G. [Forschungszentrum Karlsruhe GmbH, FZK, Karlsruhe (Germany); Landman, I.; Pestchanyi, S. [FZK-Forschungszentrum Karlsruhe, Association Euratom-FZK, Technik und Umwelt, Karlsruhe (Germany); Loarte, A. [EFDA Close Support Unit Garching, Garching bei Munchen(Germany)

    2007-07-01

    Full text of publication follows: Operation of ITER at high fusion gain is assumed to be the H-mode. A characteristic feature of this regime is the transient release of energy from the confined plasma onto divertor and the first wall by multiple ELMs (about 10{sup 4} ELMs per ITER discharge), which can play a determining role in the erosion rate and lifetime of these components. It is expected that about 50-70 % of the ELM energy releases onto divertor armour and the rest is dumped onto the First Wall (FW) armour. The expected energy heat loads on the ITER divertor and FW during Type I ELM are in range 0.5 - 4 MJ/m{sup 2} in timescales of 0.3-0.6 ms. In case of the ITER disruptions the material evaporated from the divertor expands into the SOL and generates significant radiation heating of the FW armour up to several GW/m2 during a few milliseconds that can also lead to the its melting and noticeable damage. Beryllium macro-brush armour (Be-brushes) is foreseen as plasma FW facing component (PFC) in ITER. During the intense transient events in ITER the surface melting, melt motion, melt splashing and evaporation are seen as the main mechanisms of Be-erosion. The expected erosion of the ITER plasma facing components under transient energy loads can be properly estimated by numerical simulations using the codes MEMOS and PHEMOBRID validated against experimental data obtained at the plasma gun facilities QSPA-T, MK-200UG and QSPA-Kh50 that provide a way to simulate the energy loads expected in ITER in laboratory experiments. The numerical simulations were carried out for the expected ITER ELMs for the heat loads in the range 0.5 - 3.0 MJ/m{sup 2} and the timescale up 0.6 ms and ITER disruptions for the heat loads in the range 2 - 13 MJ/m{sup 2} in timescales of 1-5 ms. Radiation heat loads at the FW armour from the vapour expanded into the SOL were calculated using the codes FOREV-2 and TOKES for both ITER ELM and ITER disruption scenarios. Melt layer damage of the Be

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

    International Nuclear Information System (INIS)

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

    2007-01-01

    Full text of publication follows: Operation of ITER at high fusion gain is assumed to be the H-mode. A characteristic feature of this regime is the transient release of energy from the confined plasma onto divertor and the first wall by multiple ELMs (about 10 4 ELMs per ITER discharge), which can play a determining role in the erosion rate and lifetime of these components. It is expected that about 50-70 % of the ELM energy releases onto divertor armour and the rest is dumped onto the First Wall (FW) armour. The expected energy heat loads on the ITER divertor and FW during Type I ELM are in range 0.5 - 4 MJ/m 2 in timescales of 0.3-0.6 ms. In case of the ITER disruptions the material evaporated from the divertor expands into the SOL and generates significant radiation heating of the FW armour up to several GW/m2 during a few milliseconds that can also lead to the its melting and noticeable damage. Beryllium macro-brush armour (Be-brushes) is foreseen as plasma FW facing component (PFC) in ITER. During the intense transient events in ITER the surface melting, melt motion, melt splashing and evaporation are seen as the main mechanisms of Be-erosion. The expected erosion of the ITER plasma facing components under transient energy loads can be properly estimated by numerical simulations using the codes MEMOS and PHEMOBRID validated against experimental data obtained at the plasma gun facilities QSPA-T, MK-200UG and QSPA-Kh50 that provide a way to simulate the energy loads expected in ITER in laboratory experiments. The numerical simulations were carried out for the expected ITER ELMs for the heat loads in the range 0.5 - 3.0 MJ/m 2 and the timescale up 0.6 ms and ITER disruptions for the heat loads in the range 2 - 13 MJ/m 2 in timescales of 1-5 ms. Radiation heat loads at the FW armour from the vapour expanded into the SOL were calculated using the codes FOREV-2 and TOKES for both ITER ELM and ITER disruption scenarios. Melt layer damage of the Be FW macro

  4. SOLPS-ITER Study of neutral leakage and drift effects on the alcator C-Mod divertor plasma

    Directory of Open Access Journals (Sweden)

    W. Dekeyser

    2017-08-01

    Full Text Available As part of an effort to validate the edge plasma model in the SOLPS-ITER code suite under ITER-relevant divertor plasma and neutral conditions, we report on progress in the modeling of the Alcator C-Mod divertor plasma with the new code. We perform simulations with a complete drifts model and kinetic neutrals, including effects of neutral viscosity, ion-molecule collisions and Lyα-opaque conditions, but assuming a pure deuterium plasma. Through a series of simulations with varying divertor geometries, we show the importance of including neutal leakage paths through the divertor substructure on the divertor plasma solution. Moreover, the impact of drifts on inner-outer target asymmetries is assessed. Including both effects, we achieve excellent agreement between simulations and upstream and outer target Langmuir Probe data. In absence of strong volumetric losses due to e.g. impurity radiation in our simulations, the strong inner target detachment observed experimentally remains elusive in our modeling at present.

  5. Design of ITER-FEAT RF heating and current drive systems

    International Nuclear Information System (INIS)

    Bosia, G.; Kobayashi, N.; Ioki, K.; Bibet, P.; Koch, R.; Chavan, R.; Tran, M.Q.; Takahashi, K.; Kuzikov, S.; Vdovin, V.

    2001-01-01

    Three radio frequency (RF) heating and current drive (H and CD) systems are being designed for ITER-FEAT: an electron cyclotron (EC), an ion cyclotron (IC) and a lower hybrid (LH) System. The launchers of the RF systems use four ITER equatorial ports and are fully interchangeable. They feature equal power outputs (20 MW/port), similar neutron shielding performance, and identical interfaces with the other machine components. An outline of the design is given in the paper. (author)

  6. Implications of ITER requirements on R and D of RF heating and current drive systems

    International Nuclear Information System (INIS)

    Bosia, G.; Agarici, G.; Beaumont, B.

    2003-01-01

    Heating and Current Drive (H and CD) systems have an essential role in ITER-FEAT operation, as all phases of ITER operation are driven and controlled by the auxiliary power flow. The RF (Electron Cyclotron and Ion Cyclotron) systems, planned to contribute for ∼ 60% of the total auxiliary power (72 MW), with Lower Hybrid used for the specialised function of current drive in the extended performance phase (20 MW), are at different level of technology development. All systems, need a significant development in order to meet ITER operation requirements In this paper these requirements are reviewed and CEA proposals for the development of the Ion cyclotron system presented. (author)

  7. 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.

  8. 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

    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...... 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...... for the future heating and transport systems with the ambition of realizing carbon neutrality....

  9. Confinement studies of neutral beam heated discharges in TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Murakami, M.; Arunasalam, V.; Bell, J.D.; Stauffer, F.; Bell, M.G.; Bitte, M.; Blanchard, W.R.; Boody, F.; Britz, N.

    1985-11-01

    The TFTR tokamak has reached its original machine design specifications (I/sub p/ = 2.5 MA and B/sub T/ = 5.2T). Recently, the D/sup 0/ neutral beam heating power has been increased to 6.3 MW. By operating at low plasma current (I/sub p/ approx. = 0.8 MA) and low density anti n/sub e/ approx. = 1 x 10/sup 19/m/sup -3/), high ion temperatures (9 +- keV) and rotation speeds (7 x 10/sup 5/ m/s) have been achieved during injection. At the opposite extreme, pellet injection into high current plasmas has been used to increase the line-average density to 8 x 10/sup 19/m/sup -3/ and the central density to 1.6 x 10/sup 20/m/sup -3// This wide range of operating conditions has enabled us to conduct scaling studies of the global energy confinement time in both ohmically and beam heated discharges as well as more detailed transport studies of the profile dependence. In ohmic discharges, the energy confinement time is observed to scale linearly with density only up to anti n/sub e/ approx. 4.5 x 10/sup 19/m/sup -3/ and then to increase more gradually, achieving a maximum value of approx. 0.45 s. In beam heated discharges, the energy confinement time is observed to decrease with beam power and to increase with plasma current. With P/sub b/ = 5.6 MW, anti n/sub e/ = 4.7 x 10/sup 19/m/sup -3/, I/sub p/ = 2.2 MA and B/sub T = 4.7T, the gross energy confinement time is 0.22 s and T/sub i/(0) = 4.8 keV. Despite shallow penetration of D/sup 0/ beams (at the beam energy less than or equal to 80 keV with low species yield), tau/sub E/(a) values are as large as those for H/sup 0/ injection, but central confinement times are substantially greater. This is a consequence of the insensitivity of the temperature and safety factor profile shapes to the heating profile. The radial variation of tau/sub E/ is even more pronounced with D/sup 0/ injection into high density pellet-injected plasmas. 25 refs.

  10. Examination of high heat flux components for the ITER divertor after thermal fatigue testing

    International Nuclear Information System (INIS)

    Missirlian, M.; Escourbiac, F.; Schmidt, A.; Riccardi, B.; Bobin-Vastra, I.

    2011-01-01

    An extensive development programme has been carried out in the EU on high heat flux components within the ITER project. In this framework, a full-scale vertical target (VTFS) prototype was manufactured with all the main features of the corresponding ITER divertor design. The fatigue cycling campaign on CFC and W armoured regions, proved the capability of such a component to meet the ITER requirements in terms of heat flux performances for the vertical target. This paper discusses metallographic observations performed on both CFC and W part after this intensive thermal fatigue testing campaign for a better understanding of thermally induced mechanical stress within the component, especially close to the armour-heat sink interface.

  11. Examination of high heat flux components for the ITER divertor after thermal fatigue testing

    Energy Technology Data Exchange (ETDEWEB)

    Missirlian, M., E-mail: marc.missirlian@cea.fr [CEA, IRFM, F-13108 Saint Paul lez Durance (France); Escourbiac, F., E-mail: frederic.escourbiac@cea.fr [CEA, IRFM, F-13108 Saint Paul lez Durance (France); Schmidt, A., E-mail: a.schmidt@fz-juelich.de [Forschungszentrum Juelich, IFE-2 (Germany); Riccardi, B., E-mail: Bruno.Riccardi@f4e.europa.eu [Fusion For Energy, E-08019 Barcelona (Spain); Bobin-Vastra, I., E-mail: isabelle.bobinvastra@areva.com [AREVA-NP, 71200 Le Creusot (France)

    2011-10-01

    An extensive development programme has been carried out in the EU on high heat flux components within the ITER project. In this framework, a full-scale vertical target (VTFS) prototype was manufactured with all the main features of the corresponding ITER divertor design. The fatigue cycling campaign on CFC and W armoured regions, proved the capability of such a component to meet the ITER requirements in terms of heat flux performances for the vertical target. This paper discusses metallographic observations performed on both CFC and W part after this intensive thermal fatigue testing campaign for a better understanding of thermally induced mechanical stress within the component, especially close to the armour-heat sink interface.

  12. Heat transfer study of water-cooled swirl tubes for neutral beam targets

    International Nuclear Information System (INIS)

    Kim, J.; Davis, R.C.; Gambill, W.R.; Haselton, H.H.

    1977-01-01

    Heat transfer considerations of water-cooled swirl-tubes including heat transfer correlations, burnout data, and 2-D considerations are presented in connection with high power neutral beam target applications. We also discuss performance results of several swirl tube targets in use at neutral beam development facilities

  13. Heat Flux Tests of the ITER FWQMs at KoHLT-1

    International Nuclear Information System (INIS)

    Bae, Young Dug; Kim, Suk Kwon; Shin, Hee Yun; Lee, Dong Won; Hong, Bong Guen

    2009-05-01

    As a party of the ITER, especially as a procurement party of the ITER blanket, we have designed the First Wall Qualification Mockup (FWQM) and fabricated five FWQMs. Two of them have been tested up to 12,690/12,020 cycles at a heat flux higher than 0.625 MW/m 2 at the KoHLT-1 facility established in the Korea Atomic Energy Research Institute (KAERI). Two KO FWQMs successfully passed the normal heat flux tests, and there was no indication of defect in the Be-to-CuCrZr joints

  14. The influence of electric fields and neutral particles on the plasma sheath at ITER divertor conditions

    NARCIS (Netherlands)

    Shumack, A.E.

    2011-01-01

    The purpose of this thesis is to support the optimization of the ‘exhaust-pipe’, or so-called ‘divertor’, of the nuclear fusion experiment ITER, a large international fusion reactor now under construction in the south of France. We focus particularly on two ‘tools’ for optimization of the plasma

  15. D III-D divertor target heat flux measurements during Ohmic and neutral beam heating

    International Nuclear Information System (INIS)

    Hill, D.N.; Petrie, T.; Mahdavi, M.A.; Lao, L.; Howl, W.

    1988-01-01

    Time resolved power deposition profiles on the D III-D divertor target plates have been measured for Ohmic and neutral beam injection heated plasmas using fast response infrared thermography (τ ≤ 150 μs). Giant Edge Localized Modes have been observed which punctuate quiescent periods of good H-mode confinement and deposit more than 5% of the stored energy of the core plasma on the divertor armour tiles on millisecond time-scales. The heat pulse associated with these events arrives approximately 0.5 ms earlier on the outer leg of the divertor relative to the inner leg. The measured power deposition profiles are displaced relative to the separatrix intercepts on the target plates, and the peak heat fluxes are a function of core plasma density. (author). Letter-to-the-editor. 11 refs, 7 figs

  16. Modeling of divertor particle and heat loads during application of resonant magnetic perturbation fields for ELM control in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Schmitz, O., E-mail: o.schmitz@fz-juelich.de [Forschungszentrum Jülich, IEK-4, Association EURATOM-FZJ, Jülich (Germany); Becoulet, M. [CEA/IRFM, Cadarache, 13108 St. Paul-lez-Durance Cedex (France); Cahyna, P. [IPP AS CR, Za Slovankou 3, 18200 Prague 8 (Czech Republic); Evans, T.E. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Feng, Y. [Max-Planck-Institut für Plasmaphysik, Greifswald (Germany); Frerichs, H.; Kirschner, A. [Forschungszentrum Jülich, IEK-4, Association EURATOM-FZJ, Jülich (Germany); Kukushkin, A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Laengner, R. [Forschungszentrum Jülich, IEK-4, Association EURATOM-FZJ, Jülich (Germany); Lunt, T. [Max-Planck-Institut für Plasmaphysik, Greifswald (Germany); Loarte, A.; Pitts, R. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Reiser, D.; Reiter, D. [Forschungszentrum Jülich, IEK-4, Association EURATOM-FZJ, Jülich (Germany); Saibene, G. [Fusion for Energy Joint Undertaking, Barcelona (Spain); Samm, U. [Forschungszentrum Jülich, IEK-4, Association EURATOM-FZJ, Jülich (Germany)

    2013-07-15

    First results from three-dimensional modeling of the divertor heat and particle flux pattern during application of resonant magnetic perturbation fields as ELM control scheme in ITER with the EMC3-Eirene fluid plasma and kinetic neutral transport code are discussed. The formation of a helical magnetic footprint breaks the toroidal symmetry of the heat and particle fluxes. Expansion of the flux pattern as far as 60 cm away from the unperturbed strike line is seen with vacuum RMP fields, resulting in a preferable heat flux spreading. Inclusion of plasma response reduces the radial extension of the heat and particle fluxes and results in a heat flux peaking closer to the unperturbed level. A strong reduction of the particle confinement is found. 3D flow channels are identified as a consistent reason due to direct parallel outflow from inside of the separatrix. Their radial inward expansion and hence the level of particle pump out is shown to be dependent on the perturbation level.

  17. Neoclassical current effects in neutral-beam-heated tokamak discharges

    International Nuclear Information System (INIS)

    Hogan, J.T.

    1981-01-01

    There is a long-standing prediction from neoclassical theory that strong contributions to the toroidal current should be driven by friction between trapped and passing particles when βsub(pol) exceeds root (R/a) in a tokamak. A number of neutral-beam heating experiments can now produce such parameters, and it is of interest to calculate the behaviour which should occur in this regime to determine the feasibility of using such a 'bootstrap' current as a steady-state tokamak current source. It is found that the neoclassical current should be large enough to reverse the external loop voltage for typical experimental parameters (ISX-B, in particular) in cases where the total current is fixed and to produce a detectable excess of total current above the pre-programmed (demand) value in cases where the loop voltage is regulated. Other manifestations of such a current should be either: a sharp rise in the central q-value (producing a cessation of internal m=1 and m=2 MHD activity), with an enhancement by two orders of magnitude of ion thermal conductivity (due to the formation of a hollow current density profile and a consequent drop in local values of the poloidal magnetic field in the central plasma region), or an enhanced tendency for disruption (arising from magnetic reconnection in hollow-profile equilibria). Since these gross manifestations are absent in a wide range of experiments on the Impurity Study Experiment (ISX-B), as reported earlier, the conclusion is that the neoclassical current, if present, can have a value no larger than 25% of its theoretically calculated value. Since the neoclassical particle (Ware) pinch is strongly related to the neoclassical current in the theory (Onsager reciprocity), the existence of the particle pinch is thus called into question. (author)

  18. 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.

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

    International Nuclear Information System (INIS)

    Youchison, D.L.; Marshall, T.D.; McDonald, J.M.; Lutz, T.J.; Watson, R.D.; Driemeyer, D.E.; Kubik, D.L.; Slattery, K.T.; Hellwig, T.H.

    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

  20. 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.

  1. Iterative convergence acceleration of neutral particle transport methods via adjacent-cell preconditioners

    International Nuclear Information System (INIS)

    Azmy, Y.Y.

    1999-01-01

    The author proposes preconditioning as a viable acceleration scheme for the inner iterations of transport calculations in slab geometry. In particular he develops Adjacent-Cell Preconditioners (AP) that have the same coupling stencil as cell-centered diffusion schemes. For lowest order methods, e.g., Diamond Difference, Step, and 0-order Nodal Integral Method (ONIM), cast in a Weighted Diamond Difference (WDD) form, he derives AP for thick (KAP) and thin (NAP) cells that for model problems are unconditionally stable and efficient. For the First-Order Nodal Integral Method (INIM) he derives a NAP that possesses similarly excellent spectral properties for model problems. The two most attractive features of the new technique are:(1) its cell-centered coupling stencil, which makes it more adequate for extension to multidimensional, higher order situations than the standard edge-centered or point-centered Diffusion Synthetic Acceleration (DSA) methods; and (2) its decreasing spectral radius with increasing cell thickness to the extent that immediate pointwise convergence, i.e., in one iteration, can be achieved for problems with sufficiently thick cells. He implemented these methods, augmented with appropriate boundary conditions and mixing formulas for material heterogeneities, in the test code APID that he uses to successfully verify the analytical spectral properties for homogeneous problems. Furthermore, he conducts numerical tests to demonstrate the robustness of the KAP and NAP in the presence of sharp mesh or material discontinuities. He shows that the AP for WDD is highly resilient to such discontinuities, but for INIM a few cases occur in which the scheme does not converge; however, when it converges, AP greatly reduces the number of iterations required to achieve convergence

  2. Heating and current drive requirements towards steady state operation in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Poli, F. M.; Kessel, C. E.; Gorelenkova, M. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Bonoli, P. T. [MIT Plasma Science and Fusion Center, Cambridge, MA 02139 (United States); Batchelor, D. B. [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6169 (United States); Harvey, B.; Petrov, Y. [CompX, Box 2672, Del Mar, CA 92014 (United States)

    2014-02-12

    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, reducing the no-wall limit. The E × B 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 weakly reversed magnetic shear profiles throughout the discharge are the focus of this work. Time-dependent transport simulations indicate that, with a trade-off of the EC equatorial and upper launcher, the formation and sustainment of quasi-steady state ITBs could be demonstrated in ITER with the baseline heating configuration. However, with proper constraints from peeling-ballooning theory on the pedestal width and height, the fusion gain and the maximum non-inductive current are below the ITER target. Upgrades of the heating and current drive system in ITER, like the use of Lower Hybrid current drive, could overcome these limitations, sustaining higher non-inductive current and confinement, more expanded ITBs which are ideal MHD stable.

  3. Heating and current drive requirements towards steady state operation in ITER

    Science.gov (United States)

    Poli, F. M.; Bonoli, P. T.; Kessel, C. E.; Batchelor, D. B.; Gorelenkova, M.; Harvey, B.; Petrov, Y.

    2014-02-01

    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, reducing the no-wall limit. The E × B 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 weakly reversed magnetic shear profiles throughout the discharge are the focus of this work. Time-dependent transport simulations indicate that, with a trade-off of the EC equatorial and upper launcher, the formation and sustainment of quasi-steady state ITBs could be demonstrated in ITER with the baseline heating configuration. However, with proper constraints from peeling-ballooning theory on the pedestal width and height, the fusion gain and the maximum non-inductive current are below the ITER target. Upgrades of the heating and current drive system in ITER, like the use of Lower Hybrid current drive, could overcome these limitations, sustaining higher non-inductive current and confinement, more expanded ITBs which are ideal MHD stable.

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

    International Nuclear Information System (INIS)

    Dhalla, F.; Mistry, S.; Turner, I.; Barrett, T.R.; Day, I.; McAdams, R.

    2015-01-01

    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.

  5. The insulation structure of the 1 MV transmission line for the ITER neutral beam injector

    International Nuclear Information System (INIS)

    De Lorenzi, A.; Grando, L.; Gobbo, R.; Pesavento, G.; Bettini, P.; Specogna, R.; Trevisan, F.

    2007-01-01

    The paper describes the studies and the tests for the development of the insulation structure of the 1 MV-50 A gas insulated (SF 6 ) line of the ITER NBI in the SinGap configuration characterized by two kinds of spacers: at least a couple of disk-shaped spacers, designed to be gas tight, and a larger number (several tens) of inner conductor post spacers. To this aim a test campaign has been carried out to assess the capability of standard epoxy spacers to withstand a high dc voltage with frequent short circuits, simulating the operational condition for the ITER NBI. Two computational tools, the first for the epoxy spacer shape optimization under electrostatic distribution and the other for the nonlinear time variant evolution of the electric field and surface charge, have been developed specifically for designing epoxy spacer under dc voltage stress. The results on the optimization of the disk spacer and on the electric field-surface charge time evolution of the post spacer are reported and discussed. The effects of the SF 6 radiation induced conductivity on the post spacer are also reported

  6. The insulation structure of the 1 MV transmission line for the ITER neutral beam injector

    Energy Technology Data Exchange (ETDEWEB)

    De Lorenzi, A. [Consorzio RFX, Associazione Euratom-Enea sulla Fusione, Corso Stati Uniti 4, I-35127 Padova (Italy)], E-mail: antonio.delorenzi@igi.cnr.it; Grando, L. [Consorzio RFX, Associazione Euratom-Enea sulla Fusione, Corso Stati Uniti 4, I-35127 Padova (Italy); Gobbo, R.; Pesavento, G. [DIE, Universita di Padova, Via Gradenigo 6A, I-35100 Padova (Italy); Bettini, P.; Specogna, R.; Trevisan, F. [DIEGM, Universita di Udine, Via delle Scienze 208, I-33100 Udine (Italy)

    2007-10-15

    The paper describes the studies and the tests for the development of the insulation structure of the 1 MV-50 A gas insulated (SF{sub 6}) line of the ITER NBI in the SinGap configuration characterized by two kinds of spacers: at least a couple of disk-shaped spacers, designed to be gas tight, and a larger number (several tens) of inner conductor post spacers. To this aim a test campaign has been carried out to assess the capability of standard epoxy spacers to withstand a high dc voltage with frequent short circuits, simulating the operational condition for the ITER NBI. Two computational tools, the first for the epoxy spacer shape optimization under electrostatic distribution and the other for the nonlinear time variant evolution of the electric field and surface charge, have been developed specifically for designing epoxy spacer under dc voltage stress. The results on the optimization of the disk spacer and on the electric field-surface charge time evolution of the post spacer are reported and discussed. The effects of the SF{sub 6} radiation induced conductivity on the post spacer are also reported.

  7. 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.

  8. 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.

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

    International Nuclear Information System (INIS)

    Boldrin, Marco; De Lorenzi, Antonio; Decamps, Hans; Grando, Luca; Simon, Muriel; Toigo, Vanni

    2013-01-01

    Highlights: ► ITER Neutral Beam Injector includes several non-standard components. ► The design status of the −1 MV 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 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 dc feedthrough, called HVD1-TL Bushing, aimed at connecting the HVD1 to the gas (SF 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

  10. 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.

  11. 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.

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

    International Nuclear Information System (INIS)

    Zaccaria, Pierluigi; Valente, Matteo; Rigato, Wladi; Dal Bello, Samuele; Marcuzzi, Diego; Agostini, Fabio Degli; Rossetto, Federico; Tollin, Marco; Masiello, Antonio; Corniani, Giorgio; Badalocchi, Matteo; Bettero, Riccardo; Rizzetto, Dario

    2015-01-01

    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.

  13. Bootstrap current of fast ions in neutral beam injection heating

    International Nuclear Information System (INIS)

    Huang Qianhong; Gong Xueyu; Yang Lei; Li Xinxia; Lu Xingqiang; Yu Jun

    2012-01-01

    The bootstrap current of fast ions produced by the neutral beam injection is investigated in a large aspect ratio tokamak with circular cross-section under specific parameters. The bootstrap current density distribution and the total bootstrap current are figured out. In addition, the beam bootstrap current always accompanies the electron return current due to the parallel momentum transfer from fast ions. With the electron return current considered, the net current density obviously decreases due to electron return current, at the same time the peak of current moves towards the centre plasma. Numerical results show that the value of the net current depends sensitively not only on the angle of the neutral beam injection but also on the ratio of the velocity of fast ions to the critical velocity: the value of net current is small for the neutral beam parallel injection but increases multipliedly for perpendicular injection, and increases with beam energy increasing. (authors)

  14. Implications of ITER requirements on R and D of RF heating and current drive systems

    International Nuclear Information System (INIS)

    Bosia, G.

    2002-01-01

    A strategic, rather than auxiliary role is assigned to H and CD systems in ITER-FEAT, as all operation phases are driven and controlled by heating and current drive (H and CD) systems. RF systems (Electron Cyclotron, Ion Cyclotron and Lower Hybrid), planned to contribute for ∼60% of ITER auxiliary power (72 MW), still require different level of pre-industrial technology development to operate in ITER at the required level of efficiency and religiosite. In this paper, RF H and CD systems technical and operational issues are reviewed and future R and D actions at CEA-Cadarache discussed, with the aim of providing a demonstration of all RF H and CD systems, within the current ITER construction time scale. The need and the economical advantage of an early on- and off- plasma design validation program for ITER-like RF devices (such as launcher and/or power sources), is also discussed with the aim of identifying and resolving operational issues. (author)

  15. 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.

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

    International Nuclear Information System (INIS)

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

    2011-01-01

    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.

  17. Status of the ITER full-tungsten divertor shaping and heat load distribution analysis

    International Nuclear Information System (INIS)

    Carpentier-Chouchana, S; Hirai, T; Escourbiac, F; Durocher, A; Fedosov, A; Ferrand, L; Kocan, M; Kukushkin, A S; Jokinen, T; Komarov, V; Lehnen, M; Merola, M; Mitteau, R; Pitts, R A; Sugihara, M; Firdaouss, M; Stangeby, P C

    2014-01-01

    In September 2011, the ITER Organization (IO) proposed to begin operation with a full-tungsten (W) armoured divertor, with the objective of taking a decision on the final target material (carbon fibre composite or W) by the end of 2013. This period of 2 years would enable the development of a full-W divertor design compatible with nuclear operations, the investigation of further several physics R and D aspects associated with the use of W targets and the completion of technology qualification. Beginning with a brief overview of the reference heat load specifications which have been defined for the full-W engineering activity, this paper will report on the current status of the ITER divertor shaping and will summarize the results of related three-dimensional heat load distribution analysis performed as part of the design validation. (paper)

  18. Tungsten erosion under plasma heat loads typical for ITER type I Elms and disruptions

    Energy Technology Data Exchange (ETDEWEB)

    Garkusha, I.E. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine)]. E-mail: garkusha@ipp.kharkov.ua; Bandura, A.N. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Byrka, O.V. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Chebotarev, V.V. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Landman, I.S. [Forschungszentrum Karlsruhe, IHM, 76021 Karlsruhe (Germany); Makhlaj, V.A. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Marchenko, A.K. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Solyakov, D.G. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Tereshin, V.I. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Trubchaninov, S.A. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Tsarenko, A.V. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine)

    2005-03-01

    The behavior of pure sintered tungsten under repetitive plasma heat loads of {approx}1 MJ/m{sup 2} (which is relevant to ITER ELMs) and 25 MJ/m{sup 2} (ITER disruptions) is studied with the quasi-steady-state plasma accelerator QSPA Kh-50. The ELM relevant heat loads have resulted in formation of two kinds of crack networks, with typical sizes of 10-20 {mu}m and {approx}1 mm, at the surface. Tungsten preheating to 600 deg. C indicates that fine intergranular cracks are probably caused by thermal stresses during fast resolidification of the melt, whereas large cracks are the result of ductile-to-brittle transition. For several hundreds of ELM-like exposures, causing surface melting, the melt motion does not dominate the profile of the melt spot. The disruption relevant experiments demonstrated that melt motion became the main factor of tungsten damage.

  19. Erosion simulation of first wall beryllium armour after ITER transient heat loads and runaway electrons action

    Energy Technology Data Exchange (ETDEWEB)

    Bazylev, B., E-mail: boris.bazylev@kit.edu [Karlsruhe Institute of Technology, IHM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Igitkhanov, Yu.; Landman, I.; Pestchanyi, S. [Karlsruhe Institute of Technology, IHM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Loarte, A. [ITER Organisation, Cadarache, 13108 Saint Paul Lez Durance Cedex (France)

    2011-10-01

    Beryllium is foreseen as plasma facing armour for the first wall (FW) in 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) and runaway electrons impact 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, evaporation, and melt motion, which determine the life-time of the plasma facing components. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the J x B forces are analyzed for bulk Be and different sizes of Be-brushes. The damage of the FW due to heat loads caused by runaway electrons is numerically simulated.

  20. Erosion simulation of first wall beryllium armour after ITER transient heat loads and runaway electrons action

    International Nuclear Information System (INIS)

    Bazylev, B.; Igitkhanov, Yu.; Landman, I.; Pestchanyi, S.; Loarte, A.

    2011-01-01

    Beryllium is foreseen as plasma facing armour for the first wall (FW) in 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) and runaway electrons impact 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, evaporation, and melt motion, which determine the life-time of the plasma facing components. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the J x B forces are analyzed for bulk Be and different sizes of Be-brushes. The damage of the FW due to heat loads caused by runaway electrons is numerically simulated.

  1. Ion orbit modelling of ELM heat loads on ITER divertor vertical targets.

    Czech Academy of Sciences Publication Activity Database

    Gunn, J. P.; Carpentier-Chouchana, S.; Dejarnac, Renaud; Escourbiac, F.; Hirai, T.; Komm, Michael; Kukushkin, A.; Panayotis, S.; Pitts, R.A.

    2017-01-01

    Roč. 12, August (2017), s. 75-83 ISSN 2352-1791. [International Conference on Plasma Surface Interactions 2016, PSI2016 /22./. Roma, 30.05.2016-03.06.2016] Institutional support: RVO:61389021 Keywords : ITER * Divertor * ELM heat loads Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) http://www.sciencedirect.com/science/article/pii/S2352179116302745

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

    Czech Academy of Sciences Publication Activity Database

    Jacquet, P.; Bobkov, V.; Colas, L.; Czarnecka, A.; Lerche, E.; Mayoral, M.-L.; Monakhov, I.; Van-Eester, D.; Arnoux, G.; Brezinsek, S.; Brix, M.; Campergue, A.-L.; Devaux, S.; Drewelow, P.; Graham, M.; Klepper, C.C.; Meigs, A.; Milanesio, D.; Mlynář, Jan; Pütterich, T.; Sirinelli, A.

    2014-01-01

    Roč. 21, č. 6 (2014), 061510-061510 ISSN 1070-664X. [Topical conference on radio frequency power in plasmas/20./. Sorrento, 25.06.2013-28.06.2013] Institutional support: RVO:61389021 Keywords : JET * ITER-like wall * ICRF heating * impurities * sawtooth * simulation * transport Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.142, year: 2014 http://scitation.aip.org/content/aip/journal/pop/21/6/10.1063/1.4884354

  3. Tritium and heat management in ITER Test Blanket Systems port cell for maintenance operations

    Energy Technology Data Exchange (ETDEWEB)

    Giancarli, L.M., E-mail: luciano.giancarli@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Cortes, P.; Iseli, M.; Lepetit, L.; Levesy, B. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Livingston, D. [Frazer-Nash Consultancy Ltd., Stonebridge House, Dorking Business Park, Dorking, Surrey RH4 1HJ (United Kingdom); Nevière, J.C. [Comex-Nucleaire, 13115 Saint Paul Lez Durance (France); Pascal, R. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ricapito, I. [Fusion for Energy, Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Shu, W. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Wyse, S. [Frazer-Nash Consultancy Ltd., Stonebridge House, Dorking Business Park, Dorking, Surrey RH4 1HJ (United Kingdom)

    2014-10-15

    Highlights: •The ITER TBM Program is one of the ITER missions. •We model a TBM port cell with CFD to optimize the design choices. •The heat and tritium releases management in TBM port cells has been optimized. •It is possible to reduce the T-concentration below one DAC in TBM port cells. •The TBM port cells can have human access within 12 h after shutdown. -- Abstract: Three ITER equatorial port cells are dedicated to the assessment of six different designs of breeding blankets, known as Test Blanket Modules (TBMs). Several high temperature components and pipework will be present in each TBM port cell and will release a significant quantity of heat that has to be extracted in order to avoid the ambient air and concrete wall temperatures to exceed allowable limits. Moreover, from these components and pipes, a fraction of the contained tritium permeates and/or leaks into the port cell. This paper describes the optimization of the heat extraction management during operation, and the tritium concentration control required for entry into the port cell to proceed with the required maintenance operations after the plasma shutdown.

  4. Bootstrap current of fast ions in neutral beam injection heating

    International Nuclear Information System (INIS)

    Huang Qianhong; Gong Xueyu; Li Xinxia; Yu Jun

    2012-01-01

    The bootstrap current of fast ions produced by neutral beam injection (NBI) is investigated in a large-aspect-ratio tokamak with circular cross-section under specific parameters. The bootstrap current density distribution and the total bootstrap current are reported. In addition, the beam bootstrap current always accompanies the electron return current due to the parallel momentum transfer from fast ions. With the electron return current taken into consideration, the net current density obviously decreases; at the same time, the peak of the current moves towards the central plasma. Numerical results show that the value of the net current depends sensitively not only on the angle of the NBI but also on the ratio of the velocity of fast ions to the critical velocity: the value of the net current is small for neutral beam parallel injection, but increases severalfold for perpendicular injection, and increases with increasing beam energy. (paper)

  5. Fast ion profiles during neutral beam and lower hybrid heating

    International Nuclear Information System (INIS)

    Heidbrink, W.W.; Strachan, J.D.; Bell, R.E.; Cavallo, A.; Motley, R.; Schilling, G.; Stevens, J.; Wilson, J.R.

    1985-07-01

    Profiles of the d(d,p)t fusion reaction are measured in the PLT tokamak using an array of collimated 3 MeV proton detectors. During deuterium neutral beam injection, the emission profile indicates that the beam deposition is at least as narrow as predicted by a bounce-averaged Fokker-Planck code. The fast ion tail formed by lower hybrid waves (at densities above the critical density for current drive) also peaks strongly near the magnetic axis

  6. The Insulation Structure of the 1 Megavolt Transmission Line for the ITER Neutral Beam Injector

    International Nuclear Information System (INIS)

    Lorenzi, A. de; Grando, L.; Gobbo, R; Pesavento, G.; Bettini, P.; Specogna, R.; Trevisan, F.

    2006-01-01

    The paper describes the studies and the tests for the development of the insulation structure of the 1 MV-50 A Gas Insulated (SF 6 ) Line of the ITER NBI in the SINGAP configuration. Basically, the Transmission Line for the SinGap configuration consists of a coaxial conductor with the inner electrode polarized at the negative value of 1 MV dc. Despite Transmission Line belongs to the Gas Insulated Switchgear (GIS) family, whose technology is nowadays mature, the design of spacers (usually epoxy post and conical type) for High Voltage DC with the same degree of reliability of spacer for High Voltage AC is far away to be attained, due to the substantial difference between AC and DC voltage electric field configuration and insulation behavior at Epoxy-SF 6 interface. Furthermore, the occurrence of very frequent short circuits during the NBI operation introduces further elements of uncertainty in the design of the insulation structure. The first step was then to assess the capability of standard epoxy spacers to withstand DC voltage with frequent short circuits, in order to determine if the choice of the material is suitable for such an application; a test circuit was set up, with the possibility to generate 200 kV DC and to produce a short-circuit every 15'. The system with the spacers was pressurized with SF 6 at 0.3 MPa. For the interpretation of the experimental results, the spacer has been modeled with commercial (ANSYS TM FEM) and research (Cell Method) numerical codes in order to evaluate both the capacitive and resistive electric field distributions. Once assessed the possibility of using epoxy compounds for the spacer construction, the post and cone spacers have been designed taking into consideration various cases, like electrostatic field configuration, resistive distribution for homogeneous and skin conductivity of the spacer, and for high SF 6 conductivity. In these cases the optimization of the triple point screening has been evaluated, leading to

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

    International Nuclear Information System (INIS)

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

    2014-01-01

    The development of a large area (A source,ITER  = 0.9 × 2 m 2 ) 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 (A source,BATMAN  ≈ 0.32 × 0.59 m 2 ) 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

  8. 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

  9. Progress in the ITER electron cyclotron heating and current drive system design

    Energy Technology Data Exchange (ETDEWEB)

    Omori, Toshimichi, E-mail: toshimichi.omori@iter.org [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Albajar, Ferran; Bonicelli, Tullio; Carannante, Giuseppe; Cavinato, Mario; Cismondi, Fabio [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Darbos, Caroline [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Denisov, Grigory [Institute of Applied Physics Russian Academy of Sciences, 46 Ulyanov Street, Nizhny Novgorod 603950 (Russian Federation); Farina, Daniela [Istituto di Fisica del Plasma, Association EURATOM-ENEA-CNR, Milano (Italy); Gagliardi, Mario [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Gandini, Franco; Gassmann, Thibault [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Goodman, Timothy [CRPP, Association EURATOM-Confédération Suisse, EPFL Ecublens, CH-1015 Lausanne (Switzerland); Hanson, Gregory [US ITER Project Office, ORNL, 055 Commerce Park, PO Box 2008, Oak Ridge, TN 37831 (United States); Henderson, Mark A. [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Kajiwara, Ken [Japan Atomic Energy Agency (JAEA), 801-1 Mukoyama, Naka-shi, Ibaraki 311-0193 (Japan); McElhaney, Karen [US ITER Project Office, ORNL, 055 Commerce Park, PO Box 2008, Oak Ridge, TN 37831 (United States); Nousiainen, Risto [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Oda, Yasuhisa [Japan Atomic Energy Agency (JAEA), 801-1 Mukoyama, Naka-shi, Ibaraki 311-0193 (Japan); Oustinov, Alexander [Institute of Applied Physics Russian Academy of Sciences, 46 Ulyanov Street, Nizhny Novgorod 603950 (Russian Federation); and others

    2015-10-15

    Highlights: • EC system is designed with an ability to upgrade in power to 28 MW then 40 MW. • The TL is capable of 3 buildings movements; ±15 mm displacements at the worst. • Improved power deposition access injecting 20 MW across nearly the entire plasma. • Ensured nuclear safety by appropriate definition of confinement boundaries. • Proposed I&C architecture for the overall EC plant was successfully reviewed. - Abstract: An electron cyclotron system is one of the four auxiliary plasma heating systems to be installed on the ITER tokamak. The ITER EC system consists of 24 gyrotrons with associated 12 high voltage power supplies, a set of evacuated transmission lines and two types of launchers. The whole system is designed to inject 20 MW of microwave power at 170 GHz into the plasma. The primary functions of the system include plasma start-up, central heating and current drive, and magneto-hydrodynamic instabilities control. The design takes present day technology and extends towards high power CW 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. The EC system is faced with significant challenges, which not only includes an advanced microwave system for plasma heating and current drive applications but also has to comply with stringent requirements associated with nuclear safety as ITER became the first fusion device licensed as basic nuclear installations as of 9 November 2012. Since conceptual design of the EC system established in 2007, the EC system has progressed to a preliminary design stage in 2012, and is now moving forward towards a final design. The majority of the subsystems have completed the detailed design and now advancing towards the final design completion.

  10. Progress in the ITER electron cyclotron heating and current drive system design

    International Nuclear Information System (INIS)

    Omori, Toshimichi; Albajar, Ferran; Bonicelli, Tullio; Carannante, Giuseppe; Cavinato, Mario; Cismondi, Fabio; Darbos, Caroline; 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; Oustinov, Alexander

    2015-01-01

    Highlights: • EC system is designed with an ability to upgrade in power to 28 MW then 40 MW. • The TL is capable of 3 buildings movements; ±15 mm displacements at the worst. • Improved power deposition access injecting 20 MW across nearly the entire plasma. • Ensured nuclear safety by appropriate definition of confinement boundaries. • Proposed I&C architecture for the overall EC plant was successfully reviewed. - Abstract: An electron cyclotron system is one of the four auxiliary plasma heating systems to be installed on the ITER tokamak. The ITER EC system consists of 24 gyrotrons with associated 12 high voltage power supplies, a set of evacuated transmission lines and two types of launchers. The whole system is designed to inject 20 MW of microwave power at 170 GHz into the plasma. The primary functions of the system include plasma start-up, central heating and current drive, and magneto-hydrodynamic instabilities control. The design takes present day technology and extends towards high power CW 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. The EC system is faced with significant challenges, which not only includes an advanced microwave system for plasma heating and current drive applications but also has to comply with stringent requirements associated with nuclear safety as ITER became the first fusion device licensed as basic nuclear installations as of 9 November 2012. Since conceptual design of the EC system established in 2007, the EC system has progressed to a preliminary design stage in 2012, and is now moving forward towards a final design. The majority of the subsystems have completed the detailed design and now advancing towards the final design completion.

  11. Modeling of the negative ions extraction from a hydrogen plasma source. Application to ITER Neutral Beam Injector

    International Nuclear Information System (INIS)

    Mochalskyy, S.

    2011-12-01

    The development of a high performance negative ion (NI) source constitutes a crucial step in the construction of a Neutral Beam Injector of the future fusion reactor ITER. NI source should deliver 40 A of H - or of D - . To address this problem in a realistic way, a 3D particles-in-cell electrostatic collisional code was developed. Binary collisions between the particles are introduced using Monte-Carlo collision scheme. This code called ONIX was used to investigate the plasma properties and the transport of the charged particles close to a typical extraction aperture. Results obtained from this code are presented in this thesis. They include negative ions and electrons 3D trajectories. The ion and electron current density profiles are shown for different local magnetic field configurations. Results of production, destruction, and transport of H - in the extraction region are also presented. The production of H - is investigated via 3 atomic processes: 1) electron dissociative attachment to the vibrationally excited molecules H 2 (v) in the volume, 2) interaction of the positive ions H + and H 2 + with the aperture wall and 3) collisions of the neutral gas H, H 2 with aperture wall. The influence of each process on the total extracted NI current is discussed. The extraction efficiency of H - from the volume is compared to the one of H - coming from the wall. Moreover, a parametric study of the H - surface production is presented. Results show the role of sheath behavior in the vicinity of the aperture developing a double layer structure responsible of the NI extraction limitations. The 2 following issues are also analysed. First the influence of the external extracted potential value on the formation of negative sheath and secondly the strength of the magnetic filter on the total extracted NI and co-extracted electron current. The suppression of the electron beam by the negative ion produced at the plasma grid wall is also discussed. Results are in good agreement

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

    International Nuclear Information System (INIS)

    Gavila, P.; Riccardi, B.; Pintsuk, G.; Ritz, G.; Kuznetsov, V.; Durocher, A.

    2015-01-01

    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"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"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 program

  13. 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

  14. External heating and current drive source requirements towards steady-state operation in ITER

    Science.gov (United States)

    Poli, F. M.; Kessel, C. E.; Bonoli, P. T.; Batchelor, D. B.; Harvey, R. W.; Snyder, P. B.

    2014-07-01

    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, reducing the no-wall limit. The E × B 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 heating and current drive (H/CD) sources that sustain reversed magnetic shear profiles throughout the discharge are the focus of this work. Time-dependent transport simulations indicate that a combination of electron cyclotron (EC) and lower hybrid (LH) waves is a promising route towards steady state operation in ITER. The LH forms and sustains expanded barriers and the EC deposition at mid-radius freezes the bootstrap current profile stabilizing the barrier and leading to confinement levels 50% higher than typical H-mode energy confinement times. Using LH spectra with spectrum centred on parallel refractive index of 1.75-1.85, the performance of these plasma scenarios is close to the ITER target of 9 MA non-inductive current, global confinement gain H98 = 1.6 and fusion gain Q = 5.

  15. Local fractional variational iteration algorithm iii for the diffusion model associated with non-differentiable heat transfer

    Directory of Open Access Journals (Sweden)

    Meng Zhi-Jun

    2016-01-01

    Full Text Available This paper addresses a new application of the local fractional variational iteration algorithm III to solve the local fractional diffusion equation defined on Cantor sets associated with non-differentiable heat transfer.

  16. The effect of electron cyclotron heating on density fluctuations at ion and electron scales in ITER baseline scenario discharges on the DIII-D tokamak

    Science.gov (United States)

    Marinoni, A.; Pinsker, R. I.; Porkolab, M.; Rost, J. C.; Davis, E. M.; Burrell, K. H.; Candy, J.; Staebler, G. M.; Grierson, B. A.; McKee, G. R.; Rhodes, T. L.; The DIII-D Team

    2017-12-01

    Experiments simulating the ITER baseline scenario on the DIII-D tokamak show that torque-free pure electron heating, when coupled to plasmas subject to a net co-current beam torque, affects density fluctuations at electron scales on a sub-confinement time scale, whereas fluctuations at ion scales change only after profiles have evolved to a new stationary state. Modifications to the density fluctuations measured by the phase contrast imaging diagnostic (PCI) are assessed by analyzing the time evolution following the switch-off of electron cyclotron heating (ECH), thus going from mixed beam/ECH to pure neutral beam heating at fixed βN . Within 20 ms after turning off ECH, the intensity of fluctuations is observed to increase at frequencies higher than 200 kHz in contrast, fluctuations at lower frequency are seen to decrease in intensity on a longer time scale, after other equilibrium quantities have evolved. Non-linear gyro-kinetic modeling at ion and electron scales scales suggest that, while the low frequency response of the diagnostic is consistent with the dominant ITG modes being weakened by the slow-time increase in flow shear, the high frequency response is due to prompt changes to the electron temperature profile that enhance electron modes and generate a larger heat flux and an inward particle pinch. These results suggest that electron heated regimes in ITER will feature multi-scale fluctuations that might affect fusion performance via modifications to profiles.

  17. Critical heat flux acoustic detection: Methods and application to ITER divertor vertical target monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Courtois, X., E-mail: xavier.courtois@cea.fr [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Escourbiac, F. [ITER Organization, Route de Vinon sur Verdon, F-13115 Saint-Paul-Lez-Durance (France); Richou, M.; Cantone, V. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Constans, S. [AREVA-NP, Le Creusot (France)

    2013-10-15

    Actively cooled plasma facing components (PFCs) have to exhaust high heat fluxes from plasma radiation and plasma–wall interaction. Critical heat flux (CHF) event may occur in the cooling channel due to unexpected heat loading or operational conditions, and has to be detected as soon as possible. Therefore it is essential to develop means of monitoring based on precursory signals providing an early detection of this destructive phenomenon, in order to be able to stop operation before irremediable damages appear. Capabilities of CHF early detection based on acoustic techniques on PFC mock-ups cooled by pressurised water were already demonstrated. This paper addresses the problem of the detection in case of flow rate reduction and of flow dilution resulting from multiple plasma facing units (PFU) which are hydraulically connected in parallel, which is the case of ITER divertor. An experimental study is launched on a dedicated mock-up submitted to heat loads up to the CHF. It shows that the measurement of the acoustic waves, generated by the cooling phenomena, allows the CHF detection in conditions similar to that of the ITER divertor, with a reasonable number of sensors. The paper describes the mock-ups and the tests sequences, and comments the results.

  18. Critical heat flux analysis and R and D for the design of the ITER divertor

    International Nuclear Information System (INIS)

    Raffray, A.R.; Chiocchio, S.; Merola, M.; Tivey, R.; Vieider, G.; Schlosser, J.; Driemeyer, D.; Escourbiac, F.; Grigoriev, S.; Youchison, D.

    1999-01-01

    The vertical target and dump target of the ITER divertor have to be designed for high heat fluxes (up to 20 MW/m 2 over ∼10 s). Accommodation of such high heat fluxes gives rise to several issues, including the critical heat flux (CHF) margin which is a key requirement influencing the choice of cooling channel geometry and coolant conditions. An R and D programme was evolved to address the overall CHF issue and to help focus the design. It involved participation of the four ITER home teams and has been very successful in substantially expanding the CHF data base for one-sided heating and in providing more accurate experimental measurements of pressure drop (and derived correlations) for these geometries. This paper describes the major R and D results and the design analysis performed in converging on a choice of reference configuration and parameters which resulted in a CHF margin of ∼1.4 or more for all divertor components. (orig.)

  19. High heat flux tests of mock-ups for ITER divertor application

    International Nuclear Information System (INIS)

    Giniatulin, R.; Gervash, A.; Komarov, V.L.; Makhankov, A.; Mazul, I.; Litunovsky, N.; Yablokov, N.

    1998-01-01

    One of the most difficult tasks in fusion reactor development is the designing, fabrication and high heat flux testing of actively cooled plasma facing components (PFCs). At present, for the ITER divertor project it is necessary to design and test components by using mock-ups which reflect the real design and fabrication technology. The cause of failure of the PFCs is likely to be through thermo-cycling of the surface with heat loads in the range 1-15 MW m -2 . Beryllium, tungsten and graphite are considered as the most suitable armour materials for the ITER divertor application. This work presents the results of the tests carried out with divertor mock-ups clad with beryllium and tungsten armour materials. The tests were carried out in an electron beam facility. The results of high heat flux screening tests and thermo-cycling tests in the heat load range 1-9 MW m -2 are presented along with the results of metallographic analysis carried out after the tests. (orig.)

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

    International Nuclear Information System (INIS)

    Gassmann, T.; Arambhadiya, B.; Beaumont, B.; Baruah, U.K.; Bonicelli, T.; Darbos, C.; Purohit, D.; Decamps, H.; Albajar, F.; Gandini, F.; Henderson, M.; Kazarian, F.; Lamalle, P.U.; Omori, T.; Parmar, D.; Patel, A.; Rathi, D.; Singh, N.P.

    2011-01-01

    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.

  1. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bazylev, B. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany)], E-mail: bazylev@ihm.fzk.de; Janeschitz, G. [Forschungszentrum Karlsruhe, Fusion, P.O. Box 3640, 76021 Karlsruhe (Germany); Landman, I.; Pestchanyi, S. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany); Loarte, A. [ITER Organisation, Cadarache, 13108 Saint Paul Lez Durance Cedex (France)

    2009-04-30

    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.

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

    International Nuclear Information System (INIS)

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

    2009-01-01

    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.

  4. Results of high heat flux tests of tungsten divertor targets under plasma heat loads expected in ITER and tokamaks (review)

    Energy Technology Data Exchange (ETDEWEB)

    Budaev, V. P., E-mail: budaev@mail.ru [National Research Centre Kurchatov Institute (Russian Federation)

    2016-12-15

    Heat loads on the tungsten divertor targets in the ITER and the tokamak power reactors reach ~10MW m{sup −2} in the steady state of DT discharges, increasing to ~0.6–3.5 GW m{sup −2} under disruptions and ELMs. The results of high heat flux tests (HHFTs) of tungsten under such transient plasma heat loads are reviewed in the paper. The main attention is paid to description of the surface microstructure, recrystallization, and the morphology of the cracks on the target. Effects of melting, cracking of tungsten, drop erosion of the surface, and formation of corrugated and porous layers are observed. Production of submicron-sized tungsten dust and the effects of the inhomogeneous surface of tungsten on the plasma–wall interaction are discussed. In conclusion, the necessity of further HHFTs and investigations of the durability of tungsten under high pulsed plasma loads on the ITER divertor plates, including disruptions and ELMs, is stressed.

  5. Heating, current drive and energetic particles studies on JET in preparation of ITER operation

    International Nuclear Information System (INIS)

    Noterdaeme, J.-M.; Budny, R.; Cardinali, A.

    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 far off axis with little or no refueling or momentum input are hollow with only small differences on whether the power deposition is located on the low field side or on the high field side. With LH current drive the magnetic shear was varied from slightly positive to negative. The improved coupling (through the use of plasma shaping and CD 4 ) allowed up to 3.4 MW of P LH in ITB plasmas with more than 15MW of combined NBI and ICRF heating. The q profile with negative magnetic shear and the ITB could be maintained for the duration of the high heating pulse (8s). Fast ions have been produced in JET with ICRF to simulate alpha particles: by using third harmonic 4 He heating, beam injected 4 He at 120 kV were accelerated to energies above 2 MeV, taking advantage of the unique capability of JET to use NBI with 4 He and to confine MeV class ions. ICRF heating was used to replicate the dynamics of alpha heating and the control of an equivalent Q=10 'burn' was simulated. (author)

  6. Application of quasi-steady-state plasma streams for simulation of ITER transient heat loads

    International Nuclear Information System (INIS)

    Bandura, A.N.; Chebotarev, V.V.; Garkusha, I.E.; Makhlaj, V.A.; Marchenko, A.K.; Solyakov, D.G.; Tereshin, V.I.; Trubchaninov, S.A.; Tsarenko, A.V.; Landman, I.

    2004-01-01

    The paper presents experimental investigations of energy characteristics of the plasma streams generated with quasi-steady-state plasma accelerator QSPA Kh-50 and adjustment of plasma parameters from the point of view its applicability for simulation of transient plasma heat loads expected for ITER disruptions and type I ELMs. Possibility of generation of high-power magnetized plasma streams with ion impact energy up to 0.6 keV, pulse length of 0.25 ms and heat loads varied in wide range from 0.5 to 30 MJ/m 2 has been demonstrated and some features of plasma interaction with tungsten targets in dependence on plasma heat loads are discussed. (author)

  7. Calculation of cracking under pulsed heat loads in tungsten manufactured according to ITER specifications

    International Nuclear Information System (INIS)

    Arakcheev, A.S.; Skovorodin, D.I.; Burdakov, A.V.; Shoshin, A.A.; Polosatkin, S.V.; Vasilyev, A.A.; Postupaev, V.V.; Vyacheslavov, L.N.; Kasatov, A.A.; Huber, A.; Mertens, Ph; Wirtz, M.; Linsmeier, Ch; Kreter, A.; Löwenhoff, Th; Begrambekov, L.; Grunin, A.; Sadovskiy, Ya

    2015-01-01

    A mathematical model of surface cracking under pulsed heat load was developed. The model correctly describes a smooth brittle–ductile transition. The elastic deformation is described in a thin-heated-layer approximation. The plastic deformation is described with the Hollomon equation. The time dependence of the deformation and stresses is described for one heating–cooling cycle for a material without initial plastic deformation. The model can be applied to tungsten manufactured according to ITER specifications. The model shows that the stability of stress-relieved tungsten deteriorates when the base temperature increases. This proved to be a result of the close ultimate tensile and yield strengths. For a heat load of arbitrary magnitude a stability criterion was obtained in the form of condition on the relation of the ultimate tensile and yield strengths.

  8. Overview of the EU small scale mock-up tests for ITER high heat flux components

    International Nuclear Information System (INIS)

    Vieider, G.; Barabash, V.; Cardella, A.

    1998-01-01

    This task within the EU R and D for ITER was aimed at the development of basic manufacturing solutions for the high heat flux plasma facing components such as the divertor targets, the baffles and limiters. More than 50 representative small-scale mock-ups have been manufactured with beryllium, carbon and tungsten armour using various joining technologies. High heat flux testing of 20 of these mock-ups showed the carbon mono-blocks to be the most robust solution, surviving 2000 cycles at absorbed heat fluxes of up to 24 MW m -2 . With flat armour tiles rapid joint failures occurred at 5-16 MW m -2 depending on joining technology and armour material. These test results serve as a basis for the selection of manufacturing options and materials for the prototypes now being ordered. (orig.)

  9. Heat flux estimation for neutral beam line components using inverse heat conduction procedures

    International Nuclear Information System (INIS)

    Bharathi, P.; Prahlad, V.; Quereshi, K.; Bansal, L.K.; Rambabu, S.; Sharma, S.K.; Parmar, S.; Patel, P.J.; Baruah, U.K.; Patel, Ravi

    2015-01-01

    In this work, we describe and compare the analytical IHCP methods such-as semi-infinite method, finite slab method and a numerical method called Stolz method for estimating the incident heat flux from the experimentally measured temperature data. In case of analytical methods, the finite time response of the sensor is needed to be accounted for an accurate power density estimations. The modified models corrected for the response time of the sensors are also discussed in this paper. Application of these methods using example temperature waveforms obtained on the SST1-NBI test stand is presented and discussed. For choosing the suitable method for the calorimetry on beam line components, the estimated results are also validated using the ANSYS analysis done on these beam Iine components. As a conclusion, the finite slab method corrected for the influence of the sensor response time found out to be the most suitable method for the inversion of temperature data in case of neutral beam line components

  10. Computational studies of the effect of magnetic field ''ripple'' on neutral beam heating of ZEPHYR

    International Nuclear Information System (INIS)

    Lister, G.G.; Gruber, O.

    1981-01-01

    The results of computations to estimate the heating efficiency of neutral injection in the proposed ZEPHYR experiment are presented. A suitably modified version of the Monte-Carlo neutral deposition and orbit following code FREYA was used for these calculations, in which particular emphasis has been placed on the effects of toroidal field ripple. We find that the ripple associated with the preliminary design of the experiment (+-6%) would result in intolerable energy losses due to ''ripple trapping'' of the fast ions produced by the neutral beam and insufficient heating of the central plasma. The necessary conditions for ignition can be obtained with a total heating power of 25 MW provided the ripple can be reduced to +-1%, in which case energy losses could be kept below 30%. These results are compatible with those found from transport code calculations of the losses to be expected due to ripple enhanced thermal conduction in the plasma

  11. Effect of neutral gas heating in argon radio frequency inductively coupled plasma

    International Nuclear Information System (INIS)

    Chin, O.H.; Jayapalan, K.K.; Wong, C.S.

    2014-01-01

    Heating of neutral gas in inductively coupled plasma (ICP) is known to result in neutral gas depletion. In this work, this effect is considered in the simulation of the magnetic field distribution of a 13.56 MHz planar coil ICP. Measured electron temperatures and densities at argon pressures of 0.03, 0.07 and 0.2 mbar were used in the simulation whilst neutral gas temperatures were heuristically fitted. The simulated results showed reasonable agreement with the measured magnetic field profile. (author)

  12. TOKES studies of the thermal quench heat load reduction in mitigated ITER disruptions

    Directory of Open Access Journals (Sweden)

    S. Pestchanyi

    2017-08-01

    Full Text Available Disruption mitigation by massive gas injection (MGI of Ne gas has been simulated using the 3D TOKES code that includes the injectors of the Disruption Mitigation System (DMS as it will be implemented in ITER. The simulations have been done using a quasi-3D approach, which gives an upper limit for the radiation heat load (notwithstanding possible asymmetries in radial heat flux associated with MHD. The heating of the first wall from the radiation flash has been assessed with respect to injection quantity, the number of injectors, and their location for an H-mode ITER discharge with 280MJ of thermal energy. Simulations for the maximum quantity of Ne (8kPam3 have shown that wall melting can be avoided by using solely the three injectors in the upper ports, whereas shallow melting occurred when the midplane injector had been added. With all four injectors, melting had been avoided for a smaller neon quantity of 250Pam3 that provides still a sufficient radiation level for thermal load mitigation.

  13. ITER...ation

    International Nuclear Information System (INIS)

    Troyon, F.

    1997-01-01

    Recurrent attacks against ITER, the new generation of tokamak are a mix of political and scientific arguments. This short article draws a historical review of the European fusion program. This program has allowed to build and manage several installations in the aim of getting experimental results necessary to lead the program forwards. ITER will bring together a fusion reactor core with technologies such as materials, superconductive coils, heating devices and instrumentation in order to validate and delimit the operating range. ITER will be a logical and decisive step towards the use of controlled fusion. (A.C.)

  14. Evaporation and vapor shielding of CFC targets exposed to plasma heat fluxes relevant to ITER ELMs

    International Nuclear Information System (INIS)

    Safronov, V.M.; Arkhipov, N.I.; Landman, I.S.; Pestchanyi, S.E.; Toporkov, D.A.; Zhitlukhin, A.M.

    2009-01-01

    Carbon fibre composite NB31 was tested at plasma gun facility MK-200UG by plasma heat fluxes relevant to Edge Localised Modes in ITER. The paper reports the results obtained on the evaporation threshold of carbon fibre composite, 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. First experimental results on investigation of the vapor shield onset conditions are presented also. The obtained experimental data are compared with the results of numerical modeling.

  15. Shut-down dose rate analyses for the ITER electron cyclotron-heating upper launcher

    Energy Technology Data Exchange (ETDEWEB)

    Weinhorst, Bastian; Serikov, Arkady; Fischer, Ulrich; Lu, Lei [Institute for Neutron Physics and Reactor Technology INR (Germany); Karlsruhe Institute of Technology KIT (Germany); Spaeh, Peter; Strauss, Dirk [Institute for Applied Materials IAM (Germany); Karlsruhe Institute of Technology KIT (Germany)

    2014-10-15

    The electron cyclotron resonance heating upper launcher (ECHUL) is going to be installed in the upper port of the ITER tokamak thermonuclear fusion reactor for plasma mode stabilization (neoclassical tearing modes and the sawtooth instability). The paper reports the latest neutronic modeling and analyses which have been performed for the ITER reference front steering launcher design. It focuses on the port accessibility after reactor shut-down for which dose rate (SDDR) distributions on a fine regular mesh grid were calculated. The results are compared to those obtained for the ITER Dummy Upper Port. The calculations showed that the heterogeneous ECHUL design gives rise to enhanced radiation streaming as compared to the homogenous dummy upper port. Therefore the used launcher geometry was upgraded to a more recent development stage. The inter-comparison shows a significant improvement of the launchers shielding properties but also the necessity to further upgrade the shielding performance. Furthermore, the analysis for the homogenous dummy upper port, which represents optimal shielding inside the launcher, demonstrates that the shielding upgrade also needs to include the launcher's environment.

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

    Science.gov (United States)

    Chapman, I. T.; Graves, J. P.; 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

    2013-06-01

    13 MW 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 neoclassical 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 α particle stabilization 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 significantly increases, the negative effect on the fusion gain is reduced, and off-axis negative-ion neutral beam injection (NNBI) can also be considered for sawtooth control. Consequently, schemes to reduce the q = 1 radius are highly desirable, such as early heating to delay the current penetration and, of course, active sawtooth destabilization to mediate small frequent sawteeth and retain a small q = 1 radius. Finally, there remains a residual risk that the ECCD + ICRH control actuators cannot keep the sawtooth period below the threshold for triggering NTMs (since this is derived only from empirical scaling and the control modelling has numerous caveats). If this is the case, a secondary control scheme of sawtooth stabilization via ECCD + ICRH + NNBI, interspersed with deliberate triggering of a crash through auxiliary power reduction and simultaneous pre-emptive NTM control by off-axis ECCD has been considered, permitting long transient periods with high fusion

  17. Modeling and analysis of alternative concept of ITER vacuum vessel primary heat transfer system

    International Nuclear Information System (INIS)

    Carbajo, Juan; Yoder, Graydon; Dell'Orco, G.; Curd, Warren; Kim, Seokho

    2010-01-01

    A RELAP5-3D model of the ITER (Latin for 'the way') vacuum vessel (VV) primary heat transfer system has been developed to evaluate a proposed design change that relocates the heat exchangers (HXs) from the exterior of the tokamak building to the interior. This alternative design protects the HXs from external hazards such as wind, tornado, and aircraft crash. The proposed design integrates the VV HXs into a VV pressure suppression system (VVPSS) tank that contains water to condense vapour in case of a leak into the plasma chamber. The proposal is to also use this water as the ultimate sink when removing decay heat from the VV system. The RELAP5-3D model has been run under normal operating and abnormal (decay heat) conditions. Results indicate that this alternative design is feasible, with no effects on the VVPSS tank under normal operation and with tank temperature and pressure increasing under decay heat conditions resulting in a requirement to remove steam generated if the VVPSS tank low pressure must be maintained.

  18. ITER Generic Diagnostic Upper Port Plug Nuclear Heating and Personnel Dose Rate Assessment

    International Nuclear Information System (INIS)

    Feder, Russell E.; Youssef, Mahmoud Z.

    2009-01-01

    Neutronics analysis to find nuclear heating rates and personnel dose rates were conducted in support of the integration of diagnostics in to the ITER Upper Port Plugs. Simplified shielding models of the Visible-Infrared diagnostic and of a large aperture diagnostic were incorporated in to the ITER global CAD model. Results for these systems are representative of typical designs with maximum shielding and a small aperture (Vis-IR) and minimal shielding with a large aperture. The neutronics discrete-ordinates code ATTILA(reg s ign) and SEVERIAN(reg s ign) (the ATTILA parallel processing version) was used. Material properties and the 500 MW D-T volume source were taken from the ITER 'Brand Model' MCNP benchmark model. A biased quadrature set equivalent to Sn=32 and a scattering degree of Pn=3 were used along with a 46-neutron and 21-gamma FENDL energy subgrouping. Total nuclear heating (neutron plug gamma heating) in the upper port plugs ranged between 380 and 350 kW for the Vis-IR and Large Aperture cases. The Large Aperture model exhibited lower total heating but much higher peak volumetric heating on the upper port plug structure. Personnel dose rates are calculated in a three step process involving a neutron-only transport calculation, the generation of activation volume sources at pre-defined time steps and finally gamma transport analyses are run for selected time steps. ANSI-ANS 6.1.1 1977 Flux-to-Dose conversion factors were used. Dose rates were evaluated for 1 full year of 500 MW DT operation which is comprised of 3000 1800-second pulses. After one year the machine is shut down for maintenance and personnel are permitted to access the diagnostic interspace after 2-weeks if dose rates are below 100 (micro)Sv/hr. Dose rates in the Visible-IR diagnostic model after one day of shutdown were 130 (micro)Sv/hr but fell below the limit to 90 (micro)Sv/hr 2-weeks later. The Large Aperture style shielding model exhibited higher and more persistent dose rates. After 1

  19. 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)

  20. Testing of high heat flux components manufactured by ENEA for ITER divertor

    International Nuclear Information System (INIS)

    Visca, Eliseo; Escourbiac, F.; Libera, S.; Mancini, A.; Mazzone, G.; Merola, M.; Pizzuto, A.

    2009-01-01

    ENEA is involved in the International Thermonuclear Experimental Reactor (ITER) R and D activities and in particular in the manufacturing of high heat flux plasma-facing components, such as the divertor targets. During the last years ENEA has manufactured actively cooled mock-ups by using different technologies, namely brazing, diffusion bonding and HIPping. A new manufacturing process that combines two main techniques PBC (Pre-Brazed Casting) and the HRP (Hot Radial Pressing) has been set up and widely tested. A full monoblock medium scale vertical target, having a straight CFC armoured part and a curved W armoured part, was manufactured using this process. The ultrasonic method was used for the non-destructive examinations performed during the manufacturing of the component, from the monoblock preparation up to the final mock-up assembling. The component was also examined by thermography on SATIR facility (CEA, France), afterwards it was thermal fatigue tested at FE200 (200 kW electron beam facility, CEA/AREVA France). The successful results of the thermal fatigue testing performed according the ITER requirements (10 MW/m 2 , 3000 cycles of 10 s on both CFC and W part, then 20/15 MW/m 2 , 2000 cycles of 10 s on CFC/W part, respectively) have confirmed that the developed process can be considerate a candidate for the manufacturing of monoblock divertor components. Furthermore, a 35-MW/m 2 Critical Heat Flux was measured at relevant thermal-hydraulics conditions at the end of the testing campaign. This paper reports the manufacturing route, the thermal fatigue testing results, the pre and post non-destructive examination and the destructive examination performed on the ITER vertical target medium scale mock-up. These activities were performed in the frame of EFDA contracts (04-1218 with CEA, 93-851 JN with AREVA and 03-1054 with ENEA).

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

    International Nuclear Information System (INIS)

    Safronov, V.; Arkhipov, N.I.; Toporkov, D.A.; Zhitlukhin, A.M.; Landman, I.

    2007-01-01

    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 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 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)

  2. High Heat Flux Test of the ITER FW Semi-prototype Mockups

    International Nuclear Information System (INIS)

    Bae, Young Dug; Lee, Dong Won; Kim, Suk Kwon; Yoon, Jae Sung; Hong, Bong Guen

    2010-01-01

    As a procurement party of the ITER (International Thermonuclear Experimental Reactor) blanket project, we have to pass the qualification program which comprises two distinct phases; Phases 1 is about qualifying joining techniques, and phase 2 covers production scaled up component which is so called 'semi-prototype' and the enhanced heat flux qualification. We have constructed the FWQMs (First Wall Qualification Mockups) and already passed the qualification thermal fatigue tests organized by the IO (ITER Organization). The tests were performed at the EB-1200 in the USA and at the BESTH /JUDITH-2 in the EU. For the phase 2 qualification, the semi-prototype design will be provided by the IO. Even though the design has not fixed because the generic design of the first wall is not fixed, it shall comprise a minimum of 3 full length pairs of first wall fingers. We designed and fabricated two semi-prototype mockups which have a full length pair of first wall finger. The two mockups were designed to have different types of the cooling channels; one has a plain rectangular cross-section and the other has a hypervapotron type cooling channel. In order to compare the cooling capability of the two types, we performed high heat flux tests of the two mockups at the KoHLT-2

  3. 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.

  4. Electron cyclotron heating and current drive: Present experiments to ITER. Revision 1

    International Nuclear Information System (INIS)

    Harvey, R.W.; Nevins, W.M.; Smith, G.R.; Lloyd, B.; O'Brien, M.R.; Warrick, C.D.

    1995-08-01

    Electron cyclotron (EC) power has technological and physics advantages for heating and current drive in a tokamak reactor, and advances in source development make it credible for applications in ITER. Strong single pass absorption makes heating to ignition particularly simple. The optimized EC current drive (ECCD) efficiency (left-angle n right-angle IR/P) shows a linear temperature scaling at temperatures up to ∼ 15 keV. For temperatures above 30 keV, the efficiency saturates at approximately 0.3·10 20 A/(m 2 W) for a frequency of 220 GHz in an ITER target plasma with toroidal field of 6 T, due primarily to harmonic overlap [G.R. Smith et al., Phys. Fluids 30 3633 (1987)] and to a lesser extent due to limitations arising from relativistic effects [N.J. Fisch, Phys. Rev. A 24 3245 (1981)]. The same efficiency can also be obtained at 170 GHz for the same plasma equilibrium except that the magnetic field is reduced to (170/220) x 6 T = 4.6 T. The ECCD efficiencies are obtained with the comprehensive 3D, bounce-averaged Fokker-Planck CQL3D codes [R.W. Harvey and M.G. McCoy, Proc. IAEA TCM/Advances in Simulation and Modeling in Thermonuclear Plasmas 1992, Montreal], and BANDIT3D [M.R. O'Brien, M. Cox, C.D. Warrick, and F. S. Zaitsev, ibid.

  5. High-power neutral-beam heating in the adiabatic toroidal compressor

    International Nuclear Information System (INIS)

    Ellis, R.A.; Eubank, H.P.; Goldston, R.; Smith, R.R.; Nagashima, T.

    1976-05-01

    Neutral-beam injection experiments on ATC have resulted in net power deposited in the plasma of up to 230 kW. The power deposited in the plasma ions is large compared to that from ohmic heating. For a variety of beam and plasma ion species, the increase in ion temperature is proportional to beam power

  6. Tests on fast heating for the regeneration process of ITER cryopumps

    International Nuclear Information System (INIS)

    Day, C.; Kammerer, B.; Mack, A.

    1996-10-01

    Within the framework of the European Fusion Technology Programme, a primary vacuum pump for the ITER reactor is being developed. As the tritium accumulated by the pumps must be limited, short pumping cycles are necessary and as a consequence to that regeneration times of about 4 min only are required by the intermittently working cryopumps; approximately 60 s are available for the heating process from LHe temperature (4.2 K) to LN 2 temperature (77 K). Methods for fast heating were tested in component tests. The heating tests were performed at the TITAN test facility. According to the basic planning, the LHe-cooled panel consisted of seven flow channels in quilted design (500 x 350 mm 2 ); the detailed planning meanwhile showed that a smaller number of channels per panel will be sufficient. The panel was mounted in a LN 2 -cooled rig, which worked as first pumping stage. After having worked out a screening study comprehensive test series with three different heating methods were performed. (orig.) [de

  7. 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.

  8. Diamagnetic measurement of JFT-2 plasma heated by neutral beam injection

    International Nuclear Information System (INIS)

    Maeno, Masaki; Sengoku, Seio; Yamamoto, Shin; Suzuki, Norio; Yamauchi, Toshihiko; Kawashima, Hisato; Miura, Yukitoshi

    1984-01-01

    A neutral beam was injected into the plasma in the JFT-2 tokamak, and the poloidal beta value βsub(p) of the plasma was determined by a diamagnetic method in which the change in the magnetic flux due to the plasma was obtained by measuring the very small perturbation of the current in the tokamak's toroidal field coil. The ratio of the perturbed to unperturbed currents in the coil was found to be (2-3) x 10 -4 . The poloidal beta value βsub(pd) determined by this method agrees within experimental error with that obtained from magnetic and energy profile analyses. βsub(pd) increases linearly with the total power Psub(net) deposited by the neutral beam in the plasma when Psub(net)=1.5 MW. The heating efficiency of the beam injection heating was found to be lower than that of Joule heating. (author)

  9. European contributions to the beam source design and R and D of the ITER neutral beam injectors

    International Nuclear Information System (INIS)

    Massmann, P.; Bayetti, P.; Bucalossi, J.

    2001-01-01

    The paper reports on the progress made by the European Home Team in strong interaction with the ITER JCT and JAERI regarding several key aspects of the beam source for the ITER injectors: integration of the SINGAP accelerator into the ITER injector design. This is a substantially simpler concept than the MAMuG accelerator of the ITER NBI 'reference design', which has potential for significant cost savings, and which avoids some of the weaknesses of the reference design such as the need for intermediate high voltage potentials from the HV power supply and pressurised gas insulation; high energy negative ion acceleration using a SINGAP accelerator; long pulse (i.e. >1000 s) negative ion source operation in deuterium; RF source development, which could reduce the scheduled maintenance of the ITER injectors (as it uses no filaments), and simplify the transmission line and the auxiliary power supplies for the ion source. (author)

  10. European contributions to the beam source design and R and D of the ITER neutral beam injectors

    International Nuclear Information System (INIS)

    Massmann, P.; Bayetti, P.; Bucalossi, J.

    1999-01-01

    The paper reports on the progress made by the European Home Team in strong interaction with the ITER JCT and JAERI regarding several key aspects of the beam source for the ITER injectors: integration of the SINGAP accelerator into the ITER injector design. This is a substantially simpler concept than the MAMuG accelerator of the ITER NBI 'reference design', which has potential for significant cost savings, and which avoids some of the weaknesses of the reference design such as the need for intermediate high voltage potentials from the HV power supply and pressurised gas insulation; high energy negative ion acceleration using a SINGAP accelerator; long pulse (i.e. >1000 s) negative ion source operation in deuterium; RF source development, which could reduce the scheduled maintenance of the ITER injectors (as it uses no filaments), and simplify the transmission line and the auxiliary power supplies for the ion source. (author)

  11. Measurements of the fast ion distribution during neutral beam injection and ion cyclotron heating in ATF [Advanced Toroidal Facility

    International Nuclear Information System (INIS)

    Wade, M.R.; Kwon, M.; Thomas, C.E.; Colchin, R.J.; England, A.C.; Gossett, J.M.; Horton, L.D.; Isler, R.C.; Lyon, J.F.; Rasmussen, D.A.; Rayburn, T.M.; Shepard, T.D.; Bell, G.L.; Fowler, R.H.; Morris, R.N.

    1990-01-01

    A neutral particle analyzer (NPA) with horizontal and vertical scanning capability has been used to make initial measurements of the fast ion distribution during neutral beam injection (NBI) and ion cyclotron heating (ICH) on the Advanced Toroidal Facility (ATF). These measurements are presented and compared with the results of modeling codes that predict the analyzer signals during these heating processes. 6 refs., 5 figs

  12. Plasma facing materials performance under ITER-relevant mitigated disruption photonic heat loads

    Science.gov (United States)

    Klimov, N. S.; Putrik, A. B.; Linke, J.; Pitts, R. A.; Zhitlukhin, A. M.; Kuprianov, I. B.; Spitsyn, A. V.; Ogorodnikova, O. V.; Podkovyrov, V. L.; Muzichenko, A. D.; Ivanov, B. V.; Sergeecheva, Ya. V.; Lesina, I. G.; Kovalenko, D. V.; Barsuk, V. A.; Danilina, N. A.; Bazylev, B. N.; Giniyatulin, R. N.

    2015-08-01

    PFMs (Plasma-facing materials: ITER grade stainless steel, beryllium, and ferritic-martensitic steels) as well as deposited erosion products of PFCs (Be-like, tungsten, and carbon based) were tested in QSPA under photonic heat loads relevant to those expected from photon radiation during disruptions mitigated by massive gas injection in ITER. Repeated pulses slightly above the melting threshold on the bulk materials eventually lead to a regular, "corrugated" surface, with hills and valleys spaced by 0.2-2 mm. The results indicate that hill growth (growth rate of ∼1 μm per pulse) and sample thinning in the valleys is a result of melt-layer redistribution. The measurements on the 316L(N)-IG indicate that the amount of tritium absorbed by the sample from the gas phase significantly increases with pulse number as well as the modified layer thickness. Repeated pulses significantly below the melting threshold on the deposited erosion products lead to a decrease of hydrogen isotopes trapped during the deposition of the eroded material.

  13. Plasma facing materials performance under ITER-relevant mitigated disruption photonic heat loads

    Energy Technology Data Exchange (ETDEWEB)

    Klimov, N.S., E-mail: klimov@triniti.ru [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoye shosse 31, Moscow 115409 (Russian Federation); Putrik, A.B. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Linke, J. [Forschungszentrum Jülich GmbH, EURATOM Association, Jülich D-52425 (Germany); Pitts, R.A. [Karlsruhe Institute of Technology, P.O. Box 3640, Karlsruhe 76021 (Germany); Zhitlukhin, A.M. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Kuprianov, I.B. [Bochvar Institute, ul. Rogova, 5a, Moscow 123098 (Russian Federation); Spitsyn, A.V. [NRC «Kurchatov Institute», Akademika Kurchatova pl., 1, Moscow 123182 (Russian Federation); Ogorodnikova, O.V. [Max-Planck-Institut für Plasmaphysik, Garching (Germany); National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoye shosse 31, Moscow 115409 (Russian Federation); Podkovyrov, V.L.; Muzichenko, A.D. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Ivanov, B.V.; Sergeecheva, Ya.V.; Lesina, I.G. [Bochvar Institute, ul. Rogova, 5a, Moscow 123098 (Russian Federation); Kovalenko, D.V.; Barsuk, V.A.; Danilina, N.A. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Bazylev, B.N. [Karlsruhe Institute of Technology, P.O. Box 3640, Karlsruhe 76021 (Germany); Giniyatulin, R.N. [Efremov Institute, Doroga na Metallostroy, 3 bld., Metallostroy, Saint-Petersburg 196641 (Russian Federation)

    2015-08-15

    PFMs (Plasma-facing materials: ITER grade stainless steel, beryllium, and ferritic–martensitic steels) as well as deposited erosion products of PFCs (Be-like, tungsten, and carbon based) were tested in QSPA under photonic heat loads relevant to those expected from photon radiation during disruptions mitigated by massive gas injection in ITER. Repeated pulses slightly above the melting threshold on the bulk materials eventually lead to a regular, “corrugated” surface, with hills and valleys spaced by 0.2–2 mm. The results indicate that hill growth (growth rate of ∼1 μm per pulse) and sample thinning in the valleys is a result of melt-layer redistribution. The measurements on the 316L(N)-IG indicate that the amount of tritium absorbed by the sample from the gas phase significantly increases with pulse number as well as the modified layer thickness. Repeated pulses significantly below the melting threshold on the deposited erosion products lead to a decrease of hydrogen isotopes trapped during the deposition of the eroded material.

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

    International Nuclear Information System (INIS)

    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 238 PuO 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 0 C. The module impacted the target at an angle of 30 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 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 μg) of 238 PuO 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

  15. Anti-alias filter in AORSA for modeling ICRF heating of DT plasmas in ITER

    Science.gov (United States)

    Berry, L. A.; Batchelor, D. B.; Jaeger, E. F.; RF SciDAC Team

    2011-10-01

    The spectral wave solver AORSA has been used extensively to model full-field, ICRF heating scenarios for DT plasmas in ITER. In these scenarios, the tritium (T) second harmonic cyclotron resonance is positioned near the magnetic axis, where fast magnetosonic waves are efficiently absorbed by tritium ions. In some cases, a fundamental deuterium (D) cyclotron layer can also be located within the plasma, but close to the high field boundary. In this case, the existence of multiple ion cyclotron resonances presents a serious challenge for numerical simulation because short-wavelength, mode-converted waves can be excited close to the plasma edge at the ion-ion hybrid layer. Although the left hand circularly polarized component of the wave field is partially shielded from the fundamental D resonance, some power penetrates, and a small fraction (typically LLC.

  16. Studies on representative disruption scenarios, associated electromagnetic and heat loads and operation window in ITER

    International Nuclear Information System (INIS)

    Fujieda, Hirobumi; Shimada, Michiya; Kawano, Yasunori; Ohmori, Junji; Neyatani, Yuzuru; Sugihara, Masayoshi; Gribov, Yuri; Ioki, Kimihiro; Khayrutdinov, Rustan; Lukash, Victor

    2007-07-01

    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 electromagnetic (EM) and heat loads on the in-vessel components and the vacuum vessel (VV). 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 (FEM) 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. It is found that the beryllium (Be) wall will not melt during the vertical movement. Significant melting is anticipated for the upper Be wall and tungsten divertor baffle due to the 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 (MGI). Some melting of the upper Be wall is anticipated at major disruptions (MD). 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 Be layer is expected to be within approx. = 30-100 μm/event out of 10 mm thick Be first wall. Various post processing programs of the results simulated with the DINA code, which are developed for the design work, are

  17. ITER SAFETY TASK NID-5D: Operational tritium loss and accident investigation for heat transport and water detritiation systems

    International Nuclear Information System (INIS)

    Kalyanam, K.M.; Fong, C.; Moledina, M.; Natalizio, A.

    1995-02-01

    The task objectives are to: a) determine major pathways for tritium loss during normal operation of the cooling systems and water detritiation system, b) estimate operational losses and environmental tritium releases from the heat transport and water detritiation systems of ITER, and c) prepare a preliminary Failure Modes and Effects Analysis (FMEA) for the ITER Water Detritiation System. The analysis will be used to estimate chronic environmental tritium releases (airborne and waterborne) for the ITER Cooling Systems and Water Detritiation System. The assessment will form the basis for demonstrating the acceptability of ITER for siting in the Early Safety and Environmental Characterization Study (ESECS), to be issued in early 1995. (author). 7 refs., 10 tabs., 11 figs

  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. Failure analysis of beryllium tile assembles following high heat flux testing for the ITER program

    International Nuclear Information System (INIS)

    Odegard, B.C. Jr.; Cadden, C. H.; Yang, N. Y. C.

    2000-01-01

    The following document describes the processing, testing and post-test analysis of two Be-Cu assemblies that have successfully met the heat load requirements for the first wall and dome sections for the ITER (International Thermonuclear Experimental Reactor) fusion reactor. Several different joint assemblies were evaluated in support of a manufacturing technology investigation aimed at diffusion bonding or brazing a beryllium armor tile to a copper alloy heat sink for fusion reactor applications. Judicious selection of materials and coatings for these assemblies was essential to eliminate or minimize interactions with the highly reactive beryllium armor material. A thin titanium layer was used as a diffusion barrier to isolate the copper heat sink from the beryllium armor. To reduce residual stresses produced by differences in the expansion coefficients between the beryllium and copper, a compliant layer of aluminum or aluminum-beryllium (AlBeMet-150) was used. Aluminum was chosen because it does not chemically react with, and exhibits limited volubility in, beryllium. Two bonding processes were used to produce the assemblies. The primary process was a diffusion bonding technique. In this case, undesirable metallurgical reactions were minimized by keeping the materials in a solid state throughout the fabrication cycle. The other process employed an aluminum-silicon layer as a brazing filler material. In both cases, a hot isostatic press (HIP) furnace was used in conjunction with vacuum-canned assemblies in order to minimize oxidation and provide sufficient pressure on the assemblies for full metal-to-metal contact and subsequent bonding. The two final assemblies were subjected to a suite of tests including: tensile tests and electron and optical metallography. Finally, high heat flux testing was conducted at the electron beam testing system (EBTS) at Sandia National Laboratories, New Mexico. Here, test mockups were fabricated and subjected to normal heat loads to

  20. Experimental evaluation of tritium permeation through stainless steel tubes of heat exchanger from primary to secondary water in ITER

    International Nuclear Information System (INIS)

    Nakamura, Hirofumi; Nishi, Masataka

    2004-01-01

    Tritium permeation through heat exchanger from primary cooling water to secondary cooling water has been investigated experimentally with SS316L heat exchanger under simulated ITER (international thermonuclear experimental reactor) operation condition in order to establish the tritium permeation evaluation method through the heat exchanger. As the result, the permeation rate of aqueous tritium was found to be about three orders smaller than that of the gaseous tritium. Tritium permeation through the heat exchanger in ITER was then evaluated, and it was revealed that total tritium permeation amount based on obtained aqueous permeability was about one order less than that with the former method with the gaseous permeability and putting the permeation reduction factor as 1000. Evaluated tritium permeation amount into secondary water during 20 years was quite small, which could be considered as negligible from the safety viewpoint

  1. Experimental investigation of ion cyclotron range of frequencies heating scenarios for ITER's half-field hydrogen phase performed in JET

    NARCIS (Netherlands)

    Lerche, E.; Van Eester, D.; Johnson, T. J.; Hellsten, T.; Ongena, J.; Mayoral, M. L.; Frigione, D.; Sozzi, C.; Calabro, G.; Lennholm, M.; Beaumont, P.; Blackman, T.; Brennan, D.; Brett, A.; Cecconello, M.; Coffey, I.; Coyne, A.; Crombe, K.; Czarnecka, A.; Felton, R.; Giroud, C.; Gorini, G.; Hellesen, C.; Jacquet, P.; Kiptily, V.; Knipe, S.; Krasilnikov, A.; Maslov, M.; Monakhov, I.; Noble, C.; Nocente, M.; Pangioni, L.; Proverbio, I.; Sergienko, G.; Stamp, M.; Studholme, W.; Tardocchi, M.; Vdovin, V.; Versloot, T.; Voitsekhovitch, I.; Whitehurst, A.; Wooldridge, E.; Zoita, V.; JET-EFDA Contributors,

    2012-01-01

    Two ion cyclotron range of frequencies (ICRF) heating schemes proposed for the half-field operation phase of ITER in hydrogen plasmas—fundamental H majority and second harmonic 3 He ICRF heating—were recently investigated in JET. Although the same magnetic field and RF frequencies ( f ≈ 42 MHz and f

  2. 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.

  3. Model validation of GAMMA code with heat transfer experiment for KO TBM in ITER

    International Nuclear Information System (INIS)

    Yum, Soo Been; Lee, Eo Hwak; Lee, Dong Won; Park, Goon Cherl

    2013-01-01

    Highlights: ► In this study, helium supplying system was constructed. ► Preparation for heat transfer experiment in KO TBM condition using helium supplying system was progressed. ► To get more applicable results, test matrix was made to cover the condition for KO TBM. ► Using CFD code; CFX 11, validation and modification for system code GAMMA was performed. -- Abstract: By considering the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He cooled molten lithium (HCML) test blanket module (TBM) for testing in ITER. A performance analysis for the thermal–hydraulics and a safety analysis for the KO TBM have been carried out using a commercial CFD code, ANSYS-CFX, and a system code, GAMMA (GAs multicomponent mixture analysis), which was developed by the gas cooled reactor in Korea. To verify the codes, a preliminary study was performed by Lee using a single TBM first wall (FW) mock-up made from the same material as the KO TBM, ferritic martensitic steel, using a 6 MPa nitrogen gas loop. The test was performed at pressures of 1.1, 1.9 and 2.9 MPa, and under various ranges of flow rate from 0.0105 to 0.0407 kg/s with a constant wall temperature condition. In the present study, a thermal–hydraulic test was performed with the newly constructed helium supplying system, in which the design pressure and temperature were 9 MPa and 500 °C, respectively. In the experiment, the same mock-up was used, and the test was performed under the conditions of 3 MPa pressure, 30 °C inlet temperature and 70 m/s helium velocity, which are almost same conditions of the KO TBM FW. One side of the mock-up was heated with a constant heat flux of 0.3–0.5 MW/m 2 using a graphite heating system, KoHLT-2 (Korea heat load test facility-2). Because the comparison result between CFX 11 and GAMMA showed a difference tendency, the modification of heat transfer correlation included in GAMMA was performed. And the modified GAMMA showed the strong parity with CFX

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

    International Nuclear Information System (INIS)

    Courtois, X.; Meunier, L.; Kuznetsov, V.; Beaumont, B.; Lamalle, P.; Conchon, D.; Languille, P.

    2016-01-01

    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.

  5. 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.

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

    International Nuclear Information System (INIS)

    Cicero, Tindaro; Jimenez, Marc; D’Amico, Gabriele; Pou, Jordi Ayneto; Dellopoulos, Georges; Alvaro, Elena; Cardenes, Sabas; Banetta, Stefano; Bellin, Boris; Zacchia, Francesco; Calcagno, Barbara; Chappuis, Philippe; Gicquel, Stefan; Mitteau, Raphael; Raffray, Rene

    2015-01-01

    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"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.

  7. Implementation of a quasi-realtime display of DIII-D neutral beam heating waveforms

    International Nuclear Information System (INIS)

    Phillips, J.C.

    1993-10-01

    The DIII-D neutral beam system employs eight 80 keV ion sources mounted on four beamlines to provide plasma heating to the DIII-D tokamak. The neutral beam system is capable of injecting over 20 MW of deuterium power with flexibility in terms of timing and modulation of the individual neutral beams. To maintain DIII-D's efficient tokamak shot cycle and make informed control decisions, it is important to be able to determine which beams fired, and exactly when, by the time the tokamak shot is over. Previously this information was available in centralized form only after a several minute wait. A cost-effective alternative to the traditional eight-channel storage oscilloscope has been implemented using off the shelf PC hardware and software. The system provides a real time display of injected neutral beam accelerator voltages and tokamak plasma current, as well an a summation waveform indicative of the total injected power as a function of time. The hardware consists of a Macintosh Centris 650 PC with a Motorola 68040 microprocessor. Data acquisition is accomplished using a National Instrument's 16-channel analog to digital conversion board for the Macintosh. The color displays and functionality were developed using National Instruments' LabView environment. Because the price of PCs has been decreasing rapidly and their capabilities increasing, this system is far less expensive than an eight-channel storage oscilloscope. As a flexible combination of PC and software, the system also provides much more capability than a dedicated oscilloscope, acting as the neutral beam coordinator's logbook, recording comments and availability statistics. Data such as shot number and neutral beam parameters are obtained over the local network from other computers and added to the display. Waveforms are easily archived to disk for future recall. Details of the implementation will be discussed along with samples of the displays and a description of the system's function and capabilities

  8. Implementation of a quasi-realtime display of DIII-D neutral beam heating waveforms

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, J.C.

    1993-10-01

    The DIII-D neutral beam system employs eight 80 keV ion sources mounted on four beamlines to provide plasma heating to the DIII-D tokamak. The neutral beam system is capable of injecting over 20 MW of deuterium power with flexibility in terms of timing and modulation of the individual neutral beams. To maintain DIII-D`s efficient tokamak shot cycle and make informed control decisions, it is important to be able to determine which beams fired, and exactly when, by the time the tokamak shot is over. Previously this information was available in centralized form only after a several minute wait. A cost-effective alternative to the traditional eight-channel storage oscilloscope has been implemented using off the shelf PC hardware and software. The system provides a real time display of injected neutral beam accelerator voltages and tokamak plasma current, as well an a summation waveform indicative of the total injected power as a function of time. The hardware consists of a Macintosh Centris 650 PC with a Motorola 68040 microprocessor. Data acquisition is accomplished using a National Instrument`s 16-channel analog to digital conversion board for the Macintosh. The color displays and functionality were developed using National Instruments` LabView environment. Because the price of PCs has been decreasing rapidly and their capabilities increasing, this system is far less expensive than an eight-channel storage oscilloscope. As a flexible combination of PC and software, the system also provides much more capability than a dedicated oscilloscope, acting as the neutral beam coordinator`s logbook, recording comments and availability statistics. Data such as shot number and neutral beam parameters are obtained over the local network from other computers and added to the display. Waveforms are easily archived to disk for future recall. Details of the implementation will be discussed along with samples of the displays and a description of the system`s function and capabilities.

  9. Confinement and βsub(p)-studies in neutral injection heated ASDEX plasmas

    International Nuclear Information System (INIS)

    Wagner, F.; Becker, G.; Behringer, K.; Campbell, D.; Eberhagen, A.; Engelhardt, W.; Fussmann, G.; Gehre, O.; Gernhardt, J.; Gierke, G. von; Haas, G.; Huang, M.; Karger, F.; Keilhacker, M.; Klueber, O.; Kornherr, M.; Lackner, K.; Lisitano, G.; Lister, G.G.; Mayer, H.M.; Meisel, D.; Mueller, E.R.; Murmann, H.; Niedermeyer, H.; Poschenrieder, W.; Rapp, H.; Roehr, H; Schneider, F.; Siller, G.; Speth, E.; Staebler, A.; Steuer, K.H.; Succi, S.; Venus, G.; Vollmer, O.

    1983-03-01

    Neutral injection experiments into limiter and divertor discharges in ASDEX are described with hydrogen and deuterium as working gas. Two operational regimes have been observed in neutral injection heated divertor discharges. One regime is characterized by deteriorated energy and particle confinement. The global energy confinement times are comparable to those of neutral injection heated limited discharges. Tthe other regime has particle and energy confinement times comparable to those of ohmic discharges with tausub(E) = 40 - 60 msec at beam powers up to 3.1 MW. This regime is further characterized by high βsub(p)-values comparable to the aspect ratio A (βsub(p) proportional 0.65 A), by good electron heating (etasub(e) proportional 2.5 x 10 13 eV cm -3 kW -1 ) and ion heating (etasub(i) proportional 4.2 x 10 13 eV cm -3 kW -1 ). In both regimes, tausub(E) increases with plasma current but there is hardly any variation with density. The differences in confinement and scaling to ohmic discharges seem to be caused by modifications of the electron loss channel. The high βsub(p)-regime develops at an injection power >=1.8 MW, anti nsub(e) >= 3 x 10 13 cm -3 and q >= 2.45, and is so far only observed in divertor discharges. There are indications that this may be due to the broad profiles with high edge temperatures which can develop in divertor discharges. An indication of broad current density profiles is given by the lack of sawtooth activity in these discharges. (orig.)

  10. MHD phenomena in a neutral beam heated high beta, low qa disruption

    International Nuclear Information System (INIS)

    Chu, M.S.; Greene, J.M.; Kim, J.S.; Lao, L.; Snider, R.T.; Stambaugh, R.D.; Strait, E.J.; Taylor, T.S.

    1988-01-01

    A neutral beam heated, β maximizing discharge at low q a in Doublet III ending in disruption is studied and correlated with theoretical models. This discharge achieved MHD β-values close to the theoretical Troyon-Sykes-Wesson limit in its evolution. The MHD phenomena of this discharge are analysed. The sequence of events leading to the high β disruptions is hypothesized as follows: the current and pressure profiles are broadened continuously by neutral beam injection. A last sawtooth internal disruption initiates an (m/n = 2/1) island through current profile steepening around the q=2 surface. The loss of plasma through stochastic field lines slows the island rotation and enhances its interaction with the limiter. The resultant enhanced island growth through island cooling or profile change enlarged the edge stochastic region. The overlapping of the edge stochastic region with the sawtooth mixing region precipitated the pressure disruption. Thus, in our hypothetical model for this discharge, β increase by neutral beam heating does not directly cause the disruption but ushers the plasma indirectly towards it through the profile broadening process and contributes to the destabilization of the 1/1 and 2/1 tearing modes. (author). 26 refs, 12 figs

  11. Use of the TFTR prototype charge exchange neutral analyzer for fast He3++ diagnostics during ICRF heating on PLT

    International Nuclear Information System (INIS)

    Medley, S.S.

    1981-07-01

    The Charge Exchange Neutral Analyzer (CENA) for TFTR is designed to measure singly charged ion species of atomic mass A = 1, 2, and 3 simultaneously with up to 75 energy channels per mass and an energy range of 0.5 3 charge exchange neutrals makes the analyzer of particular interest for recently proposed fast He 3 ++ diagnostics during ICRF heating on PLT

  12. Some estimates of mirror plasma startup by neutral beam heating of pellet and gas cloud targets

    International Nuclear Information System (INIS)

    Shearer, J.W.; Willmann, P.A.

    1978-01-01

    Hot plasma buildup by neutral beam injection into an initially cold solid or gaseous target is found to be conceivable in large mirror machine experiments such as 2XIIB or MFTF. A simple analysis shows that existing neutral beam intensities are sufficient to ablate suitable targets to form a gas or vapor cloud. An approximate rate equation model is used to follow the subsequent processes of ionization, heating, and hot plasma formation. Solutions of these rate equations are obtained by means of the ''GEAR'' techniques for solving ''stiff'' systems of differential equations. These solutions are in rough agreement with the 2XIIB stream plasma buildup experiment. They also predict that buildup on a suitable nitrogen-like target will occur in the MFTF geometry. In 2XIIB the solutions are marginal; buildup may be possible, but is not certain

  13. ITER EDA newsletter. V. 5, no. 2

    International Nuclear Information System (INIS)

    1996-02-01

    This issue of the Newsletter reports on the First Technical Meeting on ITER Neutral Beam Injection held at Naka, Japan, 13-16 February 1996, and on the First Workshop on Energetic Particles and Heating and Current Drive held at the Kurchatov Institute, Moscow, Russia, on October 2 - 6, 1995

  14. A scoping study of the application of neutral beam heating on the TCV tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Karpushov, Alexander N., E-mail: alexander.karpushov@epfl.ch [Ecole Polytechnique Federale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, Association Euratom-Confederation Suisse, CH-1015 Lausanne (Switzerland); Duval, Basil P.; Chavan, Rene [Ecole Polytechnique Federale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, Association Euratom-Confederation Suisse, CH-1015 Lausanne (Switzerland); Fable, Emiliano [Max-Planck-Institut fuer Plasmaphysik, Euratom-IPP Association, Boltzmannstrasse 2, D-85748 Garching (Germany); Mayor, Jean-Michel; Sauter, Olivier; Weisen, Henri [Ecole Polytechnique Federale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, Association Euratom-Confederation Suisse, CH-1015 Lausanne (Switzerland)

    2011-10-15

    The TCV tokamak contributes to the physics understanding of fusion plasmas, broadening the parameter range of reactor relevant regimes, by investigations based on an extensive use of the existing main experimental tools: flexible shaping and high power real time-controllable electron cyclotron heating (ECH) and current drive (ECCD) systems. A proposed implementation of direct ion heating on the TCV by the installation of a 20-35 keV neutral beam injection (NBI) with a total power of 1-3 MW would permit an extension of the accessible range of ion to electron temperatures (T{sub i}/T{sub e} {approx} 0.1-0.8) to well beyond unity, depending on the NBI/ECH mix and the plasma density. A NBI system would provide TCV with a tool for plasma study at reactor relevant T{sub i}/T{sub e} ratios {approx}1 and in investigating fast ion and MHD physics together with the effects of plasma rotation and high plasma {beta} scenarios. The feasibility studies for a NBI heating on TCV presented in this paper were undertaken to construct a specification for the neutral beam injectors together with an experimental geometry for possible operational scenarios.

  15. Heat flux to the limiter during disruptions and neutral beam injection in Doublet-III

    International Nuclear Information System (INIS)

    Hino, T.; DeGrassie, J.; Taylor, T.S.; Hopkins, G.; Meyer, C.; Petrie, T.W.; Kahn, C.L.; Ejima, S.

    1984-01-01

    The heat flux to the Doublet-III primary limiter has been monitored during plasma disruptions and during neutral beam injection. The surface temperature of the movable TiC-coated graphite limiter was measured with an Inframetrics thermal imaging system and a suitably filtered silicon photodiode spot detector. In addition, the floating electric potential of the limiter with respect to the vacuum vessel was measured. The heat pulse duration to the limiter was measured by the spot detector with a time response of x approx.= 10 μs and these times were correlated with the plasma parameters. In limiter discharges, 20% of the plasma kinetic stored energy goes to the limiter during disruptions. The power balance during disruptions is also discussed. During neutral beam injection, the limiter is not heated uniformly; the ion drift side receives much more thermal flux than the electron drift side. The fraction of beam power going to the limiter is as high as approx.= 35% in normal limiter discharges. (orig.)

  16. Simulation of tokamak armour erosion and plasma contamination at intense transient heat fluxes in ITER

    Science.gov (United States)

    Landman, I. S.; Bazylev, B. N.; Garkusha, I. E.; Loarte, A.; Pestchanyi, S. E.; Safronov, V. M.

    2005-03-01

    For ITER, the potential material damage of plasma facing tungsten-, CFC-, or beryllium components during transient processes such as ELMs or mitigated disruptions are simulated numerically using the MHD code FOREV-2D and the melt motion code MEMOS-1.5D for a heat deposition in the range of 0.5-3 MJ/m 2 on the time scale of 0.1-1 ms. Such loads can cause significant evaporation at the target surface and a contamination of the SOL by the ions of evaporated material. Results are presented on carbon plasma dynamics in toroidal geometry and on radiation fluxes from the SOL carbon ions obtained with FOREV-2D. The validation of MEMOS-1.5D against the plasma gun tokamak simulators MK-200UG and QSPA-Kh50, based on the tungsten melting threshold, is described. Simulations with MEMOS-1.5D for a beryllium first wall that provide important details about the melt motion dynamics and typical features of the damage are reported.

  17. Lower hybrid heating and current drive in Iter operation scenarios and outline system design

    International Nuclear Information System (INIS)

    1994-11-01

    Lower Hybrid Waves (LHW) are considered a valid method of plasma heating and the best demonstrated current drive method. Current drive by LHW possesses the unique feature, as compared to the other methods, to retain a good current drive efficiency in plasma regions of low to medium temperature, or in low-β phases of the discharges. This makes them an essential element to realize the so called 'advanced steady-state Tokamak scenarios' in which a hollow current density profile (deep shear reversal) - established during the ramp-up of the plasma current - offers the prospects of improved confinement and an MHD-stable route to continuous burn. This report contains both modelling and design studies of an LHW system for ITER. It aims primarily at the definition of concepts and parameters for steady-state operation using LHW combined with Fast Waves (FW), or other methods of generating a central seed current for high bootstrap current operation. However simulations addressing the use of LHW for current profile control in the high current pulsed operation scenario are also presented. The outline design of a LHW system which covers the needs for both pulsed and steady-state operation is described in detail. (author). 28 refs., 49 figs

  18. Heating efficiency of high-power perpendicular neutral-beam injection in PDX

    International Nuclear Information System (INIS)

    Hawryluk, R.J.; Arunasalam, V.; Bell, M.

    1982-03-01

    The heating efficiency of high power (up to 7.2 MW) near-perpendicular neutral beam injection in the PDX tokamak is comparable to that of tangential injection in PLT. Collisionless plasmas with central ion temperatures up to 6.5 keV and central electron temperatures greater than 2.5 keV have been obtained. The plasma pressure, including the contribution from the beam particles, increases with increasing beam power and does not appear to saturate, although the parametric dependence of the energy confinement time is different from that observed in ohmic discharges

  19. High-power ICRF and ICRF plus neutral-beam heating on PLT

    International Nuclear Information System (INIS)

    Hwang, D.; Bitter, M.; Budny, R.

    1983-01-01

    PLT ICRF experiments with RF powers up to approx.=3 MW have demonstrated efficient plasma heating in both the minority fundamental and the second harmonic ion-cyclotron regimes. In the minority 3 He regime, ion temperatures of approx.=3 keV have been produced along with approx.=1 kW of D- 3 He fusion power and substantial electron heating. In the second harmonic H regime, an equivalent averaged ion energy of approx.=4 keV has been achieved. Combined ICRF plus neutral-beam heating experiments with auxiliary powers totalling up to 4.5 MW have provided insight into auxiliary heating performance at stored plasma energy levels up to approx.=100 kJ. Values of #betta#sub(phi) in the range of 1.5-2% have been attained for Bsub(phi) approx.=17 kG. Energetic discharges with n-barsub(e) up to approx.6x10 13 cm - 3 at Bsub(phi) approx.=28 kG have also been investigated. Preliminary confinement studies suggest that energetic ion losses may contribute to a direct loss of the input RF power in the H minority heating regime but are insignificant in the 3 He minority case. The energy confinement time for the H minority regime is reduced somewhat from the Ohmic value. (author)

  20. Numerical investigation of heat transfer enhancement in ribbed channel for the first wall of DFLL-TBM in ITER

    International Nuclear Information System (INIS)

    Jin Qiang; Liu Songlin; Li Min; Wang Weihua

    2012-01-01

    As an important component of Dual Functional Lithium Lead-Test Blanket Module (DFLL-TBM), the first wall (FW) must withstand and remove the heat flux from the plasma (q″ = 0.3 MW/m 2 ) and high nuclear power deposited in the structure at normal plasma operation scenario of ITER. In this paper the transverse ribs arranged along the plasma facing inner wall surface were used to enhance the heat transfer capability. After the validation compared with empirical correlations the Standard k–ω model was employed to do the numerical simulation using FLUENT code to investigate the heat transfer efficiency and flow performance of coolant in the ribbed channel preliminarily. The perforation on the bottom of rib was proposed near the lower heat transfer area (LHTA) to improve the heat transfer performance according to results of analyses.

  1. Acceleration of calculation of nuclear heating distributions in ITER toroidal field coils using hybrid Monte Carlo/deterministic techniques

    International Nuclear Information System (INIS)

    Ibrahim, Ahmad M.; Polunovskiy, Eduard; Loughlin, Michael J.; Grove, Robert E.; Sawan, Mohamed E.

    2016-01-01

    Highlights: • Assess the detailed distribution of the nuclear heating among the components of the ITER toroidal field coils. • Utilize the FW-CADIS method to dramatically accelerate the calculation of detailed nuclear analysis. • Compare the efficiency and reliability of the FW-CADIS method and the MCNP weight window generator. - Abstract: Because the superconductivity of the ITER toroidal field coils (TFC) must be protected against local overheating, detailed spatial distribution of the TFC nuclear heating is needed to assess the acceptability of the designs of the blanket, vacuum vessel (VV), and VV thermal shield. Accurate Monte Carlo calculations of the distributions of the TFC nuclear heating are challenged by the small volumes of the tally segmentations and by the thick layers of shielding provided by the blanket and VV. To speed up the MCNP calculation of the nuclear heating distribution in different segments of the coil casing, ground insulation, and winding packs of the ITER TFC, the ITER Organization (IO) used the MCNP weight window generator (WWG). The maximum relative uncertainty of the tallies in this calculation was 82.7%. In this work, this MCNP calculation was repeated using variance reduction parameters generated by the Oak Ridge National Laboratory AutomateD VAriaNce reducTion Generator (ADVANTG) code and both MCNP calculations were compared in terms of computational efficiency and reliability. Even though the ADVANTG MCNP calculation used less than one-sixth of the computational resources of the IO calculation, the relative uncertainties of all the tallies in the ADVANTG MCNP calculation were less than 6.1%. The nuclear heating results of the two calculations were significantly different by factors between 1.5 and 2.3 in some of the segments of the furthest winding pack turn from the plasma neutron source. Even though the nuclear heating in this turn may not affect the ITER design because it is much smaller than the nuclear heating in the

  2. Acceleration of calculation of nuclear heating distributions in ITER toroidal field coils using hybrid Monte Carlo/deterministic techniques

    Energy Technology Data Exchange (ETDEWEB)

    Ibrahim, Ahmad M., E-mail: ibrahimam@ornl.gov [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Polunovskiy, Eduard; Loughlin, Michael J. [ITER Organization, Route de Vinon Sur Verdon, 13067 St. Paul Lez Durance (France); Grove, Robert E. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Sawan, Mohamed E. [University of Wisconsin-Madison, 1500 Engineering Dr., Madison, WI 53706 (United States)

    2016-11-01

    Highlights: • Assess the detailed distribution of the nuclear heating among the components of the ITER toroidal field coils. • Utilize the FW-CADIS method to dramatically accelerate the calculation of detailed nuclear analysis. • Compare the efficiency and reliability of the FW-CADIS method and the MCNP weight window generator. - Abstract: Because the superconductivity of the ITER toroidal field coils (TFC) must be protected against local overheating, detailed spatial distribution of the TFC nuclear heating is needed to assess the acceptability of the designs of the blanket, vacuum vessel (VV), and VV thermal shield. Accurate Monte Carlo calculations of the distributions of the TFC nuclear heating are challenged by the small volumes of the tally segmentations and by the thick layers of shielding provided by the blanket and VV. To speed up the MCNP calculation of the nuclear heating distribution in different segments of the coil casing, ground insulation, and winding packs of the ITER TFC, the ITER Organization (IO) used the MCNP weight window generator (WWG). The maximum relative uncertainty of the tallies in this calculation was 82.7%. In this work, this MCNP calculation was repeated using variance reduction parameters generated by the Oak Ridge National Laboratory AutomateD VAriaNce reducTion Generator (ADVANTG) code and both MCNP calculations were compared in terms of computational efficiency and reliability. Even though the ADVANTG MCNP calculation used less than one-sixth of the computational resources of the IO calculation, the relative uncertainties of all the tallies in the ADVANTG MCNP calculation were less than 6.1%. The nuclear heating results of the two calculations were significantly different by factors between 1.5 and 2.3 in some of the segments of the furthest winding pack turn from the plasma neutron source. Even though the nuclear heating in this turn may not affect the ITER design because it is much smaller than the nuclear heating in the

  3. Performance of cable-in-conduit conductors in ITER [International Thermonuclear Experimental Reactor] toroidal field coils with varying heat loads

    International Nuclear Information System (INIS)

    Kerns, J.A.; Wong, R.L.

    1989-01-01

    The toroidal field (TF) coils in the International Thermonuclear Experimental Reactor (ITER) will operate with varying heat loads generated by ac losses and nuclear heating. The total heat load is estimated to be 2 kW per TF coil under normal operation and can be higher for different operating scenarios. Ac losses are caused by ramping the poloidal field (PF) for plasma initiation, burn, and shutdown; nuclear heating results from neutrons that penetrate into the coil past the shield. Present methods to reduce or eliminate these losses lead to larger and more expensive machines, which are unacceptable with today's budget constraints. A suitable solution is to design superconductors that operate with high heat loads. The cable-in-conduit conductor (CICC) can operate with high heat loads. One CICC design is analyzed for its thermal performance using two computer codes developed at LLNL. One code calculates the steady state flow conditions along the flow path, while the other calculates the transient conditions in the flow. We have used these codes to analyze the superconductor performance during the burn phase of the ITER plasma. The results of these analyses give insight to the choice of flow rate on superconductor performance. 4 refs., 5 figs

  4. ITER central solenoid model coil heat treatment complete and assembly started

    International Nuclear Information System (INIS)

    Thome, R.J.; Okuno, K.

    1998-01-01

    A major R and D task in the ITER program is to fabricate a Superconducting Model Coil for the Central Solenoid to establish the design and fabrication methods for ITER size coils and to demonstrate conductor performance. Completion of its components is expected in 1998, to be followed by assembly with structural components and testing in a facility at JAERI

  5. Attainment of high confinement in neutral beam heated divertor discharges in the PDX tokamak

    International Nuclear Information System (INIS)

    Kaye, S.M.; Bell, M.; Bol, K.

    1983-11-01

    The PDX divertor configuration has recently been converted from an open to a closed geometry to inhibit the return of neutral gas from the divertor region to the main chamber. Since then, operation in a regime with high energy confinement in neutral beam heated discharges (ASDEX H-mode) has been routine over a wide range of operating conditions. These H-mode discharges are characterized by a sudden drop in divertor density and H/sub α/ emission and a spontaneous rise in main chamber plasma density during neutral beam injection. The confinement time is found to scale nearly linearly with plasma current, but it can be degraded due to either the presence of edge instabilities or heavy gas puffing. Detailed Thomson scattering temperature profiles show high values of Te near the plasma edge (approx. 450 eV) with sharp radial gradients (approx. 400 eV/cm) near the separatrix. Density profiles are broad and also exhibit steep gradients close to the separatrix

  6. Observation of Beam Driven Modes during Neutral Beam Heating on the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Fredrickson, E.D.; Gorelenkov, E.D.; Cheng, C.Z.; Bell, R.; Darrow, D.; Johnson, D.; Kaye, S.; LeBlanc, B.; Menard, J.; Kubota, S.; Peebles, W.

    2001-01-01

    With the first injection of neutral beams on the National Spherical Torus Experiment (NSTX), a broad and complicated spectrum of coherent modes was seen between approximately 0.4 MHz and 2.5 MHz [where f(subscript ''ci'')] for deuterium is approximately 2.2 MHz. The modes have been observed with high bandwidth magnetic pick-up coils and with a reflectometer. The parametric scaling of the mode frequency with density and magnetic field is consistent with Alfvenic modes (linear in B, inversely with the square root of density). These modes have been identified as magnetosonic waves or compressional Alfven eigenmodes (CAE) excited by a cyclotron resonance with the neutral-beam ions. Modes have also been observed in the frequency range 50-150 kHz with toroidal mode numbers n = 1-5. These lower frequency modes are thought to be related to the TAE [Toroidal Alfven Eigenmode] seen commonly in tokamaks and driven by energetic fast ion populations resulting from ICRF [ion cyclotron range of frequency] and NBI [neutral-beam injection] heating. There is no clear indication of enhanced fast ion losses associated with the modes

  7. Reconstructive schemes for variational iteration method within Yang-Laplace transform with application to fractal heat conduction problem

    Directory of Open Access Journals (Sweden)

    Liu Chun-Feng

    2013-01-01

    Full Text Available A reconstructive scheme for variational iteration method using the Yang-Laplace transform is proposed and developed with the Yang-Laplace transform. The identification of fractal Lagrange multiplier is investigated by the Yang-Laplace transform. The method is exemplified by a fractal heat conduction equation with local fractional derivative. The results developed are valid for a compact solution domain with high accuracy.

  8. Transmission of the Neutral Beam Heating Beams at TJ-II

    International Nuclear Information System (INIS)

    Fuentes Lopez, C.

    2007-01-01

    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

  9. Progress of neutral beam R and D for plasma heating and current drive at JAERI

    International Nuclear Information System (INIS)

    Ohara, Y.

    1995-01-01

    Recent progress and future plans regarding development of a high power negative ion source at the Japan Atomic Energy Research Institute (JAERI) are described. The neutral beam injection system, which is expected to play an important role not only in plasma heating but also in the plasma current drive in the fusion reactor, requires a high power negative ion source which can produce negative deuterium ion beams with current of order 20A at energy above 1MeV. In order to realize such a high power negative ion beam, intensive research and development has been carried out at JAERI since 1984. The negative hydrogen ion beam current of 10A achieved in recent years almost equals the value required for the fusion reactor. With regard to the negative ion acceleration, a high current negative ion beam of 0.2A has been accelerated up to 350keV electrostatically. On the basis of this recent progress, two development plans have been initiated as an intermediate step towards the fusion reactor. One is to develop a 500keV, 10MW negative ion based neutral beam injection system for JT-60U to demonstrate the neutral beam current drive in a high density plasma. The other is to develop a 1MeV, 1A ion source to demonstrate high current negative ion acceleration up to 1MeV. On the basis of this research and development, an efficient and reactor relevant neutral beam injection system will be developed for an experimental fusion reactor such as the International Thermonuclear Experimental Reactor. ((orig.))

  10. Design of TFTR movable limiter blades for ohmic and neutral-beam-heated plasmas

    International Nuclear Information System (INIS)

    Doll, D.W.; Ulrickson, M.A.; Cecchi, J.L.; Citrolo, J.C.; Weissenburger, D.; Bialek, J.

    1981-10-01

    A new set of movable limiter blades has been designed for TFTR that will meet both the requirements of the 4 MW ohmic heated and the 33 MW neutral beam heated plasmas. This is accomplished with three limiter blades each having and elliptical shape along the toroidal direction. Heat flux levels are acceptable for both ohmic heated and pre-strong compression plasmas. The construction consists of graphite tiles attached to cooled backing plates. The tiles have an average thickness of approx. 4.7 cm and are drawn against the backing plate with spring loaded fasteners that are keyed into the graphite. The cooled backing plate provides the structure for resisting disruption and fault induced loads. A set of rollers attached to the top and bottom blades allow them to be expanded and closed in order to vary the plasma surface for scaling experiments. Water cooling lines penetrate only the mid-plane port cover/support plate in such a way as to avoid bolted water connections inside the vacuum boundary and at the same time allow blade movement. Both the upper and lower blades are attached to the mid-plane limiter blade through pivots. Pivot connections are protected against arcing with an alumina coating and a shunt bar strap. Remote handling is considered throughout the design

  11. Radiation level analysis for the port cell of the ITER electron cyclotron-heating upper launcher

    Energy Technology Data Exchange (ETDEWEB)

    Weinhorst, Bastian, E-mail: bastian.weinhorst@kit.edu [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Fischer, Ulrich; Lu, Lei [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Strauss, Dirk; Spaeh, Peter; Scherer, Theo [KIT, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Leichtle, Dieter [F4E, Analysis & Codes/Technical Support Services, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain)

    2016-11-01

    Highlights: • First detailed neutronic modelling of the ECHUL port cell with ECHUL equipment (including beam lines with diamond windows, the beam lines mounting box, conduit boxes and rails). • Three different bioshield port plug configurations and two different neutron source configurations are investigated. • Radiation Levels are calculated in the port cell, focusing on the position of the diamond window. • The dose rate in the port cell is below the limit for maintenance in the port cell. • The radiation level at the diamond window is very low and should not influence its performance. - Abstract: The electron cyclotron-heating upper launcher (ECHUL) will be installed in four upper ports of the ITER tokamak thermonuclear fusion reactor. Each ECHUL is able to deposit 8 MW power into the plasma for plasma mode stabilization via microwave beam lines. An essential part of these beam lines are the diamond windows. They are located in the upper port cell behind the bioshield to reduce the radiation levels to a minimum. The paper describes the first detailed neutronic modelling of the ECHUL port cell with ECHUL equipment. The bioshield plug is modelled including passageways for the microwave beam lines, piping and cables looms as well as rails and openings for ventilation. The port cell is equipped with the beam lines including the diamond windows, the beam lines mounting box, conduit boxes and rails. The neutrons are transported into the port cell starting from a surface source in front of the bioshield. Neutronic results are obtained for radiation levels in the port cell at different positions, mainly focusing on the diamond windows position. It is shown that the radiation level is below the limit for maintenance in the port cell. The radiation level at the diamond window is very low and should not influence its performance.

  12. Divertor development for ITER

    International Nuclear Information System (INIS)

    Janeschitz, G.; Ando, T.; Antipenkov, A.; Barabash, V.; Chiocchio, S.; Federici, G.; Ibbott, C.; Jakeman, R.; Matera, R.; Martin, E.; Parker, R.; Tivey, R.; Pacher, H.D.

    1998-01-01

    The requirements for the ITER divertor design, i.e. power and He ash exhaust, neutral leakage control, lifetime, disruption load resistance and exchange by remote handling, are described in this paper. These requirements and the physics requirements for detached and semi-attached operation result in the vertical target configuration. This is realised by a concept incorporating 60 cassettes carrying the high heat flux components. The armour choice for these components is CFC monoblock in the strike zone near at the lower part of the vertical target, and a W brush elsewhere. Cooling is by swirl tubes or hypervapotrons depending on the component. The status of the heat sink and joining technology R and D is given. Finally, the resulting design of the high heat flux components is presented. (orig.)

  13. Mechanical design of the ITER ion cyclotron heating launcher based on in-vessel tuning system

    Energy Technology Data Exchange (ETDEWEB)

    Vulliez, K. [Association Euratom-CEA, CEA/DSM/DRFC, CEA Cadarache, F-13108 St Paul Lez Durance (France)], E-mail: karl.vulliez@cea.fr; Bosia, G. [Dipartimento di Fisica Generale, Universita di Torino (Italy); Agarici, G.; Beaumont, B.; Argouarch, A.; Mollard, P. [Association Euratom-CEA, CEA/DSM/DRFC, CEA Cadarache, F-13108 St Paul Lez Durance (France); Testoni, P. [Electrical and Electronics Engineering Department, University of Cagliari (Italy); Maggiora, R.; Milanesio, D. [Dipartimento di Elettronica Politecnico di Torino (Italy)

    2007-10-15

    Since the release of the ITER ICRH system reference design report [ITER Final Design Report: DDD 5.1 -Ion Cyclotron and Current Drive System, July 2001], further design studies have been conducted. If the base of the reference design [Final Report on EFDA contract 04/1129, ITER ICRF antenna and Matching system design (Internalmatching), April 2005] is kept unchanged, several significant modifications have been proposed for a better efficiency and reliability. The increase of the poloidal order of the array and strong modifications of the matching system concept are the main changes. Technical aspects insufficiently covered in previous studies are also now worked out in detail, like the integration on a mid-plane port satisfying the constraints of the ITER environment.

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

    International Nuclear Information System (INIS)

    Son, Hong Hyun; Jeong, Ui Ju; Seo, Gwang Hyeok; Jeun, Gyoo Dong; Kim, Sung Joong

    2016-01-01

    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.

  15. Results and analysis of high heat flux tests on a full scale vertical target prototype of ITER divertor

    International Nuclear Information System (INIS)

    Missirlian, M.; Escourbiac, F.; Schlosser, J.; Durocher, A.; Bobin-Vastra, I.

    2004-01-01

    After an extensive development program, a Full-Scale Divertor Target prototype (VTFS) manufactured with all the main features of the corresponding ITER divertor, was intensively tested in the high heat flux FE200 facility. The prototype consists of four units having a full mono-block geometry. The lower part (CFC armour) and the upper part (W armour) of each mono-block were joined to the solution annealed, quenched and cold worked CuCrZr tube by HIP technique. The CFC mono-block was successfully tested up to 1000 cycles at 23 MW/m 2 without any indication of failure. This value is well beyond the ITER design target of 300 cycles at 20 MW/m 2 . The W mono-block endured ∼600 cycles at 10 MW/m 2 . This value of flux is one order of magnitude higher than the ITER design target for the upper part of the vertical target. Fatigue damage is observed when pursuing the cycling up to 15 MW/m 2 . A first stress analysis seems to predict these factual results. However, macro-graphic examinations should bring a better damage valuation. Meanwhile, the fatigue testing will continue on the W healthy part of the VTFS prototype with castellation located on the heated surface (reducing the stresses close to the W-Cu interface). (authors)

  16. Results and analysis of high heat flux tests on a full scale vertical target prototype of ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Missirlian, M.; Escourbiac, F.; Schlosser, J.; Durocher, A. [Association Euratom-CEA, Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Merola, M. [EFDA Close Support Unit, Garching (Germany); Bobin-Vastra, I. [Framatome, 71 - Le Creusot (France)

    2004-07-01

    After an extensive development program, a Full-Scale Divertor Target prototype (VTFS) manufactured with all the main features of the corresponding ITER divertor, was intensively tested in the high heat flux FE200 facility. The prototype consists of four units having a full mono-block geometry. The lower part (CFC armour) and the upper part (W armour) of each mono-block were joined to the solution annealed, quenched and cold worked CuCrZr tube by HIP technique. The CFC mono-block was successfully tested up to 1000 cycles at 23 MW/m{sup 2} without any indication of failure. This value is well beyond the ITER design target of 300 cycles at 20 MW/m{sup 2}. The W mono-block endured {approx}600 cycles at 10 MW/m{sup 2}. This value of flux is one order of magnitude higher than the ITER design target for the upper part of the vertical target. Fatigue damage is observed when pursuing the cycling up to 15 MW/m{sup 2}. A first stress analysis seems to predict these factual results. However, macro-graphic examinations should bring a better damage valuation. Meanwhile, the fatigue testing will continue on the W healthy part of the VTFS prototype with castellation located on the heated surface (reducing the stresses close to the W-Cu interface). (authors)

  17. 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.

  18. Global energy confinement in JT-60 neutral beam heated L-mode discharges

    International Nuclear Information System (INIS)

    Naito, O.; Hosogane, N.; Tsuji, S.; Ushigusa, K.; Yoshida, H.

    1990-01-01

    The global energy confinement characteristics of neutral beam heated JT-60 discharges are presented. There is a difference in the dependence of the energy confinement time on the plasma current between limiter and divertor discharges. For limiter discharges, the energy confinement increases with plasma current up to 3.2 MA, whereas for divertor discharges this improvement saturates when the safety factor drops below 3, independent of the location of the X-point. The JT-60 L-mode results indicate that the deterioration in energy confinement for q < 3, which is also found in H-mode regimes of other devices, may be a universal characteristic of divertor discharges. Regarding the scaling with plasma size, it is shown that the global/incremental confinement time increases with plasma minor radius. For sufficiently large plasmas, however, the global/incremental confinement time is no longer a function of minor radius. (author). 13 refs, 14 figs

  19. 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.

  20. 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

  1. The ITER tritium systems

    International Nuclear Information System (INIS)

    Glugla, M.; Antipenkov, A.; Beloglazov, S.; Caldwell-Nichols, C.; Cristescu, I.R.; Cristescu, I.; Day, C.; Doerr, L.; Girard, J.-P.; Tada, E.

    2007-01-01

    ITER is the first fusion machine fully designed for operation with equimolar deuterium-tritium mixtures. The tokamak vessel will be fuelled through gas puffing and pellet injection, and the Neutral Beam heating system will introduce deuterium into the machine. Employing deuterium and tritium as fusion fuel will cause alpha heating of the plasma and will eventually provide energy. Due to the small burn-up fraction in the vacuum vessel a closed deuterium-tritium loop is required, along with all the auxiliary systems necessary for the safe handling of tritium. The ITER inner fuel cycle systems are designed to process considerable and unprecedented deuterium-tritium flow rates with high flexibility and reliability. High decontamination factors for effluent and release streams and low tritium inventories in all systems are needed to minimize chronic and accidental emissions. A multiple barrier concept assures the confinement of tritium within its respective processing components; atmosphere and vent detritiation systems are essential elements in this concept. Not only the interfaces between the primary fuel cycle systems - being procured through different Participant Teams - but also those to confinement systems such as Atmosphere Detritiation or those to fuelling and pumping - again procured through different Participant Teams - and interfaces to buildings are calling for definition and for detailed analysis to assure proper design integration. Considering the complexity of the ITER Tritium Plant configuration management and interface control will be a challenging task

  2. Transient heat loads in current fusion experiments, extrapolation to ITER and consequences for its operation

    International Nuclear Information System (INIS)

    Loarte, A; Saibene, G; Sartori, R; Riccardo, V; Andrew, P; Paley, J; Fundamenski, W; Eich, T; Herrmann, A; Pautasso, G; Kirk, A; Counsell, G; Federici, G; Strohmayer, G; Whyte, D; Leonard, A; Pitts, R A; Landman, I; Bazylev, B; Pestchanyi, S

    2007-01-01

    New experimental results on transient loads during ELMs and disruptions in present divertor tokamaks are described and used to carry out a extrapolation to ITER reference conditions and to draw consequences for its operation. In particular, the achievement of low energy/convective type I edge localized modes (ELMs) in ITER-like plasma conditions seems the only way to obtain transient loads which may be compatible with an acceptable erosion lifetime of plasma facing components (PFCs) in ITER. Power loads during disruptions, on the contrary, seem to lead in most cases to an acceptable divertor lifetime because of the relatively small plasma thermal energy remaining at the thermal quench and the large broadening of the power flux footprint during this phase. These conclusions are reinforced by calculations of the expected erosion lifetime, under these load conditions, which take into account a realistic temporal dependence of the power fluxes on PFCs during ELMs and disruptions

  3. Accounting of the Power Balance for Neutral-beam heated H-Mode Plasmas in NSTX

    International Nuclear Information System (INIS)

    Paul, S.F.; Maingi, R.; Soukhanovskii, V.; Kaye, S.M.; Kugel, H.

    2004-01-01

    A survey of the dependence of power balance on input power, shape, and plasma current was conducted for neutral-beam-heated plasmas in the National Spherical Torus Experiment (NSTX). Measurements of heat to the divertor strike plates and divertor and core radiation were taken over a wide range of plasma conditions. The different conditions were obtained by inducing a L-mode to H-mode transition, changing the divertor configuration [lower single null (LSN) vs. double-null (DND)] and conducting a NBI power scan in H-mode. 60-70% of the net input power is accounted for in the LSN discharges with 20% of power lost as fast ions, 30-45% incident on the divertor plates, up to 10% radiated in the core, and about 12% radiated in the divertor. In contrast, the power accountability in DND is 85-90%. A comparison of DND and LSN data show that the remaining power in the LSN is likely to be directed to the upper divertor

  4. Electron temperature profiles in high power neutral-beam-heated TFTR [Tokamak Fusion Test Reactor] plasmas

    International Nuclear Information System (INIS)

    Taylor, G.; Grek, B.; Stauffer, F.J.; Goldston, R.J.; Fredrickson, E.D.; Wieland, R.M.; Zarnstorff, M.C.

    1987-09-01

    In 1986, the maximum neutral beam injection (NBI) power in the Tokamak Fusion Test Reactor (TFTR) was increased to 20 MW, with three beams co-parallel and one counter-parallel to I/sub p/. TFTR was operated over a wide range of plasma parameters; 2.5 19 19 m -3 . Data bases have been constructed with over 600 measured electron temperature profiles from multipoint TV Thomson scattering which span much of this parameter space. We have also examined electron temperature profile shapes from electron cyclotron emission at the fundamental ordinary mode and second harmonic extraordinary mode for a subset of these discharges. In the light of recent work on ''profile consistency'' we have analyzed these temperature profiles in the range 0.3 < (r/a) < 0.9 to determine if a profile shape exists which is insensitive to q/sub cyl/ and beam-heating profile. Data from both sides of the temperature profile [T/sub e/(R)] were mapped to magnetic flux surfaces [T/sub e/(r/a)]. Although T/sub e/(r/a), in the region where 0.3 < r/a < 0.9 was found to be slightly broader at lower q/sub cyl/, it was found to be remarkably insensitive to β/sub p/, to the fraction of NBI power injected co-parallel to I/sub p/, and to the heating profile going from peaked on axis, to hollow. 10 refs., 8 figs

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

    Czech Academy of Sciences Publication Activity Database

    Kočan, M.; Pitts, R.A.; Lisgo, S.W.; Loarte, A.; Gunn, J. P.; Fuchs, Vladimír

    2015-01-01

    Roč. 463, July (2015), s. 709-713 ISSN 0022-3115. [International Conference on Plasma-Surface Interactions in Controlled Fusion Devices (PSI)/21./. Kanazawa, 26.05.2014-30.05.2014] Institutional support: RVO:61389021 Keywords : ELM * ITER Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 2.199, year: 2015

  6. Results and analysis of high heat flux tests on a full-scale vertical target prototype of ITER divertor

    International Nuclear Information System (INIS)

    Missirlian, M.; Escourbiac, F.; Merola, M.; Bobin-Vastra, I.; Schlosser, J.; Durocher, A.

    2005-01-01

    After an extensive R and D development program, a full-scale divertor target prototype, manufactured with all the main features of the corresponding ITER divertor, was intensively tested in the high heat flux FE200 facility. The prototype consists of four units having a full monoblock geometry. The lower part (CFC armour) and the upper part (W armour) of each monoblock were joined to the solution annealed, quenched and cold worked CuCrZr tube by HIP technique. This paper summarises and analyses the main test results obtained on this prototype

  7. Melt layer macroscopic erosion of tungsten and other metals under plasma heat loads simulating ITER off-normal events

    International Nuclear Information System (INIS)

    Garkusha, I.E.; Bandura, A.N.; Byrka, O.V.; Kulik, N.V.; Landman, I.; Wuerz, H.

    2002-01-01

    This paper is focused on experimental analysis of metal layer erosion and droplet splashing of tungsten and other metals under heat loads typical for ITER FEAT off-normal events,such as disruptions and VDE's. Plasma pressure gradient action on melt layer results in erosion crater formation with mountains of displaced material at the crater edge. It is shown that macroscopic motion of melt layer and surface cracking are the main factors responsible for tungsten damage. Weight loss measurements of all exposed materials demonstrate inessential contribution of evaporation process to metals erosion

  8. Transverse heat transfer coefficient in the dual channel ITER TF CICCs Part II. Analysis of transient temperature responses observed during a heat slug propagation experiment

    Science.gov (United States)

    Lewandowska, Monika; Herzog, Robert; Malinowski, Leszek

    2015-01-01

    A heat slug propagation experiment in the final design dual channel ITER TF CICC was performed in the SULTAN test facility at EPFL-CRPP in Villigen PSI. We analyzed the data resulting from this experiment to determine the equivalent transverse heat transfer coefficient hBC between the bundle and the central channel of this cable. In the data analysis we used methods based on the analytical solutions of a problem of transient heat transfer in a dual-channel cable, similar to Renard et al. (2006) and Bottura et al. (2006). The observed experimental and other limits related to these methods are identified and possible modifications proposed. One result from our analysis is that the hBC values obtained with different methods differ by up to a factor of 2. We have also observed that the uncertainties of hBC in both methods considered are much larger than those reported earlier.

  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. Development of a high-heat-flux target for multimegawatt, multisecond neutral beams at ORNL

    International Nuclear Information System (INIS)

    Combs, S.K.; Milora, S.L.; Bush, C.E.

    1984-01-01

    A high-heat-flux target has been developed for intercepting multimegawatt, multisecond neutral beam power at the Oak Ridge National Laboratory (ORNL). Water-cooled copper swirl tubes are used for the heat transfer medium; these tubes exhibit an enhancement in burnout heat flux over conventional axial-flow tubes. The target consists of 126 swirl tubes (each 0.95 cm in outside diameter with 0.16-cm-thick walls and approx. =1 m long) arranged in a V-shape. Two arrays of parallel tubes inclined at an angle α to the beam axis form the V-shape, and this geometry reduces the surface heat flux by a factor of 1/sin α (for the present design, α =13 0 and 21 0 ). In tests with the ORNL long-pulse ion source (13- by 43-cm grid), the target has handled up to 3-MW, 30-s beam pulses with no deleterious effects. The peak power density was estimated at approx. =15 kW/cm 2 normal to the beam axis (5.4 kW/cm 2 maximum on tube surfaces). The water flow rate through the target was 41.6 L/s (660 gpm) or 0.33 L/s (5.2 gpm) per tube (axial flow velocity = 11.6 m/s). The corresponding pressure drop across the target was 1.14 MPa (165 psi) with an inlet pressure of 1.45 MPa (210 psia). Data are also presented from backup experiments in which individual tubes were heated by a small ion source (10-cm-diam grid) to characterize tube performance. These results suggest that the target should handle peak power densities in the range 25 to 30 kW/cm 2 normal to the beam axis (approx. =10 kW/cm 2 maximum on tube surfaces) with the present flow parameters. This translates to beam power levels of 5 to 6 MW for equivalent beam optics

  11. Allowable heat load on the edge of the ITER first wall panel beryllium flat tiles

    Directory of Open Access Journals (Sweden)

    R. Mitteau

    2017-08-01

    Full Text Available Plasma facing components are usually qualified to a given heat load density applied at the top face of the armour tiles with normal incidence angle. When employed in tokamak fusion machines, heat loading on the tile sides is possible due to optimised shaping, that doesn't provide edge shadowing for all design situations. An edge heat load may occur both at the tile and component scales. The edge load needs to be controlled and quantified. The adequate control of edge heat loads is especially critical for water cooled components that uses armour tiles which are bonded to the heat sink, for ensuring the long-term integrity of the tile bonding. An edge heat load allowance criterion of 10% of the top heat load is proposed. The 10% criterion is supported by experimental heat flux tests.

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

    NARCIS (Netherlands)

    Lindsay, A.; Anderson, C.; Slikboer, E.T.; Shannon, S.; Graves, D.

    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;

  14. Optimization for steady-state and hybrid operations of ITER by using scaling models of divertor heat load

    International Nuclear Information System (INIS)

    Murakami, Yoshiki; Itami, Kiyoshi; Sugihara, Masayoshi; Fujieda, Hirobumi.

    1992-09-01

    Steady-state and hybrid mode operations of ITER are investigated by 0-D power balance calculations assuming no radiation and charge-exchange cooling in divertor region. Operation points are optimized with respect to divertor heat load which must be reduced to the level of ignition mode (∼5 MW/m 2 ). Dependence of the divertor heat load on the variety of the models, i.e., constant-χ model, Bohm-type-χ model and JT-60U empirical scaling model, is also discussed. The divertor heat load increases linearly with the fusion power (P FUS ) in all models. The possible highest fusion power much differs for each model with an allowable divertor heat load. The heat load evaluated by constant-χ model is, for example, about 1.8 times larger than that by Bohm-type-χ model at P FUS = 750 MW. Effect of reduction of the helium accumulation, improvements of the confinement capability and the current-drive efficiency are also investigated aiming at lowering the divertor heat load. It is found that NBI power should be larger than about 60 MW to obtain a burn time longer than 2000 s. The optimized operation point, where the minimum divertor heat load is achieved, does not depend on the model and is the point with the minimum-P FUS and the maximum-P NBI . When P FUS = 690 MW and P NBI = 110 MW, the divertor heat load can be reduced to the level of ignition mode without impurity seeding if H = 2.2 is achieved. Controllability of the current-profile is also discussed. (J.P.N.)

  15. 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.

  16. Impact of a narrow limiter SOL heat flux channel on the ITER first wall panel shaping

    Czech Academy of Sciences Publication Activity Database

    Kocan, M.; Pitts, R.A.; Arnoux, G.; Balboa, I.; de Vries, P.C.; Dejarnac, Renaud; Furno, I.; Goldston, R.J.; Gribov, Y.; Horáček, Jan; Komm, Michael; Labit, B.; LaBombard, B.; Lasnier, C.J.; Mitteau, R.; Nespoli, F.; Pace, D.; Pánek, Radomír; Stangeby, P.C.; Terry, J.L.; Tsui, C.; Vondráček, Petr

    2015-01-01

    Roč. 55, č. 3 (2015), 033019-033019 ISSN 0029-5515 R&D Projects: GA ČR(CZ) GAP205/12/2327; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : plasma * tokamak * ITER * first wall panel Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.040, year: 2015 http://iopscience.iop.org/0029-5515/55/3/033019/pdf/0029-5515_55_3_033019.pdf

  17. An approximate block Newton method for coupled iterations of nonlinear solvers: Theory and conjugate heat transfer applications

    Science.gov (United States)

    Yeckel, Andrew; Lun, Lisa; Derby, Jeffrey J.

    2009-12-01

    A new, approximate block Newton (ABN) method is derived and tested for the coupled solution of nonlinear models, each of which is treated as a modular, black box. Such an approach is motivated by a desire to maintain software flexibility without sacrificing solution efficiency or robustness. Though block Newton methods of similar type have been proposed and studied, we present a unique derivation and use it to sort out some of the more confusing points in the literature. In particular, we show that our ABN method behaves like a Newton iteration preconditioned by an inexact Newton solver derived from subproblem Jacobians. The method is demonstrated on several conjugate heat transfer problems modeled after melt crystal growth processes. These problems are represented by partitioned spatial regions, each modeled by independent heat transfer codes and linked by temperature and flux matching conditions at the boundaries common to the partitions. Whereas a typical block Gauss-Seidel iteration fails about half the time for the model problem, quadratic convergence is achieved by the ABN method under all conditions studied here. Additional performance advantages over existing methods are demonstrated and discussed.

  18. 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.

  19. High heat flux thermal-hydraulic analysis of ITER divertor and blanket systems

    International Nuclear Information System (INIS)

    Raffray, A.R.; Chiocchio, S.; Ioki, K.; Tivey, R.; Krassovski, D.; Kubik, D.

    1998-01-01

    Three separate cooling systems are used for the divertor and blanket components, based mainly on flow routing access and on grouping together components with the highest heat load levels and uncertainties: divertor, limiter/outboard baffle, and primary first wall/inboard baffle. The coolant parameters for these systems are set to accommodate peak heat load conditions with a reasonable critical heat flux (CHF) margin. Material temperature constraints and heat transport system space and cost requirements are also taken into consideration. This paper summarises the three cooling system designs and highlights the high heat flux thermal-hydraulic analysis carried out in converging on the design values for the coolant operating parameters. Application of results from on-going high heat flux R and D and a brief description of future R and D effort to address remaining issues are also included. (orig.)

  20. The capacitor banks for the text diagnostic neutral beam and electron cyclotron heating experiments

    International Nuclear Information System (INIS)

    Nelin, K.; Jagger, J.; Baker, M.; Ourou, A.; De Turk, P.

    1986-01-01

    The Texas Experimental Tokamak (TEXT) has been operational since November of 1980. Since that time, many experimental systems have been added to the machine. Currently, two major experiments are being added to compliment the diagnostics already online. These systems, the Diagnostic Neutral Beam (DNB) and the Electron Cyclotron Heating (ECH) experiments are described in separate papers. A set of five modular, bipolar capacitor banks are used to power both the DNB and the ECH. The total capacitance of the banks is 92μF. The stored energy is about 500kJ at+or-100kV. The banks are built as five identical, interchangeable modules. One module is adequate to run the DNB. Up to four banks are used to power the ECH. The banks are portable so that they can be moved to the open end of the laboratory for maintenance. This gives much better access for repair work and allows the experiments to continue to run with the remaining banks. Due to budgetary constraints, these banks were constructed in the most economical manner possible consistent with worker safety and long term reliability. The capacitors themselves are on loan from Los Alamos National Labs. They are rated at 1.85μF at 60kV. Our application requires that they be used in a series/parallel configuration with a peak voltage of 50kV each. This paper describes the electrical, mechanical and control design considerations required to achieve a working set of banks

  1. Measurements of the toroidal plasma rotation velocity in TFTR major-radius compression experiments with auxiliary neutral beam heating

    International Nuclear Information System (INIS)

    Bitter, M.; Wong, K.L.; Scott, S.; Hsuan, H.; Grek, B.; Johnson, D.; Tait, G.

    1990-01-01

    The time history of the central toroidal plasma rotation velocity in Tokamak Fusion Test Reactor (TFTR) experiments [Phys. Rev. Lett. 55, 2587 (1985)] with auxiliary heating by neutral deuterium beam injection and major-radius compression has been measured from the Doppler shift of the emitted Ti XXI Kα line radiation. The experiments were conducted for neutral beam powers in the range 2.1--3.8 MW and line-averaged densities in the range 1.8--3.0x10 19 m -2 . The observed rotation velocity increase during compression is consistent with theoretical estimates

  2. 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.

  3. Evaluation of first wall heat fluxes due to magnetic perturbations for a range of ITER scenarios

    Czech Academy of Sciences Publication Activity Database

    Cahyna, Pavel; Kripner, Lukáš; Loarte, A.; Huijsmans, G.; Peterka, Matěj; Pánek, Radomír

    2015-01-01

    Roč. 463, August (2015), s. 406-410 ISSN 0022-3115. [PLASMA-SURFACE INTERACTIONS 21: International Conference on Plasma-Surface Interactions in Controlled Fusion Devices. Kanazawa, 26.05.2014-30.05.2014] R&D Projects: GA ČR GAP205/11/2341 Institutional support: RVO:61389021 Keywords : ELM * plasma * ITER Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 2.199, year: 2015 http://www.sciencedirect.com/science/article/pii/S0022311514009283

  4. 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.

  5. 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.

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

    OpenAIRE

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

    2015-01-01

    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...

  7. Effect of the minority concentration on ion cyclotron resonance heating in presence of the ITER-like wall in JET

    Energy Technology Data Exchange (ETDEWEB)

    Van Eester, D.; Lerche, E.; Crombé, K.; Jachmich, S. [LPP-ERM/KMS, Association Euratom-Belgian State, TEC Partner, Brussels (Belgium); Jacquet, P.; Graham, M.; Kiptily, V.; Matthews, G.; Mayoral, M.-L.; Mc Cormick, K.; Monakhov, I.; Noble, C.; Rimini, F.; Solano, E. R. [Euratom-CCFE Fusion Association, Culham Science Centre (United Kingdom); Bobkov, V.; Maggi, C.; Neu, R.; Pütterich, T. [MPI für Plasmaphysik Euratom Assoziation, Garching (Germany); Czarnecka, A. [Institute of Plasma Physics and Laser Microfusion, Warsaw (Poland); Coenen, J. W. [IEK-4, EURATOM-FZJ, TEC Partner, Jülich (Germany); and others

    2014-02-12

    The most recent JET campaign has focused on characterizing operation with the 'ITER-like' wall. One of the questions that needed to be answered is whether the auxiliary heating methods do not lead to unacceptably high levels of impurity influx, preventing fusion-relevant operation. In view of its high single pass absorption, hydrogen minority fundamental cyclotron heating in a deuterium plasma was chosen as the reference wave heating scheme in the ion cyclotron domain of frequencies. The present paper discusses the plasma behavior as a function of the minority concentration X[H] in L-mode with up to 4MW of RF power. It was found that the tungsten concentration decreases by a factor of 4 when the minority concentration is increased from X[H] ≈ 5% to X[H] % 20% and that it remains at a similar level when X[H] is further increased to 30%; a monotonic decrease in Beryllium emission is simultaneously observed. The radiated power drops by a factor of 2 and reaches a minimum at X[H] ≈ 20%. It is discussed that poor single pass absorption at too high minority concentrations ultimately tailors the avoidance of the RF induced impurity influx. The edge density being different for different minority concentrations, it is argued that the impact ICRH has on the fate of heavy ions is not only a result of core (wave and transport) physics but also of edge dynamics and fueling.

  8. Analysis of heat transfer and erosion effects on ITER divertor plasma facing components induced by slow high-power transients

    International Nuclear Information System (INIS)

    Federici, G.; Raffray, A.R.; Chiocchio, S.; Esser, B.; Dietz, J.; Igitkhanov, Y.; Janeschitz, G.

    1995-01-01

    This paper presents the results of an analysis carried out to investigate the thermal response of ITER divertor plasma facing components (PFC's) clad with Be, W, and CFC, to high-recycling, high-power thermal transients (i.e. 10--30 MW/m 2 ) which are anticipated to last up to a few seconds. The armour erosion and surface melting are estimated for the different plasma facing materials (PFM's) together with the maximum heat flux to the coolant, and armour/heat-sink interface temperature. The analysis assumes that intense target evaporation will lead to high radiative power losses in the plasma in front of the target which self-protects the target. The cases analyzed clarify the influence of several key parameters such as the plasma heat flux to the target, the loss of the melt layer, the duration of the event, the thickness of the armour, and comparison is made with cases without vapor shielding. Finally, some implications for the performance and lifetime of divertor PFC's clad with different PFM's are discussed

  9. Radio-frequency heating and neutral atom transport in a fluid-magnetohydrodynamic treatment of burning tokamak plasmas

    International Nuclear Information System (INIS)

    Conn, R.W.; Mau, T.K.; Prinja, A.K.

    1983-01-01

    A physical model for the space and time evolution of the primary parameters of ordinary and burning tokamak plasmas is described by employing a fluid plasma treatment coupled to a magnetohydrodynamic equilibrium description, the solution to the appropriate Maxwell equations, and the solution of the linear transport equation describing neutral atom transport in plasmas. The specific problems of plasma heating by ion cyclotron radiofrequency (ICRF) waves and neutral atom transport in the plasma edge and in complicated geometrical components such as divertor channels or pumped limiter structures are analyzed. A theoretical, onedimensional slab model of ICRF heating at ω = 2ω/SUB cD/ is developed and applied to determine the space-time response of tokamak plasmas. Generally, strong single-pass absorption is found for high-density, high (β) plasmas using a low k 11 spectrum (0.05 to 0.1 cm -1 ) although for (β > 1%, electron Landau damping becomes important. Deterministic and Monte Carlo methods to solve the neutral atom transport problem are described. Specific application to determine the spectrum of neutral atoms emerging from the duct of a pump limiter shows it to be hard (mean energy > 20 eV), indicating very incomplete energy thermalization. Uncertainties are identified in the overall problem of dynamic burning plasma analysis caused by the complexity of the problem itself and by uncertainties in fundamental areas such as plasma transport coefficients, stability, and plasma edge physics

  10. Comparison between FEM and high heat flux thermal fatigue testing results of ITER divertor plasma facing mock-ups

    Energy Technology Data Exchange (ETDEWEB)

    Crescenzi, F., E-mail: fabio.crescenzi@enea.it; Roccella, S.; Visca, E.; Moriani, A.

    2014-10-15

    Highlights: • Divertor is an important part of the ITER machine. • Finite element analysis allows designers to explore multiple design options, reducing physical prototypes and optimizing design performance. • The hydraulic thermal-mechanical analysis performed by ANSYS and the test results on small-scale mock-ups manufactured by HRP were compared. • FEA results confirmed many experimental data, then it could be very useful for next design optimization. - Abstract: The divertor is one of the most challenging components of “DEMO” the next step ITER machine, so many tasks regarding modeling and experiments have been made in the past years to assess manufacturing processes, materials and thus the life-time of the components. In this context the finite element analysis (FEA) allows designers to explore multiple design options, to reduce physical prototypes and to optimize design performance. The comparison between the hydraulic thermal-mechanical analysis performed by ANSYS WORKBENCH 14.5 and the test results [1] on small-scale mock-ups manufactured with the Hot Radial Pressing (HRP) [2] technology is presented in this paper. During the thermal fatigue testing in the Efremov TSEFEY facility to assess the heat flux load-carrying capability of the mock-ups, only the surface temperature was measured, so the FEA was important because it allowed to know any other information (temperature inside the materials, local water temperature, local stress, etc.). FEA was performed coupling the thermal-hydraulic analysis, that calculated the temperature distributions on the components and the heat transfer coefficient (HTC) between water and heat sink tube, with the mechanical analysis. The comparison between analysis and testing results was based on the temperature maps of the loaded surface and on number of the cycles supported during the testing and those predicted by the mechanical analysis using the experimental fatigue curves for CuCrZr-IG, that is the structural

  11. 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

  12. 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.

  13. HELCZA-High heat flux test facility for testing ITER EU first wall components.

    Czech Academy of Sciences Publication Activity Database

    Prokůpek, J.; Samec, K.; Jílek, R.; Gavila, P.; Neufuss, S.; Entler, Slavomír

    2017-01-01

    Roč. 124, November (2017), s. 187-190 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] Institutional support: RVO:61389021 Keywords : HELCZA * High heat flux * Electron beam testing * Test facility * Plasma facing components * First wall * Divertora Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 www.sciencedirect.com/science/article/pii/S0920379617302818

  14. In-situ imaging of tungsten surface modification under ITER-like transient heat loads

    Directory of Open Access Journals (Sweden)

    A.A. Vasilyev

    2017-08-01

    Full Text Available Experimental research on behavior of rolled tungsten plates under intense transient heat loads generated by a powerful (a total power of up to 7 MW long-pulse (0.1–0.3ms electron beam with full irradiation area of 2 cm2 was carried out. Imaging of the sample by the fast CCD cameras in the NIR range and with illumination by the 532nm continuous-wave laser was applied for in-situ surface diagnostics during exposure. In these experiments tungsten plates were exposed to heat loads 0.5–1MJ/m2 with a heat flux factor (Fhf close to and above the melting threshold of tungsten at initial room temperature. Crack formation and crack propagation under the surface layer were observed during multiple exposures. Overheated areas with excessive temperature over surrounding surface of about 500K were found on severely damaged samples more than 5ms after beam ending. The application of laser illumination enables to detect areas of intense tungsten melting near crack edges and crack intersections.

  15. 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.

  16. 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.

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

    Directory of Open Access Journals (Sweden)

    Vasyl Chekurin

    2017-01-01

    Full Text Available 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. A nonlinear stationary boundary-value problem for coupled heat and radiation transfer equations for the layer, which exchanges by energy with external medium by convection and radiation, has been formulated. In the case of optically thick layer, when its thickness is much more of photon-free path, the problem becomes a singularly perturbed one. In the inverse case of optically thin layer, the problem is regularly perturbed, and it becomes a regular (unperturbed one, when the layer’s thickness is of order of several photon-free paths. An iterative method for solving of the unperturbed problem has been developed and its convergence has been tested numerically. With the use of the method, the temperature field and radiation fluxes have been studied. The model and method can be used for development of noncontact methods for temperature testing in dielectrics and for nondestructive determination of its radiation properties on the base of the data obtained by remote measuring of IR radiation emitted by the layer.

  18. 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.

  19. Recent Fast Wave Coupling and Heating Studies on NSTX, with Possible Implications for ITER

    International Nuclear Information System (INIS)

    Hosea, J.C.; Bell, R.E.; Feibush, E.; Harvey, R.W.; Jaeger, E.F.; LeBlanc, B.P; Maingi, R.; Phillips, C.K.; Roquemore, L.; Ryan, P.M.; Taylor, G.; Tritz, K.; Valeo, E.J.; Wilgen, J.; Wilson, J.R.

    2009-01-01

    The goal of the high harmonic fast wave (HHFW) research on NSTX is to maximize the coupling of RF power to the core of the plasma by minimizing the coupling of RF power to edge loss processes. HHFW core plasma heating efficiency in helium and deuterium L-mode discharges is found to improve markedly on NSTX when the density 2 cm in front of the antenna is reduced below that for the onset of perpendicular wave propagation (n onset ∝ B*k # parallel# 2 /ω). In NSTX, the observed RF power losses in the plasma edge are driven in the vicinity of the antenna as opposed to resulting from multi-pass edge damping. PDI surface losses through ion-electron collisions are estimated to be significant. Recent spectroscopic measurements suggest that additional PDI losses could be caused by the loss of energetic edge ions on direct loss orbits and perhaps result in the observed clamping of the edge rotation. Initial deuterium H-mode heating studies reveal that core heating is degraded at lower k φ (- 8 m -1 relative to 13 m -1 ) as for the Lmode case at elevated edge density. Fast visible camera images clearly indicate that a major edge loss process is occurring from the plasma scrape off layer (SOL) in the vicinity of the antenna and along the magnetic field lines to the lower outer divertor plate. Large type I ELMs, which are observed at both k φ values, appear after antenna arcs caused by precursor blobs, low level ELMs, or dust. For large ELMs without arcs, the source reflection coefficients rise on a 0.1 ms time scale, which indicates that the time derivative of the reflection coefficient can be used to discriminate between arcs and ELMs.

  20. Experiences with the ASDEX neutralizer plates and construction of water-cooled plates for long-pulse heating

    International Nuclear Information System (INIS)

    Rapp, H.; Niedermeyer, H.; Kornherr, M.

    1987-01-01

    After dismantling of the titanium neutralizer plates inspection yielded satisfactory status of flat areas whereas edges and curved shapes were heavily melted. At the inner plates of the lower divertor strongly focused melting and cutting was found which is caused by fast electrons. These electrons are continuously produced. The production mechanism is not yet clear but runaway processes can be excluded. With long-pulse additional heating of 6 MW/10s as planned for ASDEX in 1987, the total energy delivered to the plasma will increase by a factor of 30. Therefore new water-cooled neutralizer plates have been constructed which consist of a copper-steel compound. The construction principle and the topology of the cooling circuits is presented

  1. Measurements of the toroidal plasma rotation velocity in TFTR major-radius compression experiments with auxiliary neutral beam heating

    International Nuclear Information System (INIS)

    Bitter, M.; Scott, S.; Wong, K.L.

    1986-07-01

    The time history of the central toroidal plasma rotation velocity in Tokamak Fusion Test Reactor (TFTR) experiments with auxiliary heating by neutral deuterium beam injection and major-radius compression has been measured from the Doppler shift of the emitted TiXXI-Kα line radiation. The experiments were conducted for neutral beam powers in the range from 2.1 to 3.8 MW and line-averaged densities in the range from 1.8 to 3.0 x 10 19 m -2 . The observed rotation velocity increase during compression is in agreement with results from modeling calculations which assume classical slowing-down of the injected fast deuterium ions and momentum damping at the rate established in the precompression plasma

  2. The next step in a development of negative ion beam plasma neutraliser for ITER NBI

    International Nuclear Information System (INIS)

    Kulygin, V.M.; Dlougach, E.D.; Gorbunov, E.P.

    2001-01-01

    Injectors of deuterium atom beams developing for ITER plasma heating and current drive are based on the negative ion acceleration and further neutralization with a gas target. The maximal efficiency of a gas stripping process is 60%. The replacement of the gas neutralizer by plasma one must increase the neutral yield to 80%. The experimental study overview of the microwave discharge in a multi-cusp magnetic system chosen as a base device for Plasma Neutralizer realization and the design development for ITER Neutral Beam Injectors are presented. The experimental results achieved at a plasma neutralizer model PNX-U is discussed. Plasma confinement, gas flows, ionization degree were investigated. The plasma in the volume 0.5m 3 with density n e ∼ 10 18 m -3 has been achieved at power density 80kW/m 3 in operation with Argon. (author)

  3. Behavior of divertor and first wall armour materials at plasma heat fluxes relevant to ITER ELMs and disruptions

    Directory of Open Access Journals (Sweden)

    D.V. Kovalenko

    2017-08-01

    Full Text Available The paper presents the main results of numerous experiments carried out over the past 10 years at QSPA-T and QSPA-Be plasma guns in support of ITER. Special targets made of pure W, W-1%La2O3 and two types of Be (TGP-56FW and S65-C were tested under the series of repeated plasma stream and photonic flux impact. Maximum heat load on the target surface was up to 2.5MJ/m2 in the case of plasma testing and was equal to 0.5MJ/m2 in the case of photonic flux testing. Pulse waveform was rectangular with tpulse= 0.5ms. It was found that the main erosion mechanisms of W and Be under plasma stream impact are the melt layer movement, the ejection of droplets and the cracks formation. As a result of repeated photonic fluxes a regular, “corrugated” structure are eventually formed on the Be target surface. Study of erosion products of W formed under plasma stream impact on the W target has shown that the D/W atomic ratio in the deposited W films during pulsed events may be the same or even higher than that for stationary processes.

  4. Analysis of low energy neutral hydrogen fluxes using an electron cyclotron resonance heated discharge

    International Nuclear Information System (INIS)

    Cain, B.L.

    1989-01-01

    This dissertation describes the design, construction, and proof-of-principle verification of a neutral hydrogen flux detection system, based on an ECRH discharge as the neutral flux ionizer. The significant features of the ionizer are its small size and simultaneous excitation of the ECRH mode using a 30 MHz RF driver and relatively small static magnetic fields. Demonstrated is the ability of the ECRH ionizer to ionize ∼ 900 eV neutral hydrogen fluxes with subsequent detection in a high resolution energy analyzer. A versatile calibration technique is applied to determine the ionizer efficiency, which additionally gives a variety of elastic scattering and charge exchange cross section results. Also described are the details of a new low energy beam-target interaction research facility, along with the basic techniques required to calibrate many of the system components. The facility has potential applications in areas such as fundamental cross section measurement, plasma diagnostics, beam-plasma interactions, and further beam-target research. 111 refs., 82 figs

  5. ITER ITA newsletter. No. 22, May 2005

    International Nuclear Information System (INIS)

    2005-06-01

    This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about Japanese Participant Team's recent activities in the ITER Transitional Arrangements(ITA) phase and ITER related meeting the Fourth IAEA Technical Meeting (IAEA-TM) on Negative Ion Based Neutral Beam Injectors which was held in Padova, Italy from 9-11 May 2005

  6. Electron energy distribution function in the divertor region of the COMPASS tokamak during neutral beam injection heating

    Science.gov (United States)

    Hasan, E.; Dimitrova, M.; Havlicek, J.; Mitošinková, K.; Stöckel, J.; Varju, J.; Popov, Tsv K.; Komm, M.; Dejarnac, R.; Hacek, P.; Panek, R.; the COMPASS Team

    2018-02-01

    This paper presents the results from swept probe measurements in the divertor region of the COMPASS tokamak in D-shaped, L-mode discharges, with toroidal magnetic field BT = 1.15 T, plasma current Ip = 180 kA and line-average electron densities varying from 2 to 8×1019 m-3. Using neutral beam injection heating, the electron energy distribution function is studied before and during the application of the beam. The current-voltage characteristics data are processed using the first-derivative probe technique. This technique allows one to evaluate the plasma potential and the real electron energy distribution function (respectively, the electron temperatures and densities). At the low average electron density of 2×1019 m-3, the electron energy distribution function is bi-Maxwellian with a low-energy electron population with temperatures 4-6 eV and a high-energy electron group 12-25 eV. As the line-average electron density is increased, the electron temperatures decrease. At line-average electron densities above 7×1019 m-3, the electron energy distribution function is found to be Maxwellian with a temperature of 6-8.5 eV. The effect of the neutral beam injection heating power in the divertor region is also studied.

  7. Damage to tungsten macro-brush targets under multiple ELM-like heat loads. Experiments vs. numerical simulations and extrapolation to ITER

    Energy Technology Data Exchange (ETDEWEB)

    Bazylev, B.; Landman, I. [Forschungszentrum Karlsruhe (Germany). IHM; Janeschitz, G. [Forschungszentrum Karlsruhe (DE). Fusion EURATOM] (and others)

    2007-07-01

    Operation of ITER at high fusion gain is assumed to be the H-mode. A characteristic feature of this regime is the transient release of energy from the confined plasma onto PFCs by multiple ELMs (about 104 ELMs per ITER discharge), which can play a determining role in the erosion rate and lifetime of these components. The expected energy heat loads on the ITER divertor during Type I ELM are in range 0.5-4 MJ/m{sup 2} in timescales of 0.3-0.6 ms. Tungsten macro-brush armour (W-brushes) is foreseen as one of plasma facing components (PFC) for ITER divertor and dome. During the intense transient events in ITER the surface melting, melt motion, melt splashing and evaporation are seen as the main mechanisms of W erosion. The expected erosion of the ITER plasma facing components under transient energy loads can be properly estimated by numerical simulations validated against target erosion of the experiments at the plasma gun facility QSPA-T. Within the collaboration established between EU fusion programme and the Russian Federation, W-brush targets (produced either from pure tungsten or tungsten with 1% of La{sub 2}O{sub 3}) manufactured according to the EU specifications for the ITER divertor targets, have been exposed to multiple ITER ELM-like loads in plasma gun facilities at TRINITI in the range 0.5 - 2.2 MJ/m2 with pulse duration of 0.5 ms. The measured material erosion data have been used to validate the codes MEMOS and PHEMOBRID. Numerical simulations, including 3D-simulations (codes MEMOS and PHEMOBRID), carried out for the conditions of the QSPA-T experiments with heat loads in the range 0.5-2.2 MJ/m{sup 2} and the timescale 0.5 ms demonstrated a rather good agreement with the data obtained at the plasma gun facility QSPA: melting of brush edges at low heat loads, intense melt motion and bridge formation caused by the Rayleigh-Taylor instability at heat loads Q>1.3 MJ/m{sup 2}. The melt splashing generated by the Kelvin-Helmholtz, and Rayleigh

  8. Experimental study of divertor plasma-facing components damage under a combination of pulsed and quasi-stationary heat loads relevant to expected transient events at ITER

    International Nuclear Information System (INIS)

    Klimov, N S; Podkovyrov, V L; Kovalenko, D V; Zhitlukhin, A M; Barsuk, V A; Mazul, I V; Giniyatulin, R N; Kuznetsov, V Ye; Riccardi, B; Loarte, A; Merola, M; Koidan, V S; Linke, J; Landman, I S; Pestchanyi, S E; Bazylev, B N

    2011-01-01

    This paper concerns the experimental study of damage of ITER divertor plasma-facing components (PFCs) under a combination of pulsed plasma heat loads (representative of controlled ITER type I edge-localized modes (ELMs)) and quasi-stationary heat loads (representative of the high heat flux (HHF) thermal fatigue expected during ITER normal operations and slow transient events). The PFC's tungsten armor damage under pulsed plasma exposure was driven by (i) the melt layer motion, which leads to bridges formation between neighboring tiles and (ii) the W brittle failure giving rise to a stable crack pattern on the exposed surface. The crack width reaches a saturation value that does not exceed some tens of micrometers after several hundreds of ELM-like pulses. HHF thermal fatigue tests have shown (i) a peeling-off of the re-solidified material due to its brittle failure and (ii) a significant widening (up to 10 times) of the cracks and the formation of additional cracks.

  9. Neutron spectroscopy measurements and modeling of neutral beam heating fast ion dynamics

    International Nuclear Information System (INIS)

    Hellesen, C; Sunden, E Andersson; Conroy, S; Ericsson, G; Johnson, M Gatu; Hjalmarsson, A; Kaellne, J; Ronchi, E; Sjoestrand, H; Weiszflog, M; Albergante, M; Ballabio, L; Gorini, G; Tardocchi, M; Giacomelli, L; Jenkins, I; Voitsekhovitch, I

    2010-01-01

    The energy spectrum of the neutron emission from beam-target reactions in fusion plasmas at the Joint European Torus (JET) has been investigated. Different beam energies as well as injection angles were used. Both measurements and simulations of the energy spectrum were done. The measurements were made with the time-of-flight spectrometer TOFOR. Simulations of the neutron spectrum were based on first-principle calculations of neutral beam deposition profiles and the fast ion slowing down in the plasma using the code NUBEAM, which is a module of the TRANSP package. The shape of the neutron energy spectrum was seen to vary significantly depending on the energy of the beams as well as the injection angle and the deposition profile in the plasma. Cross validations of the measured and modeled neutron energy spectra were made, showing a good agreement for all investigated scenarios.

  10. Investigation of the Impact of Transient Heat Loads Applied by Laser Irradiation on ITER-Grade Tungsten

    OpenAIRE

    Huber, Alexander; Arakcheev, A.; Philipps, V.; Pintsuk, Gerald; Reinhart, Michael; Samm, Ulrich; Shoshin, A.; Schweer, Bernd; Unterberg, Bernhard; Zlobinski, M.; Sergienko, Gennady; Steudel, Isabel; Wirtz, Marius; Burdakov, A. V.; Coenen, Jan Willem

    2014-01-01

    Cracking thresholds and crack patterns in tungsten targets after repetitive ITER-like edge localized mode (ELM) pulses have been studied in recent simulation experiments by laser irradiation. The tungsten specimens were tested under selected conditions to quantify the thermal shock response. A Nd:YAG laser capable of delivering up to 32 J of energy per pulse with a duration of 1 ms at the fundamental wavelength λ = 1064 nm has been used to irradiate ITER-grade tungsten samples with repetitive...

  11. Parametric scaling studies of energy-confinement time for neutral-beam-heated Heliotron-E plasmas

    International Nuclear Information System (INIS)

    Sano, F.; Takeiri, Y.; Hanatani, K.

    1989-02-01

    Kinetic analysis of the global energy confinement time for neutral-beam-heated Heliotron-E plasmas has been performed with the 1-D, time-independent transport analysis code, PROCTR-Mod. Beam-power scans were performed by firing various number of hydrogen neutral beams, while density scans were performed by puffing gas and/or pellet fueling under the metallic or carbonized wall conditions. The wall carbonization facilitated the density increase due to the enhanced particle recycling on the walls, and also enabled long-pulse, quasi-stationary, currentless ECH + NBI operation with reduced heavy-impurity contamination. The data analysis shows that the favorable density dependence partially offsets the unfavorable power dependence, and that the anomalous electron transport loss becomes dominant in the over-all energy balance as the beam power and plasma density are increased. An alternative scaling law is also presented which is to fit τ E G [ms] by an 'offset-linear' law. The latter scaling is found to provide a better fit to the presented data sets in spite of its simple form. The parametric scaling of the local electron thermal diffusivity, χ e , is also discussed on the basis of the kinetic analysis. (J.P.N.)

  12. Impurity transport studies by means of tracer-encapsulated solid pellet injection in neutral beam heated plasmas on LHD

    International Nuclear Information System (INIS)

    Tamura, N; Sudo, S; Khlopenkov, K V; Kato, S; Sergeev, V Yu; Muto, S; Sato, K; Funaba, H; Tanaka, K; Tokuzawa, T; Yamada, I; Narihara, K; Nakamura, Y; Kawahata, K; Ohyabu, N; Motojima, O

    2003-01-01

    The quantitative properties of impurity transport in large helical device (LHD) plasmas heated by neutral beam injection have been investigated by means of tracer-encapsulated solid pellet (TESPEL) injection. In the case of a titanium (Ti) tracer, the behaviour of the emission lines from the highly ionized Ti impurity, Ti Kα(E He-like ∼ 4.7 keV) and Ti XIX (λ = 16.959 nm), has been observed clearly by a soft x-ray pulse height analyzer and a vacuum ultraviolet spectrometer, respectively. A fairly longer decay time of the Ti Kα emission lines is obtained above the value of a line-averaged electron density, 3.0x10 19 m -3 . The dependence of the behaviour of the Ti tracer impurity on the line-averaged electron density below the value of that, 3.5x10 19 m -3 is in qualitative agreement with the characteristics obtained from the observation of the behaviour of an intrinsic metallic impurity in neutral beam heated plasmas on LHD. In order to estimate the properties of the Ti impurity transport quantitatively, the one-dimensional impurity transport code, MIST has been used. As a result of the transport analysis with the MIST code, even an small inward convection should be necessary to account for the experimental results with the value of the line-averaged electron density, 3.5x10 19 m -3 . In order to examine the experimentally obtained transport coefficients, neoclassical analysis with respect to the radial impurity flux has been performed. The inferred rise of the inward convection cannot be explained solely by neoclassical impurity transport. Therefore, in order to account for the inward convection, the effect of a radial electric field and/or some other effect must be taken into account additionally

  13. Manufacturing of small-scale mock-ups and of a semi-prototype of the ITER Normal Heat Flux First Wall

    International Nuclear Information System (INIS)

    Banetta, S.; Zacchia, F.; Lorenzetto, P.; Bobin-Vastra, I.; Boireau, B.; Cottin, A.; Mitteau, R.; Eaton, R.; Raffray, R.

    2014-01-01

    This paper describes the manufacturing development and fabrication of reduced scale ITER First Wall (FW) mock-ups of the Normal Heat Flux (NHF) design, including a “semi-prototype” with a dimension of 305 mm × 660 mm, corresponding to about 1/6 of a full-scale panel. The activity was carried out in the framework of the pre-qualification of the European Domestic Agency (EU-DA or F4E) for the supply of the European share of the ITER First Wall. The hardware consists of three Upgraded (2 MW/m 2 ) Normal Heat Flux (U-NHF) small-scale mock-ups, bearing 3 beryllium tiles each, and of one Semi-Prototype, representing six full-scale fingers and bearing a total of 84 beryllium tiles. The manufacturing process makes extensive use of Hot Isostatic Pressing, which was developed over more than a decade during ITER Engineering Design Activity phase. The main manufacturing steps for the semi-prototype are described, with special reference to the lessons learned and the implications impacting the future fabrication of the full-scale prototype and the series which consists of 218 panels plus spares. In addition, a “tile-size” mock-up was manufactured in order to assess the performance of larger tiles. The use of larger tiles would be highly beneficial since it would allow a significant reduction of the panel assembly time

  14. Fast wave current drive in neutral beam heated plasmas on DIII-D

    International Nuclear Information System (INIS)

    Petty, C.C.; Forest, C.B.; Pinsker, R.I.

    1997-04-01

    The physics of non-inductive current drive and current profile control using the fast magnetosonic wave has been demonstrated on the DIII-D tokamak. In non-sawtoothing discharges formed by neutral beam injection (NBI), the radial profile of the fast wave current drive (FWCD) was determined by the response of the loop voltage profile to co, counter, and symmetric antenna phasings, and was found to be in good agreement with theoretical models. The application of counter FWCD increased the magnetic shear reversal of the plasma and delayed the onset of sawteeth, compared to co FWCD. The partial absorption of fast waves by energetic beam ions at high harmonics of the ion cyclotron frequency was also evident from a build up of fast particle pressure near the magnetic axis and a correlated increase in the neutron rate. The anomalous fast particle pressure and neutron rate increased with increasing NBI power and peaked when a harmonic of the deuterium cyclotron frequency passed through the center of the plasma. The experimental FWCD efficiency was highest at 2 T where the interaction between the fast waves and the beam ions was weakest; as the magnetic field strength was lowered, the FWCD efficiency decreased to approximately half of the maximum theoretical value

  15. Neutron diagnostic that measures Z/sub eff/ in a neutral-beam-heated Tokomak

    International Nuclear Information System (INIS)

    Slaughter, D.R.

    1979-01-01

    The rate of pitch-angle scattering in a beam-driven Tokomak is proportional to Z/sub eff/ when neutral deuterium is injected parallel or antiparallel to the toroidal field B/sub T/. The energy spectrum of neutrons produced by D--D or D--T reactions is sensitive to the angular distribution of reacting energetic deuterons so that a measurement of the spectrum may be used to infer Z/sub eff/. Energy spectra of neutrons emitted parallel to B/sub T/ during simultaneous co- and counter-injection were calculated for the case of 120-keV beams by using a PPPL code. The results were then convoluted with spectrometer lineshapes determined experimentally for a system used to measure neutron spectra during a 1.0-s source pulse. Results indicate that Z/sub eff/ in the range of 1 to 4 may be determined with uncertainties of +- 0.25 for D--D plasma and +- 0.5 for D--T plasma, provided the ion temperature T/sub i/ is well known. However, the spectrometer energy resolution is not adequate to determine T/sub i/ directly from a neutron--spectrum measurement. In the absence of accurate T/sub i/ data, the uncertainty in Z/sub eff/ is approximately +- 1. In either case, impurity identification is not established by this type of measurement

  16. Confinement of Stellarator plasmas with neutral beam and RF heating in W VII-A

    International Nuclear Information System (INIS)

    Grieger, G.; Cattanei, G.; Dorst, D.

    1986-01-01

    WENDELSTEIN VII-A has been operated for ten years. It is a low-shear, high-aspect-ratio device. The confinement properties have been thoroughly studied for both ohmically heated and net-current free plasmas. For the latter case, NBI- and ECF-maintained plasmas were of particular importance. It was found that under optimized conditions the core of high-pressure, net-current free plasmas is mainly governed by collisional effects. The experiment will now be shut down for upgrading it into the Advanced Stellarator WEDNDELSTEIN VII-AS. (author)

  17. Fusion reactivity, confinement, and stability of neutral-beam heated plasmas in TFTR and other tokamaks

    International Nuclear Information System (INIS)

    Park, Hyeon, K.

    1996-05-01

    The hypothesis that the heating beam fueling profile shape connects the edge condition and improved core confinement and fusion reactivity is extensively studied on TFTR and applied to other tokamaks. The derived absolute scalings based on beam fueling profile shape for the stored energy and neutron yield can be applied to the deuterium discharges at different major radii in TFTR. These include Supershot, High poloidal beta, L-mode, and discharges with a reversed shear (RS) magnetic configuration. These scalings are also applied to deuterium-tritium discharges. The role of plasma parameters, such as plasma current, Isdo2(p), edge safety factor, qsdo5(a), and toroidal field, Bsdo2(T), in the performance and stability of the discharges is explicitly studied. Based on practical and externally controllable plasma parameters, the limitation and optimization of fusion power production of the present TFTR is investigated and a path for a discharge condition with fusion power gain, Q > 1 is suggested based on this study. Similar physics interpretation is provided for beam heated discharges on other major tokamaks

  18. The ITER divertor concept

    International Nuclear Information System (INIS)

    Janeschitz, G.; Borrass, K.; Federici, G.; Igitkhanov, Y.; Kukushkin, A.; Pacher, H.D.; Pacher, G.W.; Sugihara, M.

    1995-01-01

    The ITER divertor must exhaust most of the alpha particle power and the He ash at acceptable erosion rates. The high recycling regime of the ITER-CDA for present parameters would yield high power loads and erosion rates on conventional targets. Improvement by radiation in the SOL at constant pressure is limited in principle. To permit a higher radiation fraction, the plasma pressure along the field must be reduced by more than a factor 10, reducing also the target ion flux. This pressure reduction can be obtained by strong plasma-neutral interaction below the X-point. Under these conditions T e in the divertor can be reduced to <5 eV along a flame like ionisation front by impurity radiation and CX losses. Downstream of the front, neutrals undergo more CX or i-n collisions than ionisation events, resulting in significant momentum loss via neutrals to the divertor chamber wall. The pressure reduction by this mechanism depends on the along-field length for neutral-plasma interaction, the parallel power flux, the neutral density, the ratio of neutral-neutral collision length to the plasma-wall distance and on the Mach number of ions and neutrals. A supersonic transition in the main plasma-neutral interaction region, expected to occur near the ionisation front, would be beneficial for momentum removal. The momentum transfer fraction to the side walls is calculated: low Knudsen number is beneficial. The impact of the different physics effects on the chosen geometry and on the ITER divertor design and the lifetime of the various divertor components are discussed. ((orig.))

  19. Plasma characteristics of long-pulse discharges heated by neutral beam injection in the Large Helical Device

    Science.gov (United States)

    Takeiri, Y.; Nakamura, Y.; Noda, N.; Osakabe, M.; Kawahata, K.; Oka, Y.; Kaneko, O.; Tsumori, K.; Sato, M.; Mutoh, T.; Shimozuma, T.; Goto, M.; Ida, K.; Inagaki, S.; Kado, S.; Masuzaki, S.; Morita, S.; Nagayama, Y.; Narihara, K.; Peterson, B. J.; Sakakibara, S.; Sato, K.; Shoji, M.; Tanaka, K.; de Vries, P. C.; Sudo, S.; Ohyabu, N.; Motojima, O.

    2000-02-01

    Long-pulse neutral beam injection heating has been achieved in the large helical device (LHD). Two different confinement states are observed for different averaged densities in the long-pulse plasmas. A quasi-steady-state plasma was sustained for 21 s with an injection power of 0.6 MW, where the central plasma temperature was around 1 keV with a line-averaged electron density of 0.3 × 1019 m-3 . The discharge duration can be so extended as to keep the plasma properties in the short-pulse discharge. The energy confinement time is nearly the same as that of the short-pulse discharge, which is 1.3 times as long as the international stellarator scaling ISS95. At higher densities, a relaxation oscillation phenomenon, observed as if the plasma would breathe, lasted for 20 s with a period of 1-2 s. The phenomenon is characterized with profile expansion and contraction of the electron temperature. The density oscillation is out of phase with the temperature oscillation and is related to the density clamping phenomenon. The observed plasma properties are shown in detail for the `breathing' oscillation phenomenon. Possible mechanisms for the breathing oscillation are also discussed, with a view of the screening effect near the last closed magnetic surface and the power balance between the heating and the radiation powers. The long-pulse heating results indicate unique characteristics of the LHD where no special feedback stabilization is required due to absence of disruption and no need for current drive.

  20. MHD stability of the ITER pedestal and SOL plasma and its influence on the heat flux width

    NARCIS (Netherlands)

    Loarte, A.; Liu, F.; Huijsmans, G.T.A.; Kukushkin, A.S.; Pitts, R.A.

    2015-01-01

    Proceedings of the 21st International Conference on Plasma-Surface Interactions in Controlled Fusion Devices Kanazawa, Japan May 26-30, 2014 MHD stability of ITER plasmas has been analyzed for QDT = 10 edge and SOL plasma conditions, showing that the SOL plasma is MHD stable down to pressure

  1. 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.

  2. Development of a plasma assisted ITER level controlled heat source and observation of novel micro/nanostructures produced upon exposure of tungsten targets

    Energy Technology Data Exchange (ETDEWEB)

    Aomoa, N.; Sarmah, Trinayan; Sah, Puspalata [CIMPLE-PSI Laboratory, Centre of Plasma Physics-Institute for Plasma Research, Sonapur 782 402 Assam (India); Chaudhuri, P.; Khirwarker, S.; Ghosh, J. [Institute for Plasma Research, Gandhinagar 382428 Gujarat (India); Satpati, B. [Saha Institute of Nuclear Physics, 1/AF Bidhannagar, Kolkata 700 064 (India); Kakati, M., E-mail: mayurkak@rediffmail.com [CIMPLE-PSI Laboratory, Centre of Plasma Physics-Institute for Plasma Research, Sonapur 782 402 Assam (India); De Temmerman, G. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046 Saint Paul Lez Durance, Cedex (France)

    2016-05-15

    Highlights: • Developed a plasma assisted ITER level high heat flux device for material testing. • The beam deposits over 10 MW/m{sup 2} flux uniformly over a remote material target. • Hopper micro-crystals were growing while exposing Plansee tungsten in the device. • CIMPLE-PSI being developed for exact reproduction of Tokomak Divertor conditions. - Abstract: This paper reports on the development of a simple, low-cost, segmented plasma torch assisted high-heat flux device for material testing, which can simulate the extreme heat flux expected in future fusion devices. Calorimetric measurements confirmed uniform heat deposition by the well collimated argon plasma beam over a target surface with power fluxes in excess of 10 MW/m{sup 2} during high current, high gas flow rate operations. To understand the outcome of possible melting of first wall material in an ITER like machine, an Plansee tungsten target was exposed in this device, which witnessed growth of micrometer level Hopper crystals and their aggregation to vertical grains in central exposed region. Increase in viscosity of the metal during high under-cooling is believed to have lead to the skeletal patterns, observed for the first time for tungsten here. Transmission electron microscopy confirmed that re-solidified grains on the target actually had crystalline substructures in the nanometer level. This laboratory is in the process of developing an exact linear Tokamak Divertor simulator, where a magnetized hydrogen/helium collimated plasma jet will be produced at higher vacuum, for plasma material interaction studies with direct relevance to modern plasma fusion machines.

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

    International Nuclear Information System (INIS)

    Bohm, Tim; Davis, Andrew; Sawan, Mohamed; Marriott, Edward; Wilson, Paul; Ulrickson, Michael; Bullock, James

    2015-01-01

    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.

  4. 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.

  5. New Ideas for Confined Alpha Diagnostics on ITER

    Science.gov (United States)

    Fisher, R. K.

    2003-10-01

    Understanding the dynamics of a burning plasma will require development of adequate alpha particle diagnostics. Three new approaches to obtain information on the confined fast alphas in ITER are proposed. The first technique measures the energetic D and T charge exchange (CX) neutrals that result from the alpha collision-induced knock-on fuel ion tails undergoing electron capture on the MeV D neutral beams planned for heating and current drive. CX neutrals with energies >1 ,MeV would be measured to avoid the background due to the large population of injected beam ions. The second technique measures the energetic knock-on neutron tail due to alphas using the lengths of the proton recoil tracks produced by neutron collisions in the film. The range of the 14 to 18 MeV recoil protons increases by ˜400 microns per MeV. The third approach would measure the CX helium neutrals resulting from confined alphas capturing two electrons in the ablation cloud surrounding a dense gas jet that has been proposed for disruption mitigation in ITER. Jet Charge Exchange (JCX) could allow measurements in the plasma core, while the Pellet Charge Exchange (PCX) technique that provided much of the data on confined alphas in TFTR, will likely be limited by pellet penetration to measurements outside r/ a , ˜ ,0.5 in ITER.

  6. Design development of heat transfer elements for characterization of neutral beam with power density of 65 MW/M2 in INTF

    International Nuclear Information System (INIS)

    Venkata Nagaraju, M.; Bandyopadhyay, Mainak; Rotti, Chandramouli; Pillai, Suraj; Singh, Mahendrajit; Joshi, Jaydeep; Chakraborty, Arun K.

    2017-01-01

    INTF Second Calorimeter is a thermal target system going to be installed in Indian Test Facility (INTF), being constructed at ITER-India laboratory in IPR. It will be placed inside the vacuum vessel on the extreme end which is at 20.6m from the exit of -ve ion beam source located on opposite end. The paper describes the design of this calorimeter including Beam power estimations, panel configuration and its optimization, HTE orientation with respect to beam axis, profiling of HTE, hydraulic calculations, thermo-mechanical and thermos-hydraulic assessments of severely loaded Heat Transfer Element in ANSYS for both normal and off-normal conditions of 5mrad beam. The design has been further validated for structural code SDC-IC, which is essentially for ITER in-vessel components

  7. Iterating skeletons

    DEFF Research Database (Denmark)

    Dieterle, Mischa; Horstmeyer, Thomas; Berthold, Jost

    2012-01-01

    a particular skeleton ad-hoc for repeated execution turns out to be considerably complicated, and raises general questions about introducing state into a stateless parallel computation. In addition, one would strongly prefer an approach which leaves the original skeleton intact, and only uses it as a building...... 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......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...

  8. Engineering study of the neutral beam and rf heating systems for DIII-D, MFTF-B, JET, JT-60 and TFTR

    International Nuclear Information System (INIS)

    Lindquist, W.B.; Staten, S.H.

    1987-01-01

    An engineering study was performed on the rf and neutral beam heating systems implemented for DIII-D, MFTF-B, JET, JT-60 and TFTR. Areas covered include: methodology used to implement the systems, technology, cost, schedule, performance, problems encountered and lessons learned. Systems are compared and contrasted in the areas studied. Summary statements were made on common problems and lessons learned. 3 refs., 6 tabs

  9. Power converters for ITER

    CERN Document Server

    Benfatto, I

    2006-01-01

    The International Thermonuclear Experimental Reactor (ITER) is a thermonuclear fusion experiment designed to provide long deuterium– tritium burning plasma operation. After a short description of ITER objectives, the main design parameters and the construction schedule, the paper describes the electrical characteristics of the French 400 kV grid at Cadarache: the European site proposed for ITER. Moreover, the paper describes the main requirements and features of the power converters designed for the ITER coil and additional heating power supplies, characterized by a total installed power of about 1.8 GVA, modular design with basic units up to 90 MVA continuous duty, dc currents up to 68 kA, and voltages from 1 kV to 1 MV dc.

  10. Evaluation of ITER MSE Viewing Optics

    International Nuclear Information System (INIS)

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

    2007-01-01

    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 formed an intermediate

  11. Spirit and prospects of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Velikhov, E.P. [Kurchatov Institute of Atomic Energy, Moscow (Russian Federation)

    2002-10-01

    ITER is the unique and the most straightforward way to study the burning plasma science in the nearest future. ITER has a firm physics ground based on the results from the world tokamaks in terms of confinement, stability, heating, current drive, divertor, energetic particle confinement to an extend required in ITER. The flexibility of ITER will allow the exploration of broad operation space of fusion power, beta, pulse length and Q values in various operational scenarios. Success of the engineering R and D programs has demonstrated that all party has an enough capability to produce all the necessary equipment in agreement with the specifications of ITER. The acquired knowledge and technologies in ITER project allow us to demonstrate the scientific and technical feasibility of a fusion reactor. It can be concluded that ITER must be constructed in the nearest future. (author)

  12. Spirit and prospects of ITER

    International Nuclear Information System (INIS)

    Velikhov, E.P.

    2002-01-01

    ITER is the unique and the most straightforward way to study the burning plasma science in the nearest future. ITER has a firm physics ground based on the results from the world tokamaks in terms of confinement, stability, heating, current drive, divertor, energetic particle confinement to an extend required in ITER. The flexibility of ITER will allow the exploration of broad operation space of fusion power, beta, pulse length and Q values in various operational scenarios. Success of the engineering R and D programs has demonstrated that all party has an enough capability to produce all the necessary equipment in agreement with the specifications of ITER. The acquired knowledge and technologies in ITER project allow us to demonstrate the scientific and technical feasibility of a fusion reactor. It can be concluded that ITER must be constructed in the nearest future. (author)

  13. 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.

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

    International Nuclear Information System (INIS)

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

    2003-01-01

    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 -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)

  15. 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)

  16. Nuclear analysis for ITER

    International Nuclear Information System (INIS)

    Santoro, R.T.; Iida, H.; Khripunov, V.; Petrizzi, L.; Sato, S.; Sawan, M.; Shatalov, G.; Schipakin, O.

    2001-01-01

    This paper summarizes the main results of nuclear analysis calculations performed during the International Thermonuclear Experimental Reactor (ITER) Engineering Design Activity (EDA). Major efforts were devoted to fulfilling the General Design Requirements to minimize the nuclear heating rate in the superconducting magnets and ensuring that radiation conditions at the cryostat are suitable for hands-on-maintenance after reactor shut-down. (author)

  17. ITER safety

    International Nuclear Information System (INIS)

    Raeder, J.; Piet, S.; Buende, R.

    1991-01-01

    As part of the series of publications by the IAEA that summarize the results of the Conceptual Design Activities for the ITER project, this document describes the ITER safety analyses. It contains an assessment of normal operation effluents, accident scenarios, plasma chamber safety, tritium system safety, magnet system safety, external loss of coolant and coolant flow problems, and a waste management assessment, while it describes the implementation of the safety approach for ITER. The document ends with a list of major conclusions, a set of topical remarks on technical safety issues, and recommendations for the Engineering Design Activities, safety considerations for siting ITER, and recommendations with regard to the safety issues for the R and D for ITER. Refs, figs and tabs

  18. ITER tokamak device

    International Nuclear Information System (INIS)

    Doggett, J.; Salpietro, E.; Shatalov, G.

    1991-01-01

    The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER; and a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fuelling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (i) magnet systems (toroidal and poloidal field coils and cryogenic systems), (ii) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (iii) first wall, (iv) divertor plate (design and materials, performance and lifetime, a.o.), (v) blanket/shield system, (vi) maintenance equipment, (vii) current drive and heating, (viii) fuel cycle system, and (ix) diagnostics. 11 refs, figs and tabs

  19. An iterative procedure for estimating areally averaged heat flux using planetary boundary layer mixed layer height and locally measured heat flux

    Energy Technology Data Exchange (ETDEWEB)

    Coulter, R. L.; Gao, W.; Lesht, B. M.

    2000-04-04

    Measurements at the central facility of the Southern Great Plains (SGP) Cloud and Radiation Testbed (CART) are intended to verify, improve, and develop parameterizations in radiative flux models that are subsequently used in General Circulation Models (GCMs). The reliability of this approach depends upon the representativeness of the local measurements at the central facility for the site as a whole or on how these measurements can be interpreted so as to accurately represent increasingly large scales. The variation of surface energy budget terms over the SGP CART site is extremely large. Surface layer measurements of the sensible heat flux (H) often vary by a factor of 2 or more at the CART site (Coulter et al. 1996). The Planetary Boundary Layer (PBL) effectively integrates the local inputs across large scales; because the mixed layer height (h) is principally driven by H, it can, in principal, be used for estimates of surface heat flux over scales on the order of tens of kilometers. By combining measurements of h from radiosondes or radar wind profiles with a one-dimensional model of mixed layer height, they are investigating the ability of diagnosing large-scale heat fluxes. The authors have developed a procedure using the model described by Boers et al. (1984) to investigate the effect of changes in surface sensible heat flux on the mixed layer height. The objective of the study is to invert the sense of the model.

  20. ITER plant systems

    International Nuclear Information System (INIS)

    Kolbasov, B.; Barnes, C.; Blevins, J.

    1991-01-01

    As part of a series of documents published by the IAEA that summarize the results of the Conceptual Design Activities for the ITER project, this publication describes the conceptual design of the ITER plant systems, in particular (i) the heat transport system, (ii) the electrical distribution system, (iii) the requirements for radioactive equipment handling, the hot cell, and waste management, (iv) the supply system for fluids and operational chemicals, (v) the qualitative analyses of failure scenarios and methods of burn stability control and emergency shutdown control, (vi) analyses of tokamak building functions and design requirements, (vii) a plant layout, and (viii) site requirements. Refs, figs and tabs

  1. 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.

  2. 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

    International Nuclear Information System (INIS)

    Bushell, J.; Sherlock, P.; Mummery, P.; Bellin, B.; Zacchia, F.

    2015-01-01

    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.

  3. ITER overview

    International Nuclear Information System (INIS)

    Shimomura, Y.; Aymar, R.; Chuyanov, V.; Huguet, M.; Parker, R.R.

    2001-01-01

    This report summarizes technical works of six years done by the ITER Joint Central Team and Home Teams under terms of Agreement of the ITER Engineering Design Activities. The major products are as follows: complete and detailed engineering design with supporting assessments, industrial-based cost estimates and schedule, non-site specific comprehensive safety and environmental assessment, and technology R and D to validate and qualify design including proof of technologies and industrial manufacture and testing of full size or scalable models of key components. The ITER design is at an advanced stage of maturity and contains sufficient technical information for a construction decision. The operation of ITER will demonstrate the availability of a new energy source, fusion. (author)

  4. ITER Overview

    International Nuclear Information System (INIS)

    Shimomura, Y.; Aymar, R.; Chuyanov, V.; Huguet, M.; Parker, R.

    1999-01-01

    This report summarizes technical works of six years done by the ITER Joint Central Team and Home Teams under terms of Agreement of the ITER Engineering Design Activities. The major products are as follows: complete and detailed engineering design with supporting assessments, industrial-based cost estimates and schedule, non-site specific comprehensive safety and environmental assessment, and technology R and D to validate and qualify design including proof of technologies and industrial manufacture and testing of full size or scalable models of key components. The ITER design is at an advanced stage of maturity and contains sufficient technical information for a construction decision. The operation of ITER will demonstrate the availability of a new energy source, fusion. (author)

  5. β-Adrenergic blockade does not impair the skin blood flow sensitivity to local heating in burned and nonburned skin under neutral and hot environments in children.

    Science.gov (United States)

    Rivas, Eric; McEntire, Serina J; Herndon, David N; Mlcak, Ronald P; Suman, Oscar E

    2017-05-01

    We tested the hypothesis that propranolol, a drug given to burn patients to reduce hypermetabolism/cardiac stress, may inhibit heat dissipation by changing the sensitivity of skin blood flow (SkBF) to local heating under neutral and hot conditions. In a randomized double-blind study, a placebo was given to eight burned children, while propranolol was given to 13 burned children with similar characteristics (mean±SD: 11.9±3 years, 147±20 cm, 45±23 kg, 56±12% Total body surface area burned). Nonburned children (n=13, 11.4±3 years, 152±15 cm, 52±13 kg) served as healthy controls. A progressive local heating protocol characterized SkBF responses in burned and unburned skin and nonburned control skin under the two environmental conditions (23 and 34°C) via laser Doppler flowmetry. Resting SkBF was greater in burned and unburned skin compared to the nonburned control (main effect: skin, Pburned; 38±36 unburned vs 9±8 control %SkBF max ). No difference was found for maximal SkBF capacity to local heating between groups. Additionally, dose-response curves for the sensitivity of SkBF to local heating were not different among burned or unburned skin, and nonburned control skin (EC 50 , P>.05) under either condition. Therapeutic propranolol does not negatively affect SkBF under neutral or hot environmental conditions and further compromise temperature regulation in burned children. © 2017 John Wiley & Sons Ltd.

  6. Reduction of ion thermal diffusivity associated with the transition of the radial electric field in neutral-beam-heated plasmas in the large helical device.

    Science.gov (United States)

    Ida, K; Funaba, H; Kado, S; Narihara, K; Tanaka, K; Takeiri, Y; Nakamura, Y; Ohyabu, N; Yamazaki, K; Yokoyama, M; Murakami, S; Ashikawa, N; deVries, P C; Emoto, M; Goto, M; Idei, H; Ikeda, K; Inagaki, S; Inoue, N; Isobe, M; Itoh, K; Kaneko, O; Kawahata, K; Khlopenkov, K; Komori, A; Kubo, S; Kumazawa, R; Liang, Y; Masuzaki, S; Minami, T; Miyazawa, J; Morisaki, T; Morita, S; Mutoh, T; Muto, S; Nagayama, Y; Nakanishi, H; Nishimura, K; Noda, N; Notake, T; Kobuchi, T; Ohdachi, S; Ohkubo, K; Oka, Y; Osakabe, M; Ozaki, T; Pavlichenko, R O; Peterson, B J; Sagara, A; Saito, K; Sakakibara, S; Sakamoto, R; Sanuki, H; Sasao, H; Sasao, M; Sato, K; Sato, M; Seki, T; Shimozuma, T; Shoji, M; Suzuki, H; Sudo, S; Tamura, N; Toi, K; Tokuzawa, T; Torii, Y; Tsumori, K; Yamamoto, T; Yamada, H; Yamada, I; Yamaguchi, S; Yamamoto, S; Yoshimura, Y; Watanabe, K Y; Watari, T; Hamada, Y; Motojima, O; Fujiwara, M

    2001-06-04

    Recent large helical device experiments revealed that the transition from ion root to electron root occurred for the first time in neutral-beam-heated discharges, where no nonthermal electrons exist. The measured values of the radial electric field were found to be in qualitative agreement with those estimated by neoclassical theory. A clear reduction of ion thermal diffusivity was observed after the mode transition from ion root to electron root as predicted by neoclassical theory when the neoclassical ion loss is more dominant than the anomalous ion loss.

  7. The ITER activity

    International Nuclear Information System (INIS)

    Glass, A.J.

    1991-01-01

    The International Thermonuclear Experimental Reactor (ITER) project is a collaboration among four parties, the United States, the Soviet Union, Japan, and the European Communities, to demonstrate the scientific and technological feasibility of fusion power for peaceful purposes. ITER will demonstrate this through the construction of a tokamak fusion reactor capable of generating 1000 megawatts of fusion power. The ITER project has three missions, as follows: (1) Physics mission -- to demonstrate ignition and controlled burn, with pulse durations from 200 to 1000 S; (2) Technology mission -- to demonstrate the technologies essential to a reactor in an integrated system, operating with high reliability and availability in pulsed operation, with steady-state operation as the ultimate goal; and (3) Testing mission -- to test nuclear and high-heat-flux components at flux levels for 1 mw/m 2 , and fluences of order 1 mw-yr/m 2

  8. 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.

  9. Physics fundamentals for ITER

    International Nuclear Information System (INIS)

    Rosenbluth, M.N.

    1999-01-01

    The design of an experimental thermonuclear reactor requires both cutting-edge technology and physics predictions precise enough to carry forward the design. The past few years of worldwide physics studies have seen great progress in understanding, innovation and integration. We will discuss this progress and the remaining issues in several key physics areas. (1) Transport and plasma confinement. A worldwide database has led to an 'empirical scaling law' for tokamaks which predicts adequate confinement for the ITER fusion mission, albeit with considerable but acceptable uncertainty. The ongoing revolution in computer capabilities has given rise to new gyrofluid and gyrokinetic simulations of microphysics which may be expected in the near future to attain predictive accuracy. Important databases on H-mode characteristics and helium retention have also been assembled. (2) Divertors, heat removal and fuelling. A novel concept for heat removal - the radiative, baffled, partially detached divertor - has been designed for ITER. Extensive two-dimensional (2D) calculations have been performed and agree qualitatively with recent experiments. Preliminary studies of the interaction of this configuration with core confinement are encouraging and the success of inside pellet launch provides an attractive alternative fuelling method. (3) Macrostability. The ITER mission can be accomplished well within ideal magnetohydrodynamic (MHD) stability limits, except for internal kink modes. Comparisons with JET, as well as a theoretical model including kinetic effects, predict such sawteeth will be benign in ITER. Alternative scenarios involving delayed current penetration or off-axis current drive may be employed if required. The recent discovery of neoclassical beta limits well below ideal MHD limits poses a threat to performance. Extrapolation to reactor scale is as yet unclear. In theory such modes are controllable by current drive profile control or feedback and experiments should

  10. Development of TiC and TiN coated molybdenum limiter system and initial results of the thermal testing in neutral beam heated JFT-2 tokamak

    International Nuclear Information System (INIS)

    Nakamura, Hiroo; Sengoku, Seio; Maeno, Masaki; Yamamoto, Shin; Seki, Masahiro; Kazawa, Minoru

    1982-06-01

    This paper describes the limiter drive system for TiC and TiN coated molybdenum limiters and the thermal testing results of the TiC coated limiter in the JFT-2 tokamak using neutral beam injection (0.7 MW). To investigate the influence of TiC coated limiter on plasma behavior and adhesion property under tokamak plasma, a full scale limiter test has been performed in the JFT-2. Reproducible plasma was obtained after the plasma conditioning. Maximum heat flux to the limiter, measured by IR camera, was 1.5 -- 6.5 kW/cm 2 in 25 msec. Cracking, exfoliation and melting on TiC coated limiter were not observed, except for a number of arc tracks. Finally, the permissible heat fluxes of TiC coated molybdenum first wall are discussed. (author)

  11. ITER-FEAT operation

    International Nuclear Information System (INIS)

    Shimomura, Y.; Huguet, M.; Mizoguchi, T.; Murakami, Y.; Polevoi, A.R.; Shimada, M.; Aymar, R.; Chuyanov, V.A.; Matsumoto, H.

    2001-01-01

    ITER is planned to be the first fusion experimental reactor in the world operating for research in physics and engineering. The first ten years of operation will be devoted primarily to physics issues at low neutron fluence and the following ten years of operation to engineering testing at higher fluence. ITER can accommodate various plasma configurations and plasma operation modes, such as inductive high Q modes, long pulse hybrid modes and non-inductive steady state modes, with large ranges of plasma current, density, beta and fusion power, and with various heating and current drive methods. This flexibility will provide an advantage for coping with uncertainties in the physics database, in studying burning plasmas, in introducing advanced features and in optimizing the plasma performance for the different programme objectives. Remote sites will be able to participate in the ITER experiment. This concept will provide an advantage not only in operating ITER for 24 hours a day but also in involving the worldwide fusion community and in promoting scientific competition among the ITER Parties. (author)

  12. Charge-exchange and fusion reaction measurements during compression experiments with neutral beam heating in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Kaita, R.; Heidbrink, W.W.; Hammett, G.W.

    1986-04-01

    Adiabatic toroidal compression experiments were performed in conjunction with high power neutral beam injection in the Tokamak Fusion Test Reactor (TFTR). Acceleration of beam ions to energies nearly twice the injection energy was measured with a charge-exchange neutral particle analyzer. Measurements were also made of 2.5 MeV neutrons and 15 MeV protons produced in fusion reactions between the deuterium beam ions and the thermal deuterium and 3 He ions, respectively. When the plasma was compressed, the d(d,n) 3 He fusion reaction rate increased a factor of five, and the 3 He(d,p) 4 He rate by a factor of twenty. These data were simulated with a bounce-averaged Fokker-Planck program, which assumed conservation of angular momentum and magnetic moment during compression. The results indicate that the beam ion acceleration was consistent with adiabatic scaling

  13. ITER blanket designs

    International Nuclear Information System (INIS)

    Gohar, Y.; Parker, R.; Rebut, P.H.

    1995-01-01

    The ITER first wall, blanket, and shield system is being designed to handle 1.5±0.3 GW of fusion power and 3 MWa m -2 average neutron fluence. In the basic performance phase of ITER operation, the shielding blanket uses austenitic steel structural material and water coolant. The first wall is made of bimetallic structure, austenitic steel and copper alloy, coated with beryllium and it is protected by beryllium bumper limiters. The choice of copper first wall is dictated by the surface heat flux values anticipated during ITER operation. The water coolant is used at low pressure and low temperature. A breeding blanket has been designed to satisfy the technical objectives of the Enhanced Performance Phase of ITER operation for the Test Program. The breeding blanket design is geometrically similar to the shielding blanket design except it is a self-cooled liquid lithium system with vanadium structural material. Self-healing electrical insulator (aluminum nitride) is used to reduce the MHD pressure drop in the system. Reactor relevancy, low tritium inventory, low activation material, low decay heat, and a tritium self-sufficiency goal are the main features of the breeding blanket design. (orig.)

  14. 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.

  15. ITER licensing

    International Nuclear Information System (INIS)

    Gordon, C.W.

    2005-01-01

    ITER was fortunate to have four countries interested in ITER siting to the point where licensing discussions were initiated. This experience uncovered the challenges of licensing a first of a kind, fusion machine under different licensing regimes and helped prepare the way for the site specific licensing process. These initial steps in licensing ITER have allowed for refining the safety case and provide confidence that the design and safety approach will be licensable. With site-specific licensing underway, the necessary regulatory submissions have been defined and are well on the way to being completed. Of course, there is still work to be done and details to be sorted out. However, the informal international discussions to bring both the proponent and regulatory authority up to a common level of understanding have laid the foundation for a licensing process that should proceed smoothly. This paper provides observations from the perspective of the International Team. (author)

  16. 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.

  17. 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.

  18. Atomic Physics in the Quest for Fusion Energy and ITER

    International Nuclear Information System (INIS)

    Skinner, Charles H.

    2008-01-01

    The urgent quest for new energy sources has led developed countries, representing over half of the world population, to collaborate on demonstrating the scientific and technological feasibility of magnetic fusion through the construction and operation of ITER. Data on high-Z ions will be important in this quest. Tungsten plasma facing components have the necessary low erosion rates and low tritium retention but the high radiative efficiency of tungsten ions leads to stringent restrictions on the concentration of tungsten ions in the burning plasma. The influx of tungsten to the burning plasma will need to be diagnosed, understood and stringently controlled. Expanded knowledge of the atomic physics of neutral and ionized tungsten will be important to monitor impurity influxes and derive tungsten concentrations. Also, inert gases such as argon and xenon will be used to dissipate the heat flux flowing to the divertor. This article will summarize the spectroscopic diagnostics planned for ITER and outline areas where additional data is needed.

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

    International Nuclear Information System (INIS)

    Marcuzzi, D.; Palma, M. Dalla; Pavei, M.; Heinemann, B.; Kraus, W.; Riedl, R.

    2009-01-01

    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.

  20. 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.

  1. Plasma auxiliary heating and current drive

    International Nuclear Information System (INIS)

    1999-01-01

    Heating and current drive systems must fulfil several roles in ITER operating scenarios: heating through the H-mode transition and to ignition; plasma burn control; current drive and current profile control in steady state scenarios; and control of MHD instabilities. They must also perform ancillary functions, such as assisting plasma start-up and wall conditioning. It is recognized that no one system can satisfy all of these requirements with the degree of flexibility that ITER will require. Four heating and current drive systems are therefore under consideration for ITER: electron cyclotron waves at a principal frequency of 170 GHz; fast waves operating in the range 40-70 MHz (ion cyclotron waves); lower hybrid waves at 5 GHz; and neutral beam injection using negative ion beam technology for operation at 1 MeV energy. It is likely that several of these systems will be employed in parallel. The systems have been chosen on the basis of the maturity of physics understanding and operating experience in current experiments and on the feasibility of applying the relevant technology to ITER. Here, the fundamental physics describing the interaction of these heating systems with the plasma is reviewed, the relevant experimental results in the exploitation of the heating and current drive capabilities of each system are discussed, key aspects of their application to ITER are outlined, and the major technological developments required in each area are summarized. (author)

  2. 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.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

    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

  4. 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.

  5. ITER-FEAT operation

    International Nuclear Information System (INIS)

    Shimomura, Y.; Huget, M.; Mizoguchi, T.; Murakami, Y.; Polevoi, A.; Shimada, M.; Aymar, R.; Chuyanov, V.; Matsumoto, H.

    2001-01-01

    ITER is planned to be the first fusion experimental reactor in the world operating for research in physics and engineering. The first 10 years' operation will be devoted primarily to physics issues at low neutron fluence and the following 10 years' operation to engineering testing at higher fluence. ITER can accommodate various plasma configurations and plasma operation modes such as inductive high Q modes, long pulse hybrid modes, non-inductive steady-state modes, with large ranges of plasma current, density, beta and fusion power, and with various heating and current drive methods. This flexibility will provide an advantage for coping with uncertainties in the physics database, in studying burning plasmas, in introducing advanced features and in optimizing the plasma performance for the different programme objectives. Remote sites will be able to participate in the ITER experiment. This concept will provide an advantage not only in operating ITER for 24 hours per day but also in involving the world-wide fusion communities and in promoting scientific competition among the Parties. (author)

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Niemczewski, Artur P. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    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.

  8. Design progress of the ITER vacuum vessel sectors and port structures

    International Nuclear Information System (INIS)

    Utin, Yu.; Ioki, K.; Alekseev, A.; Bachmann, Ch.; Cho, S.; Chuyanov, V.; Jones, L.; Kuzmin, E.; Morimoto, M.; Nakahira, M.; Sannazzaro, G.

    2007-01-01

    Recent progress of the ITER vacuum vessel (VV) design is presented. As the ITER construction phase approaches, the VV design has been improved and developed in more detail with the focus on better performance, improved manufacture and reduced cost. Based on achievements of manufacturing studies, design improvement of the typical VV Sector (no. 1) has been nearly finalized. Design improvement of other sectors is in progress-in particular, of the VV Sectors no. 2 and no. 3 which interface with tangential ports for the neutral beam (NB) injection. For all sectors, the concept for the in-wall shielding has progressed and developed in more detail. The design progress of the VV sectors has been accompanied by the progress of the port structures. In particular, design of the NB ports was advanced with the focus on the beam-facing components to handle the heat input of the neutral beams. Structural analyses have been performed to validate all design improvements

  9. Two key improvements to enhance the thermo-mechanic performances of accelerator grids for neutral beam injectors

    Energy Technology Data Exchange (ETDEWEB)

    Agostinetti, P., E-mail: piero.agostinetti@igi.cnr.it [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), Corso Stati Uniti 4, 35127 Padova (Italy); Chitarin, G. [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), Corso Stati Uniti 4, 35127 Padova (Italy); University of Padova, Department of Engineering and Management, strad. S. Nicola 3, 36100 Vicenza (Italy); Gambetta, G.; Marcuzzi, D. [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), Corso Stati Uniti 4, 35127 Padova (Italy)

    2016-11-01

    Highlights: • The grids of MITICA and ITER NBIs are subjected to huge heat loads. • With a standard design, fatigue life of the grids was below the ITER requirements. • Thanks to NICE and SRS improvements, ITER requirements are now satisfied. - Abstract: The MITICA experiment (Megavolt ITER Injector & Concept Advancement) is the prototype and the test bed of the Heating and Current Drive Neutral Beam Injectors, which will be necessary for the full-performance exploitation of ITER. MITICA injector experiments shall demonstrate the reliable and accurate emission of a 17 MW beam of neutral particles for duration up to 1 hour, fulfilling ITER specific requirements. The accelerator grids are among the most critical parts of this experiment, because they must fulfill several operational requirements and at the same time satisfy the fatigue verifications according to the ITER Structural Design Criteria for In-vessel Components (SDC-IC). After about two years of continuous development, two design improvements were found to effectively increase the fatigue life of the grids up to the requested values. The first method was to adopt a novel shape of the cooling channels inside the grids, called Nozzle Island Cooling Enhancement (NICE) and able to provide a high performance cooling without exceeding the limits on the pressure drop through the grids. The second, called Stress Relieving Slits (SRS), was to introduce suitable slits in the grids, whose design was iteratively optimized until they were able to significantly reduce the stress/strain peaks due to thermal gradients. The NICE and SRS design solutions, here described in detail, were found to be key improvements in order to obtain a design of the grids able to satisfy all the operating requirements and all the structural verifications according to the ITER criteria.

  10. Three-dimensional modeling of plasma edge transport and divertor fluxes during application of resonant magnetic perturbations on ITER

    Czech Academy of Sciences Publication Activity Database

    Schmitz, O.; Becoulet, M.; Cahyna, Pavel; 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.

    2016-01-01

    Roč. 56, č. 6 (2016), č. článku 066008. ISSN 0029-5515 Institutional support: RVO:61389021 Keywords : resonant magnetic perturbations * plasma edge physics * 3D modeling * neutral particle physics * ITER * divertor heat and particle loads * ELM control Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 3.307, year: 2016 http://iopscience.iop.org/article/10.1088/0029-5515/56/6/066008/meta

  11. ITER shielding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Strebkov, Yu [ENTEK, Moscow (Russian Federation); Avsjannikov, A [ENTEK, Moscow (Russian Federation); Baryshev, M [NIAT, Moscow (Russian Federation); Blinov, Yu [ENTEK, Moscow (Russian Federation); Shatalov, G [KIAE, Moscow (Russian Federation); Vasiliev, N [KIAE, Moscow (Russian Federation); Vinnikov, A [ENTEK, Moscow (Russian Federation); Chernjagin, A [DYNAMICA, Moscow (Russian Federation)

    1995-03-01

    A reference non-breeding blanket is under development now for the ITER Basic Performance Phase for the purpose of high reliability during the first stage of ITER operation. More severe operation modes are expected in this stage with first wall (FW) local heat loads up to 100-300Wcm{sup -2}. Integration of a blanket design with protective and start limiters requires new solutions to achieve high reliability, and possible use of beryllium as a protective material leads to technologies. The rigid shielding blanket concept was developed in Russia to satisfy the above-mentioned requirements. The concept is based on a copper alloy FW, austenitic stainless steel blanket structure, water cooling. Beryllium protection is integrated in the FW design. Fabrication technology and assembly procedure are described in parallel with the equipment used. (orig.).

  12. Status of ITER

    International Nuclear Information System (INIS)

    Aymar, R.

    2002-01-01

    At the end of engineering design activities (EDA) in July 2001, all the essential elements became available to make a decision on construction of ITER. A sufficiently detailed and integrated engineering design now exists for a generic site, has been assessed for feasibility, and costed, and essential physics and technology R and D has been carried out to underpin the design choices. Formal negotiations have now begun between the current participants--Canada, Euratom, Japan, and the Russian Federation--on a Joint Implementation Agreement for ITER which also establishes the legal entity to run ITER. These negotiations are supported on technical aspects by Coordinated Technical Activities (CTA), which maintain the integrity of the project, for the good of all participants, and concentrate on preparing for procurement by industry of the longest lead items, and for formal application for a construction license with the host country. This paper highlights the main features of the ITER design. With cryogenically-cooled magnets close to neutron-generating plasma, the design of shielding with adequate access via port plugs for auxiliaries such as heating and diagnostics, and of remote replacement and refurbishing systems for in-vessel components, are particularly interesting nuclear technology challenges. Making a safety case for ITER to satisfy potential regulators and to demonstrate, as far as possible at this stage, the environmental attractiveness of fusion as an energy source, is also important. The paper gives illustrative details on this work, and an update on the progress of technical preparations for construction, as well as the status of the above negotiations

  13. 130 kV 130 A High voltage switching mode power supply for neutral beam plasma heating: design issues

    International Nuclear Information System (INIS)

    Ganuza, D.; Del Rio, J.M.; Garcia, I.; Garcia, F.; Garcia de Madinabeitia, P.; Perez, A.; Zabaleta, J.R.

    2003-01-01

    The company JEMA has designed and manufactured two High Voltage Switching Mode Power Supplies (HVSMPS), rated at 130 kV dc and 130 A, each of which will feed the accelerator grids of two Positive Ion Neutral Injector (PINI) loads, to be installed at the Joint European Torus (EFDA-JET facility located at Culham, UK). The solution designed by JEMA includes two matching transformers which adapt the 36 kV of the JET AC power distribution network to the required 670 V at the secondary side. Additionally, such transformers provide a 30 deg.phase shift which is required by a 30000 A 12 pulse thyristor rectifier. The obtained and stabilised 650 V feed 120 IGBT invertors, which operate at 2778 Hz with modulated square waveform. Each invertor feeds a High Insulation High Frequency Transformer. The 120 transformers corresponding to one power supply are arranged in three oil filled tanks and provide the main insulation from the low voltage to the high voltage side. The square waveform obtained at the secondary of each transformer is rectified by means of a diode bridge. The connection in series of the 120 diode bridges provides the required 130 kV d.c. at the output. In order to protect the load, a redundant solid state crowbar has been designed. Such short circuiting device is composed of 26 Light Triggered Thyristors (LTTs), connected in series. Electrical simulations have been carried out in order to ensure that the system complies with the requirements of high accuracy and adequate protection of the load. The critical design of the High Voltage-High Frequency Transformers has also required electrostatic simulations of the electric field distribution

  14. 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.

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

    International Nuclear Information System (INIS)

    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.

    2011-01-01

    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 and 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.

  16. Updated safety analysis of ITER

    International Nuclear Information System (INIS)

    Taylor, Neill; Baker, Dennis; Ciattaglia, Sergio; Cortes, Pierre; Elbez-Uzan, Joelle; Iseli, Markus; Reyes, Susana; Rodriguez-Rodrigo, Lina; Rosanvallon, Sandrine; Topilski, Leonid

    2011-01-01

    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.

  17. 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.

  18. 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

    International Nuclear Information System (INIS)

    Villar Colome, J.

    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

  19. 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

  20. ITER cooling system

    International Nuclear Information System (INIS)

    Kveton, O.K.

    1990-11-01

    The present specification of the ITER cooling system does not permit its operation with water above 150 C. However, the first wall needs to be heated to higher temperatures during conditioning at 250 C and bake-out at 350 C. In order to use the cooling water for these operations the cooling system would have to operate during conditioning at 37 Bar and during bake-out at 164 Bar. This is undesirable from the safety analysis point of view, and alternative heating methods are to be found. This review suggests that superheated steam or gas heating can be used for both baking and conditioning. The blanket design must consider the use of dual heat transfer media, allowing for change from one to another in both directions. Transfer from water to gas or steam is the most intricate and risky part of the entire heating process. Superheated steam conditioning appears unfavorable. The use of inert gas is recommended, although alternative heating fluids such as organic coolant should be investigated

  1. Iterative Splitting Methods for Differential Equations

    CERN Document Server

    Geiser, Juergen

    2011-01-01

    Iterative Splitting Methods for Differential Equations explains how to solve evolution equations via novel iterative-based splitting methods that efficiently use computational and memory resources. It focuses on systems of parabolic and hyperbolic equations, including convection-diffusion-reaction equations, heat equations, and wave equations. In the theoretical part of the book, the author discusses the main theorems and results of the stability and consistency analysis for ordinary differential equations. He then presents extensions of the iterative splitting methods to partial differential

  2. Purification and characterization of a novel neutral and heat-tolerant phytase from a newly isolated strain Bacillus nealsonii ZJ0702

    Science.gov (United States)

    2013-01-01

    Background Phytic acid and phytates can interact with biomolecules, such as proteins and carbohydrates, and are anti-nutritional factors found in food and feed. Therefore, it is necessary to remove these compounds in food and feed processing. Phytase can hydrolyze phytic acid and phytates to release a series of lower phosphate esters of myoinositol and orthophosphate. Thus, the purification and characterization of novel phytases that can be used in food and feed processing is of particular interest to the food and feed industries. Results A novel neutral and heat-tolerant phytase from a newly isolated strain Bacillus nealsonii ZJ0702 was purified to homogeneity with a yield of 5.7% and a purification fold of 44. The molecular weight of the purified phytase obtained by SDS-PAGE was 43 kDa. The homology analysis based on N-terminal amino acid and DNA sequencing indicated that the purified phytase was different from other known phytases. The optimal thermal and pH activity of the phytase was observed at 55°C and 7.5, respectively. Seventy-three percent of the original activity of the phytase was maintained following incubation at 90°C for 10 min. The phytase was stable within a pH range of 6.0 − 8.0 and showed high substrate specificity for sodium phytate. Cu2+, Co2+, Zn2+, Mn2+, Ba2+ and Ni2+ ions were found to inhibit the activity of the phytase. Conclusions A novel phytase purified from B. nealsonii ZJ0702 was identified. The phytase was found to be thermally stable over a wide temperature range at neutral pH. These properties suggest that this phytase is a suitable alternative to fungal phytases for the hydrolysis of phytic acid and phytates in food and feed processing industries. PMID:24073799

  3. ITER council proceedings: 2001

    International Nuclear Information System (INIS)

    2001-01-01

    Continuing the ITER EDA, two further ITER Council Meetings were held since the publication of ITER EDA documentation series no, 20, namely the ITER Council Meeting on 27-28 February 2001 in Toronto, and the ITER Council Meeting on 18-19 July in Vienna. That Meeting was the last one during the ITER EDA. This volume contains records of these Meetings, including: Records of decisions; List of attendees; ITER EDA status report; ITER EDA technical activities report; MAC report and advice; Final report of ITER EDA; and Press release

  4. ITER EDA newsletter. V. 7, no. 5

    International Nuclear Information System (INIS)

    1998-05-01

    This newsletter contains the articles 'The materials selection in ITER and the first materials workshop', 'US fusion community discussion on fusion strategies', 'ITER central solenoid model coil heat treatment complete and assembly started' and 'Programme of the 17th IAEA fusion energy conference'. There is also a note in memoriam of Hiroschi Shibata, who died on the 5th of June 1998

  5. 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.

  6. ITER cooling systems

    International Nuclear Information System (INIS)

    Natalizio, A.; Hollies, R.E.; Sochaski, R.O.; Stubley, P.H.

    1992-06-01

    The ITER reference system uses low-temperature water for heat removal and high-temperature helium for bake-out. As these systems share common equipment, bake-out cannot be performed until the cooling system is drained and dried, and the reactor cannot be started until the helium has been purged from the cooling system. This study examines the feasibility of using a single high-temperature fluid to perform both heat removal and bake-out. The high temperature required for bake-out would also be in the range for power production. The study examines cost, operational benefits, and impact on reactor safety of two options: a high-pressure water system, and a low-pressure organic system. It was concluded that the cost savings and operational benefits are significant; there are no significant adverse safety impacts from operating either the water system or the organic system; and the capital costs of both systems are comparable

  7. Comparison of experimentally-inferred ion thermal diffusivities with neoclassical theory for neutral beam-heated discharges in the Doublet III tokamak

    International Nuclear Information System (INIS)

    Groebner, R.J.

    1986-04-01

    The study of ion transport in neutral beam-heated discharges in tokamaks is necessary to determine if neoclassical theory can reliably be used to predict the performance of future machines. Previous studies of ion tranport have generally been difficult due to the lack of information regarding the ion temperature profile. The standard procedure used to study ion transport has been to model the T/sub i/ profile with the assumption that the ion thermal diffusivity profile chi/sub i/(r) was equal to a multiplier times chi/sub i//sup neo/(r), the ion thermal diffusivity calculated from neoclassical theory. The multiplier was varied until the calculated T/sub i/ profile agreed with the available ion temperature data, usually T/sub i/(0) or the measured neutron rate. Values of the multiplier in the range of 1 to 10 have generally been obtained with few estimates of the uncertainties in these values. Furthermore, there have been few, if any, attempts to calculate chi/sub i/ by modeling the ion temperature profiles in other ways. As a result, the issue as to whether or not the ion transport in tokamaks is in agreement with neoclassical theory has not been definitively answered

  8. 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

  9. ITER EDA newsletter. V. 8, no. 10

    International Nuclear Information System (INIS)

    1999-10-01

    This ITER EDA Newsletter contains summary reports on the seventh meeting of the ITER physics expert group on energetic particles, heating and steady state operations in Frascati, Italy, on the fifth international symposium on fusion nuclear technology in Rome, Italy and on the IAEA technical committee meeting on electron cyclotron resonance heating physics and technology for fusion devices in Oh-arai, Japan. Individual abstracts are prepared for the three contributions

  10. ITER ITA newsletter. No. 23, June 2005

    International Nuclear Information System (INIS)

    2005-06-01

    This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about ITER related meeting the Eighth Meeting of the ITPA Topical Group (TG) on Diagnostics was held at the Culham Science Centre, UKAEA, from 14-18 March 2005 and the Third International Atomic Energy Agency - Technical Meeting (TM) on Electron Cyclotron Resonance Heating (ECRH) in ITER, following those in Oharai, Japan in 1999, and in Kloster Seeon, Germany in 2003, was held in Como, Italy, from May 2 to May 5, 2005, in a two-and-half day intense workshop

  11. Key enabling design features of the ITER HNB Duct Liner

    Energy Technology Data Exchange (ETDEWEB)

    Chuilon, Ben, E-mail: ben.chuilon@ccfe.ac.uk; Mistry, Sanjay; Andrews, Rodney; Verhoeven, Roel; Xue, Yongkuan

    2015-10-15

    Highlights: • Key engineering design details of the ITER HND Duct Liner are presented. • A standardised CuCrZr water cooled panel that can be remotely handled is detailed. • Bolts are protected from beam power by means of a tungsten cap to radiate heat away. • Water connections placed coaxially are protected from beam power by a tungsten ring. • Explosion-bonded CuCrZr-316L panels result in a tenfold disruption torque reduction. - Abstract: The Duct Liner (DL) for the ITER Heating Neutral Beam (HNB) is a key component in the beam transport system. Duct Liners installed into equatorial ports 4 and 5 of the Vacuum Vessel (VV) will protect the port extension from power deposition due to re-ionisation and direct interception of the HNB. Furthermore, the DL contributes towards the shielding of the VV and superconducting coils from plasma photons and neutrons. The DL incorporates a 316L(N)-IG, deep-drilled and water cooled Neutron Shield (NS) whose internal walls are lined with actively cooled CuCrZr Duct Liner Modules (DLMs). These Remote Handling Class 2 and 3 panels provide protection from neutral beam power. This paper provides an overview of the preliminary design for the ITER HNB DL and focusses on critical features that ensure compatibility with: high heat flux requirements, remote maintenance procedures, and transient magnetic fields arising from major plasma disruptions. The power deposited on a single DLM can reach 300 kW with a peak power density of 2.4 MW/m{sup 2}. Feeding coolant to the DLMs is accomplished via welded connections to the internal coolant network of the NS. These are placed coaxially to allow for thermal expansion of the DLMs without the use of deformable connections. Critically, the remote maintenance of individual DLMs necessitates access to water connections and bolts from the beam facing surface, thus subjecting them to high heat flux loads. This design challenge will become more prevalent as fusion devices become more powerful

  12. US--ITER activation analysis

    International Nuclear Information System (INIS)

    Attaya, H.; Gohar, Y.; Smith, D.

    1990-09-01

    Activation analysis has been made for the US ITER design. The radioactivity and the decay heat have been calculated, during operation and after shutdown for the two ITER phases, the Physics Phase and the Technology Phase. The Physics Phase operates about 24 full power days (FPDs) at fusion power level of 1100 MW and the Technology Phase has 860 MW fusion power and operates for about 1360 FPDs. The point-wise gamma sources have been calculated everywhere in the reactor at several times after shutdown of the two phases and are then used to calculate the biological dose everywhere in the reactor. Activation calculations have been made also for ITER divertor. The results are presented for different continuous operation times and for only one pulse. The effect of the pulsed operation on the radioactivity is analyzed. 6 refs., 12 figs., 1 tab

  13. Thermo-mechanical analysis of an acceleration grid for the international thermonuclear experimental reactor-neutral beam injection system

    International Nuclear Information System (INIS)

    Fujiwara, Yukio; Hanada, Masaya; Okumura, Yoshikazu; Suzuki, Satoshi; Watanabe, Kazuhiro

    2001-01-01

    In the engineering design of a negative-ion beam source for a high-power neutral beam injection (NBI) system, one of the most important issues is thermo-mechanical design of acceleration grids for producing several tens of MW ion beams. An acceleration grid for the international thermonuclear experimental reactor-neutral beam injection (ITER-NBI) system will be subjected to the heat loading as high as 1.5 MW. In the present paper, thermo-mechanical characteristics of the acceleration grid for the ITER-NBI system were analyzed. Numerical simulation indicated that maximum aperture-axis displacement of the acceleration grid due to thermal expansion would be about 0.7 mm for the heat loading of 1.5 MW. From the thin lens theory of beam optics, beamlet deflection angle by the aperture-axis displacement was estimated to be about 2 mrad, which is within the requirement of the engineering design of the ITER-NBI system. Numerical simulation also indicated that no melting on the acceleration grid would occur for a heat loading of 1.5 MW, while local plastic deformation would happen. To avoid the plastic deformation, it is necessary to reduce the heat loading onto the acceleration grid to less than 1 MW

  14. ITER safety challenges and opportunities

    International Nuclear Information System (INIS)

    Piet, S.J.

    1991-01-01

    Results of the Conceptual Design Activity (CDA) for the International Thermonuclear Experimental Reactor (ITER) suggest challenges and opportunities. ''ITER is capable of meeting anticipated regulatory dose limits,'' but proof is difficult because of large radioactive inventories needing stringent radioactivity confinement. We need much research and development (R ampersand D) and design analysis to establish that ITER meets regulatory requirements. We have a further opportunity to do more to prove more of fusion's potential safety and environmental advantages and maximize the amount of ITER technology on the path toward fusion power plants. To fulfill these tasks, we need to overcome three programmatic challenges and three technical challenges. The first programmatic challenge is to fund a comprehensive safety and environmental ITER R ampersand D plan. Second is to strengthen safety and environment work and personnel in the international team. Third is to establish an external consultant group to advise the ITER Joint Team on designing ITER to meet safety requirements for siting by any of the Parties. The first of the three key technical challenges is plasma engineering -- burn control, plasma shutdown, disruptions, tritium burn fraction, and steady state operation. The second is the divertor, including tritium inventory, activation hazards, chemical reactions, and coolant disturbances. The third technical challenge is optimization of design requirements considering safety risk, technical risk, and cost. Some design requirements are now too strict; some are too lax. Fuel cycle design requirements are presently too strict, mandating inappropriate T separation from H and D. Heat sink requirements are presently too lax; they should be strengthened to ensure that maximum loss of coolant accident temperatures drop

  15. ITER council proceedings: 1998

    International Nuclear Information System (INIS)

    1999-01-01

    This volume contains documents of the 13th and the 14th ITER council meeting as well as of the 1st extraordinary ITER council meeting. Documents of the ITER meetings held in Vienna and Yokohama during 1998 are also included. The contents include an outline of the ITER objectives, the ITER parameters and design overview as well as operating scenarios and plasma performance. Furthermore, design features, safety and environmental characteristics are given

  16. Telescope-based cavity for negative ion beam neutralization in future fusion reactors.

    Science.gov (United States)

    Fiorucci, Donatella; Hreibi, Ali; Chaibi, Walid

    2018-03-01

    In future fusion reactors, heating system efficiency is of the utmost importance. Photo-neutralization substantially increases the neutral beam injector (NBI) efficiency with respect to the foreseen system in the International Thermonuclear Experimental Reactor (ITER) based on a gaseous target. In this paper, we propose a telescope-based configuration to be used in the NBI photo-neutralizer cavity of the demonstration power plant (DEMO) project. This configuration greatly reduces the total length of the cavity, which likely solves overcrowding issues in a fusion reactor environment. Brought to a tabletop experiment, this cavity configuration is tested: a 4 mm beam width is obtained within a ≃1.5  m length cavity. The equivalent cavity g factor is measured to be 0.038(3), thus confirming the cavity stability.

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

    International Nuclear Information System (INIS)

    Serikov, A.; Fischer, U.; Grosse, D.; Leichtle, D.; Majerle, M.

    2011-01-01

    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. Conceptual design of cesium removal device for ITER NBI maintenance

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-03-01

    Cesium is required in order to generate a stable negative ion of hydrogen in an ion source of the neutral beam injector (NBI), which is one of the plasma-heating devices for International Thermonuclear Experimental Reactor (ITER). After long time operation of the NBI, the cesium deposits to the insulators supporting the electrode. Due to the deterioration of the insulation resistance, the continuous operation of the NBI will be difficult. In addition, the NBI device is activated by neutrons from D-T plasma, so that periodic removal and cleaning of the cesium on the insulators by remove handling is required. A study of the cesium removal scenario and the device is therefore required considering remote handling. In this report, a cesium removal procedure and conceptual design of the cesium removal device using laser ablation technique are studied, and the feasibility of the laser ablation method is shown. (author)

  19. An iterative regularization method in estimating the transient heat-transfer rate on the surface of the insulation layer of a double circular pipe

    International Nuclear Information System (INIS)

    Chen, W.-L.; Yang, Y.-C.

    2009-01-01

    In this study, a conjugate gradient method based inverse algorithm is applied to estimate the unknown space- and time-dependent heat-transfer rate on the surface of the insulation layer of a double circular pipe heat exchanger using temperature measurements. It is assumed that no prior information is available on the functional form of the unknown heat-transfer rate; hence the procedure is classified as the function estimation in inverse calculation. The temperature data obtained from the direct problem are used to simulate the temperature measurements. The accuracy of the inverse analysis is examined by using simulated exact and inexact temperature measurements. Results show that an excellent estimation on the space- and time-dependent heat-transfer rate can be obtained for the test case considered in this study.

  20. ITER Council proceedings: 1993

    International Nuclear Information System (INIS)

    1994-01-01

    Records of the third ITER Council Meeting (IC-3), held on 21-22 April 1993, in Tokyo, Japan, and the fourth ITER Council Meeting (IC-4) held on 29 September - 1 October 1993 in San Diego, USA, are presented, giving essential information on the evolution of the ITER Engineering Design Activities (EDA), such as the text of the draft of Protocol 2 further elaborated in ''ITER EDA Agreement and Protocol 2'' (ITER EDA Documentation Series No. 5), recommendations on future work programmes: a description of technology R and D tasks; the establishment of a trust fund for the ITER EDA activities; arrangements for Visiting Home Team Personnel; the general framework for the involvement of other countries in the ITER EDA; conditions for the involvement of Canada in the Euratom Contribution to the ITER EDA; and other attachments as parts of the Records of Decision of the aforementioned ITER Council Meetings

  1. ITER council proceedings: 2000

    International Nuclear Information System (INIS)

    2001-01-01

    No ITER Council Meetings were held during 2000. However, two ITER EDA Meetings were held, one in Tokyo, January 19-20, and one in Moscow, June 29-30. The parties participating in these meetings were those that partake in the extended ITER EDA, namely the EU, the Russian Federation, and Japan. This document contains, a/o, the records of these meetings, the list of attendees, the agenda, the ITER EDA Status Reports issued during these meetings, the TAC (Technical Advisory Committee) reports and recommendations, the MAC Reports and Advice (also for the July 1999 Meeting), the ITER-FEAT Outline Design Report, the TAC Reports and Recommendations both meetings), Site requirements and Site Design Assumptions, the Tentative Sequence of technical Activities 2000-2001, Report of the ITER SWG-P2 on Joint Implementation of ITER, EU/ITER Canada Proposal for New ITER Identification

  2. ITER council proceedings: 1993

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-31

    Records of the third ITER Council Meeting (IC-3), held on 21-22 April 1993, in Tokyo, Japan, and the fourth ITER Council Meeting (IC-4) held on 29 September - 1 October 1993 in San Diego, USA, are presented, giving essential information on the evolution of the ITER Engineering Design Activities (EDA), such as the text of the draft of Protocol 2 further elaborated in ``ITER EDA Agreement and Protocol 2`` (ITER EDA Documentation Series No. 5), recommendations on future work programmes: a description of technology R and D tastes; the establishment of a trust fund for the ITER EDA activities; arrangements for Visiting Home Team Personnel; the general framework for the involvement of other countries in the ITER EDA; conditions for the involvement of Canada in the Euratom Contribution to the ITER EDA; and other attachments as parts of the Records of Decision of the aforementioned ITER Council Meetings.

  3. Neutral currents

    International Nuclear Information System (INIS)

    Paschos, E.A.

    1977-01-01

    It is stated that over the past few years considerable progress has been made in the field of weak interactions. The existence of neutral currents involving leptons and hadrons has been established and some of the questions concerning their detailed structure have been answered. This imposes constraints on the gauge theories and has eliminated large classes of models. New questions have also been raised, one of which concerns the conservation laws obeyed by neutral currents. The wide range of investigations is impressive and is expected to continue with new results from particle, nuclear, and atomic physics. Headings include - various aspects of a gauge theory (choice of group, the symmetry breaking scheme, representation assignments for fermion fields); space-time structure; isospin structure; leptonic neutral currents; and atomic experiments. (U.K.)

  4. Neutral currents

    International Nuclear Information System (INIS)

    Aubert, B.

    1994-11-01

    The evidence for the existence of weak neutral current has been a very controverted topics in the early 1970's, as well as the muon did in the 1930's. The history is very rich considering the evolution of the experimental techniques in high energy particle physics. The history of the discovery and the study of weak neutral current is reviewed. Later the quest of the intermediate vector boson continues with the decision of the community to build a large proton antiproton collider. (K.A.). 14 refs., 1 fig

  5. 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.

  6. Numerical investigation on transverse heat transfer properties in cross section of full size Nb{sub 3}Sn CICC ITER conductor

    Energy Technology Data Exchange (ETDEWEB)

    Jia, Shuming; Wang, Dengming [Key Laboratory of Mechanics on Environment and Disaster in Western China, Ministry of Education, Lanzhou University, Lanzhou 730000 (China); Department of Mechanics, School of Civil Engineering and Mechanics, Lanzhou University, Lanzhou 730000 (China); Zheng, Xiaojing, E-mail: xjzheng@lzu.edu.cn [Key Laboratory of Mechanics on Environment and Disaster in Western China, Ministry of Education, Lanzhou University, Lanzhou 730000 (China); Department of Mechanics, School of Civil Engineering and Mechanics, Lanzhou University, Lanzhou 730000 (China); School of Electronical and Machanical Engineering, Xidian University, Xi’an, 710071 (China)

    2015-05-15

    The contact mechanical characteristics in the cross section of the Nb{sub 3}Sn 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Riccardo, V., E-mail: valeria.riccardo@ccfe.ac.uk [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Lomas, P.; Matthews, G.F. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Nunes, I. [Associação EURATOM-IST, IPFN – Laboratório Associado, IST, Lisbon (Portugal); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Thompson, V. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Villedieu, E. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom)

    2013-10-15

    Highlights: ► 40 m{sup 2} of plasma facing surface covered with bulk Be re-using existing supports, designed for C-based tiles (hence for much lower disruption loads). ► Optimization of power handling to allow compatibility with higher (×1.5) and longer (×2) neutral beam power. ► Beryllium re-cycling. ► Machining and cleaning to ultra high vacuum standards of <350 μm thin castellations in Be. ► Quality control to minimize installation problems (proto-types, full scale jigs, inspections). -- Abstract: 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 reduced radiative losses compared to carbon, leaving most of the plasma magnetic energy to be conducted to the wall and requiring routine disruption mitigation. The main chamber power handling has achieved and possibly exceeded the design targets.

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

    International Nuclear Information System (INIS)

    Riccardo, V.; Lomas, P.; Matthews, G.F.; Nunes, I.; Thompson, V.; Villedieu, E.

    2013-01-01

    Highlights: ► 40 m 2 of plasma facing surface covered with bulk Be re-using existing supports, designed for C-based tiles (hence for much lower disruption loads). ► Optimization of power handling to allow compatibility with higher (×1.5) and longer (×2) neutral beam power. ► Beryllium re-cycling. ► Machining and cleaning to ultra high vacuum standards of <350 μm thin castellations in Be. ► Quality control to minimize installation problems (proto-types, full scale jigs, inspections). -- Abstract: 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 reduced radiative losses compared to carbon, leaving most of the plasma magnetic energy to be conducted to the wall and requiring routine disruption mitigation. The main chamber power handling has achieved and possibly exceeded the design targets

  9. Coarse Grained Transport Model for Neutrals in Turbulent SOL Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Marandet, Y.; Mekkaoui, A.; Genesio, P.; Rosato, J.; Capes, H.; Godbert-Mouret, L.; Koubiti, M.; Stamm, R., E-mail: yannick.marandet@univ-amu.fr [PIIM, CNRS/Aix-Marseille University, Marseille (France); Reiter, D.; Boerner, P. [IEK4, FZJ, Juelich (Germany)

    2012-09-15

    Full text: Edge plasmas of magnetic fusion devices exhibit strong intermittent turbulence, which governs perpendicular transport of particles and heat. Turbulent fluxes result from the coarse graining procedure used to derive the transport equation, which entails time averaging of the underlying equations governing the turbulent evolution of the electron and ion fluids. In previous works, we have pointed out that this averaging is not carried out on the Boltzmann equation that describes the transport of neutral particles (atoms, molecules) in current edge code suites (such as SOLPS). Since fluctuations in the far SOL are of order unity, calculating the transport of neutral particles, hence the source terms in plasma fluid equations, in the average plasma background might lead to misleading results. In particular, retaining the effects of fluctuations could affect the estimation of the importance of main chamber recycling, hence first wall sputtering by charge exchange atoms, as well as main chamber impurity contamination and transport. In this contribution, we obtain an exact coarse-grained equation for the average neutral density, assuming that density fluctuations are described by multivariate Gamma statistics. This equation is a scattering free Boltzmann equation, where the ionization rate has been renormalized to account for fluctuations. The coarse grained transport model for neutrals has been implemented in the EIRENE code, and applications in 2D geometry with ITER relevant plasma parameters are presented. Our results open the way for the implementation of the effects of turbulent fluctuations on the transport of neutral particles in coupled plasma/neutral edge codes like B2-EIRENE. (author)

  10. The isotope effect on divertor conditions and neutral pumping in horizontal divertor configurations in JET-ILW Ohmic plasmas

    Directory of Open Access Journals (Sweden)

    J. Uljanovs

    2017-08-01

    Full Text Available Understanding the impact of isotope mass and divertor configuration on the divertor conditions and neutral pressures is critical for predicting the performance of the ITER divertor in DT operation. To address this need, ohmically heated hydrogen and deuterium plasma experiments were conducted in JET with the ITER-like wall in varying divertor configurations. In this study, these plasmas are simulated with EDGE2D-EIRENE outfitted with a sub-divertor model, to predict the neutral pressures in the plenum with similar fashion to the experiments. EDGE2D-EIRENE predictions show that the increased isotope mass results in up to a 25% increase in peak electron densities and 15% increase in peak ion saturation current at the outer target in deuterium when compared to hydrogen for all horizontal divertor configurations. Indicating that a change from hydrogen to deuterium as main fuel decreases the neutral mean free path, leading to higher neutral density in the divertor. Consequently, this mechanism also leads to higher neutral pressures in the sub-divertor. The experimental data provided by the hydrogen and deuterium ohmic discharges shows that closer proximity of the outer strike point to the pumping plenum results in a higher neutral pressure in the sub-divertor. The diaphragm capacitance gauge pressure measurements show that a two to three-fold increase in sub-divertor pressure was achieved in the corner and nearby horizontal configurations compared to the far-horizontal configurations, likely due to ballistic transport (with respect to the plasma facing components of the neutrals into the sub-divertor. The corner divertor configuration also indicates that a neutral expansion occurs during detachment, resulting in a sub-divertor neutral density plateau as a function of upstream density at the outer-mid plane.

  11. Targets for high power neutral beams

    International Nuclear Information System (INIS)

    Kim, J.

    1980-01-01

    Stopping high-power, long-pulse beams is fast becoming an engineering challenge, particularly in neutral beam injectors for heating magnetically confined plasmas. A brief review of neutral beam target technology is presented along with heat transfer calculations for some selected target designs

  12. ITER council proceedings: 1995

    International Nuclear Information System (INIS)

    1996-01-01

    Records of the 8. ITER Council Meeting (IC-8), held on 26-27 July 1995, in San Diego, USA, and the 9. ITER Council Meeting (IC-9) held on 12-13 December 1995, in Garching, Germany, are presented, giving essential information on the evolution of the ITER Engineering Design Activities (EDA) and the ITER Interim Design Report Package and Relevant Documents. Figs, tabs

  13. ITER council proceedings: 1999

    International Nuclear Information System (INIS)

    1999-01-01

    In 1999 the ITER meeting in Cadarache (10-11 March 1999) and the Programme Directors Meeting in Grenoble (28-29 July 1999) took place. Both meetings were exclusively devoted to ITER engineering design activities and their agendas covered all issues important for the development of ITER. This volume presents the documents of these two important meetings

  14. ITER council proceedings: 1996

    International Nuclear Information System (INIS)

    1997-01-01

    Records of the 10. ITER Council Meeting (IC-10), held on 26-27 July 1996, in St. Petersburg, Russia, and the 11. ITER Council Meeting (IC-11) held on 17-18 December 1996, in Tokyo, Japan, are presented, giving essential information on the evolution of the ITER Engineering Design Activities (EDA) and the cost review and safety analysis. Figs, tabs

  15. ITER EDA technical activities

    International Nuclear Information System (INIS)

    Aymar, R.

    1998-01-01

    Six years of technical work under the ITER EDA Agreement have resulted in a design which constitutes a complete description of the ITER device and of its auxiliary systems and facilities. The ITER Council commented that the Final Design Report provides the first comprehensive design of a fusion reactor based on well established physics and technology

  16. The remote handling systems for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Ribeiro, Isabel, E-mail: mir@isr.ist.utl.pt [Institute for Systems and Robotics/Instituto Superior Tecnico, Lisboa (Portugal); Damiani, Carlo [Fusion for Energy, Barcelona (Spain); Tesini, Alessandro [ITER Organization, Cadarache (France); Kakudate, Satoshi [ITER Tokamak Device Group, Japan Atomic Energy Agency, Ibaraki (Japan); Siuko, Mikko [VTT Systems Engineering, Tampere (Finland); Neri, Carlo [Associazione EURATOM ENEA, Frascati (Italy)

    2011-10-15

    The ITER remote handling (RH) maintenance system is a key component in ITER operation both for scheduled maintenance and for unexpected situations. It is a complex collection and integration of numerous systems, each one at its turn being the integration of diverse technologies into a coherent, space constrained, nuclearised design. This paper presents an integrated view and recent results related to the Blanket RH System, the Divertor RH System, the Transfer Cask System (TCS), the In-Vessel Viewing System, the Neutral Beam Cell RH System, the Hot Cell RH and the Multi-Purpose Deployment System.

  17. ITER-FEAT safety

    International Nuclear Information System (INIS)

    Gordon, C.W.; Bartels, H.-W.; Honda, T.; Raeder, J.; Topilski, L.; Iseli, M.; Moshonas, K.; Taylor, N.; Gulden, W.; Kolbasov, B.; Inabe, T.; Tada, E.

    2001-01-01

    Safety has been an integral part of the design process for ITER since the Conceptual Design Activities of the project. The safety approach adopted in the ITER-FEAT design and the complementary assessments underway, to be documented in the Generic Site Safety Report (GSSR), are expected to help demonstrate the attractiveness of fusion and thereby set a good precedent for future fusion power reactors. The assessments address ITER's radiological hazards taking into account fusion's favourable safety characteristics. The expectation that ITER will need regulatory approval has influenced the entire safety design and assessment approach. This paper summarises the ITER-FEAT safety approach and assessments underway. (author)

  18. ITER council proceedings: 1997

    International Nuclear Information System (INIS)

    1997-01-01

    This volume of the ITER EDA Documentation Series presents records of the 12th ITER Council Meeting, IC-12, which took place on 23-24 July, 1997 in Tampere, Finland. The Council received from the Parties (EU, Japan, Russia, US) positive responses on the Detailed Design Report. The Parties stated their willingness to contribute to fulfil their obligations in contributing to the ITER EDA. The summary discussions among the Parties led to the consensus that in July 1998 the ITER activities should proceed for additional three years with a general intent to enable an efficient start of possible, future ITER construction

  19. Chevron beam dump for ITER edge Thomson scattering system

    International Nuclear Information System (INIS)

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

    2013-01-01

    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

  20. Chevron beam dump for ITER edge Thomson scattering system

    Energy Technology Data Exchange (ETDEWEB)

    Yatsuka, E.; Hatae, T.; Bassan, M.; Itami, K. [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Vayakis, G. [ITER Organization, 13115 St Paul Lez Durance Cedex (France)

    2013-10-15

    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.

  1. 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.

  2. Qualification Test for Korean Mockups of ITER Blanket First Wall

    International Nuclear Information System (INIS)

    Kim, S. K.; Lee, D. W.; Bae, Y. D.; Hong, B. G.; Jung, H. K.; Jung, Y. I.; Park, J. Y.; Jeong, Y. H.; Choi, B. K.; Kim, B. Y.

    2009-01-01

    ITER First Wall (FW) includes the beryllium armor tiles joined to CuCrZr heat sink with stainless steel cooling tubes. This first wall panels are one of the critical components in the ITER machine with the surface heat flux of 0.5 MW/m 2 or above. So qualification program needs to be performed with the goal to qualify the joining technologies required for the ITER First Wall. Based on the results of tests, the acceptance of the developed joining technologies will be established. The results of this qualification test will affect the final selection of the manufacturers for the ITER First Wall

  3. Measurement of tokamak error fields using plasma response and its applicability to ITER

    International Nuclear Information System (INIS)

    Strait, E.J.; Buttery, R.J.; Chu, M.S.; Garofalo, A.M.; La Haye, R.J.; Schaffer, M.J.; Casper, T.A.; Gribov, Y.; Hanson, J.M.; Reimerdes, H.; Volpe, F.A.

    2014-01-01

    The nonlinear response of a low-beta tokamak plasma to non-axisymmetric fields offers an alternative to direct measurement of the non-axisymmetric part of the vacuum magnetic fields, often termed ‘error fields’. Possible approaches are discussed for determination of error fields and the required current in non-axisymmetric correction coils, with an emphasis on two relatively new methods: measurement of the torque balance on a saturated magnetic island, and measurement of the braking of plasma rotation in the absence of an island. The former is well suited to ohmically heated discharges, while the latter is more appropriate for discharges with a modest amount of neutral beam heating to drive rotation. Both can potentially provide continuous measurements during a discharge, subject to the limitation of a minimum averaging time. The applicability of these methods to ITER is discussed, and an estimate is made of their uncertainties in light of the specifications of ITER's diagnostic systems. The use of plasma response-based techniques in normal ITER operational scenarios may allow identification of the error field contributions by individual central solenoid coils, but identification of the individual contributions by the outer poloidal field coils or other sources is less likely to be feasible. (paper)

  4. β-adrenergic blockade does not impair the skin blood flow sensitivity to local heating in burned and non-burned skin under neutral and hot environments in children

    Science.gov (United States)

    Rivas, Eric; McEntire, Serina J.; Herndon, David N.; Mlcak, Ronald P.; Suman, Oscar E.

    2017-01-01

    Objective Tested the hypothesis that propranolol, a drug given to burn patients to reduce hypermetabolism/cardiac stress, may inhibit heat dissipation by changing the sensitivity of skin blood flow (SkBF) to local heating under neutral and hot conditions. Methods In a randomized double-blind study, a placebo was given to 8 burned children while propranolol was given to 13 burned children with similar characteristics (mean ± SD: 11.9±3y, 147±20cm, 45±23kg, 56±12% TBSA). Non-burned children (n=13, 11.4±3y, 152±15cm, 52±13kg) served as healthy controls. A progressive local heating protocol characterized SkBF responses in burned and unburned skin and non-burned control skin under the two environmental conditions (23°C and 34°C) via laser-Doppler flowmetry. Results Resting SkBF was greater in burned and unburned skin compared to the non-burned control (main effect: skin, Pburned; 38±36 unburned vs 9±8 control %SkBFmax). No difference was found for maximal SkBF capacity to local heating between groups. Additionally, dose response curves for the sensitivity of SkBF to local heating were not different among burned or unburned skin, and non-burned control skin (EC50, P>0.05) under either condition. Conclusion Therapeutic propranolol does not negatively affect SkBF under neutral or hot environmental conditions and further compromise temperature regulation in burned children. PMID:28071840

  5. Power and particle control in JT-60SA to support and supplement ITER and DEMO

    International Nuclear Information System (INIS)

    Sakurai, Shinji

    2007-01-01

    JT-60 is planned to be modified as a fully superconducting coil tokamak (JT-60 Super Advanced, JT-60SA). Divertor targets are water-cooled to handle heat flux up to 15 MW/m 2 . JT-60SA allows exploitation of high beta regime with stabilizing shell covered with ferritic plates and internal resistive wall mode (RWM) stabilizing coils. A remote handling system is equipped to maintain in-vessel components even for high dose rate due to a substantial annual neutron production. Divertor cassettes are introduced to be maintained by a remote handling. In the present design, a monoblock type carbon fibre composite (CFC) divertor target will be used to withstand a heat load of ∼15 MW/m 2 . CFC divertor targets and other bolted armor tiles will be mounted on the divertor cassette. All of the plasma facing components including the first wall armor are water-cooled to handle heat load during 100s or more. Divertor heat load and pumping efficiency for an ITER-like configuration has been evaluated, using 2D plasma fluid (SOLDOR) and neutral Monte-Carlo (NEUT2D) code. The pumping speed of 50 m 3 /s is specified at an albedo for neutrals in front of the in-vessel cryopanel. In the simulation for the divertor with a V -shaped corner' like as that in ITER, the plasma detachment occurs near the outer-strike point within the 'V-shaped corner', as well as near the inner-strike point, which results in low peak heat flux density 5.8 MW/m 2 for the case with additional gas puff of 5x10 21 /s compared to 11.4 MW/m 2 for the case without 'V-shaped corner'. (author)

  6. Plasma instability issues for ITER and their possible impact on plasma performance

    International Nuclear Information System (INIS)

    Houlberg, W.A.; Campbell, D.

    2009-01-01

    Full text: There are many types of plasma instabilities that may affect ITER performance. Prediction of their impact, however, is complicated by scaling relative to present plasmas. Here we summarize some of the potential impacts of a variety of instabilities on ITER performance and the uncertainties in evaluating those impacts. ELMs are one of the most significant issues because of the high localized heat loads on the plasma facing components walls caused by the filamentary structures. ITER presently plans to employ two methods to attempt to amelioriate the localized damage from large ELMS: resonant magnetic perturbations and pellet pacing. In either case, the net effect on confinement must be minimized relative to the expected confinement under natural ELMy conditions. Pacing ELMs with high frequency pellet injection raises at least two fundamental physics questions: i) how effective are very localized perturbations from the pellet cloud at triggering ELMs?, and ii) can we assure that the local perturbation does not lock the ELMs into a pattern of localized deposition? It is expected that answering these questions would require 3-D models, while present models are based on peeling-ballooning stability with 1-D models for plasma profiles. A similar set of complicating factors can be identified for other instabilities. Alfven eigenmodes in ITER are expected to be driven primarily by the energetic alphas, but MeV neutral beam injection of up to 33 MW raises the issue of synergistic effects (e.g., loss of NB fast ions from AEs driven by fast alphas), and non-linear interaction among a 'sea' of many high-n potentially unstable modes expected from ITER's large size. Instabilities that are weakened by the strong toroidal rotation (e.g. turbulence or resistive wall modes) in present NB-heated machines may be more robust under the much weaker external torque provided by ITER's high energy beams. A better understanding of extrinsic rotation driven largely by conditions at

  7. Status of the realization of the neutral beam test facility

    International Nuclear Information System (INIS)

    Toigo, Vanni

    2015-01-01

    The ITER Neutral Beam Injectors (NBI) are required to deliver 16.5 MW of additional heating power to the plasma, accelerating negative ions up to -1 MV with a beam current of 40A lasting up to 1 hour. Since these outstanding requirements were never achieved all together so far, the realization of a Neutral Beam Test Facility (NBTF), called PRIMA, currently under construction in Padova, was launched with the aim to test the operation of the NB injector and to study the relevant physical and technological issues, in advance to the implementation in ITER. Two projects are under development: MITICA and SPIDER. MITICA is a full scale prototype of the ITER NB injector; the design is based on a similar scheme and layout, with the same power supply system and also the control and protection systems are being designed according to the ITER rules and constraints. The HV components are procured by JADA; the low voltage ones and the injector are procured by F4E. SPIDER project is an ion source with the same characteristics of the ITER one, specifically addressed to study the issues related to the RF operation; for this reason, the beam energy is limited to 100keV. It can generate both Hydrogen and Deuterium Ions; the design includes provisions to filter electrons and also to allow the use of cesium to attain the high values of current density required. SPIDER is procured by F4E and INDA. The construction of PRIMA buildings and auxiliaries, started in autumn 2008, was completed in summer 2015. SPIDER plant systems procurement is well advanced and some systems are under installation or site acceptance tests. In 2016 integrated commissioning and power supply integrated tests will be performed followed by the beginning of the first experimental phase. MITICA design was completed; many procurement contracts have been signed or will be launched in the next months. Installation activity will start in December 2015 with the installation of the first HV power supply components provided

  8. Advanced scenarios for ITER operation

    Energy Technology Data Exchange (ETDEWEB)

    Sips, A.C.C. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

    2004-07-01

    In thermonuclear fusion research using magnetic confinement, the tokamak is the leading candidate for achieving conditions required for a reactor. An international experiment, ITER is proposed as the next essential and critical step on the path to demonstrating the scientific and technological feasibility of fusion energy. ITER is to produce and study plasmas dominated by self heating. This would give unique opportunities to explore, in reactor relevant conditions, the physics of {alpha}-particle heating, plasma turbulence and turbulent transport, stability limits to the plasma pressure and exhaust of power and particles. Important new results obtained in experiments, theory and modelling, enable an improved understanding of the physical processes occurring in tokamak plasmas and give enhanced confidence in ITER achieving its goals. In particular, progress has been made in research to raise the performance of tokamaks, aimed to extend the discharge pulse length towards steady-state operation (advanced scenarios). Standard tokamak discharges have a current density increasing monotonically towards the centre of the plasma. Advanced scenarios on the other hand use a modified current density profile. Different advanced scenarios range from (i) plasmas that sustain a central region with a flat current density profile (zero magnetic shear), capable of operating stationary at high plasma pressure, to (ii) discharges with an off axis maximum of the current density profile (reversed magnetic shear in the core), able to form internal transport barriers, to increase the confinement of the plasma. The physics of advanced tokamak discharges is described, together with an overview of recent results from different tokamak experiments. International collaboration between experiments aims to provide a better understanding, control and optimisation of these plasmas. The ability to explore advanced scenarios in ITER is very desirable, in order to verify the result obtained in

  9. The neutral beam test facility cryopumping operation: preliminary analysis and design of the cryogenic system

    International Nuclear Information System (INIS)

    Gravil, B.; Henry, D.; Cordier, J.J.; Hemsworth, R.; Van Houtte, D.

    2004-01-01

    The ITER neutral beam heating and current drive system is to be equipped with a cryosorption cryopump made up of 12 panels connected in parallel, refrigerated by 4.5 K 0.4 MPa supercritical helium. The pump is submitted to a non homogeneous flux of H 2 or D 2 molecules, and the absorbed flux varies from 3 Pa.m -3 .s -1 to 35 Pa.m -3 .s -1 . In the frame of the 'ITER first injector and test facility CSU-EFDA task' (TW3-THHN-IITF1), the ITER reference cryo-system and cryo-plant designs have been assessed and compared to optimised designs devoted to the Neutral Beam Test Facility (NBTF). The 4.5 K cryo-panel, which has a mass of about 1000 kg, must be periodically regenerated up to 90 K and occasionally to 470 K. The cool-down time after regeneration depends strongly on the refrigeration capacity. Fast regeneration and cool-down of the cryo-panels are not considered a priority for the test facility operation, and an analysis of the consequences of a limited cold power refrigerator on the cooling down time has been carried out and will be discussed. This paper presents a preliminary evaluation of the NBTF cryo-plant and the associated process flow diagram. (authors)

  10. The ITER divertor cassette project

    International Nuclear Information System (INIS)

    Ulrickson, M.; Tivey, R.; Akiba, M.

    2001-01-01

    The divertor ''Large Project'' was conceived with the aim of demonstrating the feasibility of meeting the lifetime requirements by employing the candidate armor materials of beryllium, tungsten (W) and carbon-fiber-composite (CFC). At the start, there existed only limited experience with constructing water-cooled high heat flux armored components for tokamaks. To this was added the complication posed by the need to use a silver-free joining technique that avoids the transmutation of n-irradiated silver to cadmium. The research project involving the four Home Teams (HTs) has focused on the design, development, manufacture and testing of full-scale Plasma Facing Components (PFCs) suitable for ITER. The task addressed all the issues facing ITER divertor design, such as providing adequate armor erosion lifetime, meeting the required armor-heat sink joint lifetime and heat sink fatigue life, sustaining thermal-hydraulic and electromechanical loads, and seeking to identify the most cost-effective manufacturing options. This paper will report the results of the divertor large project. (author)

  11. The ITER divertor cassette project

    International Nuclear Information System (INIS)

    Ulrickson, M.; Tivey, R.; Akiba, M.

    1999-01-01

    The divertor 'Large Project' was conceived with the aim of demonstrating the feasibility of meeting the lifetime requirements by employing the candidate armor materials of beryllium, tungsten (W) and carbon-fiber-composite (CFC). At the start, there existed only limited experience with constructing water-cooled high heat flux armored components for tokamaks. To this was added the complication posed by the need to use a silver-free joining technique that avoids the transmutation of n-irradiated silver to cadmium. The research project involving the four Home Teams (HTs) has focused on the design, development, manufacture and testing of full-scale Plasma Facing Components (PFCs) suitable for ITER. The task addressed all the issues facing ITER divertor design, such as providing adequate armor erosion lifetime, meeting the required armor-heat sink joint lifetime and heat sink fatigue life, sustaining thermal-hydraulic and electromechanical loads, and seeking to identify the most cost-effective manufacturing options. This paper will report the results of the divertor large project. (author)

  12. Iteration and accelerator dynamics

    International Nuclear Information System (INIS)

    Peggs, S.

    1987-10-01

    Four examples of iteration in accelerator dynamics are studied in this paper. The first three show how iterations of the simplest maps reproduce most of the significant nonlinear behavior in real accelerators. Each of these examples can be easily reproduced by the reader, at the minimal cost of writing only 20 or 40 lines of code. The fourth example outlines a general way to iteratively solve nonlinear difference equations, analytically or numerically

  13. 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.

  14. Future plan of ITER

    International Nuclear Information System (INIS)

    Kitsunezaki, Akio

    1998-01-01

    In cooperation of four countries, Japan, USA, EU and Russia, ITER plan has been proceeding as ''the conceptual design activities'' from 1988 to 1990 and ''the industrial design activities'' since 1992. To construct ITER, the legal and work side of ITER operation has been investigated by four countries. However, their economic conditions have been changed to be wrong. So that, construction of ITER can not begin after end of industrial design activities in 1998. Accordingly, they determined to continue the industrial design activities more three years in order to study low cost options and to test the superconductive model·coil. (S.Y.)

  15. ITER test programme

    International Nuclear Information System (INIS)

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

    1991-01-01

    ITER has been 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 refs, figs and tabs

  16. Selection of plasma facing materials for ITER

    International Nuclear Information System (INIS)

    Ulrickson, M.; Barabash, V.; Chiocchio, S.

    1996-01-01

    ITER will be the first tokamak having long pulse operation using deuterium-tritium fuel. The problem of designing heat removal structures for steady state in a neutron environment is a major technical goal for the ITER Engineering Design Activity (EDA). The steady state heat flux specified for divertor components is 5 MW/m 2 for normal operation with transients to 15 MW/m 2 for up to 10 s. The selection of materials for plasma facing components is one of the major research activities. Three materials are being considered for the divertor; carbon fiber composites, beryllium, and tungsten. This paper discusses the relative advantages and disadvantages of these materials. The final section of plasma facing materials for the ITER divertor will not be made until the end of the EDA

  17. Observation of an improved energy-confinement regime in neutral-beam--heated divertor discharges in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Burrell, K.H.; Ejima, S.; Schissel, D.P.

    1987-01-01

    Tokamak discharges using the expanded boundary divertor in the DIII-D device exhibit H-mode confinement. With neutral-beam power up to 6 MW, energy confinement remains comparable to the Ohmic value at a plasma current of 1 MA. Confinement is also independent of plasma density and toroidal field. Confinement increases with plasma current, but the exact functional dependence is, as yet, uncertain. These results show that the H mode can be achieved in a reactor-compatible open divertor configuration

  18. United States rejoin ITER

    International Nuclear Information System (INIS)

    Roberts, M.

    2003-01-01

    Upon pressure from the United States Congress, the US Department of Energy had to withdraw from further American participation in the ITER Engineering Design Activities after the end of its commitment to the EDA in July 1998. In the years since that time, changes have taken place in both the ITER activity and the US fusion community's position on burning plasma physics. Reflecting the interest in the United States in pursuing burning plasma physics, the DOE's Office of Science commissioned three studies as part of its examination of the option of entering the Negotiations on the Agreement on the Establishment of the International Fusion Energy Organization for the Joint Implementation of the ITER Project. These were a National Academy Review Panel Report supporting the burning plasma mission; a Fusion Energy Sciences Advisory Committee (FESAC) report confirming the role of ITER in achieving fusion power production, and The Lehman Review of the ITER project costing and project management processes (for the latter one, see ITER CTA Newsletter, no. 15, December 2002). All three studies have endorsed the US return to the ITER activities. This historical decision was announced by DOE Secretary Abraham during his remarks to employees of the Department's Princeton Plasma Physics Laboratory. The United States will be working with the other Participants in the ITER Negotiations on the Agreement and is preparing to participate in the ITA

  19. ITER CTA newsletter. No. 3

    International Nuclear Information System (INIS)

    2001-11-01

    This ITER CTA newsletter comprises reports of Dr. P. Barnard, Iter Canada Chairman and CEO, about the progress of the first formal ITER negotiations and about the demonstration of details of Canada's bid on ITER workshops, and Dr. V. Vlasenkov, Project Board Secretary, about the meeting of the ITER CTA project board

  20. 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.)

  1. Multimegawatt neutral beams for tokamaks

    International Nuclear Information System (INIS)

    Kunkel, W.B.

    1979-03-01

    Most of the large magnetic confinement experiments today and in the near future use high-power neutral-beam injectors to heat the plasma. This review briefly describes this remarkable technique and summarizes recent results as well as near term expectations. Progress has been so encouraging that it seems probable that tokamaks will achieve scientific breakeven before 1990

  2. Preliminary design of safety and interlock system for indian test facility of diagnostic neutral beam

    International Nuclear Information System (INIS)

    Tyagi, Himanshu; Soni, Jignesh; Yadav, Ratnakar; Bandyopadhyay, Mainak; Rotti, Chandramouli; Gahlaut, Agrajit; Joshi, Jaydeep; Parmar, Deepak; Bansal, Gourab; Pandya, Kaushal; Chakraborty, Arun

    2016-01-01

    Highlights: • Indian Test Facility being built to characterize DNB for ITER delivery. • Interlock system required to safeguard the investment incurred in building the facility and protecting ITER deliverable components. • Interlock levels upto 3IL-3 identified. • Safety instrumented system for occupational safety being designed. Safety I&C functions of SIL-2 identified. • The systems are based on ITER PIS and PSS design guidelines. - Abstract: Indian Test Facility (INTF) is being built in Institute For Plasma Research to characterize Diagnostic Neutral Beam in co-operation with ITER Organization. INTF is a complex system which consists of several plant systems like beam source, gas feed, vacuum, cryogenics, high voltage power supplies, high power RF generators, mechanical systems and diagnostics systems. Out of these, several INTF components are ITER deliverable, that is, beam source, beam line components and power supplies. To ensure successful operation of INTF involving integrated operation of all the constituent plant systems a matured Data Acquisition and Control System (DACS) is required. The INTF DACS is based on CODAC platform following on PCDH (Plant Control Design Handbook) guidelines. The experimental phases involve application of HV power supplies (100 KV) and High RF power (∼800 KW) which will produce energetic beam of maximum power 6MW within the facility for longer durations. Hence the entire facility will be exposed tohigh heat fluxes and RF radiations. To ensure investment protection and to provide occupational safety for working personnel a matured Safety and Interlock system is required for INTF. The Safety and Interlock systems are high-reliability I&C systems devoted completely to the specific functions. These systems will be separate from the conventional DACS of INTF which will handle the conventional control and acquisition functions. Both, the Safety and Interlock systems are based on IEC 61511 and IEC 61508 standards as

  3. Preliminary design of safety and interlock system for indian test facility of diagnostic neutral beam

    Energy Technology Data Exchange (ETDEWEB)

    Tyagi, Himanshu, E-mail: htyagi@iter-india.org [ITER-India, Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Soni, Jignesh [Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Yadav, Ratnakar; Bandyopadhyay, Mainak; Rotti, Chandramouli [ITER-India, Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Gahlaut, Agrajit [Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Joshi, Jaydeep; Parmar, Deepak [ITER-India, Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Bansal, Gourab; Pandya, Kaushal; Chakraborty, Arun [Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India)

    2016-11-15

    Highlights: • Indian Test Facility being built to characterize DNB for ITER delivery. • Interlock system required to safeguard the investment incurred in building the facility and protecting ITER deliverable components. • Interlock levels upto 3IL-3 identified. • Safety instrumented system for occupational safety being designed. Safety I&C functions of SIL-2 identified. • The systems are based on ITER PIS and PSS design guidelines. - Abstract: Indian Test Facility (INTF) is being built in Institute For Plasma Research to characterize Diagnostic Neutral Beam in co-operation with ITER Organization. INTF is a complex system which consists of several plant systems like beam source, gas feed, vacuum, cryogenics, high voltage power supplies, high power RF generators, mechanical systems and diagnostics systems. Out of these, several INTF components are ITER deliverable, that is, beam source, beam line components and power supplies. To ensure successful operation of INTF involving integrated operation of all the constituent plant systems a matured Data Acquisition and Control System (DACS) is required. The INTF DACS is based on CODAC platform following on PCDH (Plant Control Design Handbook) guidelines. The experimental phases involve application of HV power supplies (100 KV) and High RF power (∼800 KW) which will produce energetic beam of maximum power 6MW within the facility for longer durations. Hence the entire facility will be exposed tohigh heat fluxes and RF radiations. To ensure investment protection and to provide occupational safety for working personnel a matured Safety and Interlock system is required for INTF. The Safety and Interlock systems are high-reliability I&C systems devoted completely to the specific functions. These systems will be separate from the conventional DACS of INTF which will handle the conventional control and acquisition functions. Both, the Safety and Interlock systems are based on IEC 61511 and IEC 61508 standards as

  4. ITER council proceedings: 1992

    International Nuclear Information System (INIS)

    1994-01-01

    At the signing of the ITER EDA Agreement on July, 1992, each of the Parties presented to the Director General the names of their designated members of the ITER Council. Upon receiving those names, the Director General stated that the ITER Engineering Design Activities were ''ready to begin''. The next step in this process was the convening of the first meeting of the ITER Council. The first meeting of the Council, held in Vienna, was opened by Director General Hans Blix. The second meeting was held in Moscow, the formal seat of the Council. This volume presents records of these first two Council meetings and, together with the previous volumes on the text of the Agreement and Protocol 1 and the preparations for their signing respectively, represents essential information on the evolution of the ITER EDA

  5. Status of the IPP RF Negative Ion Source Development for the ITER NBI System

    International Nuclear Information System (INIS)

    Peter Franzen, P.; Falter, H.-D.; Fantz, U.

    2006-01-01

    For heating and current drive the ITER neutral beam system requires negative hydrogen ion sources capable of delivering above 40 A of D - ions from a 1.5 x 0.6 m 2 source for up to one hour pulses with an accelerated current density of 200 A/m 2 . In order to reduce the losses by electron stripping in the acceleration system and the power loading of the grids, the source pressure is required to be 0.3 Pa at an electron/ion ratio 2 H - / 230 A/m 2 D - ) in excess of the ITER requirements have been already achieved on the small test facility '' BATMAN '' (Bavarian Test Machine for Negative Ions) at the required source pressure (0.3 Pa) and electron/ion ratio ( 2 ) and limited pulse length ( 2 and the pulse length up to 3600 s, using the same source as it is used at BATMAN. In order to demonstrate the required homogeneity of a large RF plasma source as well as the operation of an ITER relevant RF circuit, a so called '' half-size source '' - with roughly the width and half the height of the ITER source - was designed and went into operation on a dedicated plasma source test bed ('' RADI ''). An extensive diagnostic and modelling programme is accompanying those activities. The paper will present as an overview a summary of the latest results of the RF source development, with an emphasis on the first results of the operation of the half size ITER source and on the status of the long pulse operation. The details will be presented in several other papers. (author)

  6. Infrared laser diagnostics for ITER

    International Nuclear Information System (INIS)

    Hutchinson, D.P.; Richards, R.K.; Ma, C.H.

    1995-01-01

    Two infrared laser-based diagnostics are under development at ORNL for measurements on burning plasmas such as ITER. The primary effort is the development of a CO 2 laser Thomson scattering diagnostic for the measurement of the velocity distribution of confined fusion-product alpha particles. Key components of the system include a high-power, single-mode CO 2 pulsed laser, an efficient optics system for beam transport and a multichannel low-noise infrared heterodyne receiver. A successful proof-of-principle experiment has been performed on the Advanced Toroidal Facility (ATF) stellerator at ORNL utilizing scattering from electron plasma frequency satellites. The diagnostic system is currently being installed on Alcator C-Mod at MIT for measurements of the fast ion tail produced by ICRH heating. A second diagnostic under development at ORNL is an infrared polarimeter for Faraday rotation measurements in future fusion experiments. A preliminary feasibility study of a CO 2 laser tangential viewing polarimeter for measuring electron density profiles in ITER has been completed. For ITER plasma parameters and a polarimeter wavelength of 10.6 microm, a Faraday rotation of up to 26 degree is predicted. An electro-optic polarization modulation technique has been developed at ORNL. Laboratory tests of this polarimeter demonstrated a sensitivity of ≤ 0.01 degree. Because of the similarity in the expected Faraday rotation in ITER and Alcator C-Mod, a collaboration between ORNL and the MIT Plasma Fusion Center has been undertaken to test this polarimeter system on Alcator C-Mod. A 10.6 microm polarimeter for this measurement has been constructed and integrated into the existing C-Mod multichannel two-color interferometer. With present experimental parameters for C-Mod, the predicted Faraday rotation was on the order of 0.1 degree. Significant output signals were observed during preliminary tests. Further experiment and detailed analyses are under way

  7. BEAMS3D Neutral Beam Injection Model

    Energy Technology Data Exchange (ETDEWEB)

    Lazerson, Samuel

    2014-04-14

    With the advent of applied 3D fi elds 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 velocity reduction, and pitch angle scattering are modeled with the ADAS atomic physics database [1]. Benchmark calculations are presented to validate the collisionless particle orbits, neutral beam injection 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.

  8. ITER towards the construction

    International Nuclear Information System (INIS)

    Shimomura, Y.

    2005-01-01

    The ITER Project has been significantly developed in the last few years in preparation for its construction. The ITER Participant's Negotiators have developed the Joint Implementation Agreement (JIA), ready for finalisation following selection of the construction site and nomination of the project's Director General. The ITER International Team and Participant Teams have continued technical and organisational preparations. Construction will be able to start immediately after the international ITER organisation is established, following signature of the JIA. The Project is strongly supported by the governments of the Participants as well as by the scientific community. The real negotiations, including siting and the final details of cost sharing, started in December 2003. The EU, with Cadarache, and Japan, with Rokkasho, have both promised large contributions to the project to strongly support their construction site proposals. Their wish to host ITER construction is too strong to allow convergence to a single site considering the ITER device in isolation. A broader collaboration among the Parties is therefore being contemplated, covering complementary activities to help accelerate fusion development towards a viable power source, and allow the Participants to reach a conclusion on ITER siting. This report reviews these preparations, and the status of negotiations

  9. 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.

  10. Mechanical design features and challenges for the ITER ICRH antenna

    Energy Technology Data Exchange (ETDEWEB)

    Borthwick, A. [UKAEA/Euratom Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)], E-mail: andy.borthwick@yahoo.co.uk; Agarici, G. [Fusion for Energy, Barcelona (Spain); Davis, A. [UKAEA/Euratom Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Dumortier, P.; Durodie, F. [LPP-ERM-KMS, Association EURATOM-Belgian State, Brussels (Belgium); Fanthome, J.; Hamlyn-Harris, C.; Hancock, A.D.; Lockley, D. [UKAEA/Euratom Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Mitteau, R. [Euratom-CEA Association, DSM/IRFM, CEA-Cadarache, 13108 St Paul lez Durance (France); Nightingale, M. [UKAEA/Euratom Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Sartori, R. [Fusion for Energy, Barcelona (Spain); Vulliez, K. [Euratom-CEA Association, DSM/IRFM, CEA-Cadarache, 13108 St Paul lez Durance (France)

    2009-06-15

    The ITER Ion Cyclotron Resonant Heating (ICRH) antenna provides plasma heating at a power of 20 MW. Operation in the ITER environment imposes significant thermal power handling capability, structural integrity, shielding and operations requirements. The design will require a step change over any predecessor in terms of power, scale and complexity. This paper reports the main mechanical design features that address the challenges and often conflicting requirements during the conceptual design phase.

  11. 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.

  12. 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.)

  13. Analysis of ferromagnetic shielding of the ITER NBI

    International Nuclear Information System (INIS)

    Roccella, M.; Lucca, F.; Roccella, R.; Cocilovo, V.; Ramogida, G.; Portone, A.; Tanga, A.; Formisano, A.; Martone, R.

    2006-01-01

    In ITER two heating and one diagnostic Neutral Beam Injectors (NBIs) are foreseen [P. L. Mondino et al., ''ITER neutral beam system '', Nucl. Fus., vol. 40, p. 501 (2000)]. Inside these components there are very stringent limits on the magnetic field (the flux density must be below some Gauss (G) along the ion path and below 20 G in the neutralizing region). To achieve these performances in an environment with high stray field due to the plasma and the poloidal field coils, both passive and active shielding systems are foreseen. The present design of the Magnetic Field Reduction System (MFRS) is made of seven active coils and of a box surrounding the NBI region, consisting of ferromagnetic plates 15 cm thick. The electromagnetic analysis of the effectiveness of these shields has been performed by a full 3D FEM model using the ANSYS code. To perform the FEM modeling of the component special care has been used to face the particular geometrical features of the component (a box of about 15 x 5 x 5 m vs. a ferromagnetic layer of only 15 cm thick). To insert an adequate number of FEM elements (at least 5) in the thickness of the ferromagnetic layer, without a prohibitive increase in the total FEM elements number, a particular modeling approach (a sort of '' Chinese boxes '' technique) has been developed. Due to this technique the FEM model enclosing the ferromagnetic box results completely independent on the fine FEM structure inside the shielding layer. It has been even possible, using this technique, introducing a thin (below 1 cm thick) slot all through the shielding plates, without perturbing the rest of the model. This slot has been used to analyze the effects of possible manufacturing lacks on the residual magnetic field inside the component. This technique has allowed the use of only structured meshes made by brick elements, much more accurate than the tetra elements, needed in the usual free meshing techniques. To have the possibility of changing the shielding

  14. Recent Progress on ECH Technology for ITER

    Science.gov (United States)

    Sirigiri, Jagadishwar

    2005-10-01

    The Electron Cyclotron Heating and Current Drive (ECH&CD) system for ITER is a critical ITER system that must be available for use on Day 1 of the ITER experimental program. The applications of the system include plasma start-up, plasma heating and suppression of Neoclassical Tearing Modes (NTMs). These applications are accomplished using 27 one megawatt continuous wave gyrotrons: 24 at a frequency of 170 GHz and 3 at a frequency of 120 GHz. There are DC power supplies for the gyrotrons, a transmission line system, one launcher at the equatorial plane and three upper port launchers. The US will play a major role in delivering parts of the ECH&CD system to ITER. The present state-of-the-art includes major advances in all areas of ECH technology. In the US, a major effort is underway to supply gyrotrons of up to 1.5 MW power level at 110 GHz to General Atomics for use in heating the DIII-D tokamak. This presentation will include a brief review of the state-of-the-art, worldwide, in ECH technology. The requirements for the ITER ECH&CD system will then be reviewed. ITER calls for gyrotrons capable of operating from a 50 kV power supply, after potential depression, with a minimum of 50% overall efficiency. This is a very significant challenge and some approaches to meeting this goal will be presented. Recent experimental results at MIT showing improved efficiency of high frequency, 1.5 MW gyrotrons will be described. These results will be incorporated into the planned development of gyrotrons for ITER. The ITER ECH&CD system will also be a challenge to the transmission lines, which must operate at high average power at up to 1000 seconds and with high efficiency. The technology challenges and efforts in the US and other ITER parties to solve these problems will be reviewed. *In collaboration with E. Choi, C. Marchewka, I. Mastovosky, M. A. Shapiro and R. J. Temkin. This work is supported by the Office of Fusion Energy Sciences of the U. S. Department of Energy.

  15. 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.

  16. ITER vacuum vessel, in vessel components and plasma facing materials

    International Nuclear Information System (INIS)

    Ioki, Kimihiro; Enoeda, M.; Federici, G.

    2007-01-01

    Design of the NB ports including duct liners under heat loads of the neutral beams has been developed. Design of the in-wall shielding has been developed in more details considering the supporting structure and the assembly method. The ferromagnetic inserts have previously not been installed in the outboard midplane region due to irregularity caused by the tangential ports for NB injection. Due to this configuration, the maximum ripple is relatively large (∝1 %) in a limited region of the plasma and the toroidal field flux lines fluctuate ∝10 mm in the FW region. To avoid these problems, additional ferromagnetic inserts are to be installed in the equatorial port region. Detailed studies were carried out on the ITER vacuum vessel to define appropriate codes and standards in the context of the ITER licensing in France. A set of draft documents regarding the ITER vacuum vessel structural code were prepared including an RCC-MR Addendum for the ITER VV with justified exceptions or modifications. The main deviation from the base Code is the extensive use of UT in lieu of radiography for the volumetric examination of all one-side access welds of the outer shell and field joint. The procurement allocation of blanket modules among 6 parties was fixed and the blanket module design has progressed in cooperation with parties. Fabrication of mock-ups for prequalification testing is under way and the tests will be performed in 2007-2008. Development of new beryllium materials is progressing in China and Russia. The ITER limiters will be installed in equatorial ports at two toroidal locations. The limiter plasma-facing surface protrudes ∝8 cm from the FW during the start-up and shutdown phase. In the new limiter concept, the limiters are retracted by ∝8 cm during the plasma flat top phase. This concept gives important advantages; (i) mitigation of the particle and heat loads due to disruptions, ELMs and blobs, (ii) improvement of the power coupling with the ICRH antenna

  17. Applications of neutral beam and rf technologies

    International Nuclear Information System (INIS)

    Haselton, H.H.

    1987-04-01

    This presentation provides an update on the applications of neutral beams and radiofrequency (rf) power in the fusion program; highlights of the ion cyclotron heating (ICH) experiments now in progress, as well as the neutral beam experiments; and heating requirements of future devices and some of the available options. Some remarks on current drive are presented because this area of technology is one that is being considered for future devices

  18. Effect of temperature and heat fluxes on the corrosion's damage nature for mild and stainless steels in neutral chloride solutions

    Energy Technology Data Exchange (ETDEWEB)

    Kaluzhina, S.A. [Voronezh State University, University Sq.1, 394006 Voronezh (Russian Federation); Malygin, A.V. [JSC Voronezhsynthezkauchuk, Leninsky Av. 2, 394014 Voronezh (Russian Federation); Vigdorovitch, V.V. [Derzhavin State University, International St. 33, 392622 Tambov (Russian Federation)

    2004-07-01

    The detail research of the corrosion-electrochemical behavior of two types steels - mild steel (0.1%C) and stainless steel 12FeCr18Ni10Ti in series chloride solutions under elevated temperature and heat flux on interface has been carried out in the present work using the special plant and the complex electrochemical and microscopic methods. The comparative data has shown that the temperature increase is stimulating as the active alloy's corrosion (mild steel), so the passive alloy's corrosion (12FeCr18Ni10Ti).However at the last case the temperature effect is being higher because the thermal de-passivation of the stainless steel which undergoes pit corrosion under t > 50 deg C. The heat-transfer role in the studied systems is ambiguous. The corrosion rate of heat-transferring electrode from mild steel exceeds the thermo-equilibrium with solution electrode's corrosion rate because of intensification of the oxygen reduction cathodic process. The opposite effect has been established for steel 12FeCr18Ni10Ti where the oxygen flux's strengthening from cold solution to the heated surface transfers the alloy to the most stable passive state and increases its resistance to general and local corrosion. The experimental results demonstrates that the thermal condition's influence on the nature and corrosion intensity of the investigated steels is being commensurable by effect's degree with their composition and showing strictly individually. (authors)

  19. An Asdex-type divertor for ITER

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1989-01-01

    An Asdex-type local divertor is proposed for ITER consisting of a copper poloidal field coil adjacent to the plasma. Estimates indicate that the power consumption is acceptable. Advantages would be a much reduced heat load not very sensitive to magnetic perturbations. A disadvantage is the finite lifetime under neutron bombardment that would require periodic replacement of the divertor coils in a reactor, but probably not in ITER because of its limited fluence. Another disadvantage would be poorer blanket coverage unless the divertor coil itself incorporates breeding material. 3 figs

  20. Perl Modules for Constructing Iterators

    Science.gov (United States)

    Tilmes, Curt

    2009-01-01

    The Iterator Perl Module provides a general-purpose framework for constructing iterator objects within Perl, and a standard API for interacting with those objects. Iterators are an object-oriented design pattern where a description of a series of values is used in a constructor. Subsequent queries can request values in that series. These Perl modules build on the standard Iterator framework and provide iterators for some other types of values. Iterator::DateTime constructs iterators from DateTime objects or Date::Parse descriptions and ICal/RFC 2445 style re-currence descriptions. It supports a variety of input parameters, including a start to the sequence, an end to the sequence, an Ical/RFC 2445 recurrence describing the frequency of the values in the series, and a format description that can refine the presentation manner of the DateTime. Iterator::String constructs iterators from string representations. This module is useful in contexts where the API consists of supplying a string and getting back an iterator where the specific iteration desired is opaque to the caller. It is of particular value to the Iterator::Hash module which provides nested iterations. Iterator::Hash constructs iterators from Perl hashes that can include multiple iterators. The constructed iterators will return all the permutations of the iterations of the hash by nested iteration of embedded iterators. A hash simply includes a set of keys mapped to values. It is a very common data structure used throughout Perl programming. The Iterator:: Hash module allows a hash to include strings defining iterators (parsed and dispatched with Iterator::String) that are used to construct an overall series of hash values.

  1. Physical performance analysis and progress of the development of the negative ion RF source for the ITER NBI system

    International Nuclear Information System (INIS)

    Fantz, U.; Franzen, P.; Kraus, W.; Berger, M.; Christ-Koch, S.; Falter, H.; Froeschle, M.; Gutser, R.; Heinemann, B.; Martens, C.; McNeely, P.; Riedl, R.; Speth, E.; Staebler, A.; Wuenderlich, D.

    2009-01-01

    For heating and current drive the neutral beam injection (NBI) system for ITER requires a 1 MeV deuterium beam for up to 1 h pulse length. In order to inject the required 17 MW the large area source (1.9 m x 0.9 m) has to deliver 40 A of negative ion current at the specified source pressure of 0.3 Pa. In 2007, the IPP RF driven negative hydrogen ion source was chosen by the ITER board as the new reference source for the ITER NBI system due to, in principle, its maintenance free operation and the progress in the RF source development. The performance analysis of the IPP RF sources is strongly supported by an extensive diagnostic program and modelling of the source and beam extraction. The control of the plasma chemistry and the processes in the plasma region near the extraction system are the most critical topics for source optimization both for long pulse operation as well as for the source homogeneity. The long pulse stability has been demonstrated at the test facility MANITU which is now operating routinely at stable pulses of up to 10 min with parameters near the ITER requirements. A quite uniform plasma illumination of a large area source (0.8 m x 0.8 m) has been demonstrated at the ion source test facility RADI. The new test facility ELISE presently planned at IPP is being designed for long pulse plasma operation and short pulse, but large-scale extraction from a half-size ITER source which is an important intermediate step towards ITER NBI.

  2. ITER definition phase

    International Nuclear Information System (INIS)

    1989-01-01

    The International Thermonuclear Experimental Reactor (ITER) is envisioned as a fusion device which would demonstrate the scientific and technological feasibility of fusion power. As a first step towards achieving this goal, the European Community, Japan, the Soviet Union, and the United States of America have entered into joint conceptual design activities under the auspices of the International Atomic Energy Agency. A brief summary of the Definition Phase of ITER activities is contained in this report. Included in this report are the background, objectives, organization, definition phase activities, and research and development plan of this endeavor in international scientific collaboration. A more extended technical summary is contained in the two-volume report, ''ITER Concept Definition,'' IAEA/ITER/DS/3. 2 figs, 2 tabs

  3. The physics role of ITER

    International Nuclear Information System (INIS)

    Rutherford, P.H.

    1997-04-01

    Experimental research on the International Thermonuclear Experimental Reactor (ITER) will go far beyond what is possible on present-day tokamaks to address new and challenging issues in the physics of reactor-like plasmas. First and foremost, experiments in ITER will explore the physics issues of burning plasmas--plasmas that are dominantly self-heated by alpha-particles created by the fusion reactions themselves. Such issues will include (i) new plasma-physical effects introduced by the presence within the plasma of an intense population of energetic alpha particles; (ii) the physics of magnetic confinement for a burning plasma, which will involve a complex interplay of transport, stability and an internal self-generated heat source; and (iii) the physics of very-long-pulse/steady-state burning plasmas, in which much of the plasma current is also self-generated and which will require effective control of plasma purity and plasma-wall interactions. Achieving and sustaining burning plasma regimes in a tokamak necessarily requires plasmas that are larger than those in present experiments and have higher energy content and power flow, as well as much longer pulse length. Accordingly, the experimental program on ITER will embrace the study of issues of plasma physics and plasma-materials interactions that are specific to a reactor-scale fusion experiment. Such issues will include (i) confinement physics for a tokamak in which, for the first time, the core-plasma and the edge-plasma are simultaneously in a reactor-like regime; (ii) phenomena arising during plasma transients, including so-called disruptions, in regimes of high plasma current and thermal energy; and (iii) physics of a radiative divertor designed for handling high power flow for long pulses, including novel plasma and atomic-physics effects as well as materials science of surfaces subject to intense plasma interaction. Experiments on ITER will be conducted by researchers in control rooms situated at major

  4. ITER convertible blanket evaluation

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.

    1995-01-01

    Proposed International Thermonuclear Experimental Reactor (ITER) convertible blankets were reviewed. Key design difficulties were identified. A new particle filter concept is introduced and key performance parameters estimated. Results show that this particle filter concept can satisfy all of the convertible blanket design requirements except the generic issue of Be blanket lifetime. If the convertible blanket is an acceptable approach for ITER operation, this particle filter option should be a strong candidate

  5. ITER EDA and technology

    International Nuclear Information System (INIS)

    Baker, C.C.

    2001-01-01

    The year 1998 was the culmination of the six-year Engineering Design Activities (EDA) of the International Thermonuclear Experimental Reactor (ITER) Project. The EDA results in design and validating technology R and D, plus the associated effort in voluntary physics research, is a significant achievement and major milestone in the history of magnetic fusion energy development. Consequently, the ITER EDA was a major theme at this Conference, contributing almost 40 papers

  6. 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.

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

    International Nuclear Information System (INIS)

    Escourbiac, F.; Missirlian, M.; Schlosser, J.; Bobin-Vastra, I.; Kuznetsov, V.; Schedler, B.

    2004-01-01

    The use of flat tile technology to handle heat fluxes in the range of 20 MW/m 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 2 . These results highlight the high potential of this technology for ITER divertor application

  8. Toward construction of ITER

    International Nuclear Information System (INIS)

    Shimomura, Yasuo

    2005-01-01

    The ITER Project has been significantly developed in the past years in preparation for its construction. The ITER Negotiators have developed a draft Joint Implementation Agreement (JIA), ready for completion following the nomination of the Project's Director General (DG). The ITER International Team and Participant Teams have continued technical and organizational preparations. The actual construction will be able to start immediately after the international ITER organization will be established, following signature of the JIA. The Project is now strongly supported by all the participants as well as by the scientific community with the final high-level negotiations, focused on siting and the concluding details of cost sharing, started in December 2003. The EU, with Cadarache, and Japan, with Rokkasho, have both promised large contributions to the project to strongly support their construction site proposals. The extent to which they both wish to host the ITER facility is such that large contributions to a broader collaboration among the Parties are also proposed by them. This covers complementary activities to help accelerate fusion development towards a viable power source, as well as may allow the Participants to reach a conclusion on ITER siting. (author)

  9. ITER Status and Plans

    Science.gov (United States)

    Greenfield, Charles M.

    2017-10-01

    The US Burning Plasma Organization is pleased to welcome Dr. Bernard Bigot, who will give an update on progress in the ITER Project. Dr. Bigot took over as Director General of the ITER Organization in early 2015 following a distinguished career that included serving as Chairman and CEO of the French Alternative Energies and Atomic Energy Commission and as High Commissioner for ITER in France. During his tenure at ITER the project has moved into high gear, with rapid progress evident on the construction site and preparation of a staged schedule and a research plan leading from where we are today through all the way to full DT operation. In an unprecedented international effort, seven partners ``China, the European Union, India, Japan, Korea, Russia and the United States'' have pooled their financial and scientific resources to build the biggest fusion reactor in history. ITER will open the way to the next step: a demonstration fusion power plant. All DPP attendees are welcome to attend this ITER town meeting.

  10. Negative-ion-based neutral beams for fusion

    International Nuclear Information System (INIS)

    Cooper, W.S.; Anderson, O.A.; Chan, C.F.

    1987-10-01

    To maximize the usefulness of an engineering test reactor (e.g., ITER, TIBER), it is highly desirable that it operate under steady-state conditions. The most attractive option for maintaining the circulating current needed in the center of the plasma is the injection of powerful beams of neutral deuterium atoms. The beam simultaneously heats the plasma. At the energies required, in excess of 500 keV, such beams can be made by accelerating D - ions and then removing the electron. Sources are being developed that generate the D - ions in the volume of a specially constructed plasma discharge, without the addition of cesium. These sources must operate with minimum gas flow, to avoid stripping the D - beam, and with minimum electron output. We are designing at LBL highly efficient electrostatic accelerators that combine electric strong-focusing with dc acceleration and offer the possibility of varying the beam energy at constant current while minimizing breakdown. Some form of rf acceleration may also be required. To minimize irradiation of the ion sources and accelerators, the D - beam can be transported through a maze in the neutron shielding. The D - ions can be converted to neutrals in a gas or plasma target, but advances in laser and mirror technology may make possible very efficient photodetachment systems by the time an ETR becomes operational. 9 refs., 4 figs

  11. Divertor heat flux mitigation in the National Spherical Torus Experimenta)

    Science.gov (United States)

    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.; NSTX Team

    2009-02-01

    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 4-6MWm-2to0.5-2MWm-2 in small-ELM 0.8-1.0MA, 4-6MW neutral beam injection-heated H-mode discharges. A self-consistent picture of the 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. Physics basis of ITER-FEAT

    International Nuclear Information System (INIS)

    Shimada, M.; Campbell, D.J.; Wakatani, M.; Ninomiya, H.; Ivanov, N.V.; Mukhovatov, V.

    2001-01-01

    This paper reviews Physics R and D results obtained since the publication of the ITER Physics Basis document. The heating power required for the LH transition has been re-assessed, including recent results from C-Mod and JT-60U and it has been found that the predicted power is a factor of two lower than the previous projection. For predicting ITER-FEAT performance, a conservative scaling IPB98(y,2) has been adopted for the energy confinement, producing confinement times ∼20% lower than those derived from the IPB98(y,1) law. While energy confinement degradation at high density remains a serious issue, recent experiments suggest that good confinement is achievable in ITER at n/n G ∼0.85 with high triangularity. The estimated runaway electron energy has been reduced to ∼20MJ, since recent experiments show that runaway electrons disappear for q 95 leq2. (author)

  13. ITER Experts' meeting on density limits

    International Nuclear Information System (INIS)

    Borrass, K.; Igitkhanov, Y.L.; Uckan, N.A.

    1989-12-01

    The necessity of achieving a prescribed wall load or fusion power essentially determines the plasma pressure in a device like ITER. The range of operation densities and temperatures compatible with this condition is constrained by the problems of power exhaust and the disruptive density limit. The maximum allowable heat loads on the divertor plates and the maximum allowable sheath edge temperature practically impose a lower limit on the operating densities, whereas the disruptive density limit imposes an upper limit. For most of the density limit scalings proposed in the past an overlap of the two constraints or at best a very narrow accessible density range is predicted for ITER. Improved understanding of the underlying mechanisms is therefore a crucial issue in order to provide a more reliable basis for extrapolation to ITER and to identify possible ways of alleviating the problem

  14. ITER CTA newsletter. No. 6

    International Nuclear Information System (INIS)

    2002-01-01

    This ITER CTA Newsletter issue comprises information about the following ITER Meetings: The second negotiation meeting on the joint implementation of ITER, held in Tokyo(Japan) on 22-23 January 2002, and an international ITER symposium on burning plasma science and technology, held the day later after the second negotiation meeting at the same place

  15. ITER CTA newsletter. No. 2

    International Nuclear Information System (INIS)

    2001-10-01

    This ITER CTA newsletter contains results of the ITER toroidal field model coil project presented by ITER EU Home Team (Garching) and an article in commemoration of the late Dr. Charles Maisonnier, one of the former leaders of ITER who made significant contributions to its development

  16. Neutral beam in ALVAND IIC tokamak

    International Nuclear Information System (INIS)

    Ghrannevisse, M.; Moradshahi, M.; Avakian, M.

    1992-01-01

    Neutral beams have a wide application in tokamak experiments. It used to heat; fuel; adjust electric potentials in plasmas and diagnose particles densities and momentum distributions. It may be used to sustain currents in tokamaks to extend the pulse length. A 5 KV; 500 mA ion source has been constructed by plasma physics group, AEOI and it used to produce plasma and study the plasma parameters. Recently this ion source has been neutralized and it adapted to a neutral beam source; and it used to heat a cylindrical DC plasma and the plasma of ALVAND IIC Tokamak which is a small research tokamak with a minor radius of 12.6 cm, and a major radius of 45.5 cm. In this paper we report the neutralization of the ion beam and the results obtained by injection of this neutral beam into plasmas. (author) 2 refs., 4 figs

  17. ITER Dynamic Tritium Inventory Modeling Code

    International Nuclear Information System (INIS)

    Cristescu, Ioana-R.; Doerr, L.; Busigin, A.; Murdoch, D.

    2005-01-01

    A tool for tritium inventory evaluation within each sub-system of the Fuel Cycle of ITER is vital, with respect to both the process of licensing ITER and also for operation. It is very likely that measurements of total tritium inventories may not be possible for all sub-systems, however tritium accounting may be achieved by modeling its hold-up within each sub-system and by validating these models in real-time against the monitored flows and tritium streams between the systems. To get reliable results, an accurate dynamic modeling of the tritium content in each sub-system is necessary. In order to optimize the configuration and operation of the ITER fuel cycle, a dynamic fuel cycle model was developed progressively in the decade up to 2000-2001. As the design for some sub-systems from the fuel cycle (i.e. Vacuum pumping, Neutral Beam Injectors (NBI)) have substantially progressed meanwhile, a new code developed under a different platform to incorporate these modifications has been developed. The new code is taking over the models and algorithms for some subsystems, such as Isotope Separation System (ISS); where simplified models have been previously considered, more detailed have been introduced, as for the Water Detritiation System (WDS). To reflect all these changes, the new code developed inside EU participating team was nominated TRIMO (Tritium Inventory Modeling), to emphasize the use of the code on assessing the tritium inventory within ITER

  18. Induction of neutral trehalase Nth1 by heat and osmotic stress is controlled by STRE elements and Msn2/Msn4 transcription factors: variations of PKA effect during stress and growth.

    Science.gov (United States)

    Zähringer, H; Thevelein, J M; Nwaka, S

    2000-01-01

    Saccharomyces cerevisiae neutral trehalase, encoded by NTH1, controls trehalose hydrolysis in response to multiple stress conditions, including nutrient limitation. The presence of three stress responsive elements (STREs, CCCCT) in the NTH1 promoter suggested that the transcriptional activator proteins Msn2 and Msn4, as well as the cAMP-dependent protein kinase (PKA), control the stress-induced expression of Nth1. Here, we give direct evidence that Msn2/Msn4 and the STREs control the heat-, osmotic stress- and diauxic shift-dependent induction of Nth1. Disruption of MSN2 and MSN4 abolishes or significantly reduces the heat- and NaCl-induced increases in Nth1 activity and transcription. Stress-induced increases in activity of a lacZ reporter gene put under control of the NTH1 promoter is nearly absent in the double mutant. In all instances, basal expression is also reduced by about 50%. The trehalose concentration in the msn2 msn4 double mutant increases less during heat stress and drops more slowly during recovery than in wild-type cells. This shows that Msn2/Msn4-controlled expression of enzymes of trehalose synthesis and hydrolysis help to maintain trehalose concentration during stress. However, the Msn2/Msn4-independent mechanism exists for heat control of trehalose metabolism. Site-directed mutagenesis of the three STREs (CCCCT changed to CATCT) in NTH1 promoter fused to a reporter gene indicates that the relative proximity of STREs to each other is important for the function of NTH1. Elimination of the three STREs abolishes the stress-induced responses and reduces basal expression by 30%. Contrary to most STRE-regulated genes, the PKA effect on the induction of NTH1 by heat and sodium chloride is variable. During diauxic growth, NTH1 promoter-controlled reporter activity strongly increases, as opposed to the previously observed decrease in Nth1 activity, suggesting a tight but opposite control of the enzyme at the transcriptional and post-translational levels

  19. Assessment of vanadium alloys for ITER application

    International Nuclear Information System (INIS)

    Borgstedt, H.U.; Clemens, H.; Ehrlich, K.; Fromm, E.; Kelzenberg, S.; Moeslang, A.; Pick, M.; Ruehle, M.; Schaaf, B. van der; Schaefer, L.; Schiller, P.; Schirra, M.; Witwer, M.; Witzenburg, W. van; Zolti, E.; Zucchetti, M.

    1993-09-01

    The assessment effort concerned required evaluation of various relevant properties of vanadium alloys. The outcome predictably shows that these properties, as well as timing, funding, manufacturing and licensing aspects, each set their own specific boundary conditions for application of these alloys in ITER. Some of these are not really felt as constraints. Their capacity to accommodate high heat loads, for example, is better than other candidate materials and appears to be the main reason for the present interest in these alloys. Other favourable properties include neutronic properties (low nuclear heating rates, good tritium breeding performance and low helium generation rates), intrinsically low activation, excellent tensile and creep properties up to high temperatures and high strength-to-density ratio. Not all of these properties necessarily are relevant for ITER, but they would be important for longer term application. (orig.)

  20. Earthly sun called ITER

    International Nuclear Information System (INIS)

    Pozdeyev, Mikhail

    2002-01-01

    Full text: Participating in the film are Academicians Velikhov and Glukhikh, Mr. Filatof, ITER Director from Russia, Mr. Sannikov from Kurchatov Institute. The film tells about the starting point of the project (Mr. Lavrentyev), the pioneers of the project (Academicians Tamme, Sakharov, Artsimovich) and about the situation the project is standing now. Participating in [ITER now are the US, Russia, Japan and the European Union. There are two associated members as well - Kazakhstan and Canada. By now the engineering design phase has been finished. Computer animation used in the video gives us the idea how the first thermonuclear reactor based on famous Russian TOKOMAK works. (author)

  1. Iterated multidimensional wave conversion

    International Nuclear Information System (INIS)

    Brizard, A. J.; Tracy, E. R.; Johnston, D.; Kaufman, A. N.; Richardson, A. S.; Zobin, N.

    2011-01-01

    Mode conversion can occur repeatedly in a two-dimensional cavity (e.g., the poloidal cross section of an axisymmetric tokamak). We report on two novel concepts that allow for a complete and global visualization of the ray evolution under iterated conversions. First, iterated conversion is discussed in terms of ray-induced maps from the two-dimensional conversion surface to itself (which can be visualized in terms of three-dimensional rooms). Second, the two-dimensional conversion surface is shown to possess a symplectic structure derived from Dirac constraints associated with the two dispersion surfaces of the interacting waves.

  2. Tungsten impurity transport experiments in Alcator C-Mod to address high priority research and development for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Loarte, A.; Polevoi, A. R.; Hosokawa, M. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Reinke, M. L. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Chilenski, M.; Howard, N.; Hubbard, A.; Hughes, J. W.; Rice, J. E.; Walk, J. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Köchl, F. [Technische Universität Wien, Atominstitut, Stadionallee 2, 1020 Vienna (Austria); Pütterich, T.; Dux, R. [Max-Planck-Institut für Plasmaphysik, Boltzmanstraße 2, D-85748 Garching (Germany); Zhogolev, V. E. [NRC “Kurchatov Institute,” Kurchatov Square 1, 123098 Moscow (Russian Federation)

    2015-05-15

    Experiments in Alcator C-Mod tokamak plasmas in the Enhanced D-alpha H-mode regime with ITER-like mid-radius plasma density peaking and Ion Cyclotron Resonant heating, in which tungsten is introduced by the laser blow-off technique, have demonstrated that accumulation of tungsten in the central region of the plasma does not take place in these conditions. The measurements obtained are consistent with anomalous transport dominating tungsten transport except in the central region of the plasma where tungsten transport is neoclassical, as previously observed in other devices with dominant neutral beam injection heating, such as JET and ASDEX Upgrade. In contrast to such results, however, the measured scale lengths for plasma temperature and density in the central region of these Alcator C-Mod plasmas, with density profiles relatively flat in the core region due to the lack of core fuelling, are favourable to prevent inter and intra sawtooth tungsten accumulation in this region under dominance of neoclassical transport. Simulations of ITER H-mode plasmas, including both anomalous (modelled by the Gyro-Landau-Fluid code GLF23) and neoclassical transport for main ions and tungsten and with density profiles of similar peaking to those obtained in Alcator C-Mod show that accumulation of tungsten in the central plasma region is also unlikely to occur in stationary ITER H-mode plasmas due to the low fuelling source by the neutral beam injection (injection energy ∼ 1 MeV), which is in good agreement with findings in the Alcator C-Mod experiments.

  3. Physics research needs for ITER

    International Nuclear Information System (INIS)

    Sauthoff, N.R.

    1995-01-01

    Design of ITER entails the application of physics design tools that have been validated against the world-wide data base of fusion research. In many cases, these tools do not yet exist and must be developed as part of the ITER physics program. ITER's considerable increases in power and size demand significant extrapolations from the current data base; in several cases, new physical effects are projected to dominate the behavior of the ITER plasma. This paper focuses on those design tools and data that have been identified by the ITER team and are not yet available; these needs serve as the basis for the ITER Physics Research Needs, which have been developed jointly by the ITER Physics Expert Groups and the ITER design team. Development of the tools and the supporting data base is an on-going activity that constitutes a significant opportunity for contributions to the ITER program by fusion research programs world-wide

  4. Overview of the JET neutral beam enhancement project

    International Nuclear Information System (INIS)

    Ciric, D.; Brown, D.P.D.; Challis, C.D.; Chuilon, B.; Cox, S.J.; Crowley, B.; Day, I.E.; Edwards, D.C.; Evison, G.; Hackett, L.J.; Hotchin, S.; Hudson, Z.; Jenkins, I.; Jones, T.T.C.; King, R.; Kovari, M.; Martin, D.; Milnes, J.; Parkin, A.; Puma, A. Li

    2007-01-01

    The JET neutral beam (NB) heating system is being upgraded as a part of the ongoing JET Enhancement Programme. This is one of the largest upgrades of the JET machine carried out within the EFDA-JET framework. The main goals of the project are to increase the NB power delivered to JET plasma, to increase the beam pulse duration and to improve the availability and reliability of the JET NB system. The upgrade of the system is being carried out through the modification of the two existing neutral injector boxes (NIBs), each equipped with up to eight positive ion neutral injectors (PINIs). Significant changes of the JET NB system will be carried out within the next few years and will include modification of all PINIs, modification or replacement of various beamline components and corresponding instrumentation, procurement and installation of new high voltage power supply (HVPS) units and corresponding control systems and refurbishment of the 36 kV power distribution. Various physics, engineering and planning issues related to this project, as well as the current status of the project are discussed in detail. Particular attention is given to the results of a PINI prototype test, which are of crucial importance for the successful completion of the entire enhancement programme. Upon the completion of the project in 2009/2010, JET NB system should be capable of delivering more than 34 MW of deuterium beam power into the JET plasma for a duration of up to 20 s with improved reliability. This will significantly enhance overall capabilities of the JET machine in support of ITER development

  5. ICRF Review: From ERASMUS To ITER

    Science.gov (United States)

    Weynants, R. R.

    2009-11-01

    This is a personal account of how I saw ICRF evolve since 1974, with a presentation that is ordered according to the topics: heating, antenna coupling, impurity generation/mitigation and system technology. The nature of the main issues is each time reviewed, recent findings are incorporated, and it is shown how the ICRF community has been able to react to sometimes rapidly changing demands and is indeed resolutely preparing ITER.

  6. Supply of a prototype component for the ITER divertor baffle

    International Nuclear Information System (INIS)

    Bobin-Vastra, I.; Febvre, M.; Schedler, B.; Ploechl, L.; Bouveret, Y.; Cauvin, D.; Raisson, G.; Merola, M.

    2001-01-01

    The ITER divertor baffle is one of the Plasma facing components which are developed in the frame of the ITER concept. The supply consisted in the manufacturing of four panels with four First Wall geometries using macroblock or heat sink+armour concepts. DS-Copper, and CuCrZr were the materials for the heat sink, and CFC or Tungsten Plasma spray were the armour. The panels included two Copper-based tubes each. The final purpose is the comparison of the fabricability of each type and the performances of each panel under heat fluxes

  7. 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....

  8. ITER power electrical networks

    International Nuclear Information System (INIS)

    Sejas Portela, S.

    2011-01-01

    The ITER project (International Thermonuclear Experimental Reactor) is an international effort to research and development to design, build and operate an experimental facility to demonstrate the scientific and technological possibility of obtaining useful energy from the physical phenomenon known as nuclear fusion.

  9. ITER conceptual design report

    International Nuclear Information System (INIS)

    1991-01-01

    Results of the International Thermonuclear Experimental Reactor (ITER) Conceptual Design Activity (CDA) are reported. This report covers the Terms of Reference for the project: defining the technical specifications, defining future research needs, define site requirements, and carrying out a coordinated research effort coincident with the CDA. Refs, figs and tabs

  10. ITER at Cadarache

    International Nuclear Information System (INIS)

    2005-06-01

    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.)

  11. ITER conceptual design

    International Nuclear Information System (INIS)

    Tomabechi, K.; Gilleland, J.R.; Sokolov, Yu.A.; Toschi, R.

    1991-01-01

    The Conceptual Design Activities of the International Thermonuclear Experimental Reactor (ITER) were carried out jointly by the European Community, Japan, the Soviet Union and the United States of America, under the auspices of the International Atomic Energy Agency. The European Community provided the site for joint work sessions at the Max-Planck-Institut fuer Plasmaphysik in Garching, Germany. The Conceptual Design Activities began in the spring of 1988 and ended in December 1990. The objectives of the activities were to develop the design of ITER, to perform a safety and environmental analysis, to define the site requirements as well as the future research and development needs, to estimate the cost and manpower, and to prepare a schedule for detailed engineering design, construction and operation. On the basis of the investigation and analysis performed, a concept of ITER was developed which incorporated maximum flexibility of the performance of the device and allowed a variety of operating scenarios to be adopted. The heart of the machine is a tokamak having a plasma major radius of 6 m, a plasma minor radius of 2.15 m, a nominal plasma current of 22 MA and a nominal fusion power of 1 GW. The conceptual design can meet the technical objectives of the ITER programme. Because of the success of the Conceptual Design Activities, the Parties are now considering the implementation of the next phase, called the Engineering Design Activities. (author). Refs, figs and tabs

  12. US ITER Management Plan

    International Nuclear Information System (INIS)

    1991-12-01

    This US ITER Management Plan is the plan for conducting the Engineering Design Activities within the US. The plan applies to all design, analyses, and associated physics and technology research and development (R ampersand D) required to support the program. The plan defines the management considerations associated with these activities. The plan also defines the management controls that the project participants will follow to establish, implement, monitor, and report these activities. The activities are to be conducted by the project in accordance with this plan. The plan will be updated to reflect the then-current management approach required to meet the project objectives. The plan will be reviewed at least annually for possible revision. Section 2 presents the ITER objectives, a brief description of the ITER concept as developed during the Conceptual Design Activities, and comments on the Engineering Design Activities. Section 3 discusses the planned international organization for the Engineering Design Activities, from which the tasks will flow to the US Home Team. Section 4 describes the US ITER management organization and responsibilities during the Engineering Design Activities. Section 5 describes the project management and control to be used to perform the assigned tasks during the Engineering Design Activities. Section 6 presents the references. Several appendices are provided that contain detailed information related to the front material

  13. ITER fuel cycle

    International Nuclear Information System (INIS)

    Leger, D.; Dinner, P.; Yoshida, H.

    1991-01-01

    Resulting from the Conceptual Design Activities (1988-1990) by the parties involved in the International Thermonuclear Experimental Reactor (ITER) project, this document summarizes the design requirements and the Conceptual Design Descriptions for each of the principal subsystems and design options of the ITER Fuel Cycle conceptual design. The ITER Fuel Cycle system provides for the handling of all tritiated water and gas mixtures on ITER. The system is subdivided into subsystems for fuelling, primary (torus) vacuum pumping, fuel processing, blanket tritium recovery, and common processes (including isotopic separation, fuel management and storage, and processes for detritiation of solid, liquid, and gaseous wastes). After an introduction describing system function and conceptual design procedure, a summary of the design is presented including a discussion of scope and main parameters, and the fuel design options for fuelling, plasma chamber vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary and common processes. Design requirements are defined and design descriptions are given for the various subsystems (fuelling, plasma vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary/common processes). The document ends with sections on fuel cycle design integration, fuel cycle building layout, safety considerations, a summary of the research and development programme, costing, and conclusions. Refs, figs and tabs

  14. Advances in iterative methods

    International Nuclear Information System (INIS)

    Beauwens, B.; Arkuszewski, J.; Boryszewicz, M.

    1981-01-01

    Results obtained in the field of linear iterative methods within the Coordinated Research Program on Transport Theory and Advanced Reactor Calculations are summarized. The general convergence theory of linear iterative methods is essentially based on the properties of nonnegative operators on ordered normed spaces. The following aspects of this theory have been improved: new comparison theorems for regular splittings, generalization of the notions of M- and H-matrices, new interpretations of classical convergence theorems for positive-definite operators. The estimation of asymptotic convergence rates was developed with two purposes: the analysis of model problems and the optimization of relaxation parameters. In the framework of factorization iterative methods, model problem analysis is needed to investigate whether the increased computational complexity of higher-order methods does not offset their increased asymptotic convergence rates, as well as to appreciate the effect of standard relaxation techniques (polynomial relaxation). On the other hand, the optimal use of factorization iterative methods requires the development of adequate relaxation techniques and their optimization. The relative performances of a few possibilities have been explored for model problems. Presently, the best results have been obtained with optimal diagonal-Chebyshev relaxation

  15. 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`.

  16. ITER Safety and Licensing

    International Nuclear Information System (INIS)

    Girard, J-.P; Taylor, N.; Garin, P.; Uzan-Elbez, J.; GULDEN, W.; Rodriguez-Rodrigo, L.

    2006-01-01

    The site for the construction of ITER has been chosen in June 2005. The facility will be implemented in Europe, south of France close to Marseille. The generic safety scheme is now under revision to adapt the design to the host country regulation. Even though ITER will be an international organization, it will have to comply with the French requirements in the fields of public and occupational health and safety, nuclear safety, radiation protection, licensing, nuclear substances and environmental protection. The organization of the central team together with its partners organized in domestic agencies for the in-kind procurement of components is a key issue for the success of the experimentation. ITER is the first facility that will achieve sustained nuclear fusion. It is both important for the experimental one-of-a-kind device, ITER itself, and for the future of fusion power plants to well understand the key safety issues of this potential new source of energy production. The main safety concern is confinement of the tritium, activated dust in the vacuum vessel and activated corrosion products in the coolant of the plasma-facing components. This is achieved in the design through multiple confinement barriers to implement the defence in depth approach. It will be demonstrated in documents submitted to the French regulator that these barriers maintain their function in all postulated incident and accident conditions. The licensing process started by examination of the safety options. This step has been performed by Europe during the candidature phase in 2002. In parallel to the final design, and taking into account the local regulations, the Preliminary Safety Report (RPrS) will be drafted with support of the European partner and others in the framework of ITER Task Agreements. Together with the license application, the RPrS will be forwarded to the regulatory bodies, which will launch public hearings and a safety review. Both processes must succeed in order to

  17. Advanced fuelling system for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Raman, Roger [University of Washington, Seattle, WA (United States)], E-mail: raman@aa.washington.edu

    2008-12-15

    Steady-state high-performance discharges in reactors, such as the Advanced Tokamak (AT) scenarios would rely on optimized density and pressure profiles that must be maintained. This maximizes the bootstrap current fraction, reduces reactor recycling power and reduces thermal stresses. Other than a system for the balance of current drive not provided by bootstrap current drive, no other sources of input power, such as from neutral beams, are allowed. For these systems, a precision fuelling system would be the ideal way to control the fusion burn by controlling and maintaining the required pressure profile. This requires a fuelling system that is capable of depositing fuel at any radial location within the plasma while at the same time not altering the density profile to a level that degrades the required pressure profile. Present fuelling systems are incapable of meeting these requirements. An advanced fuelling system based on Compact Toroid injection has the potential to meet these needs while simultaneously providing a source of toroidal momentum input. Description of a conceptual Compact Toroid fueller for ITER is presented in conjunction with a plan for developing this much needed technology.

  18. Design progress of the VV sectors and ports towards the ITER construction

    International Nuclear Information System (INIS)

    Utin, Yu.; Ioki, K.; Bachmann, C.

    2007-01-01

    The ITER vacuum vessel (VV) is an all-welded torus-shaped double-wall structure with stiffening ribs between the shells. The VV main function is to provide the high-vacuum and primary confinement boundary. The vessel also supports in-vessel components such as the blanket modules and the divertor cassettes. Along with these components, the VV provides radiation shielding - the neutron heat is removed by water circulating between the shells. To satisfy the manufacture and assembly needs, the VV consists of nine sectors. To provide access inside the vessel for auxiliary plasma heating, diagnostics, vacuum pumping and other needs, the VV is equipped with upper, equatorial and lower ports. The upper and regular equatorial ports are occupied with the port plugs. In addition, there are three ports at the equatorial level dedicated for neutral beam (NB) injection. As the ITER construction phase app