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Sample records for iter reference scenarios

  1. Free-boundary simulations of ITER advanced scenarios

    International Nuclear Information System (INIS)

    Besseghir, K.

    2013-06-01

    The successful operation of ITER advanced scenarios is likely to be a major step forward in the development of controlled fusion as a power production source. ITER advanced scenarios raise specific challenges that are not encountered in presently-operated tokamaks. In this thesis, it is argued that ITER advanced operation may benefit from optimal control techniques. Optimal control ensures high performance operation while guaranteeing tokamak integrity. The application of optimal control techniques for ITER operation is assessed and it is concluded that robust optimisation is appropriate for ITER operation of advanced scenarios. Real-time optimisation schemes are discussed and it is concluded that the necessary conditions of optimality tracking approach may potentially be appropriate for ITER operation, thus offering a viable closed-loop optimal control approach. Simulations of ITER advanced operation are necessary in order to assess the present ITER design and uncover the main difficulties that may be encountered during advanced operation. The DINA-CH and CRONOS full tokamak simulator is used to simulate the operation of the ITER hybrid and steady-state scenarios. It is concluded that the present ITER design is appropriate for performing a hybrid scenario pulse lasting more than 1000 sec, with a flat-top plasma current of 12 MA, and a fusion gain of Q ≅ 8. Similarly, a steady-state scenario without internal transport barrier, with a flat-top plasma current of 10 MA, and with a fusion gain of Q ≅ 5 can be realised using the present ITER design. The sensitivity of the advanced scenarios with respect to transport models and physical assumption is assessed using CRONOS. It is concluded that the hybrid scenario and the steady-state scenario are highly sensitive to the L-H transition timing, to the value of the confinement enhancement factor, to the heating and current drive scenario during ramp-up, and, to a lesser extent, to the density peaking and pedestal

  2. Free-boundary simulations of ITER advanced scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Besseghir, K.

    2013-06-15

    The successful operation of ITER advanced scenarios is likely to be a major step forward in the development of controlled fusion as a power production source. ITER advanced scenarios raise specific challenges that are not encountered in presently-operated tokamaks. In this thesis, it is argued that ITER advanced operation may benefit from optimal control techniques. Optimal control ensures high performance operation while guaranteeing tokamak integrity. The application of optimal control techniques for ITER operation is assessed and it is concluded that robust optimisation is appropriate for ITER operation of advanced scenarios. Real-time optimisation schemes are discussed and it is concluded that the necessary conditions of optimality tracking approach may potentially be appropriate for ITER operation, thus offering a viable closed-loop optimal control approach. Simulations of ITER advanced operation are necessary in order to assess the present ITER design and uncover the main difficulties that may be encountered during advanced operation. The DINA-CH and CRONOS full tokamak simulator is used to simulate the operation of the ITER hybrid and steady-state scenarios. It is concluded that the present ITER design is appropriate for performing a hybrid scenario pulse lasting more than 1000 sec, with a flat-top plasma current of 12 MA, and a fusion gain of Q ≅ 8. Similarly, a steady-state scenario without internal transport barrier, with a flat-top plasma current of 10 MA, and with a fusion gain of Q ≅ 5 can be realised using the present ITER design. The sensitivity of the advanced scenarios with respect to transport models and physical assumption is assessed using CRONOS. It is concluded that the hybrid scenario and the steady-state scenario are highly sensitive to the L-H transition timing, to the value of the confinement enhancement factor, to the heating and current drive scenario during ramp-up, and, to a lesser extent, to the density peaking and pedestal

  3. Demonstration of ITER operational scenarios on DIII-D

    International Nuclear Information System (INIS)

    Doyle, E.J.; DeBoo, J.C.; Ferron, J.R.; Jackson, G.L.; Luce, T.C.; Osborne, T.H.; Politzer, P.A.; Groebner, R.J.; Hyatt, A.W.; La Haye, R.J.; Petrie, T.W.; Petty, C.C.; Murakami, M.; Park, J.-M.; Reimerdes, H.; Budny, R.V.; Casper, T.A.; Holcomb, C.T.; Challis, C.D.; McKee, G.R.

    2010-01-01

    The DIII-D programme has recently initiated an effort to provide suitably scaled experimental evaluations of four primary ITER operational scenarios. New and unique features of this work are that the plasmas incorporate essential features of the ITER scenarios and anticipated operating characteristics; e.g. the plasma cross-section, aspect ratio and value of I/aB of the DIII-D discharges match the ITER design, with size reduced by a factor of 3.7. Key aspects of all four scenarios, such as target values for β N and H 98 , have been replicated successfully on DIII-D, providing an improved and unified physics basis for transport and stability modelling, as well as for performance extrapolation to ITER. In all four scenarios, normalized performance equals or closely approaches that required to realize the physics and technology goals of ITER, and projections of the DIII-D discharges are consistent with ITER achieving its goals of ≥400 MW of fusion power production and Q ≥ 10. These studies also address many of the key physics issues related to the ITER design, including the L-H transition power threshold, the size of edge localized modes, pedestal parameter scaling, the impact of tearing modes on confinement and disruptivity, beta limits and the required capabilities of the plasma control system. An example of direct influence on the ITER design from this work is a modification of the physics requirements for the poloidal field coil set at 15 MA, based on observations that the inductance in the baseline scenario case evolves to a value that lies outside the original ITER specification.

  4. Demonstration of ITER Operational Scenarios on DIII-D

    International Nuclear Information System (INIS)

    Doyle, E.J.; Budny, R.V.; DeBoo, J.C.; Ferron, J.R.; Jackson, G.L.; Luce, T.C.; Murakami, M.; Osborne, T.H.; Park, J.; Politzer, P.A.; Reimerdes, H.; Casper, T.A.; Challis, C.D.; Groebner, R.J.; Holcomb, C.T.; Hyatt, A.W.; La Haye, R.J.; McKee, G.R.; Petrie, T.W.; Petty, C.C.; Rhodes, T.L.; Shafer, M.W.; Snyder, P.B.; Strait, E.J; Wade, M.R.; Wang, G.; West, W.P.; Zeng, L.

    2008-01-01

    The DIII-D program has recently initiated an effort to provide suitably scaled experimental evaluations of four primary ITER operational scenarios. New and unique features of this work are that the plasmas incorporate essential features of the ITER scenarios and anticipated operating characteristics; e.g., the plasma cross-section, aspect ratio and value of I/aB of the DIII-D discharges match the ITER design, with size reduced by a factor of 3.7. Key aspects of all four scenarios, such as target values for β N and H 98 , have been replicated successfully on DIII-D, providing an improved and unified physics basis for transport and stability modeling, as well as for performance extrapolation to ITER. In all four scenarios normalized performance equals or closely approaches that required to realize the physics and technology goals of ITER, and projections of the DIII-D discharges are consistent with ITER achieving its goals of (ge) 400 MW of fusion power production and Q (ge) 10. These studies also address many of the key physics issues related to the ITER design, including the L-H transition power threshold, the size of ELMs, pedestal parameter scaling, the impact of tearing modes on confinement and disruptivity, beta limits and the required capabilities of the plasma control system. An example of direct influence on the ITER design from this work is a modification of the specified operating range in internal inductance at 15 MA for the poloidal field coil set, based on observations that the measured inductance in the baseline scenario case lay outside the original ITER specification

  5. 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

  6. Development of ITER 15 MA ELMy H-mode Inductive Scenario

    International Nuclear Information System (INIS)

    C. E. Kessel, D. Campbell, Y. Gribov, G. Saibene, G. Ambrosino, T. Casper, M. Cavinato, H. Fujieda, R. Hawryluk, L. D. Horton, A. Kavin, R. Kharyrutdinov, F. Koechl, J. Leuer, A. Loarte, P. J. Lomas, T. Luce, V. Lukash, M. Mattei, I.Nunes, V. Parail, A. Polevoi, A. Portone, R. Sartori, A.C.C. Sips, P. R. Thomas, A. Welander and J. Wesley

    2008-01-01

    The poloidal field (PF) coil system on ITER, which provides both feedforward and feedback control of plasma position, shape, and current, is a critical element for achieving mission performance. Analysis of PF capabilities has focused on the 15 MA Q = 10 scenario with a 300-500 s flattop burn phase. The operating space available for the 15 MA ELMy H-mode plasma discharges in ITER and upgrades to the PF coils or associated systems to establish confidence that ITER mission objectives can be reached have been identified. Time dependent self-consistent free-boundary calculations were performed to examine the impact of plasma variability, discharge programming, and plasma disturbances. Based on these calculations a new reference scenario was developed based upon a large bore initial plasma, early divertor transition, low level heating in L-mode, and a late H-mode onset. Equilibrium analyses for this scenario indicate that the original PF coil limitations do not allow low li (<0.8) operation or lower flux states, and the flattop burn durations were predicted to be less than the desired 400 s. This finding motivates the expansion of the operating space, considering several upgrade options to the PF coils. Analysis was also carried out to examine the feedback current reserve required in the CS and PF coils during a series of disturbances and a feasibility assessment of the 17 MA scenario was undertaken. Results of the studies show that the new scenario and modified PF system will allow a wide range of 15 MA 300-500 s operation and more limited but finite 17 MA operation

  7. ITER safety and operational scenario

    International Nuclear Information System (INIS)

    Shimomura, Y.; Saji, G.

    1998-01-01

    The safety and environmental characteristics of ITER and its operational scenario are described. Fusion has built-in safety characteristics without depending on layers of safety protection systems. Safety considerations are integrated in the design by making use of the intrinsic safety characteristics of fusion adequate to the moderate hazard inventories. In addition to this, a systematic nuclear safety approach has been applied to the design of ITER. The safety assessment of the design shows how ITER will safely accommodate uncertainties, flexibility of plasma operations, and experimental components, which is fundamental in ITER, the first experimental fusion reactor. The operation of ITER will progress step by step from hydrogen plasma operation with low plasma current, low magnetic field, short pulse and low duty factor without fusion power to deuterium-tritium plasma operation with full plasma current, full magnetic field, long pulse and high duty factor with full fusion power. In each step, characteristics of plasma and optimization of plasma operation will be studied which will significantly reduce uncertainties and frequency/severity of plasma transient events in the next step. This approach enhances reliability of ITER operation. (orig.)

  8. Iterative Object Localization Algorithm Using Visual Images with a Reference Coordinate

    Directory of Open Access Journals (Sweden)

    We-Duke Cho

    2008-09-01

    Full Text Available We present a simplified algorithm for localizing an object using multiple visual images that are obtained from widely used digital imaging devices. We use a parallel projection model which supports both zooming and panning of the imaging devices. Our proposed algorithm is based on a virtual viewable plane for creating a relationship between an object position and a reference coordinate. The reference point is obtained from a rough estimate which may be obtained from the preestimation process. The algorithm minimizes localization error through the iterative process with relatively low-computational complexity. In addition, nonlinearity distortion of the digital image devices is compensated during the iterative process. Finally, the performances of several scenarios are evaluated and analyzed in both indoor and outdoor environments.

  9. Predictive integrated modelling for ITER scenarios

    International Nuclear Information System (INIS)

    Artaud, J.F.; Imbeaux, F.; Aniel, T.; Basiuk, V.; Eriksson, L.G.; Giruzzi, G.; Hoang, G.T.; Huysmans, G.; Joffrin, E.; Peysson, Y.; Schneider, M.; Thomas, P.

    2005-01-01

    The uncertainty on the prediction of ITER scenarios is evaluated. 2 transport models which have been extensively validated against the multi-machine database are used for the computation of the transport coefficients. The first model is GLF23, the second called Kiauto is a model in which the profile of dilution coefficient is a gyro Bohm-like analytical function, renormalized in order to get profiles consistent with a given global energy confinement scaling. The package of codes CRONOS is used, it gives access to the dynamics of the discharge and allows the study of interplay between heat transport, current diffusion and sources. The main motivation of this work is to study the influence of parameters such plasma current, heat, density, impurities and toroidal moment transport. We can draw the following conclusions: 1) the target Q = 10 can be obtained in ITER hybrid scenario at I p = 13 MA, using either the DS03 two terms scaling or the GLF23 model based on the same pedestal; 2) I p = 11.3 MA, Q = 10 can be reached only assuming a very peaked pressure profile and a low pedestal; 3) at fixed Greenwald fraction, Q increases with density peaking; 4) achieving a stationary q-profile with q > 1 requires a large non-inductive current fraction (80%) that could be provided by 20 to 40 MW of LHCD; and 5) owing to the high temperature the q-profile penetration is delayed and q = 1 is reached about 600 s in ITER hybrid scenario at I p = 13 MA, in the absence of active q-profile control. (A.C.)

  10. The ITER poloidal field configuration and operation scenario

    International Nuclear Information System (INIS)

    Gribov, Y.; Portone, A.; Mondino, P.L.

    1995-01-01

    The ITER Poloidal Field (PF) system must satisfy the following requirements. (1) ITER must have a well-controlled, single null divertor magnetic configuration with nominal plasma current 21MA and moderate plasma elongation k95 < 1.65. (2) For a variety of plasma scenarios the ITER PF system must provide: inductive breakdown and start-up in an expanding-aperture limiter configuration near the outboard first wall; an inductive current ramp-up to the nominal plasma current with a reasonable assumption of resistive loss during current ramp-up; a pulse length of 1,000s for ignition and inductively-sustained burn at nominal plasma current; plasma shutdown (following fusion power termination) in a similar contracting-aperture limiter configuration. The present design of the PF system can satisfy the ITER requirements within specified limitations

  11. Optimizing the current ramp-up phase for the hybrid ITER scenario

    NARCIS (Netherlands)

    Hogeweij, G.M.D.; Artaud, J.F.; Casper, T.A.; Citrin, J.; Imbeaux, F.; Köchl, F.; Litaudon, X.; Voitsekhovitch, I.

    2013-01-01

    The current ramp-up phase for the ITER hybrid scenario is analysed with the CRONOS integrated modelling suite. The simulations presented in this paper show that the heating systems available at ITER allow, within the operational limits, the attainment of a hybrid q profile at the end of the current

  12. Optimizing the current ramp-up phase for the hybrid ITER scenario

    International Nuclear Information System (INIS)

    Hogeweij, G.M.D.; Citrin, J.; Artaud, J.-F.; Imbeaux, F.; Litaudon, X.; Casper, T.A.; Köchl, F.; Voitsekhovitch, I.

    2013-01-01

    The current ramp-up phase for the ITER hybrid scenario is analysed with the CRONOS integrated modelling suite. The simulations presented in this paper show that the heating systems available at ITER allow, within the operational limits, the attainment of a hybrid q profile at the end of the current ramp-up. A reference ramp-up scenario is reached by a combination of NBI, ECCD (UPL) and LHCD. A heating scheme with only NBI and ECCD can also reach the target q profile; however, LHCD can play a crucial role in reducing the flux consumption during the ramp-up phase. The optimum heating scheme depends on the chosen transport model, and on assumptions of parameters like n e peaking, edge T e,i and Z eff . The sensitivity of the current diffusion on parameters that are not easily controlled, shows that development of real-time control is important to reach the target q profile. A first step in that direction has been indicated in this paper. Minimizing resistive flux consumption and optimizing the q profile turn out to be conflicting requirements. A trade-off between these two requirements has to be made. In this paper it is shown that fast current ramp with L-mode current overshoot is at the one extreme, i.e. the optimum q profile at the cost of increased resistive flux consumption, whereas early H-mode transition is at the other extreme. (paper)

  13. Development of RF Tools and Scenarios for ITER on JET

    International Nuclear Information System (INIS)

    Noterdaeme, J.M.; Bobkov, V.; Mantsinen, M.; Salmi, A.; Santala, M.; Rantamaki, K.; Ekedahl, A.; Eriksson, L.G.; Lamalle, P.U.; Lyssoivan, A.; Van Eester, A.D.; Mailloux, J.; Monakhov, I.; Sharapov, S.; Mayoral, M.L.; Meo, F.

    2005-01-01

    The improvement of lower hybrid (LH) coupling with local puffing of D 2 gas, which made operation at ITER relevant distances (10 cm) and with ELMs (edge localized modes) a reality, has been extended to ITER- like plasma shapes with higher triangularity. With ICRF(ion cyclotron resonance frequency), we developed 4 tools such as -1) localized direct electron heating using the He 3 mode conversion scenario for electron heat transport studies, -2) the production of He 4 ions with energies in the MeV range by 3 ω c acceleration of beam injected ions at 120 keV to investigate Alfven instabilities and test α diagnostics, -3) the stabilisation and destabilization of sawteeth and -4) ICRF as as a wall conditioning. Several ITER relevant scenarios were tested. The (He 3 )H minority heating scenario, considered for the non-activated start-up phase of ITER, produces at very low concentration energetic He 3 which heat the electrons indirectly. For n(He 3 )/n e > 2%, the scenario is transformed to a mode conversion scenario where the electrons are heated directly. The (D)H minority heating is not accessible as the concentration of C 6+ dominates the wave propagation and always leads to a mode conversion. The minority heating of T in D is very effective heating for ions and producing neutrons. New results were obtained in several areas of ICRF physics. Experimental evidence confirmed the theoretical prediction that, as the Larmor radius increases beyond 0.5 times the perpendicular wavelength of the wave, the second harmonic acceleration of the ions decreases to very small levels. An exotic fusion reaction (pT) must be taken into account when evaluating neutron rates. The contribution of fast particles accelerated by ICRF to the plasma rotation was clearly identified, but it is only part of an underlying, and not yet understood, co-current plasma rotation. Progress was made in the physics of ELMs while their effect on the ICRF coupling could be minimized with the conjugate

  14. Modelling controlled VDE's and ramp-down scenarios in ITER

    Science.gov (United States)

    Lodestro, L. L.; Kolesnikov, R. A.; Meyer, W. H.; Pearlstein, L. D.; Humphreys, D. A.; Walker, M. L.

    2011-10-01

    Following the design reviews of recent years, the ITER poloidal-field coil-set design, including in-vessel coils (VS3), and the divertor configuration have settled down. The divertor and its material composition (the latter has not been finalized) affect the development of fiducial equilibria and scenarios together with the coils through constraints on strike-point locations and limits on the PF and control systems. Previously we have reported on our studies simulating controlled vertical events in ITER with the JCT 2001 controller to which we added a PID VS3 circuit. In this paper we report and compare controlled VDE results using an optimized integrated VS and shape controller in the updated configuration. We also present our recent simulations of alternate ramp-down scenarios, looking at the effects of ramp-down time and shape strategies, using these controllers. This work performed under the auspices of the U.S. Department of Energy by LLNL under Contract DE-AC52-07NA27344.

  15. Numerical optimization of actuator trajectories for ITER hybrid scenario profile evolution

    International Nuclear Information System (INIS)

    Dongen, J van; Hogeweij, G M D; Felici, F; Geelen, P; Maljaars, E

    2014-01-01

    Optimal actuator trajectories for an ITER hybrid scenario ramp-up are computed using a numerical optimization method. For both L-mode and H-mode scenarios, the time trajectory of plasma current, EC heating and current drive distribution is determined that minimizes a chosen cost function, while satisfying constraints. The cost function is formulated to reflect two desired properties of the plasma q profile at the end of the ramp-up. The first objective is to maximize the ITG turbulence threshold by maximizing the volume-averaged s/q ratio. The second objective is to achieve a stationary q profile by having a flat loop voltage profile. Actuator and physics-derived constraints are included, imposing limits on plasma current, ramp rates, internal inductance and q profile. This numerical method uses the fast control-oriented plasma profile evolution code RAPTOR, which is successfully benchmarked against more complete CRONOS simulations for L-mode and H-mode mode ITER hybrid scenarios. It is shown that the optimized trajectories computed using RAPTOR also result in an improved ramp-up scenario for CRONOS simulations using the same input trajectories. Furthermore, the optimal trajectories are shown to vary depending on the precise timing of the L–H transition. (paper)

  16. The Reference Scenarios for the Swiss Emergency Planning

    International Nuclear Information System (INIS)

    Hanspeter Isaak; Navert, Stephan B.; Ralph Schulz

    2006-01-01

    For the purpose of emergency planning and preparedness, realistic reference scenarios and corresponding accident source terms have been defined on the basis of common plant features. Three types of representative reference scenarios encompass the accident sequences expected to be the most probable. Accident source terms are assumed to be identical for all Swiss nuclear power plants, although the plants differ in reactor type and power. Plant-specific probabilistic safety analyses were used to justify the reference scenarios and the postulated accident source terms. From the full spectrum of release categories available, those categories were selected which would be covered by the releases and time frames assumed in the reference scenarios. For each nuclear power plant, the cumulative frequency of accident sequences not covered by the reference scenarios was determined. It was found that the cumulative frequency for such accident sequences does not exceed about 1 x 10 -6 per year. The Swiss Federal Nuclear Safety Inspectorate concludes that the postulated accident source terms for the reference scenarios are consistent with the current international approach in emergency planning, where one should concentrate on the most probable accident sequences. (N.C.)

  17. Assessment and modeling of inductive and non-inductive scenarios for ITER

    International Nuclear Information System (INIS)

    Boucher, D.; Vayakis, G.; Moreau, D.

    1999-01-01

    This paper presents recent developments in modeling and simulations of ITER performances and scenarios. The first part presents an improved modeling of coupled divertor/main plasma operation including the simulation of the measurements involved in the control loop. The second part explores the fusion performances predicted under non-inductive operation with internal transport barrier. The final part covers a detailed scenario for non-inductive operation using a reverse shear configuration with lower hybrid and fast wave current drive. (author)

  18. Time parallelization of advanced operation scenario simulations of ITER plasma

    International Nuclear Information System (INIS)

    Samaddar, D; Casper, T A; Kim, S H; Houlberg, W A; Berry, L A; Elwasif, W R; Batchelor, D

    2013-01-01

    This work demonstrates that simulations of advanced burning plasma operation scenarios can be successfully parallelized in time using the parareal algorithm. CORSICA -an advanced operation scenario code for tokamak plasmas is used as a test case. This is a unique application since the parareal algorithm has so far been applied to relatively much simpler systems except for the case of turbulence. In the present application, a computational gain of an order of magnitude has been achieved which is extremely promising. A successful implementation of the Parareal algorithm to codes like CORSICA ushers in the possibility of time efficient simulations of ITER plasmas.

  19. ITER-FEAT magnetic configuration and plasma position/shape control in the nominal PF scenario

    International Nuclear Information System (INIS)

    Gribov, Y.V.; Albanese, R.; Ambrosino, G.

    2001-01-01

    The capability of the ITER-FEAT poloidal field system to support the four 'design' scenarios and the high current 'assessed' scenario have been studied. To operate with highly elongated plasma, the system has segmentation of the central solenoid and a separate fast feedback loop for plasma vertical stabilisation. Within the limits imposed on the coil currents, voltages and power, the poloidal field system provides the required plasma scenario and control capabilities. The separatrix deviation from the required position, in scenarios with minor disruptions is within less than about 100 mm. (author)

  20. Development of steady-state scenarios compatible with ITER-like wall conditions

    Energy Technology Data Exchange (ETDEWEB)

    Litaudon, X [Association Euratom-CEA, CEA/DSM/DRFC-Cadarache 13108, St Paul Durance (France); Arnoux, G [Association Euratom-CEA, CEA/DSM/DRFC-Cadarache 13108, St Paul Durance (France); Beurskens, M [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)] (and others)

    2007-12-15

    A key issue for steady-state tokamak operation is to determine the edge conditions that are compatible both with good core confinement and with the power handling and plasma exhaust capabilities of the plasma facing components (PFCs) and divertor systems. A quantitative response to this open question will provide a robust scientific basis for reliable extrapolation of present regimes to an ITER compatible steady-state scenario. In this context, the JET programme addressing steady-state operation is focused on the development of non-inductive, high confinement plasmas with the constraints imposed by the PFCs. A new beryllium main chamber wall and tungsten divertor together with an upgrade of the heating/fuelling capability are currently in preparation at JET. Operation at higher power with this ITER-like wall will impose new constraints on non-inductive scenarios. Recent experiments have focused on the preparation for this new phase of JET operation. In this paper, progress in the development of advanced tokamak (AT) scenarios at JET is reviewed keeping this long-term objective in mind. The approach has consisted of addressing various critical issues separately during the 2006-2007 campaigns with a view to full scenario integration when the JET upgrades are complete. Regimes with internal transport barriers (ITBs) have been developed at q{sub 95} {approx} 5 and high triangularity, {delta} (relevant to the ITER steady-state demonstration) by applying more than 30 MW of additional heating power reaching {beta}{sub N} {approx} 2 at B{sub o} {approx} 3.1 T. Operating at higher {delta} has allowed the edge pedestal and core densities to be increased pushing the ion temperature closer to that of the electrons. Although not yet fully integrated into a performance enhancing ITB scenario, Neon seeding has been successfully explored to increase the radiated power fraction (up to 60%), providing significant reduction of target tile power fluxes (and hence temperatures) and

  1. Assistance tools for generic definition of ITER maintenance tasks and scenarios in advanced supervisory control systems

    International Nuclear Information System (INIS)

    Zieba, Stéphane; Russotto, François-Xavier; Da Silva Simoes, Max; Measson, Yvan

    2013-01-01

    Highlights: ► Improve supervisory control systems for ITER in-vessel and hot cell maintenance. ► Optimize remote handling operations effectiveness, reliability and safety. ► Provide a generic description of the maintenance tasks and scenarios. ► Development of context-based assistances for operators and supervisor. ► Improvement of operator's situation awareness. -- Abstract: This paper concerns the improvement of supervisory control systems in the context of remote handling for the maintenance tasks in ITER. This work aims at providing a single formalism and tools to define in a generic way the ITER maintenance tasks and scenarios for in-vessel and hot cell operations. A three-layered approach is proposed to model these tasks and scenarios. Physical actions are defined for the scene elements. From these physical actions, behaviours are defined to represent high-level functionalities. Finally, interaction modes define the way that behaviours are achieved in terms of human–machine interactions. Case study concerning the blanket maintenance procedure is discussed concerning the contributions of the descriptive model and the context-based assistances to the activities of supervisory control

  2. Control advances for achieving the ITER baseline scenario on KSTAR

    Science.gov (United States)

    Eidietis, N. W.; Barr, J.; Hahn, S. H.; Humphreys, D. A.; in, Y. K.; Jeon, Y. M.; Lanctot, M. J.; Mueller, D.; Walker, M. L.

    2017-10-01

    Control methodologies developed to enable successful production of ITER baseline scenario (IBS) plasmas on the superconducting KSTAR tokamak are presented: decoupled vertical control (DVC), real-time feedforward (rtFF) calculation, and multi-input multi-output (MIMO) X-point control. DVC provides fast vertical control with the in-vessel control coils (IVCC) while sharing slow vertical control with the poloidal field (PF) coils to avoid IVCC saturation. rtFF compensates for inaccuracies in offline PF current feedforward programming, allowing reduction or removal of integral gain (and its detrimental phase lag) from the shape controller. Finally, MIMO X-point control provides accurate positioning of the X-point despite low controllability due to the large distance between coils and plasma. Combined, these techniques enabled achievement of IBS parameters (q95 = 3.2, βN = 2) with a scaled ITER shape on KSTAR. n =2 RMP response displays a strong dependence upon this shaping. Work supported by the US DOE under Award DE-SC0010685 and the KSTAR project.

  3. FAST Plasma Scenarios and Equilibrium Configurations

    International Nuclear Information System (INIS)

    Calabro, G.; Crisanti, F.; Ramogida, G.; Cardinali, A.; Cucchiaro, A.; Maddaluno, G.; Pizzuto, A.; Pericoli Ridolfini, V.; Tuccillo, A.A.; Zonca, F.; Albanese, R.; Granucci, G.; Nowak, S.

    2008-01-01

    In this paper we present the Fusion Advanced Studies Torus (FAST) plasma scenarios and equilibrium configurations, designed to reproduce the ITER ones (with scaled plasma current) and suitable to fulfil plasma conditions for integrated studies of burning plasma physics, Plasma Wall interaction, ITER relevant operation problems and Steady State scenarios. The attention is focused on FAST flexibility in terms of both performance and physics that can be investigated: operations are foreseen at a wide range of parameters from high performance H-Mode (toroidal field, B T , up to 8.5 T; plasma current, I P , up to 8 MA) to advanced tokamak (AT) operation (I P =3 MA) as well as full non inductive current scenario (I P =2 MA). The coupled heating power is provided with 30MW delivered by an Ion Cyclotron Resonance Heating (ICRH) system (30-90MHz), 6 MW by a Lower Hybrid (LH) system (3.7 or 5 GHz) for the long pulse AT scenario, 4 MW by an Electron Cyclotron Resonant Heating (ECRH) system (170 GHz-B T =6T) for MHD and electron heating localized control and, eventually, with 10 MW by a Negative Ion Beam (NNBI), which the ports are designed to accommodate. In the reference H-mode scenario FAST preserves (with respect to ITER) fast ions induced as well as turbulence fluctuation spectra, thus, addressing the cross-scale couplings issue of micro- to meso-scale physics. The noninductive scenario at I P =2MA is obtained with 60-70 % of bootstrap and the remaining by LHCD. Predictive simulations of the H-mode scenarios described above have been performed by means of JETTO code, using a semi-empirical mixed Bohm/gyro-Bohm transport model. Plasma position and Shape Control studies are also presented for the reference scenario

  4. Investigation of key parameters for the development of reliable ITER baseline operation scenarios using CORSICA

    Science.gov (United States)

    Kim, S. H.; Casper, T. A.; Snipes, J. A.

    2018-05-01

    ITER will demonstrate the feasibility of burning plasma operation by operating DT plasmas in the ELMy H-mode regime with a high ratio of fusion power gain Q ~ 10. 15 MA ITER baseline operation scenario has been studied using CORSICA, focusing on the entry to burn, flat-top burning plasma operation and exit from burn. The burning plasma operation for about 400 s of the current flat-top was achieved in H-mode within the various engineering constraints imposed by the poloidal field coil and power supply systems. The target fusion gain (Q ~ 10) was achievable in the 15 MA ITER baseline operation with a moderate amount of the total auxiliary heating power (~50 MW). It has been observed that the tungsten (W) concentration needs to be maintained low level (n w/n e up to the order of 1.0  ×  10-5) to avoid the radiative collapse and uncontrolled early termination of the discharge. The dynamic evolution of the density can modify the H-mode access unless the applied auxiliary heating power is significantly higher than the H-mode threshold power. Several qualitative sensitivity studies have been performed to provide guidance for further optimizing the plasma operation and performance. Increasing the density profile peaking factor was quite effective in increasing the alpha particle self-heating power and fusion power multiplication factor. Varying the combination of auxiliary heating power has shown that the fusion power multiplication factor can be reduced along with the increase in the total auxiliary heating power. As the 15 MA ITER baseline operation scenario requires full capacity of the coil and power supply systems, the operation window for H-mode access and shape modification was narrow. The updated ITER baseline operation scenarios developed in this work will become a basis for further optimization studies necessary along with the improvement in understanding the burning plasma physics.

  5. Multi-scenario evaluation and specification of electromagnetic loads on ITER vacuum vessel

    International Nuclear Information System (INIS)

    Rozov, Vladimir; Martinez, J.-M.; Portafaix, C.; Sannazzaro, G.

    2014-01-01

    Highlights: • We present the results of multi-scenario analysis of EM loads on ITER vacuum vessel (VV). • The differentiation of models provides the economic way to perform big amount of calculations. • Functional approximation is proposed for distributed data/FE/numerical results specification. • Examples of specification of the load profiles by trigonometric polynomials (DHT) are given. • Principles of accounting for toroidal asymmetry at EM interactions in tokamak are considered. - Abstract: The electro-magnetic (EM) transients cause mechanical forces, which represent one of the most critical loads for the ITER vacuum vessel (VV). The paper is focused on the results of multi-scenario analysis and systematization of these EM loads, including specifically addressed pressures on shells and the net vertical force. The proposed mathematical model and computational technology, based on the use of integral parameters and operational analysis methods, enabled qualitative and quantitative analysis of the problem, time-efficient computations and systematic assessment of a large number of scenarios. The obtained estimates, found envelopes and peak values exemplify the principal loads on the VV and provide a database to support engineering load specifications. Special attention is given to the challenge of specification and documenting of the results in a form, suitable for using the data in engineering applications. The practical aspects of specification of distributed data, such as experimental and finite-element (FE) results, by analytical interpolants are discussed. The example of functional approximation of the load profiles by trigonometric polynomials based on discrete Hartley transform (DHT) is given

  6. Multi-scenario evaluation and specification of electromagnetic loads on ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Rozov, Vladimir, E-mail: vladimir.rozov@iter.org; Martinez, J.-M.; Portafaix, C.; Sannazzaro, G.

    2014-10-15

    Highlights: • We present the results of multi-scenario analysis of EM loads on ITER vacuum vessel (VV). • The differentiation of models provides the economic way to perform big amount of calculations. • Functional approximation is proposed for distributed data/FE/numerical results specification. • Examples of specification of the load profiles by trigonometric polynomials (DHT) are given. • Principles of accounting for toroidal asymmetry at EM interactions in tokamak are considered. - Abstract: The electro-magnetic (EM) transients cause mechanical forces, which represent one of the most critical loads for the ITER vacuum vessel (VV). The paper is focused on the results of multi-scenario analysis and systematization of these EM loads, including specifically addressed pressures on shells and the net vertical force. The proposed mathematical model and computational technology, based on the use of integral parameters and operational analysis methods, enabled qualitative and quantitative analysis of the problem, time-efficient computations and systematic assessment of a large number of scenarios. The obtained estimates, found envelopes and peak values exemplify the principal loads on the VV and provide a database to support engineering load specifications. Special attention is given to the challenge of specification and documenting of the results in a form, suitable for using the data in engineering applications. The practical aspects of specification of distributed data, such as experimental and finite-element (FE) results, by analytical interpolants are discussed. The example of functional approximation of the load profiles by trigonometric polynomials based on discrete Hartley transform (DHT) is given.

  7. Scenario-based fitted Q-iteration for adaptive control of water reservoir systems under uncertainty

    Science.gov (United States)

    Bertoni, Federica; Giuliani, Matteo; Castelletti, Andrea

    2017-04-01

    Over recent years, mathematical models have largely been used to support planning and management of water resources systems. Yet, the increasing uncertainties in their inputs - due to increased variability in the hydrological regimes - are a major challenge to the optimal operations of these systems. Such uncertainty, boosted by projected changing climate, violates the stationarity principle generally used for describing hydro-meteorological processes, which assumes time persisting statistical characteristics of a given variable as inferred by historical data. As this principle is unlikely to be valid in the future, the probability density function used for modeling stochastic disturbances (e.g., inflows) becomes an additional uncertain parameter of the problem, which can be described in a deterministic and set-membership based fashion. This study contributes a novel method for designing optimal, adaptive policies for controlling water reservoir systems under climate-related uncertainty. The proposed method, called scenario-based Fitted Q-Iteration (sFQI), extends the original Fitted Q-Iteration algorithm by enlarging the state space to include the space of the uncertain system's parameters (i.e., the uncertain climate scenarios). As a result, sFQI embeds the set-membership uncertainty of the future inflow scenarios in the action-value function and is able to approximate, with a single learning process, the optimal control policy associated to any scenario included in the uncertainty set. The method is demonstrated on a synthetic water system, consisting of a regulated lake operated for ensuring reliable water supply to downstream users. Numerical results show that the sFQI algorithm successfully identifies adaptive solutions to operate the system under different inflow scenarios, which outperform the control policy designed under historical conditions. Moreover, the sFQI policy generalizes over inflow scenarios not directly experienced during the policy design

  8. Simulations of ITER disruption and VDE scenarios with TSC and comparison with DINA results

    International Nuclear Information System (INIS)

    Bandyopadhyay, I.

    2008-01-01

    Vertical Displacement Events (VDEs) and plasma current disruptions pose one of the major concerns for the lifetime of in-vessel components in ITER, as well as for machine robustness, as large electromagnetic and thermal loads will induced at such events. Hence, accurate modelling of such events is crucial for estimating disruption induced forces. In the past, ITER disruption modelling has been carried out for ITER using the DINA code. However, since predictive simulations of such events depend on a large number of model assumptions, there exists chances of large error bars on the model predictions. As such it is imperative to validate the code results with other models. Towards this objective, we have carried out the VDE and Disruption simulations using the TSC code and the results are compared with the earlier DINA predictions. A detailed electromagnetic model of the ITER vessel, blankets and the first wall components has been created in TSC. In both VDE and disruption cases, the initial plasma is taken as ITER reference scenario 2 end of burn (EOB) specifications with I p = 15 MA, B t = 5 .3 T, e > 8.8 keV, e > = 1.1 x 10 20 m -3 . The plasma current disruption is initiated by dropping the plasma β in 1 msec, so that after the β crash e > = 6 eV, following which the plasma position control is switched off, resulting in a plasma current quench in about 65 msec. On the other hand, in the VDE case, the plasma control is switched off which results in either upward or downward VDE depending on the initial position of the plasma current centroid. In this case the plasma current remains close to 15 MA for a much longer time, about 700 msec in the simulations till the edge safety factor (q) becomes less than 1.5, following which the β is crashed resulting in plasma current quench. Significant differences exist in the DINA and TSC models, for example, even though the plasma current quench rate predicted by the models matches well in till the halo currents start flowing

  9. The iTREN-2030 reference scenario until 2030. Deliverable D4

    Energy Technology Data Exchange (ETDEWEB)

    Fiorello, Davide; De Stasio, Claudia; Koehler, Jonathan; Kraft, Markus; Netwon, Sean; Purwanto, Joko; Schade, Burkhard; Schade, Wolfgang; Szimba, Eckhard

    2009-07-01

    The basic objective of iTREN-2030 is to extend the forecasting and assessment capabilities of the TRANS-TOOLS transport model to the new policy issues arising from the technology, environment and energy fields. This is achieved by couplin the TRANS-TOOLS model with three other models, ASTRA, POLES and TREMOVE covering these new policy issues. The TRANS-TOOLS transport network model has been developed to constitute the reference tool for supporting transport policy in the EU and currently is being developed in several European projects. The scenario set-up to be developed in iTREN-2030 has been modified, so that the projects develops a reference scenario and an integrated scenario. For the reference scenario, the three other modelling tools are harmonised with TRANS-TOOLS and made consistent with each other. This results in a coherent scenario for Europe until 2030 for technology, transport, energy, environment and economic development. The integrated scenario will consider the changing framework conditions until 2030, inparticular the policy pressure coming from climate policy and the increasing scarcity of fossil fuels as well as the impact of the financial and economic crisis. Within the iTREN-2030 project, the overall objective of Work Package 4 (WP4) producing tis deliverable is to develop the reference scenario for the quantitative projections using the four modelling tools involved in the project. The main aims of WP4 are to (a) define a consistent framework for using the different tools in an integrated way; (b) calibrate models with exchanged input to a coherent joint reference; (c) implement external input from WP3 and running models for projections; (d) produce output procedures and templates to facilitate assessment in WP5.

  10. Power and particle control for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Cohen, S A; Cummings, J; Post, D E; Redi, M H [Princeton Univ., NJ (USA). Plasma Physics Lab.; Braams, B J [New York Univ., NY (USA). Courant Inst. of Mathematical Sciences; Brooks, J [Argonne National Lab., IL (USA); Engelmann, F; Pacher, G W; Pacher, H D [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany, F.R.). NET Design Team; Harrison, M; Hotston, E [AEA Fusion, Culham (UK).

    1990-12-15

    Achievement of ITER's objectives, long-pulse ignited operation and nuclear component testing in quasi-steady-state, requires exhaust of power and helium ash, control of impurity content, and long lifetimes for plasma-facing components. In this paper we describe the data base and modeling results used to extrapolate present edge plasma parameters to ITER. Particular emphasis has been given to determining the uncertainties in predicted divertor performance. These analyses have been applied to four typical scenarios: A1 (ignited, reference Physics Phase), B1 (long pulse, hybrid, Technology Phase), B6 (steady-state, Technology Phase, impurity seeded) and B4 (steady-state, Technology Phase). 43 refs., 3 tabs.

  11. 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.

  12. Electromagnetic study on HCCR TBM for ITER major disruption scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Ku, Duck Young; Lee, Youngmin; Cho, Seungyon; Ahn, Muyoung [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) has been developed in Korea in order to experiment a breeding blanket module in ITER. This TBM will verify the feasibility of tritium self-sufficiency in reactor and the extraction of high-grade heat suitable for electricity generation. Since various loads such as seismic load, electromagnetic (EM) load and heat load significantly affect the soundness of the TBM, a variety of analyses were carried out for design optimization. The EM load is particularly one of main design drivers because large amount of magnetic energy in the plasma are transferred to in-vessel components including the TBM during plasma disruption. Because the TBM is located in equatorial port, major disruption (MD) among various plasma disruption scenarios causes the largest EM loads on the TBM.

  13. Electromagnetic study on HCCR TBM for ITER major disruption scenarios

    International Nuclear Information System (INIS)

    Ku, Duck Young; Lee, Youngmin; Cho, Seungyon; Ahn, Muyoung

    2014-01-01

    Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) has been developed in Korea in order to experiment a breeding blanket module in ITER. This TBM will verify the feasibility of tritium self-sufficiency in reactor and the extraction of high-grade heat suitable for electricity generation. Since various loads such as seismic load, electromagnetic (EM) load and heat load significantly affect the soundness of the TBM, a variety of analyses were carried out for design optimization. The EM load is particularly one of main design drivers because large amount of magnetic energy in the plasma are transferred to in-vessel components including the TBM during plasma disruption. Because the TBM is located in equatorial port, major disruption (MD) among various plasma disruption scenarios causes the largest EM loads on the TBM

  14. 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...

  15. Overview of JET results in support of the ITER physics basis

    International Nuclear Information System (INIS)

    Gormezano, C.

    2001-01-01

    The JET experimental campaign has focused on studies in support of the ITER physics basis. An overview of the results obtained is given both for the reference ITER scenario, the ELMy H-mode, and for advanced scenarios which in JET are based on Internal Transport Barriers. JET studies for the ELMy H-mode have been instrumental for the definition of ITER-FEAT. Positive elongation and current scaling in the ITER scaling law have been confirmed, but the observed density scaling fits better a two term (core and edge) model. Significant progress in neo-classical tearing mode limits has been made showing that ITER operation seems to be optimised. Effective helium pumping and divertor enrichment is found to be well within ITER requirements. Target asymmetries and H-isotope retention are well simulated by modelling codes taking into account drift flows in the scrape-off plasmas. Striking improvements in fuelling effectiveness have been found with the new high field pellet launch facility. Good progress has been made on scenarios for achieving good confinement at high densities, both with RI modes and with high field side pellets. Significant development of advanced scenarios in view of their application to ITER has been achieved. Integrated advanced scenarios are in good progress with edge pressure control (impurity radiation). An access domain has been explored showing in particular that the power threshold increases with magnetic field but can be significantly reduced when Lower Hybrid current drive is used to produce target plasma with negative shear. The role of ion pressure peaking on MHD has been well documented. Lack of sufficient additional heating power and interaction with the septum at high beta prevents assessment of beta limits (steady plasmas achieved with β N up to 2.6). Plasmas with non-inductive current (I NI /Ip=60%), well aligned with plasma current, high beta and good confinement have also been obtained. (author)

  16. Identification of reference accident scenarios in SEVESO establishments

    International Nuclear Information System (INIS)

    Delvosalle, C.; Fievez, C.; Pipart, A.; Fabrega, J. Casal; Planas, E.; Christou, M.; Mushtaq, F.

    2005-01-01

    In the frame of the ESREL special session on ARAMIS project, this paper aims at presenting the work carried out in the first Work Package, devoted to the definition of accident scenarios. This topic is a key-point in risk assessment, and serves as basis for the whole risk quantification. A first part of the work aims at building a Methodology for the Identification of Major Accident Hazards (MIMAH), which is carried out with the development of generic fault and event trees based on a typology of equipment and substances. This work is coupled with an historical analysis of accidents. In a second part, influence of safety devices and policies will be considered, in order to build a Methodology for the Identification of Reference Accident Scenarios (MIRAS). This last one will take into account safety systems and lead to obtain more realistic scenarios

  17. iterClust: a statistical framework for iterative clustering analysis.

    Science.gov (United States)

    Ding, Hongxu; Wang, Wanxin; Califano, Andrea

    2018-03-22

    In a scenario where populations A, B1 and B2 (subpopulations of B) exist, pronounced differences between A and B may mask subtle differences between B1 and B2. Here we present iterClust, an iterative clustering framework, which can separate more pronounced differences (e.g. A and B) in starting iterations, followed by relatively subtle differences (e.g. B1 and B2), providing a comprehensive clustering trajectory. iterClust is implemented as a Bioconductor R package. andrea.califano@columbia.edu, hd2326@columbia.edu. Supplementary information is available at Bioinformatics online.

  18. 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

  19. Social Foundation of Scenario Planning

    DEFF Research Database (Denmark)

    Rowland, Nicholas James; Spaniol, Matthew Jon

    2017-01-01

    In this article, the authors establish that models of scenario planning typically involve a series of phases, stages, or steps that imply a sequenced (i.e., linear or chronological) process. Recursive models, in contrast, allow phases to repeat, thus, incorporating iteration. The authors acknowle......In this article, the authors establish that models of scenario planning typically involve a series of phases, stages, or steps that imply a sequenced (i.e., linear or chronological) process. Recursive models, in contrast, allow phases to repeat, thus, incorporating iteration. The authors...... from science and technology studies (STS) on knowledge production, the authors explain transition from one phase to the next and iteration between and within phases based on social negotiation. To this end, the authors examine the interplay between the “scenario development” phase and the “scenario use......” phase of a planning process with a non-governmental organization in Denmark. The upshot for facilitators is practical insight into how transition between phases and phase iteration in scenario planning can be identified, leveraged, and, thus, managed. The upshot for scholars is a related insight...

  20. Ideal MHD stability and performance of ITER steady-state scenarios with ITBs

    Science.gov (United States)

    Poli, F. M.; Kessel, C. E.; Chance, M. S.; Jardin, S. C.; Manickam, J.

    2012-06-01

    Non-inductive steady-state scenarios on ITER will need to operate with internal transport barriers (ITBs) in order to reach adequate fusion gain at typical currents of 9 MA. The large pressure gradients at the location of the internal barrier are conducive to the development of ideal MHD instabilities that may limit the plasma performance and may lead to plasma disruptions. Fully non-inductive scenario simulations with five combinations of heating and current drive sources are presented in this work, with plasma currents in the range 7-10 MA. For each configuration the linear, ideal MHD stability is analysed for variations of the Greenwald fraction and of the pressure peaking factor around the operating point, aiming at defining an operational space for stable, steady-state operations at optimized performance. It is shown that plasmas with lower hybrid heating and current drive maintain the minimum safety factor above 1.5, which is desirable in steady-state operations to avoid neoclassical tearing modes. Operating with moderate ITBs at 2/3 of the minor radius, these plasmas have a minimum safety factor above 2, are ideal MHD stable and reach Q ≳ 5 operating above the ideal no-wall limit.

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

    Science.gov (United States)

    Poli, Francesca

    2012-10-01

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

  2. Progress of the ITER NBI acceleration grid power supply reference design

    International Nuclear Information System (INIS)

    Toigo, Vanni; Zanotto, Loris; Bigi, Marco; Decamps, Hans; Ferro, Alberto; Gaio, Elena; Gutiérrez, Daniel; Tsuchida, Kazuki; Watanabe, Kazuhiro

    2013-01-01

    Highlights: ► This paper reports the progress in the reference design of the Acceleration Grid Power Supply (AGPS) of the ITER Neutral Beam Injector (NBI) ► A critical revision of the main design choices is presented in light of the definition of some key interface parameters between the two AGPS subsystems. ► The verification of the fulfillment of the requirements in any operational conditions is reported and discussed. -- Abstract: This paper reports the progress in the reference design of the Acceleration Grid Power Supply (AGPS) of the ITER Neutral Beam Injector (NBI). The design of the AGPS is very challenging, as it shall be rated to provide about 55 MW at 1 MV dc in quasi steady-state conditions; moreover, the procurement of the system is shared between the European Domestic Agency (F4E) and the Japanese Domestic Agency (JADA), resulting in additional design complication due to the need of a common definition of the interface parameters. A critical revision of the main design choices is presented also in light of the definition of some key interface parameters between the two AGPS subsystems. Moreover, the verification of the fulfillment of the requirements in any operational conditions taking into account the tolerance of the different parameters is also reported and discussed

  3. 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

  4. No-reference image quality assessment for horizontal-path imaging scenarios

    Science.gov (United States)

    Rios, Carlos; Gladysz, Szymon

    2013-05-01

    There exist several image-enhancement algorithms and tasks associated with imaging through turbulence that depend on defining the quality of an image. Examples include: "lucky imaging", choosing the width of the inverse filter for image reconstruction, or stopping iterative deconvolution. We collected a number of image quality metrics found in the literature. Particularly interesting are the blind, "no-reference" metrics. We discuss ways of evaluating the usefulness of these metrics, even when a fully objective comparison is impossible because of the lack of a reference image. Metrics are tested on simulated and real data. Field data comes from experiments performed by the NATO SET 165 research group over a 7 km distance in Dayton, Ohio.

  5. Automatic reference selection for quantitative EEG interpretation: identification of diffuse/localised activity and the active earlobe reference, iterative detection of the distribution of EEG rhythms.

    Science.gov (United States)

    Wang, Bei; Wang, Xingyu; Ikeda, Akio; Nagamine, Takashi; Shibasaki, Hiroshi; Nakamura, Masatoshi

    2014-01-01

    EEG (Electroencephalograph) interpretation is important for the diagnosis of neurological disorders. The proper adjustment of the montage can highlight the EEG rhythm of interest and avoid false interpretation. The aim of this study was to develop an automatic reference selection method to identify a suitable reference. The results may contribute to the accurate inspection of the distribution of EEG rhythms for quantitative EEG interpretation. The method includes two pre-judgements and one iterative detection module. The diffuse case is initially identified by pre-judgement 1 when intermittent rhythmic waveforms occur over large areas along the scalp. The earlobe reference or averaged reference is adopted for the diffuse case due to the effect of the earlobe reference depending on pre-judgement 2. An iterative detection algorithm is developed for the localised case when the signal is distributed in a small area of the brain. The suitable averaged reference is finally determined based on the detected focal and distributed electrodes. The presented technique was applied to the pathological EEG recordings of nine patients. One example of the diffuse case is introduced by illustrating the results of the pre-judgements. The diffusely intermittent rhythmic slow wave is identified. The effect of active earlobe reference is analysed. Two examples of the localised case are presented, indicating the results of the iterative detection module. The focal and distributed electrodes are detected automatically during the repeating algorithm. The identification of diffuse and localised activity was satisfactory compared with the visual inspection. The EEG rhythm of interest can be highlighted using a suitable selected reference. The implementation of an automatic reference selection method is helpful to detect the distribution of an EEG rhythm, which can improve the accuracy of EEG interpretation during both visual inspection and automatic interpretation. Copyright © 2013 IPEM

  6. Dose assessments for Greifswald and Cadarache with new source terms from ITER NSSR-1

    International Nuclear Information System (INIS)

    Raskob, W.; Forschungszentrum Karlsruhe GmbH Technik und Umwelt; Hasemann, I.

    1997-08-01

    Probabilistic dose assessments for accidental atmospheric releases of various ITER source terms which contain tritium and/or activation products were performed for the sites of Greifswald, Germany, and Cadarache, France. No country specific rules were applied and the input parameters were adapted as far as possible to those used within former ITER studies to achieve a better comparability with site independent dose assessments performed in the frame of ITER. The calculations were based on source terms which, at the first time, contain a combination of tritium and activation products. This allowed a better judgement of the contribution of the individual fusion relevant materials to the total dose. The results were compared to site independent dose limits defined in the frame of ITER. Source terms for two different categories, representing 'extremely unlikely events' (CAT-IV) and 'hypothetical sequences' (CAT-V), were investigated. In no cases, the release scenarios of category CAT-IV exceeded the ITER limits. In addition, early doses from the hypothetical scenarios of type CAT-V were still below 50 mSv or 100 mSv, values which are commonly used as lower reference values for evacuation in many potential home countries of ITER. Only the banning of food products was found to be a potential countermeasure which may affect larger areas. (orig.) [de

  7. ITER operational space for full plasma current H-mode operation

    Energy Technology Data Exchange (ETDEWEB)

    Mattei, M. [Assoc. Euratom-ENEA-CREATE, Seconda University di Napoli, Aversa (Italy)], E-mail: massimiliano.mattei@unirc.it; Cavinato, M.; Saibene, G.; Portone, A. [Fusion for Energy Joint Undertaking, 08019 Barcelona (Spain); Albanese, R.; Ambrosino, G. [Assoc. Euratom-ENEA-CREATE, University Napoli Federico II, Napoli (Italy); Horton, L.D. [Max Planck-Institut fur Plasmaphysik, EURATOM-Association, Garching (Germany); Kessel, C. [Princeton Plasma Physics Laboratory, Princeton University (United States); Koechl, F. [Assoc. EURATOM-OAW/ATI, Vienna (Austria); Lomas, P.J. [Euratom/UKAEA Fusion Assoc., Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Nunes, I. [Assoc. EURATOM/IST, Centro de Fusao Nuclear, Lisbon (Portugal); Parail, V. [Max Planck-Institut fur Plasmaphysik, EURATOM-Association, Garching (Germany); Sartori, R. [Fusion for Energy Joint Undertaking, 08019 Barcelona (Spain); Sips, A.C.C. [Max Planck-Institut fur Plasmaphysik, EURATOM-Association, Garching (Germany); Thomas, P.R. [Fusion for Energy Joint Undertaking, 08019 Barcelona (Spain)

    2009-06-15

    Sensitivity studies performed as part of the ITER IO design review highlighted a very stiff dependence of the maximum Q attainable on the machine parameters. In particular, in the considered range, the achievable Q scales with I{sub p}{sup 4}. As a consequence, the achievement of the ITER objective of Q = 10 requires the machine to be routinely operated at a nominal current I{sub p} of 15 MA, and at full toroidal field BT of 5.3 T. This paper analyses the capabilities of the poloidal field (PF) system (including the central solenoid) of ITER against realistic full current plasma scenarios. An exploration of the ITER operational space for the 15 and 17 MA inductive scenario is carried out. An extensive analysis includes the evaluation of margins for the closed loop shape control action. The overall results of this analysis indicate that the control of a 15 MA plasma in ITER is likely to be adequate in the range of li 0.7-0.9 whereas, for a 17 MA plasma, control capabilities are strongly reduced. The ITER operational space, provided by the reference pre-2008 PF system, was rather limited if compared to the range of parameters normally observed in present experiment. Proposals for increasing the current and field limits on PF2, PF5 and PF6, adjustment on the number of turns in some of the PF coils, changes to the divertor dome geometry, to the conductor of PF6 to Nb3Sn, moving PF6 radially and/or vertically are described and evaluated in the paper. Some of them have been included in 2008 ITER revised configuration.

  8. Analysis and validation center for ITER RH maintenance scenarios in a virtual environment

    International Nuclear Information System (INIS)

    Elzendoorn, B.S.Q.; Baar, M. de; Hamilton, D.; Heemskerk, C.J.M.; Koning, J.F.; Ronden, D.M.S.

    2012-01-01

    A facility for detailed simulation of maintenance processes in the ITER Hot Cell Facility (HCF) has been taken into operation. The facility mimics the Remote Handling (RH) work-cells as are presently foreseen. Novel virtual reality (VR) technology, extended with a physics engine is used to create a realistic setting in which a team of Remote Handling (RH) operators can interact with a virtual Hot Cell environment. The physics engine is used to emulate the Hot Cell behavior and to provide tactile feed-back of the (virtual) slave. Multi-operator maintenance scenarios can be developed and tested in virtual reality. Complex interactions between the RH operators and the HCF control system software will be tested. Task performance will be quantified and operational resource consumption will be estimated.

  9. Analysis and validation center for ITER RH maintenance scenarios in a virtual environment

    Energy Technology Data Exchange (ETDEWEB)

    Elzendoorn, B.S.Q., E-mail: B.S.Q.Elzendoorn@rijnhuizen.nl [FOM-Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster and ITER-NL, PO Box 1207, 3430 BE, Nieuwegein (Netherlands); Baar, M. de [FOM-Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster and ITER-NL, PO Box 1207, 3430 BE, Nieuwegein (Netherlands); Hamilton, D. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul-lez-Durance Cedex (France); Heemskerk, C.J.M. [Heemskerk Innovative Technology, Sassenheim (Netherlands); Koning, J.F.; Ronden, D.M.S. [FOM-Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster and ITER-NL, PO Box 1207, 3430 BE, Nieuwegein (Netherlands)

    2012-03-15

    A facility for detailed simulation of maintenance processes in the ITER Hot Cell Facility (HCF) has been taken into operation. The facility mimics the Remote Handling (RH) work-cells as are presently foreseen. Novel virtual reality (VR) technology, extended with a physics engine is used to create a realistic setting in which a team of Remote Handling (RH) operators can interact with a virtual Hot Cell environment. The physics engine is used to emulate the Hot Cell behavior and to provide tactile feed-back of the (virtual) slave. Multi-operator maintenance scenarios can be developed and tested in virtual reality. Complex interactions between the RH operators and the HCF control system software will be tested. Task performance will be quantified and operational resource consumption will be estimated.

  10. Development of internal transport barrier scenarios at ITER-relevant high triangularity in JET

    International Nuclear Information System (INIS)

    Rimini, F.G.; Becoulet, M.; Giovannozzi, E.; Lomas, P.J.; Tudisco, O.; Alper, B.; Crisanti, F.; Baar, M. de; Luna, E. de La; Vries, P. de; Ekedahl, A.; Hawkes, N.; Huysmans, G.; Litaudon, X.; Parail, V.; Saibene, G.; Tuccillo, A.A.; Zastrow, K.D.

    2005-01-01

    The development of scenarios characterized by H-mode confinement and internal transport barriers (ITBs) in high triangularity, δ ∼ 0.4-0.5, discharges is of particular interest for ITER advanced tokamak operation. Previous JET experiments have shown that high triangularity favours H-modes which are ELM-free or develop type I edge localized mode (ELM) activity, which inhibits long lasting ITBs. The recent experiments reported here concentrate on integrated optimization of edge and core conditions. The stability of the edge pedestal was controlled using gas injection, deuterium or light impurities, and plasma current ramps. Both methods yield more ITB-friendly edge pedestal conditions, varying from small type I to type III ELMs and, in extreme cases, resulting in L-mode. In parallel, the conditions for triggering and sustaining ITBs encompassing a large proportion of the plasma volume (outer ITBs) were optimized, as opposed to less performing ITBs located closer to the plasma centre (inner ITB). These plasmas have deeply reversed target current profiles with q min ∼ 3 and a narrow inner ITB, located typically at a small normalized radius ρ E , at q 95 = 7.5, H 89 β N ∼ 3.5-4 and ∼60% of the Greenwald density limit. In summary, a high triangularity scenario has been developed, which combines the desirable characteristics of controlled edge, long lasting wide ITBs and high performance at density higher than the low triangularity JET scenarios

  11. 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

  12. Development of internal transport barrier scenarios at ITER-relevant high triangularity in Jet

    Energy Technology Data Exchange (ETDEWEB)

    Rimini, F.G.; Becoulet, M.; Ekedahl, A.; Huysmans, G.; Joffrin, E.; Litaudon, X. [Association Euratom-CEA, Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Giovannozzi, E.; Tudisco, O.; Crisanti, F. [Association Euratol/ENEA/CNR sulla Fusione, Frascali, Rome (Italy); Lomas, P.J.; Alper, B.; Hawkes, N.; Parail, V.; Zastrow, K.D. [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon (United Kingdom); Baar, M. de; Vries, P. de [Association Euratom-Fom, TEC Cluster, Nieuwegein (Netherlands); La Luna, E. de [Association Euratom-Ciemat, Madrid (Spain); Saibene, G. [EFDA CSU, Garching (Germany)

    2004-07-01

    The development of ITB s(Internal Transport Barrier) scenarios in high triangularity discharges is of particular interest for ITER advanced tokamak operation. Previous JET experiments have shown that high triangularity favours ELM (Edge Localized Mode)-Free or type I ELMs, which inhibit long lasting ITBs. The recent experiments reported here concentrate on integrated optimisation of edge and core conditions. Edge pedestal was controlled using gas injection, Deuterium or light impurities, and plasma current ramps. Both methods yield more ITB-friendly edge pedestal conditions, varying from small type I to type III ELMs and, in extreme cases, to L-mode edge. In parallel, the conditions for triggering and sustaining a wide ITB were optimised. This plasmas have deeply reversed target current profiles with g{sub min} 3. A narrow inner ITB, located in the reversed shear region, is routinely observed. Large radius ITBs are only triggered when the input power exceeds 20-22 MW, but they do not usually survive the transition into H-mode. The best results, in terms of sustained high performance, have been obtained with Neon injection: a wide ITB is triggered during the phase with L-mode edge and survives into H-mode for about 2 s at H{sub 89}{beta}{sub N} {approx} 3.5 and {approx} 60% of the Greenwald density limit. In summary, a high triangularity scenario has been developed, which combines the desirable I characteristics of controlled edge, long lasting wide ITBs and high performance at density higher than the low triangularity JET scenarios. (authors)

  13. ECE diagnostics for RTO/RC ITER

    International Nuclear Information System (INIS)

    Vayakis, G.; Bartlett, D.V.; Costley, A.E.

    2001-01-01

    This paper presents the current status of the Electron Cyclotron Emission (ECE) diagnostic on the Reduced Technical Objectives/Reduced Cost International Thermonuclear Experimental Reactor (RTO/RC ITER). It discusses the implications of the new machine design on the measurement requirements, the ability of the diagnostic technique to meet these, and the changes in the implementation imposed by the new layout. Finally, it outlines the physics studies, design and R and D work required prior to the detailed design and construction of the diagnostic. Key results are: (i) that the localisation of the measurement is similar to that in ITER-FDR (40-100 mm in X-mode, 60-200 mm in O-mode for the reference scenario), so that the relative spatial resolution degrades in this, smaller, machine, and (ii) the expected effect of transport barriers on the temperature profile in the high temperature region will be poorly resolved, because the effect of the temperature gradient on the outboard side is to degrade the resolution to (∼250 mm in X-mode, ∼350 mm in O-mode). Nevertheless ECE will be able to make a unique and useful contribution to the RTO/RC ITER measurement set

  14. Experiment and Modeling of ITER Demonstration Discharges in the DIII-D Tokamak

    International Nuclear Information System (INIS)

    Park, Jin Myung; Doyle, E. J.; Ferron, J.R.; Holcomb, C.T.; Jackson, G.L.; Lao, L.L.; Luce, T.C.; Owen, Larry W.; Murakami, Masanori; Osborne, T.H.; Politzer, P.A.; Prater, R.; Snyder, P.B.

    2011-01-01

    DIII-D is providing experimental evaluation of 4 leading ITER operational scenarios: the baseline scenario in ELMing H-mode, the advanced inductive scenario, the hybrid scenario, and the steady state scenario. The anticipated ITER shape, aspect ratio and value of I/αB were reproduced, with the size reduced by a factor of 3.7, while matching key performance targets for β N and H 98 . Since 2008, substantial experimental progress was made to improve the match to other expected ITER parameters for the baseline scenario. A lower density baseline discharge was developed with improved stationarity and density control to match the expected ITER edge pedestal collisionality (ν* e ∼ 0.1). Target values for β N and H 98 were maintained at lower collisionality (lower density) operation without loss in fusion performance but with significant change in ELM characteristics. The effects of lower plasma rotation were investigated by adding counter-neutral beam power, resulting in only a modest reduction in confinement. Robust preemptive stabilization of 2/1 NTMs was demonstrated for the first time using ECCD under ITER-like conditions. Data from these experiments were used extensively to test and develop theory and modeling for realistic ITER projection and for further development of its optimum scenarios in DIII-D. Theory-based modeling of core transport (TGLF) with an edge pedestal boundary condition provided by the EPED1 model reproduces T e and T i profiles reasonably well for the 4 ITER scenarios developed in DIII-D. Modeling of the baseline scenario for low and high rotation discharges indicates that a modest performance increase of ∼ 15% is needed to compensate for the expected lower rotation of ITER. Modeling of the steady-state scenario reproduces a strong dependence of confinement, stability, and noninductive fraction (f NI ) on q 95 , as found in the experimental I p scan, indicating that optimization of the q profile is critical to simultaneously achieving the

  15. Biomass Scenario Model Documentation: Data and References

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Y.; Newes, E.; Bush, B.; Peterson, S.; Stright, D.

    2013-05-01

    The Biomass Scenario Model (BSM) is a system dynamics model that represents the entire biomass-to-biofuels supply chain, from feedstock to fuel use. The BSM is a complex model that has been used for extensive analyses; the model and its results can be better understood if input data used for initialization and calibration are well-characterized. It has been carefully validated and calibrated against the available data, with data gaps filled in using expert opinion and internally consistent assumed values. Most of the main data sources that feed into the model are recognized as baseline values by the industry. This report documents data sources and references in Version 2 of the BSM (BSM2), which only contains the ethanol pathway, although subsequent versions of the BSM contain multiple conversion pathways. The BSM2 contains over 12,000 total input values, with 506 distinct variables. Many of the variables are opportunities for the user to define scenarios, while others are simply used to initialize a stock, such as the initial number of biorefineries. However, around 35% of the distinct variables are defined by external sources, such as models or reports. The focus of this report is to provide insight into which sources are most influential in each area of the supply chain.

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

    International Nuclear Information System (INIS)

    1996-07-01

    This issue of the Newsletter on the Engineering Design Activities (EDA) for the ITER Tokamak project contains a report on the Tenth ITER Council Meeting, held July 24-25, 1996, in St. Petersburg, Russia; a description of the Status of the ITER EDA by the ITER Director, Dr. R. Aymar; and a report on the so-called Task Number One by the ITER Special Working Group (Basis for the Start of Explorations, presenting possible scenarios toward siting, licensing and host support)

  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. 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)

  19. 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)

  20. Analysis of ITER upper port plug remote handling maintenance scenarios

    International Nuclear Information System (INIS)

    Koning, J.F.; Baar, M.R. de; Elzendoorn, B.S.Q.; Heemskerk, C.J.M.; Ronden, D.M.S.; Schuth, W.J.

    2012-01-01

    Highlights: ► Remote Handling Study Centre: providing RH compatibility analysis. ► Simulation: virtual reality including kinematics and realtime physics simulator. ► Applied on analysis of RH compatibility of Upper Launcher component replacement. ► Resulting in lowered maintenance procedure time and lessons learned. - Abstract: The ITER tokamak has a modular design, with port plugs, blanket modules and divertor cassettes. This set-up allows for maintenance of diagnostics, heating systems and first wall elements. The maintenance can be done in situ, or in the Hot Cell. Safe and effective remote handling (RH) will be ensured by the RH requirements and standards. Compliance is verified through remote handling compatibility assessments at the ITER Design Review milestones. The Remote Handling Study Centre at FOM Institute DIFFER is created to study ITER RH maintenance processes at different levels of complexity, from relatively simple situational awareness checks using snap-shots in the CAD system, time studies using virtual reality (VR) animations, to extensive operational sequence validation with multiple operators in real-time. The multi-operator facility mimics an RH work-cell as presently foreseen in the ITER RH control room. Novel VR technology is used to create a realistic setting in which a team of RH operators can interact with virtual ITER environments. A physics engine is used to emulate real-time contact interaction as to provide realistic haptic feed-back. Complex interactions between the RH operators and the control room system software are tested. RH task performance is quantified and operational resource usage estimated. The article provides a description and lessons learned from a recent study on replacement of the Steering Mirror Assembly on the ECRH (Electron Cyclotron Resonance Heating) Upper Launcher port plug.

  1. Analysis of ITER upper port plug remote handling maintenance scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Koning, J.F., E-mail: j.f.koning@heemskerk-innovative.nl [FOM Institute DIFFER - Dutch Institute for Fundamental Energy Research, Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster and ITER-NL, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Baar, M.R. de; Elzendoorn, B.S.Q. [FOM Institute DIFFER - Dutch Institute for Fundamental Energy Research, Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster and ITER-NL, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Heemskerk, C.J.M. [Heemskerk Innovative Technology, Noordwijk (Netherlands); Ronden, D.M.S.; Schuth, W.J. [FOM Institute DIFFER - Dutch Institute for Fundamental Energy Research, Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster and ITER-NL, PO Box 1207, 3430 BE Nieuwegein (Netherlands)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer Remote Handling Study Centre: providing RH compatibility analysis. Black-Right-Pointing-Pointer Simulation: virtual reality including kinematics and realtime physics simulator. Black-Right-Pointing-Pointer Applied on analysis of RH compatibility of Upper Launcher component replacement. Black-Right-Pointing-Pointer Resulting in lowered maintenance procedure time and lessons learned. - Abstract: The ITER tokamak has a modular design, with port plugs, blanket modules and divertor cassettes. This set-up allows for maintenance of diagnostics, heating systems and first wall elements. The maintenance can be done in situ, or in the Hot Cell. Safe and effective remote handling (RH) will be ensured by the RH requirements and standards. Compliance is verified through remote handling compatibility assessments at the ITER Design Review milestones. The Remote Handling Study Centre at FOM Institute DIFFER is created to study ITER RH maintenance processes at different levels of complexity, from relatively simple situational awareness checks using snap-shots in the CAD system, time studies using virtual reality (VR) animations, to extensive operational sequence validation with multiple operators in real-time. The multi-operator facility mimics an RH work-cell as presently foreseen in the ITER RH control room. Novel VR technology is used to create a realistic setting in which a team of RH operators can interact with virtual ITER environments. A physics engine is used to emulate real-time contact interaction as to provide realistic haptic feed-back. Complex interactions between the RH operators and the control room system software are tested. RH task performance is quantified and operational resource usage estimated. The article provides a description and lessons learned from a recent study on replacement of the Steering Mirror Assembly on the ECRH (Electron Cyclotron Resonance Heating) Upper Launcher port plug.

  2. Remote maintenance development for ITER

    International Nuclear Information System (INIS)

    Tada, Eisuke; Shibanuma, Kiyoshi

    1998-01-01

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

  3. Remote maintenance development for ITER

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-04-01

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

  4. ITER and world chaos; Iter ou le bouleversement du monde

    Energy Technology Data Exchange (ETDEWEB)

    Pourcel, Eric

    2012-02-15

    ITER is the International Thermonuclear Experimental Reactor: the author here develops three scenarios linked to the control of nuclear fusion as a method of producing electrical energy that could take over from fossil fuels in the twenty-First century. His expose shows the likely strategic disarray that might result

  5. Complex amplitude reconstruction by iterative amplitude-phase retrieval algorithm with reference

    Science.gov (United States)

    Shen, Cheng; Guo, Cheng; Tan, Jiubin; Liu, Shutian; Liu, Zhengjun

    2018-06-01

    Multi-image iterative phase retrieval methods have been successfully applied in plenty of research fields due to their simple but efficient implementation. However, there is a mismatch between the measurement of the first long imaging distance and the sequential interval. In this paper, an amplitude-phase retrieval algorithm with reference is put forward without additional measurements or priori knowledge. It gets rid of measuring the first imaging distance. With a designed update formula, it significantly raises the convergence speed and the reconstruction fidelity, especially in phase retrieval. Its superiority over the original amplitude-phase retrieval (APR) method is validated by numerical analysis and experiments. Furthermore, it provides a conceptual design of a compact holographic image sensor, which can achieve numerical refocusing easily.

  6. Detailed mechanical design and manufacturing study for the ITER reference breeding blanket

    International Nuclear Information System (INIS)

    Zacchia, F.; Daenner, W.; Stefanis, L. de; Ferrari, M.; Gerber, A.; Mustoe, J.

    1998-01-01

    This papers relates on the detailed mechanical design, manufacturing feasibility and assembly analysis of a water-cooled solid breeding blanket concept, selected as the ITER reference design. This breeding blanket design is characterised by: i) pressurised water flowing inside flat steel panels for cooling of the internals; each panel is welded along its contour onto the first wall structure and to the rear shield plate after closure of the module (last assembly step). ii) Beryllium (neutronic multiplier) in the form of micro-spheres filling the volume between parallel flat coolant panels. iii) Breeder pebbles enclosed in rods, which form bundles and are themselves embedded inside the Beryllium micro-spheres. (authors)

  7. Calculations of NTM stabilization in ITER-FEAT by ECCD with realistic antenna geometry

    International Nuclear Information System (INIS)

    Ramponi, G.; Nowak, S.; Lazzaro, E.; Giruzzi, G.; Bosia, G.

    2001-01-01

    Neoclassical Tearing Modes stabilization is one of the main purposes for the implementation of an Electron Cyclotron Current Drive system on ITER-FEAT. Previous estimates have shown that a wave power level of 20-30 MW should be appropriate for a substantial reduction of the (3,2) and/or (2,1) modes. Here detailed calculations are presented combining, for the first time, the following elements: i) realistic antenna geometry resulting from detailed study of the implementation in an ITER upper port; ii) Gaussian beam-tracing calculations; iii) 3D Fokker-Planck calculations of the driven current density profile; iv) island evolution calculation, including island rotation effects. The power level necessary for complete stabilization of NTMs is evaluated for the ITER FEAT reference scenarios and the chosen wave frequency of 170 GHz. Optimization as a function of the injection poloidal and toroidal angles is discussed

  8. Dose assessments for Greifswald and Cadarache with updated source terms from ITER NSSR-2

    International Nuclear Information System (INIS)

    Raskob, W.; Hasemann, I.

    1998-08-01

    The International Thermonuclear Experimental Reactor ITER is in its late engineering phase. One of the most important safety aspects - in particular for achieving public acceptance - is to assure that the releases of harzardous material are minimal during normal operation and for accidental events, even if very unlikely. To this purpose probabilistic dose assessments for accidental atmospheric releases of various ITER source terms which contain tritium and/or activation products were performed for the sites of Greifswald, Germany, and Cadarache, France. In addition, routine releases into the atmosphere and hydrosphere have been evaluated. No country specific rules were applied and the input parameters were adapted as far as possible to those used within former studies to achieve a better comparability with site independent dose assessments performed in the frame of ITER. The calculations were based on source terms which, for the first time, contain a combination of tritium and activation products. This allowed a better judgment of the contribution to the total dose of the individual fusion relevant materials. The results were compared to site independent dose limits defined in the frame of ITER. Annual doses from routine releases (CAT-I) are below 0.1 μSv for the aquatic scenarios and are close to 1 μSv for the atmospheric source terms. Source terms for two different categories of accidental releases, representing 'extremely unlikely events' (CAT-IV) and 'hypothetical sequences' (CAT-V), were investigated. In none of these cases, the release scenarios of category CAT-IV exceed the ITER limits. In addition, relevant characteristic quantities of the early dose distribution from the hypothetical scenarios of type CAT-V are still below 50 mSv or 100 mSv, values which are commonly used as lower reference values for evacuation in many potential home countries of ITER. These site specific assessments confirmed that the proposed release limits and thus the derived dose

  9. ARAMIS project: A comprehensive methodology for the identification of reference accident scenarios in process industries

    International Nuclear Information System (INIS)

    Delvosalle, Christian; Fievez, Cecile; Pipart, Aurore; Debray, Bruno

    2006-01-01

    In the frame of the Accidental Risk Assessment Methodology for Industries (ARAMIS) project, this paper aims at presenting the work carried out in the part of the project devoted to the definition of accident scenarios. This topic is a key-point in risk assessment and serves as basis for the whole risk quantification. The first result of the work is the building of a methodology for the identification of major accident hazards (MIMAH), which is carried out with the development of generic fault and event trees based on a typology of equipment and substances. The term 'major accidents' must be understood as the worst accidents likely to occur on the equipment, assuming that no safety systems are installed. A second methodology, called methodology for the identification of reference accident scenarios (MIRAS) takes into account the influence of safety systems on both the frequencies and possible consequences of accidents. This methodology leads to identify more realistic accident scenarios. The reference accident scenarios are chosen with the help of a tool called 'risk matrix', crossing the frequency and the consequences of accidents. This paper presents both methodologies and an application on an ethylene oxide storage

  10. Calibration of ITER Instant Power Neutron Monitors: Recommended Scenario of Experiments at the Reactor

    Science.gov (United States)

    Borisov, A. A.; Deryabina, N. A.; Markovskij, D. V.

    2017-12-01

    Instant power is a key parameter of the ITER. Its monitoring with an accuracy of a few percent is an urgent and challenging aspect of neutron diagnostics. In a series of works published in Problems of Atomic Science and Technology, Series: Thermonuclear Fusion under a common title, the step-by-step neutronics analysis was given to substantiate a calibration technique for the DT and DD modes of the ITER. A Gauss quadrature scheme, optimal for processing "expensive" experiments, is used for numerical integration of 235U and 238U detector responses to the point sources of 14-MeV neutrons. This approach allows controlling the integration accuracy in relation to the number of coordinate mesh points and thus minimizing the number of irradiations at the given uncertainty of the full monitor response. In the previous works, responses of the divertor and blanket monitors to the isotropic point sources of DT and DD neutrons in the plasma profile and to the models of real sources were calculated within the ITER model using the MCNP code. The neutronics analyses have allowed formulating the basic principles of calibration that are optimal for having the maximum accuracy at the minimum duration of in situ experiments at the reactor. In this work, scenarios of the preliminary and basic experimental ITER runs are suggested on the basis of those principles. It is proposed to calibrate the monitors only with DT neutrons and use correction factors to the DT mode calibration for the DD mode. It is reasonable to perform full calibration only with 235U chambers and calibrate 238U chambers by responses of the 235U chambers during reactor operation (cross-calibration). The divertor monitor can be calibrated using both direct measurement of responses at the Gauss positions of a point source and simplified techniques based on the concepts of equivalent ring sources and inverse response distributions, which will considerably reduce the amount of measurements. It is shown that the monitor

  11. Simulations of KSTAR high performance steady state operation scenarios

    International Nuclear Information System (INIS)

    Na, Yong-Su; Kessel, C.E.; Park, J.M.; Yi, Sumin; Kim, J.Y.; Becoulet, A.; Sips, A.C.C.

    2009-01-01

    We report the results of predictive modelling of high performance steady state operation scenarios in KSTAR. Firstly, the capabilities of steady state operation are investigated with time-dependent simulations using a free-boundary plasma equilibrium evolution code coupled with transport calculations. Secondly, the reproducibility of high performance steady state operation scenarios developed in the DIII-D tokamak, of similar size to that of KSTAR, is investigated using the experimental data taken from DIII-D. Finally, the capability of ITER-relevant steady state operation is investigated in KSTAR. It is found that KSTAR is able to establish high performance steady state operation scenarios; β N above 3, H 98 (y, 2) up to 2.0, f BS up to 0.76 and f NI equals 1.0. In this work, a realistic density profile is newly introduced for predictive simulations by employing the scaling law of a density peaking factor. The influence of the current ramp-up scenario and the transport model is discussed with respect to the fusion performance and non-inductive current drive fraction in the transport simulations. As observed in the experiments, both the heating and the plasma current waveforms in the current ramp-up phase produce a strong effect on the q-profile, the fusion performance and also on the non-inductive current drive fraction in the current flattop phase. A criterion in terms of q min is found to establish ITER-relevant steady state operation scenarios. This will provide a guideline for designing the current ramp-up phase in KSTAR. It is observed that the transport model also affects the predictive values of fusion performance as well as the non-inductive current drive fraction. The Weiland transport model predicts the highest fusion performance as well as non-inductive current drive fraction in KSTAR. In contrast, the GLF23 model exhibits the lowest ones. ITER-relevant advanced scenarios cannot be obtained with the GLF23 model in the conditions given in this work

  12. 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

  13. 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

  14. 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

  15. Development and test of the ITER conductor joints

    Energy Technology Data Exchange (ETDEWEB)

    Martovetsky, N., LLNL

    1998-05-14

    Joints for the ITER superconducting Central Solenoid should perform in rapidly varying magnetic field with low losses and low DC resistance. This paper describes the design of the ITER joint and presents its assembly process. Two joints were built and tested at the PTF facility at MIT. Test results are presented, losses in transverse and parallel field and the DC performance are discussed. The developed joint demonstrates sufficient margin for baseline ITER operating scenarios.

  16. ITER project and fusion technology

    International Nuclear Information System (INIS)

    Takatsu, H.

    2011-01-01

    In the sessions of ITR, FTP and SEE of the 23rd IAEA Fusion Energy Conference, 159 papers were presented in total, highlighted by the remarkable progress of the ITER project: ITER baseline has been established and procurement activities have been started as planned with a target of realizing the first plasma in 2019; ITER physics basis is sound and operation scenarios and operational issues have been extensively studied in close collaboration with the worldwide physics community; the test blanket module programme has been incorporated into the ITER programme and extensive R and D works are ongoing in the member countries with a view to delivering their own modules in a timely manner according to the ITER master schedule. Good progress was also reported in the areas of a variety of complementary activities to DEMO, including Broader Approach activities and long-term technology. This paper summarizes the highlights of the papers presented in the ITR, FTP and SEE sessions with a minimum set of background information.

  17. High-β steady-state advanced tokamak regimes for ITER and FIRE

    International Nuclear Information System (INIS)

    Meade, D.M.; Sauthoff, N.R.; Kessel, C.E.; Budny, R.V.; Gorelenkov, N.; Jardin, S.C.; Schmidt, J.A.; Navratil, G.A.; Bialek, J.; Ulrickson, M.A.; Rognlein, T.; Mandrekas, J.

    2005-01-01

    An attractive tokamak-based fusion power plant will require the development of high-β steady-state advanced tokamak regimes to produce a high-gain burning plasma with a large fraction of self-driven current and high fusion-power density. Both ITER and FIRE are being designed with the objective to address these issues by exploring and understanding burning plasma physics both in the conventional H-mode regime, and in advanced tokamak regimes with β N ∼ 3 - 4, and f bs ∼50-80%. ITER has employed conservative scenarios, as appropriate for its nuclear technology mission, while FIRE has employed more aggressive assumptions aimed at exploring the scenarios envisioned in the ARIES power-plant studies. The main characteristics of the advanced scenarios presently under study for ITER and FIRE are compared with advanced tokamak regimes envisioned for the European Power Plant Conceptual Study (PPCS-C), the US ARIES-RS Power Plant Study and the Japanese Advanced Steady-State Tokamak Reactor (ASSTR). The goal of the present work is to develop advanced tokamak scenarios that would fully exploit the capability of ITER and FIRE. This paper will summarize the status of the work and indicate critical areas where further R and D is needed. (author)

  18. Progress in physics basis and its impact on ITER

    International Nuclear Information System (INIS)

    Shimada, M.; Campbell, D.; Stambaugh, R.; Ide, S.; Kamada, Y.; Leonard, A.; Polevoi, A.; Mukhovatov, V.; Costley, A.E.; Gribov, Y.; Oikawa, T.; Sugihara, M.; Asakura, N.; Donne, A.J.H.; Doyle, E.J.; Federici, G.; Kukushkin, A.S.; Gormezano, C.; Gruber, O.; Houlberg, W.; Lipschultz, B.; Medvedev, S.

    2005-01-01

    This paper summarises recent progress in the physics basis and its impact on the expected performance of ITER. Significant progress has been made in many outstanding issues and in the development of hybrid and steady state operation scenarios, leading to increased confidence of achieving ITER's goals. Experiments show that tailoring the current profile can improve confinement over the standard H-mode and allow an increase in beta up to the no-wall limit at safety factors ∼ 4. Extrapolation to ITER suggests that at the reduced plasma current of ∼ 12MA, high Q > 10 and long pulse (>1000 s) operation is possible with benign ELMs. Analysis of disruption scenarios has been performed based on guidelines on current quench rates and halo currents, derived from the experimental database. With conservative assumptions, estimated electromagnetic forces on the in-vessel components are below the design target values, confirming the robustness of the ITER design against disruption forces. (author)

  19. The ITER poloidal field system

    Energy Technology Data Exchange (ETDEWEB)

    Wesley, J [General Atomics, San Diego, CA (USA); Beljakov, V; Kavin, A; Korshakov, V; Kostenko, A; Roshal, A; Zakharov, L [Kurchatov Inst. of Atomic Energy, Moscow (USSR); Bulmer, R; Kaiser, T; Miller, J R; Pearlstein, L D [Lawrence Livermore National Lab., CA (USA); Hogan, J [Oak Ridge National Lab., TN (USA); Kurihara, K; Shimomura, Y; Sugihara, M; Yoshino, R [Japan Atomic Energy Resea

    1990-12-15

    The ITER poloidal field (PF) system uses superconducting coils to provide the plasma equilibrium fields, slow equilibrium control and plasma flux linkage (V-s) needed for the ITER Operations and Research Program. Double-null (DN) divertor plasmas and operation scenarios for 22 MA Physics (high-Q/ignition) and 15 MA Technology (high-fluence testing) phases are provided. For 22 MA plasmas, total PF flux swing is 333 V-s. This provides inductive current drive (CD) for start-up with 66 V-s of resistive loss and 440-s (330-s minimum) sustained burn. The PF system also allows plasma start-up and shutdown scenarios, and can maintain the plasma configuration during burn over a range of current and pressure profiles. Other capabilities include increased plasma current (25 MA with inductive CD; 28 MA with non-inductive CD assist), divertor separatrix sweeping, and semi-DN and single-null plasmas.

  20. 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.

  1. 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.

  2. Implementation of the multireference Brillouin-Wigner and Mukherjee’s coupled cluster methods with non-iterative triple excitations utilizing reference-level parallelism

    Energy Technology Data Exchange (ETDEWEB)

    Bhaskaran-Nair, Kiran; Brabec, Jiri; Apra, Edoardo; van Dam, Hubertus JJ; Pittner, Jiri; Kowalski, Karol

    2012-09-07

    In this paper we discuss the performance of the non-iterative State-Specific Mul- tireference Coupled Cluster (SS-MRCC) methods accounting for the effect of triply excited cluster amplitudes. The corrections to the Brillouin-Wigner and Mukherjee MRCC models based on the manifold of singly and doubly excited cluster amplitudes (BW-MRCCSD and Mk-MRCCSD, respectively) are tested and compared with the exact full configuration interaction results (FCI) for small systems (H2O, N2, and Be3). For larger systems (naphthyne isomers and -carotene), the non-iterative BW-MRCCSD(T) and Mk-MRCCSD(T) methods are compared against the results obtained with the single reference coupled cluster methods. We also report on the parallel performance of the non-iterative implementations based on the use of pro- cessor groups.

  3. Neutron cameras for ITER

    International Nuclear Information System (INIS)

    Johnson, L.C.; Barnes, C.W.; Batistoni, P.

    1998-01-01

    Neutron cameras with horizontal and vertical views have been designed for ITER, based on systems used on JET and TFTR. The cameras consist of fan-shaped arrays of collimated flight tubes, with suitably chosen detectors situated outside the biological shield. The sight lines view the ITER plasma through slots in the shield blanket and penetrate the vacuum vessel, cryostat, and biological shield through stainless steel windows. This paper analyzes the expected performance of several neutron camera arrangements for ITER. In addition to the reference designs, the authors examine proposed compact cameras, in which neutron fluxes are inferred from 16 N decay gammas in dedicated flowing water loops, and conventional cameras with fewer sight lines and more limited fields of view than in the reference designs. It is shown that the spatial sampling provided by the reference designs is sufficient to satisfy target measurement requirements and that some reduction in field of view may be permissible. The accuracy of measurements with 16 N-based compact cameras is not yet established, and they fail to satisfy requirements for parameter range and time resolution by large margins

  4. Magnetic configuration control of ITER plasmas

    International Nuclear Information System (INIS)

    Albanese, R.; Mattei, M.; Portone, A.; Ambrosino, G.; Artaserse, G.; Crisanti, F.; De Tommasi, G.; Fresa, R.; Sartori, F.; Villone, F.

    2007-01-01

    The aim of this paper is to present some new tools used to review the capability of the ITER Poloidal Field (PF) system in controlling the broad range of plasma configurations presently forecasted during ITER operation. The attention is focused on the axi-symmetric aspects of plasma magnetic configuration control since they pose the greatest challenges in terms of control power and they have the largest impact on machine capital cost. Some preliminary results obtained during ongoing activities in collaboration between ENEA/CREATE and EFDA are presented. The paper is divided in two main parts devoted, respectively, to the presentation of a procedure for the PF current optimisation during the scenario, and of a software environment for the study of the PF system capabilities using the plasma linearized response. The proposed PF current optimisation procedure is then used to assess Scenario 2 design, also taking into account the presence of axisymmetric eddy currents and possible variations of poloidal beta and internal inductance. The numerical linear model based tool derived from the JET oriented eXtreme Shape Controller (XSC) tools is finally used to obtain results on the strike point sweeping in ITER

  5. 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.

  6. 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

  7. The reference energy scenario of the DGEMP for 2030,there's many a slip twixt the cup and the lip; Le scenario energetique de reference de la DGEMP pour 2030, ou il y a loin de la coupe aux levres

    Energy Technology Data Exchange (ETDEWEB)

    Acket, C.; Nifenecker, H

    2008-05-15

    The authors discuss the energy reference scenario for 2030, proposed by the DGEMP. This scenario is published every 4 years. It represents the french energy situation in 2030, if no new energy policy was decided, in particularly no new measure from the Grenelle of the environment. The scenario supposes also no energy crisis in the fossil fuels supply and the fuels prices. In this topic, the objective of a contribution of 20% of renewable energies in the final energy consumption, seems possible. (A.L.B.)

  8. ITER Council proceedings: April 1988 - August 1989

    International Nuclear Information System (INIS)

    1990-01-01

    The ITER documentation series, of which this is the sixth report, began with a concise record of the decisions and actions taken in establishing ITER. The contents of that first report include the Terms of Reference Concerning Conceptual Design Activities for an International Thermonuclear Experimental Reactor. The Terms of Reference, which were part of the IAEA's invitation to the prospective parties, formally describe how the co-operative work of the four Parties in the specified activities is directed and managed. The first report in the series also covered activities from the initial meeting of the ITER Quadripartite Initiative Committee in March 1987 through March 1988. The present report is intended to make available in convenient form the essential information on ''landmark'' events in the direction of the ITER activities from the first meeting of the ITER Council (IC), in April 1988, through the letter report by the Council following their fourth meeting in July 1989. This report therefore covers approximately the first half of the Conceptual Design Activities, which are to be concluded in December 1990. THe next section of this report provides, for convenient reference, an overview of the organization and schedule that were adopted for the ITER activities through 1990. The sections that follow contain, for each of the four IC meetings during the period covered by this report, a copy of the official record of the transactions. The written reports of the ITER Management Committee (IMC) and the ITER Scientific and Technical Advisory Committee (ISTAC) to the IC in connection with the meetings are represented by summaries, prepared by the ITER Secretariat. These summaries include direct quotations of especially significant statements in the reports. In general, other supporting material is included only if it is of more than transitory significance. 1 fig and tabs

  9. Neutronic calculations in support of the design of the ITER High Resolution Neutron Spectrometer

    International Nuclear Information System (INIS)

    Moro, F.; Esposito, B.; Marocco, D.; Villari, R.; Petrizzi, L.; Sunden, E. Andersson; Conroy, S.; Ericsson, G.; Johnson, M. Gatu; Dapena, M.

    2011-01-01

    This paper presents the results of neutronic calculations performed to address important issues related to the optimization of the ITER HRNS (High resolution Neutron Spectrometer) design, in particular concerning the definition of the collimator and the choice of the detector system. The calculations have been carried out using the MCNP5 Monte Carlo code in a full 3-D geometry. The HRNS collimation system has been included in the latest MCNP ITER 40 o model (Alite-4). The ITER scenario 2 reference DT plasma fusion neutron source peaked at 14.1 MeV with Gaussian energy distribution has been used. Neutron fluxes and energy spectra (>1 MeV) have been evaluated at different positions along the HRNS collimator and at the detector location. The noise-to-signal ratio (i.e. the ratio of collided to uncollided neutrons), the breakdown of the collided spectrum into its components, the dependency on the first wall aperture and the gamma-ray spectra at the detector position have also been analyzed. The impact of the results on the design of the HRNS diagnostic system is discussed.

  10. Simulation and Analysis of the Hybrid Operating Mode in ITER

    International Nuclear Information System (INIS)

    Kessel, C.E.; Budny, R.V.; Indireshkumar, K.

    2005-01-01

    The hybrid operating mode in ITER is examined with 0D systems analysis, 1.5D discharge scenario simulations using TSC and TRANSP, and the ideal MHD stability is discussed. The hybrid mode has the potential to provide very long pulses and significant neutron fluence if the physics regime can be produced in ITER. This paper reports progress in establishing the physics basis and engineering limitation for the hybrid mode in ITER

  11. Hybrid integral-differential simulator of EM force interactions/scenario-assessment tool with pre-computed influence matrix in applications to ITER

    Science.gov (United States)

    Rozov, V.; Alekseev, A.

    2015-08-01

    A necessity to address a wide spectrum of engineering problems in ITER determined the need for efficient tools for modeling of the magnetic environment and force interactions between the main components of the magnet system. The assessment of the operating window for the machine, determined by the electro-magnetic (EM) forces, and the check of feasibility of particular scenarios play an important role for ensuring the safety of exploitation. Such analysis-powered prevention of damages forms an element of the Machine Operations and Investment Protection strategy. The corresponding analysis is a necessary step in preparation of the commissioning, which finalizes the construction phase. It shall be supported by the development of the efficient and robust simulators and multi-physics/multi-system integration of models. The developed numerical model of interactions in the ITER magnetic system, based on the use of pre-computed influence matrices, facilitated immediate and complete assessment and systematic specification of EM loads on magnets in all foreseen operating regimes, their maximum values, envelopes and the most critical scenarios. The common principles of interaction in typical bilateral configurations have been generalized for asymmetry conditions, inspired by the plasma and by the hardware, including asymmetric plasma event and magnetic system fault cases. The specification of loads is supported by the technology of functional approximation of nodal and distributed data by continuous patterns/analytical interpolants. The global model of interactions together with the mesh-independent analytical format of output provides the source of self-consistent and transferable data on the spatial distribution of the system of forces for assessments of structural performance of the components, assemblies and supporting structures. The numerical model used is fully parametrized, which makes it very suitable for multi-variant and sensitivity studies (positioning, off

  12. Overview of magnetic control in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Zabeo, L., E-mail: luca.zabeo@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul Lez Durance (France); Ambrosino, G., E-mail: ambrosin@unina.it [CREATE/Universitá di Napoli Federico II, Dip. Ingegneria Elettrica e delle Tecnologie dell’informazione, Naples (Italy); Cavinato, M., E-mail: mario.cavinato@f4e.europa.eu [Fusion for Energy (F4E), Josep Pla 2, Torres Diagonal Litoral - B3, 08019 Barcelona (Spain); Gribov, Y., E-mail: yuri.gribov@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul Lez Durance (France); Kavin, A., E-mail: kavina@sintez.niiefa.spb.su [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation); Lukash, V., E-mail: lukash@nfi.kiae.ru [Kurchatov Institute, Moscow (Russian Federation); Mattei, M., E-mail: massimiliano.mattei@unina2.it [CREATE/Seconda Universitá di Napoli, Dip. Ingegneria Industriale e dell’informazione, Naples (Italy); Pironti, A., E-mail: pironti@unina.it [CREATE/Seconda Universitá di Napoli, Dip. Ingegneria Industriale e dell’informazione, Naples (Italy); Snipes, J.A., E-mail: joseph.snipes@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul Lez Durance (France); Vayakis, G., E-mail: george.vayakis@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul Lez Durance (France); Winter, A., E-mail: axel.winter@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul Lez Durance (France)

    2014-05-15

    ITER is targeting Q = 10 with 500 MW of fusion power. To meet this target, the plasma needs to be controlled and shaped for a period of hundreds of seconds, avoiding contact with internal components, and acting against instabilities that could result in the loss of control of the plasma and in its disruptive termination. Axisymmetric magnetic control is a well-understood area being the basic control for any tokamak device. ITER adds more stringent constraints to the control primarily due to machine protection and engineering limits. The limits on the actuators by means of the maximum current and voltage at the coils and the few hundred ms time response of the vacuum vessel requires optimization of the control strategies and the validation of the capabilities of the machine in controlling the designed scenarios. Scenarios have been optimized with realistic control strategies able to guarantee robust control against plasma behavior and engineering limits due to recent changes in the ITER design. Technological issues such as performance changes associated with the optimization of the final design of the central solenoid, control of fast transitions like H to L mode to avoid plasma-wall contact, and optimization of the plasma ramp-down have been modeled to demonstrate the successful operability of ITER and compatibility with the latest refinements in the magnetic system design. Validation and optimization of the scenarios refining the operational space available for ITER and associated control strategies will be proposed. The present capabilities of magnetic control will be assessed and the remaining critical aspects that still need to be refined will be presented. The paper will also demonstrate the capabilities of the diagnostic system for magnetic control as a basic element for control. In fact, the noisy environment (affecting primarily vertical stability), the non-axisymmetric elements in the machine structure (affecting the accuracy of the identification of the

  13. DGEMP-OE (2008) Energy Baseline Scenario. Synthesis report; Scenario energetique de reference DGEMP-OE(2008). Rapport de synthese

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-07-01

    A 'Business as usual' or 'Baseline' scenario of energy trends to 2020-2030 is produced by France every four years, as requested by the International Energy Agency in order to update the global scenarios published in its World Energy Outlook. Since the most recent scenario of this type was drawn up in 2003-2004, the time has come to renew the effort for the IEA's next in-depth review of French energy policy. Specifically, the DGEMP seeks to predict the future of France's energy situation assuming that no policies or new measures are taken affecting (i.e. improving or deteriorating) the situation other than those already in place or adopted as of 1 January 2008 (in other words, before measures such as those stemming from the Grenelle Environment Forum). On the other hand, it is assumed that change in the energy system is guided by 'conventional wisdom' according to which political options and behaviours by economic units are expected to be 'reasonable'. As a result, even should its projections prove inappropriate, this cannot be considered a 'worst-case' scenario. Indeed, beyond the IEA, this scenario can be used to establish an MEA (Multilateral Environment Agreement) scenario (based on existing measures) for national communications submitted under the U.N. Climate Convention. The scenarios by the 'Energy' Commission, part of the Centre d'Analyse Strategique (CAS), could have been used, particularly since the consultant who worked with the CAS to develop its scenarios was also commissioned by the DGEMP. However, several considerations argued in favour of proceeding separately: - The CAS scenarios drew on the DGEMP's 2004 baseline scenario, even though certain parameters were updated (in particular energy prices). - Moreover, the concept underpinning the DGEMP baseline scenario is that it should to every extent possible remain constant over time to secure continued consensus on this 'reference

  14. Remote maintenance development for ITER

    International Nuclear Information System (INIS)

    Tada, Eisuke; Shibanuma, Kiyoshi

    1997-01-01

    This paper both describes the overall design concept of the ITER remote maintenance system, which has been developed mainly for use with in-vessel components such as divertor and blanket, and outlines of the ITER R and D program, which has been established to develop remote handling equipment/tools and radiation hard components. In ITER, the reactor structures inside cryostat have to be maintained remotely because of activation due to DT operation. Therefore, remote-handling technology is fundamental, and the reactor-structure design must be made consistent with remote maintainability. The overall maintenance scenario and design concepts of the required remote handling equipment/tools have been developed according to their maintenance classification. Technologies are also being developed to verify the feasibility of the maintenance design and include fabrication and testing of a fullscale remote-handling equipment/tools for in-vessel maintenance. (author)

  15. Scenario evolution: Interaction between event tree construction and numerical analyses

    International Nuclear Information System (INIS)

    Barr, G.E.; Barnard, R.W.; Dockery, H.A.; Dunn, E.; MacIntyre, A.T.

    1991-01-01

    Construction of well-posed scenarios for the range of conditions possible at any proposed repository site is a critical first step to assessing total system performance. Even tree construction is the method that is being used to develop potential failure scenarios for the proposed nuclear waste repository at Yucca Mountain. An event tree begins with an initial event or condition. Subsequent events are listed in a sequence, leading eventually to release of radionuclides to the accessible environment. Ensuring the validity of the scenarios requires iteration between problems constructed using scenarios contained in the event tree sequence, experimental results, and numerical analyses. Details not adequately captured within the tree initially may become more apparent as a result of analyses. To illustrate this process, we discuss the iterations used to develop numerical analyses for PACE-90 using basaltic igneous activity and human-intrusion event trees

  16. Fusion radioactivity confinement and application to postulated ITER accidents

    International Nuclear Information System (INIS)

    Piet, S.J.; Brereton, S.J.

    1991-01-01

    An assessment of the ITER radioactivity confinement shows reduction of potential accidental releases to the environment by two orders of magnitude. Important credits are the 1% volume/day confinement leakage rate, radioactivity decay for short-lived isotopes, resumption of detritiation/negative pressure within seven days of the accident, and wind meander during the slow confinement leakage. Achieving this two order of magnitude credit in practice requires appropriate design details, especially the leakage rate and detritiation/negative pressure equipment, and research to validate some key assumptions. The confinement maximizes dependence on passive safety features, thereby working toward using fusion's potential safety advantages. The confinement includes several confinement zones with varying human access allowances. Some confinement areas are normally isolated from the environment, the closed ventilation zone. Some areas have an inert cover gas to inhibit combustion. If future assessments of accidental overpressure show the need, we propose a filter/vent system. This report documents our work for the ITER Conceptual Design Activity (CDA). The report is consistent with the final CDA design reports and descriptions, except that our analysis includes a filter/vent. For gaseous or vapor tritium and for most activated aerosols, the reference release fraction is about 2%. For short-lived tungsten-rhenium aerosols, the reference release fraction is somewhat lower, as low as 0.5% for some accident scenarios. Even without resumption of detritiation/decontamination or negative pressure within seven days of the accident, the release fraction for stays below 4%

  17. High voltage investigations for ITER coils

    International Nuclear Information System (INIS)

    Fink, S.; Fietz, W.H.

    2006-01-01

    The superconducting ITER magnets will be excited with high voltage during operation and fast discharge. Because the coils are complex systems the internal voltage distribution can differ to a large extent from the ideal linear voltage distribution. In case of fast excitations internal voltages between conductor and radial plate of a TF coil can be even higher than the terminal voltage of 3.5 kV to ground which appears during a fast discharge without a fault. Hence the determination of the transient voltage distribution is important for a proper insulation co-ordination and will provide a necessary basis for the verification of the individual insulation design and the choice of test voltages and waveforms. Especially the extent of internal overvoltages in case of failures, e. g. malfunction of discharge units and / or arcing is of special interest. Transient calculations for the ITER TF coil system have been performed for fast discharge and fault scenarios to define test voltages for ITER TF. The conductor and radial plate insulation of the ITER TF Model Coil were exposed at room temperature to test voltages derived from the results from these calculations. Breakdown appeared during the highest AC voltage step. A fault scenario for the TF fast discharge system is presented where one fault triggers a second fault, leading to considerable voltage stress. In addition a FEM model of Poloidal Field Coil 3 for the determination of the parameters of a detailed network model is presented in order to prepare detailed investigations of the transient voltage behaviour of the PF coils. (author)

  18. Assessment of post closure radioactive safety for the Korean reference disposal system: development of scenarios and quantitative assessment

    International Nuclear Information System (INIS)

    Kang, C. H.; Hwang, Y. S.; Lee, Y. M.

    2005-01-01

    The total system performance assessment (TSPA) on the Korean reference disposal system has been performed for different types of scenarios. Firstly two reference scenarios, the natural discharge and well ones are developed assessed. The natural discharge scenario assumes that a radionuclide is released from a waste container with an average lifetime of 1,000 years by intruding groundwater to a biosphere through a bentonite buffer and a natural barrier composed of a fractured porous rock and a major water conducting feature (MWCF). The well scenario describes that a radionuclide passing through a buffer enters a fractured rock which is intersected with a drinking well. Two scenarios are named as R1 and R2 respectively. The third scenario is for the initial waste container failure case. A waste container is apt to have initial defects during manufacturing and transportation to a deposition hole. The probability function of the ratio of waste container failure is assumed based on the engineering sense. The rest of waste containers are assumed to have full function of isolation of hazardous nuclides during the lifetime. This initial container failure scenario (ICF) has two different variations: one with a lifetime of 1,000 years ana the other with 10,000 years. Two variations are assessed for two different biosphere, natural discharge and well. The forth one is to assess the impact of excavation disturbed zones. Deposition tunnels are excavated by tunnel boring machine (TBM) or controlled blast (CB). The disturbed zone in assumed to be 30 cm and 1 meter for TBN and CB respectively. Six cases are developed for the EDZ scenarios considering all possible combination of changes in permeability a fracture aperture, and a porosity of a fractured rock. The fifth scenario stipulates the change of long term climate (LTC). The ice age assumed to be prevailed again after a few tens of thousand years. The advent of the ice age alters groundwater composition, pathways, and most

  19. DEMONSTRATION IN THE DIII-D TOKAMAK OF AN ALTERNATE BASELINE SCENARIO FOR ITER AND OTHER BURNING PLASMA EXPERIMENTS

    International Nuclear Information System (INIS)

    LUCE, T.C.; WADE, M.R.; FERRON, J.R.; HYATT, A.W.; KELLMAN, A.G.; KINSEY, J.E.; LAHAY, R.J.; LASNIER, C.J.; MURAKAMI, M.; POLITZER, P.A.; SCOVILLE, J.T.

    2002-01-01

    OAK A271 DEMONSTRATION IN THE DIII-D TOKAMAK OF AN ALTERNATE BASELINE SCENARIO FOR ITER AND OTHER BURNING PLASMA EXPERIMENTS. Discharges which can satisfy the high gain goals of burning plasma experiments have been demonstrated in the DIII-D tokamak in stationary conditions with relatively low plasma current (q 95 > 4). A figure of merit for fusion gain Β N H 89 /q 95 2 has been maintained at values corresponding to Q = 10 operation in a burning plasma for > 6 s or 36 τ E and 2 τ R . The key element is the relaxation of the current profile to a stationary state with q min > 1, which allows stable operation up to the no-wall ideal β limit. These plasmas maintain particle balance by active pumping rather than transient wall conditions. The reduced current lessens significantly the potential for structural damage in the event of a major disruption

  20. Scenario evolution: Interaction between event tree construction and numerical analyses

    International Nuclear Information System (INIS)

    Barr, G.E.; Barnard, R.W.; Dockery, H.A.; Dunn, E.; MacIntyre, A.T.

    1990-01-01

    Construction of well-posed scenarios for the range of conditions possible at any proposed repository site is a critical first step to assessing total system performance. Event tree construction is the method that is being used to develop potential failure scenarios for the proposed nuclear waste repository at Yucca Mountain. An event tree begins with an initial event or condition. Subsequent events are listed in a sequence, leading eventually to release of radionuclides to the accessible environment. Ensuring the validity of the scenarios requires iteration between problems constructed using scenarios contained in the event tree sequence, experimental results, and numerical analyses. Details not adequately captured within the tree initially may become more apparent as a result of analyses. To illustrate this process, the authors discuss the iterations used to develop numerical analyses for PACE-90 (Performance Assessment Calculational Exercises) using basaltic igneous activity and human-intrusion event trees

  1. Reference scenarios for deforestation and forest degradation in support of REDD: a review of data and methods

    International Nuclear Information System (INIS)

    Olander, Lydia P; Murray, Brian C; Gibbs, Holly K; Steininger, Marc; Swenson, Jennifer J

    2008-01-01

    Global climate policy initiatives are now being proposed to compensate tropical forest nations for reducing carbon emissions from deforestation and forest degradation (REDD). These proposals have the potential to include developing countries more actively in international greenhouse gas mitigation and to address a substantial share of the world's emissions which come from tropical deforestation. For such a policy to be viable it must have a credible benchmark against which emissions reduction can be calculated. This benchmark, sometimes termed a baseline or reference emissions scenario, can be based directly on historical emissions or can use historical emissions as input for business as usual projections. Here, we review existing data and methods that could be used to measure historical deforestation and forest degradation reference scenarios including FAO (Food and Agricultural Organization of the United Nations) national statistics and various remote sensing sources. The freely available and corrected global Landsat imagery for 1990, 2000 and soon to come for 2005 may be the best primary data source for most developing countries with other coarser resolution high frequency or radar data as a valuable complement for addressing problems with cloud cover and for distinguishing larger scale degradation. While sampling of imagery has been effectively useful for pan-tropical and continental estimates of deforestation, wall-to-wall (or full coverage) allows more detailed assessments for measuring national-level reference emissions. It is possible to measure historical deforestation with sufficient certainty for determining reference emissions, but there must be continued calls at the international level for making high-resolution imagery available, and for financial and technical assistance to help countries determine credible reference scenarios. The data available for past years may not be sufficient for assessing all forms of forest degradation, but new data sources

  2. Strong convergence of modified Ishikawa iterations for nonlinear ...

    Indian Academy of Sciences (India)

    interval [0, 1]. The second iteration process is referred to as Ishikawa's iteration process [11] which is .... Let E be a smooth Banach space with dual E∗ ..... and applications, in: Theory and Applications of Nonlinear Operators of Accretive and.

  3. 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

  4. Development of burning plasma and advanced scenarios in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Luce, T.C.

    2005-01-01

    Significant progress in the development of burning plasma scenarios, steady-state scenarios at high fusion performance, and basic tokamak physics has been made by the DIII-D Team. Discharges similar to the ITER baseline scenario have demonstrated normalized fusion performance nearly 50% higher than required for Q = 10 in ITER, under stationary conditions. Discharges that extrapolate to Q ∼ 10 for longer than one hour in ITER at reduced current have also been demonstrated in DIII-D under stationary conditions. Proof of high fusion performance with full noninductive operation has been obtained. Underlying this work are studies validating approaches to confinement extrapolation, disruption avoidance and mitigation, tritium retention, ELM avoidance, and operation above the no-wall pressure limit. In addition, the unique capabilities of the DIII-D facility have advanced studies of the sawtooth instability with unprecedented time and space resolution, threshold behavior in the electron heat transport, and rotation in plasmas in the absence of external torque. (author)

  5. Parametric assessments on hydrogenic species transport in CVD-diamond vacuum windows used in ITER ECRH

    Energy Technology Data Exchange (ETDEWEB)

    Moreno, C.; Sedano, L.A.; Fernandez, A. [EURATOM-CIEMAT Association, Madrid (Spain)

    2007-07-01

    Insulators used as H and CD and Diagnostic vacuum windows (VW) in ITER may become modified by surface intake and bulk transport of hydrogenic species. VW, operating under severe radiation levels, have a primary safety role as tritium confinement barriers. Ionizing radiation enhances the (H') uptake and release at surfaces and diffusion rates in the bulk. Radiation damage modifies the material's bulk trapped inventories by increasing steady state trapping centre concentrations. An experimental programme is ongoing at CIEMAT, to quantify radiation effects on H transport characteristics and also the possible impact on the VW. The reference material for ECRH VW is CVD diamond. As a parallel activity, parametric transport assessments are being made in order to obtain a wide evaluation of permeation fluxes, ranges, and soluted/trapped inventories in CVD diamond. Transport models have been developed based on extended capabilities of finite differences integrator tool TMAP7. Special attention is paid to radiation parameters defining inputs acting on transport magnitudes. These inputs have been analysed by using ionizing/damage radiation transport tools such as MCNPX/SRIM. VW operational scenarios are discussed with special attention being paid to the ITER design assumptions for the values of H-species source terms (neutrals and implanted) in the ECRH system. The available material transport database with and without radiation is discussed and taken as reference for this parametric exercise. Permeation fluxes through base materials are shown to be below DRG limits established for ITER. (orig.)

  6. The Fusion Advanced Studies Torus (FAST): a Proposal for an ITER Satellite Facility in Support of the Development of Fusion Energy

    International Nuclear Information System (INIS)

    Pizzuto, A.

    2008-01-01

    FAST is the conceptual design for a new machine proposed to support ITER experimental exploitation as well as to anticipate DEMO relevant physics and technology. FAST is aimed at integrated investigations of fast particle physics, plasma operations and plasma wall interaction in burning plasma relevant conditions. In Deuterium plasma operations, FAST has the capability to simultaneously approach relevant dimensionless physical parameters in all the ITER scenarios. The necessity of achieving ITER relevant power densities and performance with moderate cost has led to a compact Tokamak design (R=1.82 m , a= 0.64 m), with a high toroidal field (BT up to 8.5 T) and plasma current (Ip up to 8 MA). In order to study fast particle behaviours with dimensionless parameters similar to ITER, the project is based on a dominant Ion Cyclotron Resonance Heating system (ICRH; 30 MW coupled to the plasma). Moreover, the experiment foresees 6 MW of Lower Hybrid (LH), essentially for plasma control and for non-inductive current drive, and of Electron Cyclotron Resonance Heating (ECRH; 4MW) for localized electron heating and plasma control. Ports have been designed to also accommodate up to 10 MW of negative neutral beam injection (NNBI) in the energy range of 0.5-1 MeV. The total power input is in the 30-40 MW range in the different plasma scenarios, with a wall power load comparable with that of ITER (P/R∼22 MW/m). All ITER scenarios can be studied: starting from the reference H-mode, with plasma edge and ELMs characteristics similar to those of ITER (Q up to ∼ 2.5), and arriving to full non-inductive current drive scenarios lasting ∼ 160 s, Under these conditions, first wall as well as divertor plates will be made of tungsten. The divertor itself is designed to be completely removable by remote handling. This will allow studying, in view of DEMO, the behaviour of innovative divertor concepts, such as those foreseeing the use of liquid lithium. FAST is capable to operate with

  7. Overview of physics basis for ITER

    International Nuclear Information System (INIS)

    Mukhovatov, V; Shimada, M; Chudnovskiy, A N; Costley, A E; Gribov, Y; Federici, G; Kardaun, O; Kukushkin, A S; Polevoi, A; Pustovitov, V D; Shimomura, Y; Sugie, T; Sugihara, M; Vayakis, G

    2003-01-01

    ITER will be the first magnetic confinement device with burning DT plasma and fusion power of about 0.5 GW. Parameters of ITER plasma have been predicted using methodologies summarized in the ITER Physics Basis (1999 Nucl. Fusion 39 2175). During the past few years, new results have been obtained that substantiate confidence in achieving Q>=10 in ITER with inductive H-mode operation. These include achievement of a good H-mode confinement near the Greenwald density at high triangularity of the plasma cross section; improvements in theory-based confinement projections for the core plasma, even though further studies are needed for understanding the transport near the plasma edge; improvement in helium ash removal due to the elastic collisions of He atoms with D/T ions in the divertor predicted by modelling; demonstration of feedback control of neoclassical tearing modes and resultant improvement in the achievable beta values; better understanding of edge localized mode (ELM) physics and development of ELM mitigation techniques; and demonstration of mitigation of plasma disruptions. ITER will have a flexibility to operate also in steady-state and intermediate (hybrid) regimes. The 'advanced tokamak' regimes with weak or negative central magnetic shear and internal transport barriers are considered as potential scenarios for steady-state operation. The paper concentrates on inductively driven plasma performance and discusses requirements for steady-state operation in ITER

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

    Science.gov (United States)

    Covele, B.; Valanju, P.; Kotschenreuther, M.; Mahajan, S.

    2014-07-01

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

  9. IAEA activities related to safety indicators, time frames and reference scenarios

    International Nuclear Information System (INIS)

    Batandjieva, B.; Hioki, K.; Metcalf, P.

    2002-01-01

    which was completed last year. The project proposed a methodology for the development of reference biospheres which is in line with suggestions of the International Commission for Radiological Protection 81 (ICRP). Its application is illustrated by way of three example biospheres. At present the Agency is continuing the work on development of reference scenarios for waste disposal in order to assist in the development of international agreement on the main principles and concepts for development and use of reference scenarios for use in the development of safety cases and decision making related to waste repositories. (authors)

  10. Use of MCAM in creating 3D neutronics model for ITER building

    International Nuclear Information System (INIS)

    Zeng Qin; Wang Guozhong; Dang Tongqiang; Long Pengcheng; Loughlin, Michael

    2012-01-01

    Highlights: ► We created a 3D neutronics model of the ITER building. ► The model was produced from the engineering CAD model by MCAM software. ► The neutron flux map in the ITER building was calculated. - Abstract: The three dimensional (3D) neutronics reference model of International Thermonuclear Experimental Reactor (ITER) only defines the tokamak machine and extends to the bio-shield. In order to meet further 3D neutronics analysis needs, it is necessary to create a 3D reference model of the ITER building. Monte Carlo Automatic Modeling Program for Radiation Transport Simulation (MCAM) was developed as a computer aided design (CAD) based bi-directional interface program between general CAD systems and Monte Carlo radiation transport simulation codes. With the help of MCAM version 4.8, the 3D neutronics model of ITER building was created based on the engineering CAD model. The calculation of the neutron flux map in ITER building during operation showed the correctness and usability of the model. This model is the first detailed ITER building 3D neutronics model and it will be made available to all international organization collaborators as a reference model.

  11. Electromagnetic scattering using the iterative multi-region technique

    CERN Document Server

    Al Sharkawy, Mohamed H

    2007-01-01

    In this work, an iterative approach using the finite difference frequency domain method is presented to solve the problem of scattering from large-scale electromagnetic structures. The idea of the proposed iterative approach is to divide one computational domain into smaller subregions and solve each subregion separately. Then the subregion solutions are combined iteratively to obtain a solution for the complete domain. As a result, a considerable reduction in the computation time and memory is achieved. This procedure is referred to as the iterative multiregion (IMR) technique.Different enhan

  12. ITER final design report, cost review and safety analysis (FDR) and relevant documents

    International Nuclear Information System (INIS)

    1999-01-01

    This volume contains the fourth major milestone report and documents associated with its acceptance, review and approval. This ITER Final Design Report, Cost Review and Safety Analysis was presented to the ITER Council at its 13th meeting in February 1998 and was approved at its extraordinary meeting on 25 June 1998. 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 and schedule and cost estimates are given

  13. ITER parametric analysis and operational performance

    International Nuclear Information System (INIS)

    Perkins, L.J.; Spears, W.R.; Galambos, J.D.

    1991-01-01

    One of the key components of the ITER Conceptual Design Activities (CDA) is the determination of optimum design, investigation of operation in various modes, recommendation of baseline performance specifications, studies of sensitivity of ITER design to uncertainties in physics, investigation of operational flexibility, assessment of alternative designs, and determination of implications for extrapolation to prospective DEMO reactors. These terms of reference are reported in this document. Refs, figs and tabs

  14. 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.

  15. 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.

  16. Canada's energy future : reference case and scenarios to 2030

    International Nuclear Information System (INIS)

    2007-01-01

    Energy is essential to the comfort and economic prosperity of Canadians. This report highlighted some of the issues that Canada faces with respect to its energy future. The report focused on emerging trends in energy supply and demand, and examined various energy futures that may be available to Canadians up to the year 2030. Three different scenarios were presented: (1) a continuing trends scenario; (2) a triple E scenario in which economic, environmental and energy objectives are balanced; and (3) a fortified islands scenario in which security concerns were coupled with international unrest and protectionist governments. The report determined that energy demand will remain a function of population and economic growth. Automobiles will continue to rely on fossil fuels. Energy efficiency will improve in relation to the effectiveness of government policies, and a move towards natural gas alternatives will occur. However, fossil fuels will remain a dominant source of energy supply. Oil sands production grew in all 3 of the evaluated scenarios. It is expected that total natural gas production will decline and imports of liquefied natural gas (LNG) will increase. In all 3 scenarios greenhouse gas (GHG) emissions increased or only slightly declined. A full spectrum of GHG mitigation strategies will need to be implemented so that Canada can meet its target of a 20 per cent reduction in GHGs by 2020. It was concluded that effective policies are needed to optimize Canada's multiple objectives of economic growth, environment sustainability, and development of energy resources. 6 tabs., 118 figs

  17. Estimation of POL-iteration methods in fast running DNBR code

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyuk; Kim, S. J.; Seo, K. W.; Hwang, D. H. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, various root finding methods are applied to the POL-iteration module in SCOMS and POLiteration efficiency is compared with reference method. On the base of these results, optimum algorithm of POL iteration is selected. The POL requires the iteration until present local power reach limit power. The process to search the limiting power is equivalent with a root finding of nonlinear equation. POL iteration process involved in online monitoring system used a variant bisection method that is the most robust algorithm to find the root of nonlinear equation. The method including the interval accelerating factor and escaping routine out of ill-posed condition assured the robustness of SCOMS system. POL iteration module in SCOMS shall satisfy the requirement which is a minimum calculation time. For this requirement of calculation time, non-iterative algorithm, few channel model, simple steam table are implemented into SCOMS to improve the calculation time. MDNBR evaluation at a given operating condition requires the DNBR calculation at all axial locations. An increasing of POL-iteration number increased a calculation load of SCOMS significantly. Therefore, calculation efficiency of SCOMS is strongly dependent on the POL iteration number. In case study, the iterations of the methods have a superlinear convergence for finding limiting power but Brent method shows a quardratic convergence speed. These methods are effective and better than the reference bisection algorithm.

  18. Use of MCAM in creating 3D neutronics model for ITER building

    Energy Technology Data Exchange (ETDEWEB)

    Zeng Qin [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China); Wang Guozhong, E-mail: mango33@mail.ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China); Dang Tongqiang [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China); Long Pengcheng [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China); Loughlin, Michael [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lz-Durance (France)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer We created a 3D neutronics model of the ITER building. Black-Right-Pointing-Pointer The model was produced from the engineering CAD model by MCAM software. Black-Right-Pointing-Pointer The neutron flux map in the ITER building was calculated. - Abstract: The three dimensional (3D) neutronics reference model of International Thermonuclear Experimental Reactor (ITER) only defines the tokamak machine and extends to the bio-shield. In order to meet further 3D neutronics analysis needs, it is necessary to create a 3D reference model of the ITER building. Monte Carlo Automatic Modeling Program for Radiation Transport Simulation (MCAM) was developed as a computer aided design (CAD) based bi-directional interface program between general CAD systems and Monte Carlo radiation transport simulation codes. With the help of MCAM version 4.8, the 3D neutronics model of ITER building was created based on the engineering CAD model. The calculation of the neutron flux map in ITER building during operation showed the correctness and usability of the model. This model is the first detailed ITER building 3D neutronics model and it will be made available to all international organization collaborators as a reference model.

  19. MELCOR 1.8.2 Analyses in Support of ITER's RPrS

    International Nuclear Information System (INIS)

    Brad J Merrill

    2008-01-01

    The International Thermonuclear Experimental Reactor (ITER) Program is performing accident analyses for ITER's 'Rapport Preliminaire de Surete' (Report Preliminary on Safety - RPrS) with a modified version of the MELCOR 1.8.2 code. The RPrS is an ITER safety document required in the ITER licensing process to obtain a 'Decret Autorisation de Construction' (a Decree Authorizing Construction - DAC) for the ITER device. This report documents the accident analyses performed by the US with the MELCOR 1.8.2 code in support of the ITER RPrS effort. This work was funded through an ITER Task Agreement for MELCOR Quality Assurance and Safety Analyses. Under this agreement, the US was tasked with performing analyses for three accident scenarios in the ITER facility. Contained within the text of this report are discussions that identify the cause of these accidents, descriptions of how these accidents are likely to proceed, the method used to analyze the consequences of these accidents, and discussions of the transient thermal hydraulic and radiological release results for these accidents

  20. Disruptions in ITER and strategies for their control and mitigation

    Energy Technology Data Exchange (ETDEWEB)

    Lehnen, M., E-mail: michael.lehnen@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Aleynikova, K.; Aleynikov, P.B.; Campbell, D.J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Drewelow, P. [Max-Planck-Institut für Plasmaphysik, Greifswald branch, EURATOM Ass., D-17491 Greifswald (Germany); Eidietis, N.W. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Gasparyan, Yu. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoe sh. 31, Moscow 115409 (Russian Federation); Granetz, R.S. [MIT Plasma Science and Fusion Center, Cambridge, MA 02139 (United States); Gribov, Y. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Hartmann, N. [Forschungszentrum Jülich GmbH, Institute of Energy and Climate Research—Plasma Physics, Association EURATOM-FZJ, Trilateral Euregio Cluster, 52425 Jülich (Germany); Hollmann, E.M. [University of California-San Diego, La Jolla, CA 92093 (United States); Izzo, V.A. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Jachmich, S. [Laboratory for Plasma Physics, ERM/KMS, Association EURATOM – Belgian State, B-1000 Brussels (Belgium); Kim, S.-H.; Kočan, M. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Koslowski, H.R. [Forschungszentrum Jülich GmbH, Institute of Energy and Climate Research—Plasma Physics, Association EURATOM-FZJ, Trilateral Euregio Cluster, 52425 Jülich (Germany); Kovalenko, D. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Kruezi, U. [CCFE, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom); and others

    2015-08-15

    The thermal and electromagnetic loads related to disruptions in ITER are substantial and require careful design of tokamak components to ensure they reach the projected lifetime and to ensure that safety relevant components fulfil their function for the worst foreseen scenarios. The disruption load specifications are the basis for the design process of components like the full-W divertor, the blanket modules and the vacuum vessel and will set the boundary conditions for ITER operations. This paper will give a brief overview on the disruption loads and mitigation strategies for ITER and will discuss the physics basis which is continuously refined through the current disruption R&D programs.

  1. A Fast-Track Path to DEMO Enabled by ITER and FNSF-AT

    Energy Technology Data Exchange (ETDEWEB)

    Garofalo, A. M.; Choi, M.; Humphreys, D. A.; Kinsey, J. E.; Lao, L. L.; Snyder, P. B.; John, H. E.St.; Turnbull, A. D.; Taylor, T.S., E-mail: garofalo@fusion.gat.com [General Atomics, San Diego (United States); Chan, V. S.; Canik, J. M. [Oak Ridge National Laboratory, Oak Ridge (United States); Sawan, M. E. [University of Wisconsin, Madison (United States); Stangeby, P. C. [University of Toronto Institute for Aerospace Studies, Toronto (Canada)

    2012-09-15

    Full text: A Fusion Nuclear Science Facility based on the Advanced Tokamak concept (FNSF-AT) [1] is a key element of a fast track plan to a commercially attractive fusion DEMO. The next step forward on the path towards fusion commercialization must be a device that complements ITER in addressing the community identified science and technology gaps to DEMO, and that enables a DEMO construction decision triggered by the achievement of Q = 10 in ITER, presently scheduled for the year 2030. This paper elucidates the logic flow leading to the FNSF-AT approach for such a next step forward, and presents the results of recent analysis resolving key physics and engineering issues. A FNSF-AT will show fusion can make its own fuel, provide a materials irradiation facility, show fusion can produce high-grade process heat and electricity. In order to accomplish these goals, the FNSF has to operate steady-state with significant duty cycle and significant neutron fluence. In FNSF-AT, advanced tokamak physics enables steady-state burning plasmas with the high fluence required for FNSF's nuclear science development objective, in the compact size required to demonstrate Tritium fuel self-sufficiency using only a moderate quantity of the limited supply of Tritium. Physics based integrated modeling has found a steady-state baseline equilibrium with good stability and controllability properties. 2-D analysis assuming ITER heat and particle diffusion coefficients in the SOL predicts peak heat flux < 10 MW/m{sup 2} at the outer divertor targets. High fidelity and high-resolution 3D neutronics calculations have also been carried out, showing acceptable cumulative end-of-life organic insulator dose levels in all the device coils, and TBR > 1 for two blanket concepts considered. This FNSF-AT baseline plasma scenario has significant margin to meet the FNSF nuclear science mission. Moreover, the facility allows the development of more advanced scenarios to close the physics gaps to DEMO

  2. Activation and afterheat analyses for the HCPB test blanket module in ITER

    International Nuclear Information System (INIS)

    Pereslavtsev, P.; Fischer, U.

    2008-01-01

    To provide a sound data basis for the safety analyses of the HCPB TBM system in ITER, the afterheat and activity inventories were assessed making use of a code system that allows performing 3D activation calculations by linking the Monte Carlo transport code MCNP and the fusion inventory code FISPACT through an appropriate interface. A suitable MCNP model of a 20 deg. ITER torus sector with an integrated TBM of the HCPB PI (plant integration) type in the horizontal test blanket port was developed and adapted to the requirements for coupled 3D neutron transport and activation calculations. Two different irradiation scenarios were considered in the coupled 3D neutron transport and activation calculations. The first one is representative for the TBM irradiation in ITER with a total of 9000 neutron pulses over a 3 (calendar) years period. The second (conservative) irradiation scenario assumes an extended irradiation time over the full anticipated lifetime of ITER. The radioactivity inventories, the afterheat and the contact gamma dose were calculated as function of the decay time. Data were processed for the total activity, afterheat and contact dose rates of the TBM, its constituting components and materials

  3. Safety-factor profile tailoring by improved electron cyclotron system for sawtooth control and reverse shear scenarios in ITER

    International Nuclear Information System (INIS)

    Zucca, C.; Sauter, O.; Fable, E.; Henderson, M. A.; Polevoi, A.; Farina, D.; Ramponi, G.; Saibene, G.; Zohm, H.

    2008-01-01

    The effect of the predicted local electron cyclotron current driven by the optimized electron cyclotron system on ITER is discussed. A design variant was recently proposed to enlarge the physics program covered by the upper and equatorial launchers. By extending the functionality range of the upper launcher, significant control capabilities of the sawtooth period can be obtained. The upper launcher improvement still allows enough margin to exceed the requirements for neoclassical tearing mode stabilization, for which it was originally designed. The analysis of the sawtooth control is carried on with the ASTRA transport code, coupled with the threshold model by Por-celli, to study the control capabilities of the improved upper launcher on the sawtooth instability. The simulations take into account the significant stabilizing effect of the fusion alpha particles. The sawtooth period can be increased by a factor of 1.5 with co-ECCD outside the q = 1 surface, and decreased by at least 30% with co-ECCD inside q = 1. The present ITER base-line design has the electron cyclotron launchers providing only co-ECCD. The variant for the equatorial launcher proposes the possibility to drive counter-ECCD with 1 of the 3 rows of mirrors: the counter-ECCD can then be balanced with co-ECCD and provide pure ECH with no net driven current. The difference between full co-ECCD off-axis using all 20MW from the equatorial launcher and 20MW co-ECCD driven by 2/3 from the equatorial launcher and 1/3 from the upper launcher is shown to be negligible. Cnt-ECCD also offers greater control of the plasma current density, therefore this analysis addresses the performance of the equatorial launcher to control the central q profile. The equatorial launcher is shown to control very efficiently the value of q 0.2 -q min in advanced scenarios, if one row provides counter-ECCD.

  4. VDE/disruption EM analysis for ITER in-vessel components

    International Nuclear Information System (INIS)

    Miki, N.; Ioki, K.; Ilio, F.; Kodama, T.; Chiocchio, S.; Williamson, D.; Roccella, M.; Barabaschi, P.; Sayer, R.S.

    1998-01-01

    This paper summarises the results of EM analyses for ITER in-vessel components, such as blanket modules, backplate and divertor modules. In the ITER design the following two disruption scenarios are taken into account: centered or radial disruption, and vertical displacement event (VDE). Eddy currents and forces due to plasma disruption were calculated using the 3D shell element code EDDYCUFF and the 3D solid element code EMAS. The plasma motion and current decay used in the EM analysis was supplied by 2-D axisymmetric plasma equilibrium codes, TSC and MAXFEA. (authors)

  5. 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

  6. ITER site selection studies in Spain

    International Nuclear Information System (INIS)

    Medrano, M.; Alejaldre, C.; Doncel, J.; Garcia, A.; Ibarra, A.; Jimenez, J.A.; Sanchez de Mora, M.A.; Alcala, F.; Diez, J.E.; Dominguez, M.; Albisu, F.

    2003-01-01

    The studies carried out to evaluate and select a candidate site for International Thermonuclear Experimental Reactor (ITER) construction in Spain are presented in this paper. The ITER design, completed in July 2001, considered a number of technical requirements that must be fulfilled by the selected site. Several assumptions concerning the ITER site were made in order to carry on the design before final site selection. In the studies undertaken for ITER site selection in Spain, the referred technical requirements and assumptions were applied across the whole of Spain and two areas were identified as being preferential. These areas are on the Mediterranean coast and are situated in the Catalan and Valencian regions. A comparative evaluation based on technical characteristics for the concrete plots, proposed within the preferential areas, has been done. The result of these studies was the selection of a site that was deemed to be the most competitive--Vandellos (Tarragona)--and it was proposed to the European Commission for detailed studies in order to be considered as a possible European site for ITER construction. Another key factor for hosting ITER in Spain, is the licensing process. The present status is summarised in this paper

  7. ITER driver blanket, European Community design

    International Nuclear Information System (INIS)

    Simbolotti, G.; Zampaglione, V.; Ferrari, M.; Gallina, M.; Mazzone, G.; Nardi, C.; Petrizzi, L.; Rado, V.; Violante, V.; Daenner, W.; Lorenzetto, P.; Gierszewski, P.; Grattarola, M.; Rosatelli, F.; Secolo, F.; Zacchia, F.; Caira, M.; Sorabella, L.

    1993-01-01

    Depending on the final decision on the operation time of ITER (International Thermonuclear Experimental Reactor), the Driver Blanket might become a basic component of the machine with the main function of producing a significant fraction (close to 0.8) of the tritium required for the ITER operation, the remaining fraction being available from external supplies. The Driver Blanket is not required to provide reactor relevant performance in terms of tritium self-sufficiency. However, reactor relevant reliability and safety are mandatory requirements for this component in order not to significantly afftect the overall plant availability and to allow the ITER experimental program to be safely and successfully carried out. With the framework of the ITER Conceptual Design Activities (CDA, 1988-1990), a conceptual design of the ITER Driver Blanket has been carried out by ENEA Fusion Dept., in collaboration with ANSALDO S.p.A. and SRS S.r.l., and in close consultation with the NET Team and CFFTP (Canadian Fusion Fuels Technology Project). Such a design has been selected as EC (European Community) reference design for the ITER Driver Blanket. The status of the design at the end of CDA is reported in the present paper. (orig.)

  8. 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.

  9. Lower hybrid current drive at ITER-relevant high plasma densities

    International Nuclear Information System (INIS)

    Cesario, R.; Amicucci, L.; Cardinali, A.; Castaldo, C.; Marinucci, M.; Panaccione, L.; Pericoli-Ridolfini, V.; Tuccillo, A. A.; Tudisco, O.; Calabro, G.

    2009-01-01

    Recent experiments indicated that a further non-inductive current, besides bootstrap, should be necessary for developing advanced scenario for ITER. The lower hybrid current drive (LHCD) should provide such tool, but its effectiveness was still not proved in operations with ITER-relevant density of the plasma column periphery. Progress of the LH deposition modelling is presented, performed considering the wave physics of the edge, and different ITER-relevant edge parameters. Operations with relatively high edge electron temperatures are expected to reduce the LH || spectral broadening and, consequently, enabling the LH power to propagate also in high density plasmas ( || is the wavenumber component aligned to the confinement magnetic field). New results of FTU experiments are presented, performed by following the aforementioned modeling: they indicate that, for the first time, the LHCD conditions are established by operating at ITER-relevant high edge densities.

  10. 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.)

  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. Biomass Scenario Model Scenario Library: Definitions, Construction, and Description

    Energy Technology Data Exchange (ETDEWEB)

    Inman, D.; Vimmerstedt, L.; Bush, B.; Peterson, S.

    2014-04-01

    Understanding the development of the biofuels industry in the United States is important to policymakers and industry. The Biomass Scenario Model (BSM) is a system dynamics model of the biomass-to-biofuels system that can be used to explore policy effects on biofuels development. Because of the complexity of the model, as well as the wide range of possible future conditions that affect biofuels industry development, we have not developed a single reference case but instead developed a set of specific scenarios that provide various contexts for our analyses. The purpose of this report is to describe the scenarios that comprise the BSM scenario library. At present, we have the following policy-focused scenarios in our library: minimal policies, ethanol-focused policies, equal access to policies, output-focused policies, technological diversity focused, and the point-of-production- focused. This report describes each scenario, its policy settings, and general insights gained through use of the scenarios in analytic studies.

  13. The ITER EC H and CD upper launcher: EM disruption analyses

    International Nuclear Information System (INIS)

    Vaccaro, A.; Aiello, G.; Grossetti, G.; Meier, A.; Scherer, T.A.; Schreck, S.; Späh, P.; Strauß, D.; Saibene, G.; Cavinato, M.

    2013-01-01

    In the frame of the new grant started in November 2011 between Fusion for Energy (F4E) and the ECHUL-CA consortium, the development process of the Electron Cyclotron Heating and Current Drive (EC H and CD) upper launcher (UL) in ITER has moved a step toward the final design phase. Based on the 2009 preliminary design review version, the new configuration of the UL now features a thicker single-wall mainframe (up to 90 mm), a recessed first wall panel (100 mm, to reduce the impact of halo currents) and a new arrangement of the internal shield blocks. The main design drivers for the structural components are still the electromagnetic (EM) loads, which need to be reassessed for the new configuration of the UL. In this paper the results of a new EM 20° sector model of ITER, specialized for the UL, are shown. Six different disruption scenarios are considered in this work: upward linear (36 ms) and exponential (36 ms) vertical displacement events (VDE), upward linear (36 ms) and exponential (16 ms) major disruptions (MD), category II upward slow and slow–fast VDEs. Comparing the analyses’ results allowed to define a set of structural loads to be used as a reference for the forthcoming structural calculations

  14. RF modeling of the ITER-relevant lower hybrid antenna

    International Nuclear Information System (INIS)

    Hillairet, J.; Ceccuzzi, S.; Belo, J.; Marfisi, L.; Artaud, J.F.; Bae, Y.S.; Berger-By, G.; Bernard, J.M.; Cara, Ph.; Cardinali, A.; Castaldo, C.; Cesario, R.; Decker, J.; Delpech, L.; Ekedahl, A.; Garcia, J.; Garibaldi, P.; Goniche, M.; Guilhem, D.; Hoang, G.T.

    2011-01-01

    In the frame of the EFDA task HCD-08-03-01, a 5 GHz Lower Hybrid system which should be able to deliver 20 MW CW on ITER and sustain the expected high heat fluxes has been reviewed. The design and overall dimensions of the key RF elements of the launcher and its subsystem has been updated from the 2001 design in collaboration with ITER organization. Modeling of the LH wave propagation and absorption into the plasma shows that the optimal parallel index must be chosen between 1.9 and 2.0 for the ITER steady-state scenario. The present study has been made with n || = 2.0 but can be adapted for n || = 1.9. Individual components have been studied separately giving confidence on the global RF design of the whole antenna.

  15. Feedback and rotational stabilization of resistive wall modes in ITER

    International Nuclear Information System (INIS)

    Liu Yueqiang; Bondeson, A.; Chu, M.S.; La Haye, R.J.; Favez, J.-Y.; Lister, J.B.; Gribov, Y.; Gryaznevich, M.; Hender, T.C.; Howell, D.F.

    2005-01-01

    Different models have been introduced in the stability code MARS-F in order to study the damping effect of resistive wall modes (RWM) in rotating plasmas. Benchmark of MARS-F calculations with RWM experiments on JET and D3D indicates that the semi-kinetic damping model is a good candidate for explaining the damping mechanisms. Based on these results, the critical rotation speeds required for RWM stabilization in an advanced ITER scenario are predicted. Active feedback control of the n = 1 RWM in ITER is also studied using the MARS-F code. (author)

  16. Electron cyclotron power management for control of neoclassical tearing modes in the ITER baseline scenario

    Science.gov (United States)

    Poli, F. M.; Fredrickson, E. D.; Henderson, M. A.; Kim, S.-H.; Bertelli, N.; Poli, E.; Farina, D.; Figini, L.

    2018-01-01

    Time-dependent simulations are used to evolve plasma discharges in combination with a modified Rutherford equation for calculation of neoclassical tearing mode (NTM) stability in response to electron cyclotron (EC) feedback control in ITER. The main application of this integrated approach is to support the development of control algorithms by analyzing the plasma response with physics-based models and to assess how uncertainties in the detection of the magnetic island and in the EC alignment affect the ability of the ITER EC system to fulfill its purpose. Simulations indicate that it is critical to detect the island as soon as possible, before its size exceeds the EC deposition width, and that maintaining alignment with the rational surface within half of the EC deposition width is needed for stabilization and suppression of the modes, especially in the case of modes with helicity (2, 1) . A broadening of the deposition profile, for example due to wave scattering by turbulence fluctuations or not well aligned beams, could even be favorable in the case of the (2, 1)- NTM, by relaxing an over-focussing of the EC beam and improving the stabilization at the mode onset. Pre-emptive control reduces the power needed for suppression and stabilization in the ITER baseline discharge to a maximum of 5 MW, which should be reserved and available to the upper launcher during the entire flattop phase. Assuming continuous triggering of NTMs, with pre-emptive control ITER would be still able to demonstrate a fusion gain of Q=10 .

  17. Analysis and validation center for ITER RH maintenance scenarios in a virtual environment

    NARCIS (Netherlands)

    Elzendoorn, B. S. Q.; M.R. de Baar,; Hamilton, D.; Heemskerk, C. J. M.; Koning, J. F.; Ronden, D. M. S.

    2012-01-01

    A facility for detailed simulation of maintenance processes in the ITER Hot Cell Facility (HCF) has been taken into operation. The facility mimics the Remote Handling (RH) work-cells as are presently foreseen. Novel virtual reality (VR) technology, extended with a physics engine is used to create a

  18. Milestones in the Development of Iterative Solution Methods

    Directory of Open Access Journals (Sweden)

    Owe Axelsson

    2010-01-01

    Full Text Available Iterative solution methods to solve linear systems of equations were originally formulated as basic iteration methods of defect-correction type, commonly referred to as Richardson's iteration method. These methods developed further into various versions of splitting methods, including the successive overrelaxation (SOR method. Later, immensely important developments included convergence acceleration methods, such as the Chebyshev and conjugate gradient iteration methods and preconditioning methods of various forms. A major strive has been to find methods with a total computational complexity of optimal order, that is, proportional to the degrees of freedom involved in the equation. Methods that have turned out to have been particularly important for the further developments of linear equation solvers are surveyed. Some of them are presented in greater detail.

  19. Thermal effects of divertor sweeping in ITER

    International Nuclear Information System (INIS)

    Wesley, J.C.

    1992-01-01

    In this paper, thermal effects of magnetically sweeping the separatrix strike point on the outer divertor target of the International Thermonuclear Fusion Reactor (ITER) are calculated. For the 0. 2 Hz x ± 12 cm sweep scenario proposed for ITER operations, the thermal capability of a generic target design is found to be slightly inadequate (by ∼ 5%) to accommodate the full degree of plasma scrape-off peaking postulated as a design basis. The principal problem identified is that the 5 s sweep period is long relative to the 1. 4 s thermal time constant of the divertor target. An increase of the sweep frequency to ∼ 1 Hz is suggested: this increase would provide a power handling margin of ∼ 25% relative to present operational criteria

  20. 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

  1. 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

  2. USING UML SCENARIOS IN B2B SYSTEMS

    Directory of Open Access Journals (Sweden)

    A. Jakimi

    2010-05-01

    Full Text Available Scenarios has become a popular technique for requirements elicitation and specification building. Since scenarios capture only partial descriptions of the system behavior, an approach for scenario composition and/or integration is needed to produce more complete specifications. The Unified Modeling Language (UML, which has become a standard notation for object-oriented modeling, provides a suitable framework for scenario acquisition using Use Case diagrams and Sequence or Collaboration diagrams. In this paper, we suggest an algorithmic and tool support for composing and integrating scenarios that are represented in form of sequence diagrams. We suggest four operators (;: sequential operator, ||: concurrent operator, ?: conditional operator and  * :iteration operator to compose a set of scenarios that describe a use case of a given system. In this paper, we suggest also to apply the scenario approach to B2B systems (Business to Business. We propose to develop B2B systems as a three activities process deriving formal specifications and code skeletons from UML scenarios. Activities of this proposed process are generally automatic and are supported by a set of developed algorithms and tools.

  3. Active beam spectroscopy for ITER

    International Nuclear Information System (INIS)

    Von Hellermann, M.; Giroud, C.; Jaspers, R.; Hawkes, N.C.; Mullane, M.O.; Zastrow, K.D.; Krasilnikov, A.; Tugarinov, S.; Lotte, P.; Malaquias, A.; Rachlew, E.

    2003-01-01

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

  4. Carbon-constrained scenarios. Final report

    International Nuclear Information System (INIS)

    2009-05-01

    This report provides the results of the study entitled 'Carbon-Constrained Scenarios' that was funded by FONDDRI from 2004 to 2008. The study was achieved in four steps: (i) Investigating the stakes of a strong carbon constraint for the industries participating in the study, not only looking at the internal decarbonization potential of each industry but also exploring the potential shifts of the demand for industrial products. (ii) Developing an hybrid modelling platform based on a tight dialog between the sectoral energy model POLES and the macro-economic model IMACLIM-R, in order to achieve a consistent assessment of the consequences of an economy-wide carbon constraint on energy-intensive industrial sectors, while taking into account technical constraints, barriers to the deployment of new technologies and general economic equilibrium effects. (iii) Producing several scenarios up to 2050 with different sets of hypotheses concerning the driving factors for emissions - in particular the development styles. (iv) Establishing an iterative dialog between researchers and industry representatives on the results of the scenarios so as to improve them, but also to facilitate the understanding and the appropriate use of these results by the industrial partners. This report provides the results of the different scenarios computed in the course of the project. It is a partial synthesis of the work that has been accomplished and of the numerous exchanges that this study has induced between modellers and stakeholders. The first part was written in April 2007 and describes the first reference scenario and the first mitigation scenario designed to achieve stabilization at 450 ppm CO 2 at the end of the 21. century. This scenario has been called 'mimetic' because it has been build on the assumption that the ambitious climate policy would coexist with a progressive convergence of development paths toward the current paradigm of industrialized countries: urban sprawl, general

  5. Emissions in 2001 conform the reference scenario (GE WLO with high oil price) and including Clean and Efficient

    International Nuclear Information System (INIS)

    Kroon, P.; Menkveld, M.

    2008-08-01

    This memo shows the calculation of an estimate for the total greenhouse gas emissions in 2011 in the reference scenario (GE WLO is the Dutch abbreviation for Global Economy and Welfare and Environment), including the impact of the Clean and Efficient programme from the assessment of ECN (Energy research Centre of the Netherlands) and MNP (Netherlands Environmental Assessment Agency) [nl

  6. 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.

  7. Simulation of LLCB TBM in-vessel first wall coolant break into ITER vacuum vessel by using RELAP/MOD3.4

    International Nuclear Information System (INIS)

    Tony Sandeep, K.; Chaudhari, Vilas; Rajendra Kumar, E.; Dutta, Anu; Singh, R.K.

    2013-06-01

    To prove Test Blanket Module (TBM) safety in International Thermonuclear Experimental Reactor (ITER), various accident scenarios are postulated . One of the postulated initiating events to be analysed is TBM First wall (FW) coolant leak in ITER Vacuum vessel (VV). This accident has been classified as a reference event for the TBM (probability of occurrence >1 E -06 /a). The postulated accident occurs as a result of small leak of TBM FW helium into ITER vacuum vessel (VV), caused by the TBM weld failure. The ingress of this TBM FW helium into ITER plasma induces intense plasma disruption that deposits 1.8 MJ/m 2 of plasma stored thermal energy onto the TBM FW over a period of 1 sec in duration (assumption). Runaway electrons in this process are lost from plasma current channel and cause multiple TBM and ITER FW cooling tube failures within 10 cm torriodal strip. The size of the break is identified as double ended rupture of all coolant channels within this strip around the reactor. For LLCB TBM this represents failure of 4 FW channels. The size of ITER FW break is 0.02 m 2 . Consequently, a simultaneous blow down of TBM FW helium and ITER FW water occurs, injecting helium and water into VV. This pressurisation causes the activation of VV pressure suppressions system and ingress of water into VV. This pressurisation causes the VV pressure suppressions system (VVPSS) to open in an attempt to contain the pressure below the safety limit of 0.2 MPa. This report is intended to do the above accident analysis and assessment of active components of TBM using RELAP code and hence prove its safety in ITER environment. (author)

  8. Isotope and fast ions turbulence suppression effects: Consequences for high-β ITER plasmas

    Science.gov (United States)

    Garcia, J.; Görler, T.; Jenko, F.

    2018-05-01

    The impact of isotope effects and fast ions on microturbulence is analyzed by means of non-linear gyrokinetic simulations for an ITER hybrid scenario at high beta obtained from previous integrated modelling simulations with simplified assumptions. Simulations show that ITER might work very close to threshold, and in these conditions, significant turbulence suppression is found from DD to DT plasmas. Electromagnetic effects are shown to play an important role in the onset of this isotope effect. Additionally, even external ExB flow shear, which is expected to be low in ITER, has a stronger impact on DT than on DD. The fast ions generated by fusion reactions can additionally reduce turbulence even more although the impact in ITER seems weaker than in present-day tokamaks.

  9. Initial report of the ITER Council's Ways and Means Working Party

    International Nuclear Information System (INIS)

    1990-01-01

    Under the auspices of the IAEA, the ITER Conceptual Design Activities began in April, 1988, and are scheduled to be completed in December, 1990. As part of the Terms of Reference of the ITER Conceptual Design Activities, the ITER Council chartered a Working Party which plans to define needed elements to conduct an Engineering Design of ITER, should the members wish to go on with such design. This document is an initial report of the Working Party on Ways and Means; it seeks to define the Engineering Design Activities, and to list the practical topics to be considered by the ITER EDA. 2 figs, 4 tabs

  10. A new approach of equilibrium reconstruction for ITER

    International Nuclear Information System (INIS)

    Imazawa, R.; Kawano, Y.; Kusama, Y.

    2011-01-01

    We have proposed a new approach for equilibrium reconstruction that can be applied to ITER-like burning plasmas. In this study, we have focused on carrying out equilibrium reconstruction using polarimetry, which is feasible for ITER-like burning plasmas. Polarimetry in burning plasmas is different from that in the existing tokamaks in two regards: (1) increased importance of the relativistic effects and (2) significant coupling with the Faraday and Cotton–Mouton effects. We found that when polarimetric data (orientation angle, θ, and ellipticity angle, ε, of a polarization state) are used as the constraints in the equilibrium reconstruction, the optimum weighting factors for θ and ε depend on the magnetic surfaces through which the viewing chord of polarimetry passes. We applied our approach to the operation scenarios II (S2) and IV (S4) in ITER. In the case where the viewing chords are via both the equatorial and upper ports, the measurement requirements for the accuracy of the q-profile in ITER (±10%) were satisfied in S2 and S4 when the measuring errors of θ and ε were less than 0.5° and 3°, respectively.

  11. A comparison between the example reference biosphere model ERB 2B and a process-based model: simulation of a natural release scenario.

    Science.gov (United States)

    Almahayni, T

    2014-12-01

    The BIOMASS methodology was developed with the objective of constructing defensible assessment biospheres for assessing potential radiological impacts of radioactive waste repositories. To this end, a set of Example Reference Biospheres were developed to demonstrate the use of the methodology and to provide an international point of reference. In this paper, the performance of the Example Reference Biosphere model ERB 2B associated with the natural release scenario, discharge of contaminated groundwater to the surface environment, was evaluated by comparing its long-term projections of radionuclide dynamics and distribution in a soil-plant system to those of a process-based, transient advection-dispersion model (AD). The models were parametrised with data characteristic of a typical rainfed winter wheat crop grown on a sandy loam soil under temperate climate conditions. Three safety-relevant radionuclides, (99)Tc, (129)I and (237)Np with different degree of sorption were selected for the study. Although the models were driven by the same hydraulic (soil moisture content and water fluxes) and radiological (Kds) input data, their projections were remarkably different. On one hand, both models were able to capture short and long-term variation in activity concentration in the subsoil compartment. On the other hand, the Reference Biosphere model did not project any radionuclide accumulation in the topsoil and crop compartments. This behaviour would underestimate the radiological exposure under natural release scenarios. The results highlight the potential role deep roots play in soil-to-plant transfer under a natural release scenario where radionuclides are released into the subsoil. When considering the relative activity and root depth profiles within the soil column, much of the radioactivity was taken up into the crop from the subsoil compartment. Further improvements were suggested to address the limitations of the Reference Biosphere model presented in this paper

  12. Progress in LHCD: a tool for advanced regimes on ITER

    Energy Technology Data Exchange (ETDEWEB)

    Tuccillo, A. A.

    2005-07-01

    Since the early 80s, Lower Hybrid (LH) waves have driven plasma current non-inductively in tokamak experiments with efficiency far higher than other auxiliary systems, especially at low plasma temperature. The latter aspect makes LH the natural candidate for off axis current drive (CD) in ITER where current profile control will be required to maintain burning performance on a long time scale. The difficulties in coupling LH waves in high performance H-mode plasmas led in the 90s to the decision, by the ITER JCT, of not supporting directly the development of a LH system, with subsequent exclusion of LH from ITER initial Hn and CD systems. Nevertheless, the LH activity continued in all areas: theory, experiments, modelling and R and D. Recently, in JET, the problem of coupling LH waves in severe edge conditions (ELMy plasmas and very low SOL densities) has been solved by localised gas injection. LH has then boosted the research on advanced scenarios allowing a variety of current profiles, from peaked to deeply hollow, to be used in generating Internal Transport Barriers (ITBs). In JET, long ITBs (11s) have been sustained for times close to the resistive time scale in full CD conditions, additionally pulses lasting minutes, relying on LHCD, are routinely obtained in Tore Supra. Advanced scenarios have also been obtained in conditions of dominant electron heating (FTU, Tore Supra) and efficient CD has been demonstrated at ITER relevant densities on FTU. In these experiments, an improvement of the Electron Cyclotron (EC) CD efficiency (up to 4?) due to synergy with LH is also observed both in normal operation and with EC resonating at a down shifted frequency. Modelling all these experiments has been a good test bench for LH codes and has increased the confidence in their use to predict future experiments. The use of LH as an actuator for Real Time Control of the plasma current profile has become a powerful tool to optimise ITB dynamics in JET and JT-60U. An LH system

  13. Verifying elementary ITER maintenance actions with the MS2 benchmark product

    NARCIS (Netherlands)

    Heemskerk, C. J. M.; Elzendoorn, B. S. Q.; Magielsen, A. J.; Schropp, G. Y. R.

    2011-01-01

    A new facility has been taken in operation to investigate the influence of visual and haptic feedback on the performance of remotely executed ITER RH maintenance tasks. A reference set of representative ITER remote handling maintenance tasks was included the master slave manipulator system (MS2)

  14. 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

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

    Directory of Open Access Journals (Sweden)

    Pinsker R.I.

    2015-01-01

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

  16. The First Benchmarking of ITER BR Nb3Sn Strand of CNDA

    International Nuclear Information System (INIS)

    Long Feng; Liu Fang; Wu Yu; Ni Zhipeng

    2012-01-01

    According to the International Thermonuclear Experimental Reactor (ITER) Procurement Arrangement (PA) of Cable-In-Conduit Conductor (CICC) unit lengths for the Toroidal Field (TF) and Poloidal Field (PF) magnet systems of ITER, at the start of process qualification, the Domestic Agency (DA) shall be required to conduct a benchmarking of the room and low temperature acceptance tests carried out at the Strand Suppliers and/or at its Reference Laboratories designated by the ITER Organization (IO). The first benchmarking was carried out successfully in 2009. Nineteen participants from six DAs (China, European Union, Japan, South Korea, Russia, and the United States) participated in the first benchmarking. Bronze-route (BR) Nb 3 Sn strand and samples prepared by the ITER reference lab (CERN) were sent out to each participant by CERN. In this paper, the test facility and test results of the first benchmarking by the Chinese DA (CNDA) are presented.

  17. Joining technologies for the plasma facing components of ITER

    International Nuclear Information System (INIS)

    Barabash, V.; Kalinin, G.; Matera, R.

    1998-01-01

    An extensive R and D program on the development of the joining technologies between armour (beryllium, tungsten and carbon fibre composites)/copper alloys heat sink and copper alloys/ stainless steel has been carried out by ITER Home Teams. A brief review of this R and D program is presented in this paper. Based on the results, reference technologies for use in ITER have been selected and recommended for further development. (author)

  18. Scenario development for the Waste Isolation Pilot Plant: Building confidence in the assessment

    International Nuclear Information System (INIS)

    Galson, D.A.; Swift, P.N.

    1994-01-01

    Scenario developments is part of the iterative performance assessment (PA) process for the Waste Isolation Pilot Plant (WIPP). Scenario development for the WIPP has been the subject of intense external review, and is certain to be the subject of continued scrutiny as the project proceeds toward regulatory compliance. The principal means of increasing confidence is this aspect of the PA will be through the use of a systematic and thorough procedure toward developing the scenarios and conceptual models on which the assessment is to be based. Early and ongoing interaction with project reviewers can assist with confidence building. Quality of argument and clarity of presentation in PA will be of key concern. Appropriate tools are required for documenting and tracking assumptions, through a single assessment phase, and between iterative assessment phases. Risks associated with future human actions are of particular concern to the WIPP project, and international consensus on the principles for incorporation of future human actions in assessments would be valuable

  19. Scenario development for the Waste Isolation Pilot Plant: Building confidence in the assessment

    International Nuclear Information System (INIS)

    Galson, D.A.; Swift, P.N.

    1994-07-01

    Scenario development is part of the iterative performance assessment (PA) process for the Waste Isolation Pilot Plant (WIPP). Scenario development for the WIPP has been the subject of intense external review and is certain to be the subject of continued scrutiny as the project proceeds toward regulatory compliance. The principal means of increasing confidence in this aspect of the PA will be through the use of the systematic and thorough procedure toward developing the scenarios and conceptual models on which the assessment is to be based. Early and ongoing interaction with project reviewers can assist with confidence building. Quality of argument and clarity of presentation in PA will be of key concern. Appropriate tools are required for documenting and tracking assumptions, through a single assessment phase, and between iterative assessment phases. Risks associated with future human actions are of particular concern to the WIPP project, and international consensus on the principles for incorporation of future human actions in assessments would be valuable

  20. RAMI analysis of the ITER Central Safety System

    Energy Technology Data Exchange (ETDEWEB)

    Kitazawa, Sin-iti, E-mail: kitazawa.siniti@jaea.go.jp [ITER Project Unit, Japan Atomic Energy Agency (JAEA), Naka, 311-0193 Ibaraki (Japan); Okayama, Katsumi [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Neyatani, Yuzuru [ITER Project Unit, Japan Atomic Energy Agency (JAEA), Naka, 311-0193 Ibaraki (Japan); Sagot, Francois; Houtte, Didier van [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2014-06-15

    Highlights: • We performed the functional analysis of the ITER CSS. • We performed a failure mode analysis of the ITER CSS. • We estimated the reliability and availability of the ITER CSS. • The ITER RAMI approach was applied to the ITER CSS for technical risk control in the design phase. - Abstract: ITER is the first worldwide international project aiming to design a facility to produce nuclear fusion energy. The technical requirements of its plant systems have been established in the ITER Project Baseline. In the project, the Reliability, Availability, Maintainability and Inspectability (RAMI) approach has been adopted for technical risk control to help aid the design of the components in preparation for operation and maintenance. A RAMI analysis was performed on the conceptual design of the ITER Central Safety System (CSS). A functional breakdown was prepared in a bottom-up approach, resulting in the system being divided into 2 main functions and 20 sub-functions. These functions were described using the IDEF0 method. Reliability block diagrams were prepared to estimate the reliability and availability of each function under the stipulated operating conditions. Initial and expected scenarios were analyzed to define risk-mitigation actions. The inherent availability of the ITER CSS expected after implementation of mitigation actions was calculated to be 99.80% over 2 years, which is the typical interval of the scheduled maintenance cycles. This is consistent with the project required value of 99.9 ± 0.1%. A Failure Modes, Effects and Criticality Analysis was performed with criticality charts highlighting the risk level of the different failure modes with regard to their probability of occurrence and their effects on the availability of the plasma operation. This analysis defined when risk mitigation actions were required in terms of design, testing, operation procedures and/or maintenance to reduce the risk levels and increase the availability of the

  1. Iterated Gate Teleportation and Blind Quantum Computation.

    Science.gov (United States)

    Pérez-Delgado, Carlos A; Fitzsimons, Joseph F

    2015-06-05

    Blind quantum computation allows a user to delegate a computation to an untrusted server while keeping the computation hidden. A number of recent works have sought to establish bounds on the communication requirements necessary to implement blind computation, and a bound based on the no-programming theorem of Nielsen and Chuang has emerged as a natural limiting factor. Here we show that this constraint only holds in limited scenarios, and show how to overcome it using a novel method of iterated gate teleportations. This technique enables drastic reductions in the communication required for distributed quantum protocols, extending beyond the blind computation setting. Applied to blind quantum computation, this technique offers significant efficiency improvements, and in some scenarios offers an exponential reduction in communication requirements.

  2. Decision support systems in nuclear emergencies: A scenario-based comparison of domestic and reference tools

    International Nuclear Information System (INIS)

    Vamanu, D.; Slavnicu, S. D.; Slavnicu, E.; Vamanu, B.

    2004-01-01

    The article reports selective results of a comparison between RODOS - an emerging decision support system for the management of nuclear emergencies in Europe developed by an international research consortium under EEC aegis, and a resident software package developed and maintained for similar purposes at IFIN-HH, Bucharest. Reproducible similarity patterns obtained in the output data distributions provide for simple normalising procedures that may ensure convergent radiological assessments. When properly consolidated on a sufficient scenario casuistry, such procedures could lend a certain resilience to domestic decision support tools over the interim lead time required by the full implementation of RODOS, or other major league, internationally accepted reference systems. (authors)

  3. Smart-DS: Synthetic Models for Advanced, Realistic Testing: Distribution Systems and Scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Krishnan, Venkat K [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Palmintier, Bryan S [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Hodge, Brian S [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Hale, Elaine T [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Elgindy, Tarek [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Bugbee, Bruce [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Rossol, Michael N [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Lopez, Anthony J [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Krishnamurthy, Dheepak [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Vergara, Claudio [MIT; Domingo, Carlos Mateo [IIT Comillas; Postigo, Fernando [IIT Comillas; de Cuadra, Fernando [IIT Comillas; Gomez, Tomas [IIT Comillas; Duenas, Pablo [MIT; Luke, Max [MIT; Li, Vivian [MIT; Vinoth, Mohan [GE Grid Solutions; Kadankodu, Sree [GE Grid Solutions

    2017-08-09

    The National Renewable Energy Laboratory (NREL) in collaboration with Massachusetts Institute of Technology (MIT), Universidad Pontificia Comillas (Comillas-IIT, Spain) and GE Grid Solutions, is working on an ARPA-E GRID DATA project, titled Smart-DS, to create: 1) High-quality, realistic, synthetic distribution network models, and 2) Advanced tools for automated scenario generation based on high-resolution weather data and generation growth projections. Through these advancements, the Smart-DS project is envisioned to accelerate the development, testing, and adoption of advanced algorithms, approaches, and technologies for sustainable and resilient electric power systems, especially in the realm of U.S. distribution systems. This talk will present the goals and overall approach of the Smart-DS project, including the process of creating the synthetic distribution datasets using reference network model (RNM) and the comprehensive validation process to ensure network realism, feasibility, and applicability to advanced use cases. The talk will provide demonstrations of early versions of synthetic models, along with the lessons learnt from expert engagements to enhance future iterations. Finally, the scenario generation framework, its development plans, and co-ordination with GRID DATA repository teams to house these datasets for public access will also be discussed.

  4. Joint DIII-D/EAST research on the development of a high poloidal beta scenario for the steady state missions of ITER and CFETR

    Science.gov (United States)

    Garofalo, A. M.; Gong, X. Z.; Ding, S. Y.; Huang, J.; McClenaghan, J.; Pan, C. K.; Qian, J.; Ren, Q. L.; Staebler, G. M.; Chen, J.; Cui, L.; Grierson, B. A.; Hanson, J. M.; Holcomb, C. T.; Jian, X.; Li, G.; Li, M.; Pankin, A. Y.; Peysson, Y.; Zhai, X.; Bonoli, P.; Brower, D.; Ding, W. X.; Ferron, J. R.; Guo, W.; Lao, L. L.; Li, K.; Liu, H.; Lyv, B.; Xu, G.; Zang, Q.

    2018-01-01

    Experimental and modeling investigations on the DIII-D and EAST tokamaks show the attractive transport and stability properties of fully noninductive, high poloidal-beta (β P ) plasmas, and their suitability for steady-state operating scenarios in ITER and CFETR. A key feature of the high-β P regime is the large-radius (ρ > 0.6) internal transport barrier (ITB), often observed in all channels (ne, Te, Ti, rotation), and responsible for both excellent energy confinement quality and excellent stability properties. Experiments on DIII-D have shown that, with a large-radius ITB, very high β N and β P values (both ≥ 4) can be reached by taking advantage of the stabilizing effect of a nearby conducting wall. Synergistically, higher plasma pressure provides turbulence suppression by Shafranov shift, leading to ITB sustainment independent of the plasma rotation. Experiments on EAST have been used to assess the long pulse potential of the high-β P regime. Using RF-only heating and current drive, EAST achieved minute-long fully noninductive steady state H-mode operation with strike points on an ITER-like tungsten divertor. Improved confinement (relative to standard H-mode) and steady state ITB features are observed with a monotonic q-profile with q min ˜ 1.5. Separately, experiments have shown that increasing the density in plasmas driven by lower hybrid wave broadens the q-profile, a technique that could enable a large radius ITB. These experimental results have been used to validate MHD, current drive, and turbulent transport models, and to project the high-β P regime to a burning plasma. These projections suggest the Shafranov shift alone will not suffice to provide improved confinement (over standard H-mode) without rotation and rotation shear. However, increasing the negative magnetic shear (higher q on axis) provides a similar turbulence suppression mechanism to Shafranov shift, and can help devices such as ITER and CFETR achieve their steady-state fusion

  5. Preliminary failure modes and effects analysis on Korean HCCR TBS to be tested in ITER

    International Nuclear Information System (INIS)

    Ahn, Mu-Young; Cho, Seungyon; Jin, Hyung Gon; Lee, Dong Won; Park, Yi-Hyun; Lee, Youngmin

    2015-01-01

    Highlights: • Postulated initiating events are identified through failure modes and effects analysis on the current HCCR TBS design. • A set of postulated initiating events are selected for consideration of deterministic analysis. • Accident evolutions on the selected postualted initiating events are qualitatively described for deterministic analysis. - Abstract: Korean Helium cooled ceramic reflector (HCCR) Test blanket system (TBS), which comprises Test blanket module (TBM) and ancillary systems in various locations of ITER building, is operated at high temperature and pressure with decay heat. Therefore, safety is utmost concern in design process and it is required to demonstrate that the HCCR TBS is designed to comply with the safety requirements and guidelines of ITER. Due to complexity of the system with many interfaces with ITER, a systematic approach is necessary for safety analysis. This paper presents preliminary failure modes and effects analysis (FMEA) study performed for the HCCR TBS. FMEA is a systematic methodology in which failure modes for components in the system and their consequences are studied from the bottom-up. Over eighty failure modes have been investigated on the HCCR TBS. The failure modes that have similar consequences are grouped as postulated initiating events (PIEs) and total seven reference accident scenarios are derived from FMEA study for deterministic accident analysis. Failure modes not covered here due to evolving design of the HCCR TBS and uncertainty in maintenance procedures will be studied further in near future.

  6. Preliminary failure modes and effects analysis on Korean HCCR TBS to be tested in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Mu-Young, E-mail: myahn74@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of); Jin, Hyung Gon; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Yi-Hyun; Lee, Youngmin [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Highlights: • Postulated initiating events are identified through failure modes and effects analysis on the current HCCR TBS design. • A set of postulated initiating events are selected for consideration of deterministic analysis. • Accident evolutions on the selected postualted initiating events are qualitatively described for deterministic analysis. - Abstract: Korean Helium cooled ceramic reflector (HCCR) Test blanket system (TBS), which comprises Test blanket module (TBM) and ancillary systems in various locations of ITER building, is operated at high temperature and pressure with decay heat. Therefore, safety is utmost concern in design process and it is required to demonstrate that the HCCR TBS is designed to comply with the safety requirements and guidelines of ITER. Due to complexity of the system with many interfaces with ITER, a systematic approach is necessary for safety analysis. This paper presents preliminary failure modes and effects analysis (FMEA) study performed for the HCCR TBS. FMEA is a systematic methodology in which failure modes for components in the system and their consequences are studied from the bottom-up. Over eighty failure modes have been investigated on the HCCR TBS. The failure modes that have similar consequences are grouped as postulated initiating events (PIEs) and total seven reference accident scenarios are derived from FMEA study for deterministic accident analysis. Failure modes not covered here due to evolving design of the HCCR TBS and uncertainty in maintenance procedures will be studied further in near future.

  7. Variational Iterative Refinement Source Term Estimation Algorithm Assessment for Rural and Urban Environments

    Science.gov (United States)

    Delle Monache, L.; Rodriguez, L. M.; Meech, S.; Hahn, D.; Betancourt, T.; Steinhoff, D.

    2016-12-01

    It is necessary to accurately estimate the initial source characteristics in the event of an accidental or intentional release of a Chemical, Biological, Radiological, or Nuclear (CBRN) agent into the atmosphere. The accurate estimation of the source characteristics are important because many times they are unknown and the Atmospheric Transport and Dispersion (AT&D) models rely heavily on these estimates to create hazard assessments. To correctly assess the source characteristics in an operational environment where time is critical, the National Center for Atmospheric Research (NCAR) has developed a Source Term Estimation (STE) method, known as the Variational Iterative Refinement STE algorithm (VIRSA). VIRSA consists of a combination of modeling systems. These systems include an AT&D model, its corresponding STE model, a Hybrid Lagrangian-Eulerian Plume Model (H-LEPM), and its mathematical adjoint model. In an operational scenario where we have information regarding the infrastructure of a city, the AT&D model used is the Urban Dispersion Model (UDM) and when using this model in VIRSA we refer to the system as uVIRSA. In all other scenarios where we do not have the city infrastructure information readily available, the AT&D model used is the Second-order Closure Integrated PUFF model (SCIPUFF) and the system is referred to as sVIRSA. VIRSA was originally developed using SCIPUFF 2.4 for the Defense Threat Reduction Agency and integrated into the Hazard Prediction and Assessment Capability and Joint Program for Information Systems Joint Effects Model. The results discussed here are the verification and validation of the upgraded system with SCIPUFF 3.0 and the newly implemented UDM capability. To verify uVIRSA and sVIRSA, synthetic concentration observation scenarios were created in urban and rural environments and the results of this verification are shown. Finally, we validate the STE performance of uVIRSA using scenarios from the Joint Urban 2003 (JU03

  8. A fatigue lifetime assessment of WEST ITER Like Plasma Facing Unit

    International Nuclear Information System (INIS)

    Languille, P.; Missirlian, M.; Guilhem, D.; Ferlay, F.; Batal, T.; Bucalossi, J.; Firdaouss, M.; Larroque, S.; Martinez, A.; Richou, M.

    2016-01-01

    Highlights: • ITER plasma facing component divertor technology is integrated in WEST. • ITER Like attachments in WEST has been optimised. • The ITER Like PFU is compatible with a wide range of plasma scenarios. - Abstract: Based on a monoblock concept (e.g. a tube-in-tile concept), each elementary tungsten plasma facing component (called Plasma-Facing Unit PFU) of the WEST lower divertor follows as closely as possible the same monoblock geometry, materials and bonding technology that is envisaged for ITER. A fatigue simulation of W PFU was used to validate its specific integration into WEST. The complex design, the material heterogeneities and the usage outside operational load design envelope are all possible causes of fatigue failure. This paper shows how the ITER like monoblocks and its U-shaped attachments technology are integrated into the WEST divertor by performing finite element analysis. The WEST lower divertor is designed to withstand 15 MW steady-state of injected power, with peaked heat fluxes up to 20 MW/m 2 . The integration and the design choices of a WEST ITER Like Plasma Facing Unit inside the WEST vacuum chamber is valid for an “expected life time” of repeated inter ELMs thermal steady state (>10 s) cycles and for 300 off-normal vertical displacement events.

  9. A fatigue lifetime assessment of WEST ITER Like Plasma Facing Unit

    Energy Technology Data Exchange (ETDEWEB)

    Languille, P., E-mail: pascal.languille@gmail.com; Missirlian, M.; Guilhem, D.; Ferlay, F.; Batal, T.; Bucalossi, J.; Firdaouss, M.; Larroque, S.; Martinez, A.; Richou, M.

    2016-11-01

    Highlights: • ITER plasma facing component divertor technology is integrated in WEST. • ITER Like attachments in WEST has been optimised. • The ITER Like PFU is compatible with a wide range of plasma scenarios. - Abstract: Based on a monoblock concept (e.g. a tube-in-tile concept), each elementary tungsten plasma facing component (called Plasma-Facing Unit PFU) of the WEST lower divertor follows as closely as possible the same monoblock geometry, materials and bonding technology that is envisaged for ITER. A fatigue simulation of W PFU was used to validate its specific integration into WEST. The complex design, the material heterogeneities and the usage outside operational load design envelope are all possible causes of fatigue failure. This paper shows how the ITER like monoblocks and its U-shaped attachments technology are integrated into the WEST divertor by performing finite element analysis. The WEST lower divertor is designed to withstand 15 MW steady-state of injected power, with peaked heat fluxes up to 20 MW/m{sup 2}. The integration and the design choices of a WEST ITER Like Plasma Facing Unit inside the WEST vacuum chamber is valid for an “expected life time” of repeated inter ELMs thermal steady state (>10 s) cycles and for 300 off-normal vertical displacement events.

  10. Conceptual design of an electron cyclotron wave system for NET/ITER

    International Nuclear Information System (INIS)

    Kasparek, W.; Kumric, H.; Mueller, G.A.; Pretterebner, J.; Schueller, P.G.; Wagner, D.

    1991-07-01

    Electron Cyclotron waves (ECWs) provide a scheme for electron heating, which, owing to the strong localization of the resonant interaction with the plasma, allows an efficient tailoring of the power deposition profile. In the proposed ITER reference scenario for current drive and heating, ECWs are considered to assist plasma formation, pre-heating, local current profile control near the q=2 surface and possibly for baking the first wall tiles. For these functions, a total power of 20 MW, CW, at a frequency around 120 GHz is needed. A higher frequency system (140 GHz, 20 MW, CW), is also considered to heat the plasma centre and provide burn control. The same system at increased power could be used for plasma heating to ignition. For NET, due to the higher magnetic field, the frequencies needed for the tasks mentioned above are approximately 140 GHz and 160 GHz, respectively. ECWs are also envisaged for bulk heating of the NET plasma. Here, frequencies of about 160 GHz are necessary. A detailed study for the 120 GHz/20 MW ITER reference system has been performed. Scaling rules as well as additional antenna designs for higher frequency systems have been developed. The design principle was to offer a high degree of flexibility for the wide range of envisaged uses of the ECWs. The ECW system should satisfy the physics requirements, advanced requirements of reliability and availability, and must be compatible with the nuclear environment (which requires radiation resistance as well as remote maintenance of at least the antenna part). Therefore, it has been tried to place the most critical components as far away from the machine as possible. To improve the availability, the installation of 15% spare tubes and transmission systems is proposed. (orig.)

  11. Another year of successful collaboration between ITER and CERN

    CERN Multimedia

    Lucio Rossi and Frederick Bordry

    The 4th meeting of the Steering Committee of the CERN-ITER Collaboration Agreement was held on 19 November at CERN. It marked the end of a second year of successful collaboration between ITER and CERN on superconducting magnets and associated technologies and the establishment of CERN as the ITERreference laboratory” for superconducting strand testing for the next five years.   From left to right: Luca Bottura, Head of the CERN Superconductor and Devices Section, Neil Mitchell, Head of the ITER Magnet Division, Frederick Bordry, Head of the CERN Technology Department, Arnaud Devred, Head of the ITER Superconductor Systems and Auxiliaries Section, and Lucio Rossi, Head of the CERN Magnet, Superconductors and Cryostats Group, standing in front of a sample holder used for critical current measurements of Nb3Sn strands in the Superconductor Laboratory (bldg 163) at CERN. The implementation agreement for 2009 encompassed a wide range of topics ranging from expertise in stainless ...

  12. ITER EDA status

    International Nuclear Information System (INIS)

    Aymar, R.

    2001-01-01

    The Project has focused on drafting the Plant Description Document (PDD), which will be published as the Technical Basis for the ITER Final Design Report (FDR), and its related documentation in time for the ITER review process. The preparations have involved continued intensive detailed design work, analyses and assessments by the Home Teams and the Joint Central Team, who have co-operated closely and efficiently. The main technical document has been completed in time for circulation, as planned, to TAC members for their review at TAC-17 (19-22 February 2001). Some of the supporting documents, such as the Plant Design Specification (PDS), Design Requirements and Guidelines (DRG1 and DRG2), and the Plant Safety Requirement (PSR) are also available for reference in draft form. A summary paper of the PDD for the Council's information is available as a separate document. A new documentation structure for the Project has been established. This hierarchical structure for documentation facilitates the entire organization in a way that allows better change control and avoids duplications. The initiative was intended to make this documentation system valid for the construction and operation phases of ITER. As requested, the Director and the JCT have been assisting the Explorations to plan for future joint technical activities during the Negotiations, and to consider technical issues important for ITER construction and operation for their introduction in the draft of a future joint implementation agreement. As charged by the Explorers, the Director has held discussions with the Home Team Leaders in order to prepare for the staffing of the International Team and Participants Teams during the Negotiations (Co-ordinated Technical Activities, CTA) and also in view of informing all ITER staff about their future directions in a timely fashion. One important element of the work was the completion by the Parties' industries of costing studies of about 83 ''procurement packages

  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. Optimization of Iter with Iter-89P scaling

    International Nuclear Information System (INIS)

    Johner, J.

    1991-10-01

    Ignition in the ITER baseline machine is studied in the frame of a 1/2-D model using the ITER-89P scaling of the energy confinement time. The required value of the enhancement factor f L with respect to the L-mode, allowing ignition with a total fusion power of 1100 MW, is found to be 1.9 at an optimum operating temperature of 11 keV. A sensitivity analysis shows that the critical f L =2 value can be exceeded with relatively small changes in the physical assumptions. It is concluded that the safety margin is not sufficient for this project. Optimization of a thermonuclear plasma in a tokamak is then performed with constraints of given maximum magnetic field B in the superconducting windings, given distance between the plasma and the maximum magnetic field point, imposed safety factor at the plasma edge, and given averaged neutron flux at the plasma surface. The minimum enhancement factor f L with respect to the L-mode, allowing ignition at a given value of the total fusion power P fus , is only a function of the torus aspect ratio A. Taking the ITER reference values for the above constraints, the required value of f L is practically independent of the aspect ratio but can be sensibly improved by increasing the total fusion power P fus . With P fus =1700 MW, a reasonable safety margin (f L ≅ 1.5) is obtained. Analytical expressions of the conditions resulting from the above optimization are also derived for an arbitrary monomial scaling of the energy confinement time, and shown to give excellent agreement with the numerical results

  15. ITER Operating Limits and Conditions

    International Nuclear Information System (INIS)

    Ciattaglia, S.; Barabaschi, P.; Carretero, J.A.

    2006-01-01

    The Operating Limits and Conditions (OLCs) are operating parameters and conditions, chosen among all system/components, which together define the domain of the safe operation of ITER in all foreseen ITER status (operation, maintenance, commissioning). At the same time they are selected to guarantee the required operation flexibility which is a critical factor for the success of an experimental machine such as ITER. System and components important for personnel or public safety (Safety Important Class, SIC) are identified from the overall plant safety analysis on functional importance to safety of the components. SIC classification has to be presented already inside the preliminary safety analysis report and approved by the licensing safety authority before the relevant construction. OLCs comprise the safety limits, i.e. that if exceeded could result in a potential safety hazard, the relevant settings that determine the intervention of SIC systems and the operational limits on equipment which warn from or stop a functional departure from a planned operational status that could challenge equipment and functions. The safety limits have to indicate clearly states that leave the nominal safety state of ITER; they are derived from the safety analysis of ITER. OLCs can represent in some cases few parameters grouping together. Some operational conditions, e.g. inventories, will be controlled through no real time measurements and procedures. Operating experience from present tokamaks, in particular JET, and from nuclear plants is considered at the maximum possible extent. This paper presents the guidelines to develop the ITER OLCs with particular reference to safety limits. A few examples are reported as well as open issues on some OLCs control and measurement and the relevant R-and-D planned to solve the issues. (author)

  16. Introduction condition of a tokamak fusion power plant as an advanced technology in world energy scenario

    International Nuclear Information System (INIS)

    Hiwatari, R.; Tokimatsu, K.; Asaoka, Y.; Okano, K.; Konishi, S.; Ogawa, Y.

    2005-01-01

    The present study reveals the following two introduction conditions of a tokamak fusion power plant in a long term world energy scenario. The first condition is the electric breakeven condition, which is required for the fusion energy to be recognized as a suitable candidate of an alternative energy source in the long term world energy scenario. As for the plasma performance (normalized beta value β N , confinement improvement factor for H-mode HH, the ratio of plasma density to Greenwald density limit fn GW ), the electric breakeven condition requires the simultaneous achievement of 1.2 N GW tmax =16 T, thermal efficiency η e =30%, and current drive power P NBI N ∼1.8, HH∼1.0, and fn GW ∼0.9, which correspond to the ITER reference operation parameters, have a strong potential to achieve the electric breakeven condition. The second condition is the economic breakeven condition, which is required to be selected as an alternative energy source. By using a long term world energy and environment model, the potential of the fusion energy in the long term world energy scenario is being investigated. Under the constraint of 550 ppm CO 2 concentration in the atmosphere, a breakeven price for introduction of the fusion energy in the year 2050 is estimated from 65mill/kWh to 135mill/kWh, which is considered as the economic breakeven condition in the present study. Under the conditions of B tmax =16T, η e =40%, plant availability 60%, and a radial build with/without CS coil, the economic breakeven condition requires β N ∼2.5 for 135mill/kWh of higher breakeven price case and β N ∼6.0 for 65mill/kWh of lower breakeven price case. Finally, the demonstration of steady state operation with β N ∼3.0 in the ITER project leads to the prospect to achieve the upper region of breakeven price in the world energy scenario. (author)

  17. Impact of ITER liquid metal design options on safety level and licensing - Sweden

    International Nuclear Information System (INIS)

    Harfors, C.; Devell, L.; Johansson, Kjell; Lundell, B.; Rolandsson, S.

    1993-01-01

    The safety level and licensability of five design options for ITER coolant, breeding material and structural material are assessed, with emphasis on some specified accident scenarios. The safety level is assessed in terms of barrier requirements and the feasibility to construct and qualify such a barrier. The licensability in Sweden of each design option is assessed based on the indicated safety level and on a judgement of the technical feasibility to construct and qualify the ITER tokamak itself, based on the selected design option. 20 refs

  18. Current generation by helicons and LH waves in modern tokamaks and reactors FNSF-AT, ITER and DEMO. Scenarios, modeling and antennae

    Science.gov (United States)

    Vdovin, V.

    2014-02-01

    The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20-40) IC frequency harmonics) at frequencies of 500-1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure βN > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D - Kurchatov Institute experiment on helicons CD [1].

  19. Current generation by helicons and LH waves in modern tokamaks and reactors FNSF-AT, ITER and DEMO. Scenarios, modeling and antennae

    Energy Technology Data Exchange (ETDEWEB)

    Vdovin, V. [NRC Kurchatov Institute Tokamak Physics Institute, Moscow (Russian Federation)

    2014-02-12

    The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20–40) IC frequency harmonics) at frequencies of 500–1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure β{sub N} > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D – Kurchatov Institute experiment on helicons CD [1].

  20. 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.

  1. ITER operating limit definition criteria

    International Nuclear Information System (INIS)

    Ciattaglia, S.; Barabaschi, P.; Carretero, J.A.; Chiocchio, S.; Hureau, D.; Girard, J.Ph.; Gordon, C.; Portone, A.; Rodrigo, L. Rodriguez; Roldan, C.; Saibene, G.; Uzan-Elbez, J.

    2009-01-01

    The operating limits and conditions (OLCs) are operating parameters and conditions, chosen among all system/components, which, together, define the domain of the safe operation of ITER in all foreseen ITER states (operation, maintenance, commissioning). At the same time they are selected to guarantee the required operation flexibility which is a critical factor for the success of an experimental machine such as ITER. System and components that are important for personnel or public safety (safety important class, SIC) are identified considering their functional importance in the overall plant safety analysis. SIC classification has to be presented already in the preliminary safety analysis report and approved by the licensing authority before manufacturing and construction. OLCs comprise the safety limits that, if exceeded, could result in a potential safety hazard, the relevant settings that determine the intervention of SIC systems, and the operational limits on equipment which warn against or stop a functional deviation from a planned operational status that could challenge equipment and functions. Some operational conditions, e.g. in-Vacuum Vessel (VV) radioactive inventories, will be controlled through procedures. Operating experience from present tokamaks, in particular JET, and from nuclear plants, is considered to the maximum possible extent. This paper presents the guidelines for the development of the ITER OLCs with particular reference to safety limits.

  2. ITER: Chronicle of a bankruptcy foretold

    International Nuclear Information System (INIS)

    Petit, Jean-Pierre

    2011-01-01

    By notably referring to Cedric Reux's research thesis and to Andrew Thornton's research thesis, the author shows that the ITER project is a very dangerous one as the use of a tokamak as an energy plant could lead to a catastrophe. He first recalls and describes the ITER's operation principle (use of nuclear fusion instead of nuclear fission) by recalling historical developments of this approach. He describes a tokamak operation, evokes some lessons learned from the experience, discusses tritium reactions, and the problem of plasma pollution. He then refers to Reux and Thornton, and also to an other author (Boozer from Columbia University) for their theoretical works on tokamaks, and quotes some important statements of their works. As quoted in this document, all these authors highlight the severe problem related to disruptions and to their consequences. The author quotes another article presented during a conference of the risk and consequences of disruptions in large tokamaks, and notably discusses the risk related to fatigue. He also illustrates the involved processes by commenting processes occurring at the Sun surface

  3. A comparison theorem for the SOR iterative method

    Science.gov (United States)

    Sun, Li-Ying

    2005-09-01

    In 1997, Kohno et al. have reported numerically that the improving modified Gauss-Seidel method, which was referred to as the IMGS method, is superior to the SOR iterative method. In this paper, we prove that the spectral radius of the IMGS method is smaller than that of the SOR method and Gauss-Seidel method, if the relaxation parameter [omega][set membership, variant](0,1]. As a result, we prove theoretically that this method is succeeded in improving the convergence of some classical iterative methods. Some recent results are improved.

  4. Conformance Testing, the Elixer within the Chain for Learning Scenarios and Objects

    NARCIS (Netherlands)

    Nadolski, Rob; O'Neill, Owen; Vegt van der, Wim; Koper, Rob

    2006-01-01

    The chain for learning scenarios and learning objects includes five iterative links: (i) development, (ii) publication, (iii) making resources searchable and reusable and (iv) facilitating their arrangement (v) towards a runnable unit of learning. The use of e-learning specifications and

  5. RAMI analysis for ITER radial X-ray camera system

    Energy Technology Data Exchange (ETDEWEB)

    Qin, Shijun, E-mail: sjqin@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Hu, Liqun; Chen, Kaiyun [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Barnsley, Robin; Sirinelli, Antoine [ITER Organization, Route Vinon sur Verdon, CS 90046, 13067, St. Paul lez Durance, Cedex (France); Song, Yuntao; Lu, Kun; Yao, Damao; Chen, Yebin; Li, Shi; Cao, Hongrui; Yu, Hong; Sheng, Xiuli [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2016-11-15

    Highlights: • The functional analysis of the ITER RXC system was performed. • A failure modes, effects and criticality analysis of the ITER RXC system was performed. • The reliability and availability of the ITER RXC system and its main functions were calculated. • The ITER RAMI approach was applied to the ITER RXC system for technical risk control in the preliminary design phase. - Abstract: ITER is the first international experimental nuclear fusion device. In the project, the RAMI approach (reliability, availability, maintainability and inspectability) has been adopted for technical risk control to mitigate all the possible failure of components in preparation for operation and maintenance. RAMI analysis of the ITER Radial X-ray Camera diagnostic (RXC) system during preliminary design phase was required, which insures the system with a very high performance to measure the X-ray emission and research the MHD of plasma with high accuracy on the ITER machine. A functional breakdown was prepared in a bottom-up approach, resulting in the system being divided into 3 main functions, 6 intermediate functions and 28 basic functions which are described using the IDEFØ method. Reliability block diagrams (RBDs) were prepared to calculate the reliability and availability of each function under assumption of operating conditions and failure data. Initial and expected scenarios were analyzed to define risk-mitigation actions. The initial availability of RXC system was 92.93%, while after optimization the expected availability was 95.23% over 11,520 h (approx. 16 months) which corresponds to ITER typical operation cycle. A Failure Modes, Effects and Criticality Analysis (FMECA) was performed to the system initial risk. Criticality charts highlight the risks of the different failure modes with regard to the probability of their occurrence and impact on operations. There are 28 risks for the initial state, including 8 major risks. No major risk remains after taking into

  6. SCENARIO PLANNING AS LEARNING

    Directory of Open Access Journals (Sweden)

    Antonio Lourenço Junior

    2010-10-01

    Full Text Available Scenario Planning has been increasingly used, from its introduction to the decision process as effective tools to test decisions, and improve performance in a dynamic environment (Chermack, 2005. The purpose of this article is to demonstrate the potential of an experimental Scenario Planning Model to mobilize, encourage and add more content to the organization’s decision making process – mainly with respect to Strategic Plans of two governmental institutions, a pharmaceutical company and a technology education foundation.  This study describes the application stages of a hybrid scenario-planning model – herein referred to as Planning as Learning – via action-research, showing the scenarios resulting from the experiment and describes the main results of an assessment of such practice. In order to do that, two well-established Scenario Planning models (Prospective school and Shell’s model were analyzed. They were used as a reference for the proposition and application of an experimental model in the two study objects. A questionnaire was used to assess the technique impact. It was possible to obtain high levels of reliability. In-depth interviews were also conducted with the participants. At the end, the results confirmed the model efficiency as a basis for decision making in the competitive environment in which the two institutions are inserted, also to encourage the learning process as a group, as observed throughout the work.

  7. Electromagnetic analysis, structural integrity and progress on mechanical design of the ITER ferromagnetic insert

    Energy Technology Data Exchange (ETDEWEB)

    Morimoto, M. [Mitsubishi Heavy Industries, Ltd., 1-1 Wadasaki-cho 1-chome, Hyogo-ku, Kobe 652-8585 (Japan)], E-mail: masaaki_morimoto@maia.eonet.ne.jp; Ioki, K.; Terasawa, A.; Utin, Yu.; Barabash, V.; Gribov, Y. [ITER Organization, 13108 St. Paul lez Durance (France)

    2009-12-15

    Ferromagnetic material is used to reduce the toroidal field ripple in JFT-2M and JT-60U . In ITER, since the ferromagnetic material is inserted in the space between the double walls of ITER Vacuum Vessel (VV), it is called 'ferromagnetic inserts'. Suitable material is selected to satisfy the design requirements of ITER. The proper location and amount of the ferromagnetic inserts are optimized with the goal of reduction of the toroidal field ripple. The ferromagnetic inserts are designed to minimize electromagnetic forces acting on them. The electromagnetic forces have been calculated with the latest disruption scenarios. Magnetization forces due to magnetic fields have also been calculated. Structural integrity has been validated by a structural analysis.

  8. Study of neutron spectrometers for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Kaellne, Jan

    2005-11-15

    A review is presented of the developments in the field of neutron emission spectrometry (NES) which is of relevance for identifying the role of NES diagnostics on ITER and selecting suitable instrumentation. Neutron spectrometers will be part of the ITER neutron diagnostic complement and this study makes a special effort to examine which performance characteristics the spectrometers should possess to provide the best burning plasma diagnostic information together with neutron cameras and neutron yield monitors. The performance of NES diagnostics is coupled to how much interface space can be provided which has lead to an interest to find compact instruments and their NES capabilities. This study assesses all known spectrometer types of potential interest for ITER and makes a ranking of their performance (as demonstrated or projected), which, in turn, are compared with ITER measurement requirements as a reference; the ratio of diagnostic performance to interface cost for different spectrometers is also discussed for different spectrometer types. The overall result of the study is an assessment of which diagnostic functions neutron measurements can provide in burning plasma fusion experiments on ITER and the role that NES can play depending on the category of instrument installed. Of special note is the result that much higher quality diagnostic information can be obtained from neutron measurements with total yield monitors, profile flux cameras and spectrometers when the synergy in the data is considered in the analysis and interpretation.

  9. STL pocket reference

    CERN Document Server

    Lischner, Ray

    2003-01-01

    The STL Pocket Reference describes the functions, classes, and templates in that part of the C++ standard library often referred to as the Standard Template Library (STL). The STL encompasses containers, iterators, algorithms, and function objects, which collectively represent one of the most important and widely used subsets of standard library functionality. The C++ standard library, even the subset known as the STL, is vast. It's next to impossible to work with the STL without some sort of reference at your side to remind you of template parameters, function invocations, return types--ind

  10. Verification of SuperMC with ITER C-Lite neutronic model

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Shu [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230027 (China); Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Yu, Shengpeng [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); He, Peng, E-mail: peng.he@fds.org.cn [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China)

    2016-12-15

    Highlights: • Verification of the SuperMC Monte Carlo transport code with ITER C-Lite model. • The modeling of the ITER C-Lite model using the latest SuperMC/MCAM. • All the calculated quantities are consistent with MCNP well. • Efficient variance reduction methods are adopted to accelerate the calculation. - Abstract: In pursit of accurate and high fidelity simulation, the reference model of ITER is becoming more and more detailed and complicated. Due to the complexity in geometry and the thick shielding of the reference model, the accurate modeling and precise simulaion of fusion neutronics are very challenging. Facing these difficulties, SuperMC, the Monte Carlo simulation software system developed by the FDS Team, has optimized its CAD interface for the automatic converting of more complicated models and increased its calculation efficiency with advanced variance reduction methods To demonstrate its capabilites of automatic modeling, neutron/photon coupled simulation and visual analysis for the ITER facility, numerical benchmarks using the ITER C-Lite neutronic model were performed. The nuclear heating in divertor and inboard toroidal field (TF) coils and a global neutron flux map were evaluated. All the calculated nuclear heating is compared with the results of the MCNP code and good consistencies between the two codes is shown. Using the global variance reduction methods in SuperMC, the average speed-up is 292 times for the calculation of inboard TF coils nuclear heating, and 91 times for the calculation of global flux map, compared with the analog run. These tests have shown that SuperMC is suitable for the design and analysis of ITER facility.

  11. Methodological review of UK and international low carbon scenarios

    International Nuclear Information System (INIS)

    Hughes, Nick; Strachan, Neil

    2010-01-01

    Scenarios have a long history in business, politics and military planning, as a tool for strategic planning to inform protective, proactive or consensus-based decision making in the face of uncertain futures. Recent years have seen a growth in scenarios for assessing the implications of low carbon futures, but relatively little work has linked these energy scenarios to the broader literature on scenario development. This paper undertakes a methodological review of a selection of UK and international low carbon scenario studies, using a typology of 'trend based', 'technical feasibility' and 'modelling' studies. Dominant methodologies in such studies have been the 2x2 axis and the 'back-casting' approach. Strengths of the studies reviewed include technological detail, and identification of key economic and social constraints. Weaknesses include the over-reliance on constructs such as exogenous emissions constraints, and high level trends, which diminish the ability to understand how the various future scenarios could be brought about or avoided. This is compounded by the lack of depiction of specific system actors; the tendency to abstract policy from the scenarios; and the resulting failure to consider policy, technology and behaviour in an iterative, 'co-evolving' fashion.

  12. New iteration of decommissioning program for NPP Krsko

    International Nuclear Information System (INIS)

    Lokner, V.; Levanat, I.; Rapic, A.; Zeleznik, N.; Mele, I.; Jenko, T.

    2004-01-01

    As required by the paragraph 10 of the Agreement between the governments of Slovenia and Croatia on status and other legal issues related to investment, exploitation, and decommissioning of Nuclear power plant Krsko, Decommissioning program for Krsko NPP including LILW and spent fuel management was drafted. The Intergovernmental body required that the Program should be extensive revision of existing program as one of several iterations to be prepared before the final version. The purpose of the Program is to estimate the expenses of the future decommissioning, radioactive waste and spent fuel management for Krsko NPP. Costing estimation would be the basis for establishment of a special fund in Croatia and for adjustment of the annual rates for the existing decommissioning fund in Slovenia. The Program development was entrusted to specialized organizations both in Croatia and Slovenia, which formed the Project team as the operative body. Consulting firms from Croatia and Slovenia were involved as well as experts from the International Atomic Energy Agency (through short visits to Zagreb and Ljubljana) for specialized fields (e.g. economic aspects of decommissioning, pre-feasibility study for spent fuel repository in crystalline rock, etc.). The analysis was performed in several steps. The first step was to develop rational and feasible integral scenarios (strategies) of decommissioning and LILW and spent fuel management on the basis of detailed technical analysis and within defined boundary conditions. Based on technological data, every scenario was attributed with time distribution of expenses for all main activities. In the second step, financial analysis of the scenarios was undertaken aiming at estimation of total discounted expense and the related annuity (19 installments to the single fund, empty in 2003) for each of the scenarios. The third step involves additional analysis of the chosen scenarios aiming at their (technical or financial) improvements even at

  13. Development of design Criteria for ITER In-vessel Components

    International Nuclear Information System (INIS)

    Sannazzaro, G.; Barabash, V.; Kang, S.C.; Fernandez, E.; Kalinin, G.; Obushev, A.; Martínez, V.J.; Vázquez, I.; Fernández, F.; Guirao, J.

    2013-01-01

    Absrtract: The components located inside the ITER vacuum chamber (in-vessel components – IC), due to their specific nature and the environments they are exposed to (neutron radiation, high heat fluxes, electromagnetic forces, etc.), have specific design criteria which are, in this paper, referred as Structural Design Criteria for In-vessel Components (SDC-IC). The development of these criteria started in the very early phase of the ITER design and followed closely the criteria of the RCC-MR code. Specific rules to include the effect of neutron irradiation were implemented. In 2008 the need of an update of the SDC-IC was identified to add missing specifications, to implement improvements, to modernise rules including recent evolutions in international codes and regulations (i.e. PED). Collaboration was set up between ITER Organization (IO), European (EUDA) and Russian Federation (RFDA) Domestic Agencies to generate a new version of SDC-IC. A Peer Review Group (PRG) composed by members of the ITER Organization and all ITER Domestic Agencies and code experts was set-up to review the proposed modifications, to provide comments, contributions and recommendations

  14. Summary of beryllium qualification activity for ITER first-wall applications

    International Nuclear Information System (INIS)

    Barabash, V; Eaton, R; Hirai, T; Kupriyanov, I; Nikolaev, G; Wang Zhanhong; Liu Xiang; Roedig, M; Linke, J

    2011-01-01

    Beryllium is considered as an armor material for the ITER first wall. The ITER Final Design Report 2001 identified the reference grades S-65C vacuum hot pressed (VHP) from Brush Wellman and DShG-200 from the Russian Federation. These grades have been selected based on excellent thermal fatigue/shock behavior and the available comprehensive database. Later, Chinese and Russian ITER Parties proposed their new grades: CN-G01 (from China) and TGP-56FW (from Russia). To assess the performance of these new grades, the ITER Organization, Chinese and Russian Parties established a program for the characterization of these materials. A summary of the published data and new results are presented in the paper. It was concluded that the proposed Chinese (CN-G01) and Russian (TGP-56FW) beryllium grades can be accepted. Three grades of beryllium are now available for the armor application for the ITER first wall: S-65, CN-G01 and TGP-56FW.

  15. Summary of beryllium qualification activity for ITER first-wall applications

    Science.gov (United States)

    Barabash, V.; Eaton, R.; Hirai, T.; Kupriyanov, I.; Nikolaev, G.; Wang, Zhanhong; Liu, Xiang; Roedig, M.; Linke, J.

    2011-12-01

    Beryllium is considered as an armor material for the ITER first wall. The ITER Final Design Report 2001 identified the reference grades S-65C vacuum hot pressed (VHP) from Brush Wellman and DShG-200 from the Russian Federation. These grades have been selected based on excellent thermal fatigue/shock behavior and the available comprehensive database. Later, Chinese and Russian ITER Parties proposed their new grades: CN-G01 (from China) and TGP-56FW (from Russia). To assess the performance of these new grades, the ITER Organization, Chinese and Russian Parties established a program for the characterization of these materials. A summary of the published data and new results are presented in the paper. It was concluded that the proposed Chinese (CN-G01) and Russian (TGP-56FW) beryllium grades can be accepted. Three grades of beryllium are now available for the armor application for the ITER first wall: S-65, CN-G01 and TGP-56FW.

  16. Physics of the conceptual design of the ITER plasma control system

    Energy Technology Data Exchange (ETDEWEB)

    Snipes, J.A., E-mail: Joseph.Snipes@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Bremond, S. [CEA-IRFM, 13108 St Paul-lez-Durance (France); Campbell, D.J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Casper, T. [1166 Bordeaux St, Pleasanton, CA 94566 (United States); Douai, D. [CEA-IRFM, 13108 St Paul-lez-Durance (France); Gribov, Y. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Humphreys, D. [General Atomics, San Diego, CA 92186 (United States); Lister, J. [Association EURATOM-Confédération Suisse, Ecole Polytechnique Fédérale de Lausanne (EPFL), CRPP, Lausanne CH-1015 (Switzerland); Loarte, A.; Pitts, R. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Sugihara, M., E-mail: Sugihara_ma@yahoo.co.jp [Japan (Japan); Winter, A.; Zabeo, L. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France)

    2014-05-15

    Highlights: • ITER plasma control system conceptual design has been finalized. • ITER's plasma control system will evolve with the ITER research plan. • A sophisticated actuator sharing scheme is being developed to apply multiple coupled control actions simultaneously with a limited set of actuators. - Abstract: The ITER plasma control system (PCS) will play a central role in enabling the experimental program to attempt to sustain DT plasmas with Q = 10 for several hundred seconds and also support research toward the development of steady-state operation in ITER. The PCS is now in the final phase of its conceptual design. The PCS relies on about 45 diagnostic systems to assess real-time plasma conditions and about 20 actuator systems for overall control of ITER plasmas. It will integrate algorithms required for active control of a wide range of plasma parameters with sophisticated event forecasting and handling functions, which will enable appropriate transitions to be implemented, in real-time, in response to plasma evolution or actuator constraints. In specifying the PCS conceptual design, it is essential to define requirements related to all phases of plasma operation, ranging from early (non-active) H/He plasmas through high fusion gain inductive plasmas to fully non-inductive steady-state operation, to ensure that the PCS control functionality and architecture will be capable of satisfying the demands of the ITER research plan. The scope of the control functionality required of the PCS includes plasma equilibrium and density control commonly utilized in existing experiments, control of the plasma heat exhaust, control of a range of MHD instabilities (including mitigation of disruptions), and aspects such as control of the non-inductive current and the current profile required to maintain stable plasmas in steady-state scenarios. Control areas are often strongly coupled and the integrated control of the plasma to reach and sustain high plasma

  17. Physics of the conceptual design of the ITER plasma control system

    International Nuclear Information System (INIS)

    Snipes, J.A.; Bremond, S.; Campbell, D.J.; Casper, T.; Douai, D.; Gribov, Y.; Humphreys, D.; Lister, J.; Loarte, A.; Pitts, R.; Sugihara, M.; Winter, A.; Zabeo, L.

    2014-01-01

    Highlights: • ITER plasma control system conceptual design has been finalized. • ITER's plasma control system will evolve with the ITER research plan. • A sophisticated actuator sharing scheme is being developed to apply multiple coupled control actions simultaneously with a limited set of actuators. - Abstract: The ITER plasma control system (PCS) will play a central role in enabling the experimental program to attempt to sustain DT plasmas with Q = 10 for several hundred seconds and also support research toward the development of steady-state operation in ITER. The PCS is now in the final phase of its conceptual design. The PCS relies on about 45 diagnostic systems to assess real-time plasma conditions and about 20 actuator systems for overall control of ITER plasmas. It will integrate algorithms required for active control of a wide range of plasma parameters with sophisticated event forecasting and handling functions, which will enable appropriate transitions to be implemented, in real-time, in response to plasma evolution or actuator constraints. In specifying the PCS conceptual design, it is essential to define requirements related to all phases of plasma operation, ranging from early (non-active) H/He plasmas through high fusion gain inductive plasmas to fully non-inductive steady-state operation, to ensure that the PCS control functionality and architecture will be capable of satisfying the demands of the ITER research plan. The scope of the control functionality required of the PCS includes plasma equilibrium and density control commonly utilized in existing experiments, control of the plasma heat exhaust, control of a range of MHD instabilities (including mitigation of disruptions), and aspects such as control of the non-inductive current and the current profile required to maintain stable plasmas in steady-state scenarios. Control areas are often strongly coupled and the integrated control of the plasma to reach and sustain high plasma

  18. Development and ITER relevant application of a user friendly interface (TEM) for use with the TMAP4 code

    International Nuclear Information System (INIS)

    Tanaka, M.R.; Fong, C.; Kalyanam, K.M.; Sood, S.K.; Delisle, M.; Natalizio, A.

    1995-01-01

    The Tritium Enclosure Model (TEM) has been developed as a user friendly interface to facilitate the application of the previously validated, verified and ITER approved TMAP4 Code. TEM (and TMAP4) dynamically analyzes the movement of tritium through structures, between structures and adjoining enclosures. Credible ITER relevant accident scenarios were developed and analyzed. The analyses considered the scenario with the cleanup system active or inactive, with and without the surface interactions. For surface interaction cases, the epoxy characteristics reported in the TMAP4 User Manual were used. Typical applications for TEM are the estimation of time-dependent tritium inventories in enclosures, as well as emissions to the environment following an accidental spill into any set of enclosures connected to cleanup systems. This paper outlines the various features of TEM and reports on the application of TEM to determine environmental source terms for the ITER Fuel Cycle and Cooling Systems, under chronic and accidental tritium releases. 3 refs., 2 figs., 1 tab

  19. 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...

  20. Experimental test campaign on an ITER divertor mock-up

    Energy Technology Data Exchange (ETDEWEB)

    Dell' Orco, G. E-mail: giovanni.dellorco@brasimone.enea.it; Malavasi, A.; Merola, M.; Polazzi, G.; Simoncini, M.; Zito, D

    2002-11-01

    In 1998, in the frame of the European R and D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body (CB), designed with some mechanical and hydraulic simplifications with respect to the reference body and its actively cooled Dummy Armour Prototype (DAP). This DAP consists of a Vertical Target (VT), a Wing (WI) and a Dump Target (DT), manufactured by European industries, which are integrated to the Gas Box Liner (GBL) supplied by the Russian Federation ITER Home Team. In 1999, in parallel with the manufacturing activity, the ITER European Home Team decided to assign to ENEA a Task for checking the component integration and performing the thermal-hydraulic and thermal mechanical testing of the DAP and CB. In 1999-2000, ENEA performed the experimental campaign at Brasimone Labs. The present work presents the experimental results of the component integration and the thermal-hydraulic and thermo-mechanical fatigue tests.

  1. Experimental test campaign on an ITER divertor mock-up

    International Nuclear Information System (INIS)

    Dell'Orco, G.; Malavasi, A.; Merola, M.; Polazzi, G.; Simoncini, M.; Zito, D.

    2002-01-01

    In 1998, in the frame of the European R and D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body (CB), designed with some mechanical and hydraulic simplifications with respect to the reference body and its actively cooled Dummy Armour Prototype (DAP). This DAP consists of a Vertical Target (VT), a Wing (WI) and a Dump Target (DT), manufactured by European industries, which are integrated to the Gas Box Liner (GBL) supplied by the Russian Federation ITER Home Team. In 1999, in parallel with the manufacturing activity, the ITER European Home Team decided to assign to ENEA a Task for checking the component integration and performing the thermal-hydraulic and thermal mechanical testing of the DAP and CB. In 1999-2000, ENEA performed the experimental campaign at Brasimone Labs. The present work presents the experimental results of the component integration and the thermal-hydraulic and thermo-mechanical fatigue tests

  2. Metrology for ITER Assembly

    International Nuclear Information System (INIS)

    Bogusch, E.

    2006-01-01

    The overall dimensions of the ITER Tokamak and the particular assembly sequence preclude the use of conventional optical metrology, mechanical jigs and traditional dimensional control equipment, as used for the assembly of smaller, previous generation, fusion devices. This paper describes the state of the art of the capabilities of available metrology systems, with reference to the previous experience in Fusion engineering and in other industries. Two complementary procedures of transferring datum from the primary datum network on the bioshield to the secondary datum s inside the VV with the desired accuracy of about 0.1 mm is described, one method using the access directly through the ports and the other using transfer techniques, developed during the co-operation with ITER/EFDA. Another important task described is the development of a method for the rapid and easy measurement of the gaps between sectors, required for the production of the customised splice plates between them. The scope of the paper includes the evaluation of the composition and cost of the systems and team of technical staff required to meet the requirements of the assembly procedure. The results from a practical, full-scale demonstration of the methodologies used, using the proposed equipment, is described. This work has demonstrated the feasibility of achieving the necessary accuracies for the successful building of ITER. (author)

  3. Erosion evaluation capability of the IVVS for ITER applications

    Energy Technology Data Exchange (ETDEWEB)

    Pollastrone, Fabio, E-mail: fabio.pollastrone@enea.it [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Ferri de Collibus, Mario; Florean, Marco; Francucci, Massimo; Mugnaini, Giampiero; Neri, Carlo; Rossi, Paolo [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Dubus, Gregory; Damiani, Carlo [Fusion For Energy c/Josep Pla, 2 Torres Diagonal Litoral, 08019 Barcelona (Spain)

    2014-10-15

    Highlights: •High resolution laser radar range images for hostile environment (IVVS). •Evaluation of the erosion on the surface scanned by IVVS laser radar. •Erosion evaluation procedure and software. •Test and results of the erosion evaluation procedure. -- Abstract: In ITER it is foreseen the use of the In Vessel Viewing System (IVVS), whose scanning head is a 3D laser imaging system able to obtain high-resolution intensity and range images in hostile environments. The IVVS will be permanently installed into a port extension, therefore it has to be compliant with ITER primary vacuum requirements. In the frame of a Fusion for Energy Grant, an investigation of the expected IVVS metrology performances was required to evaluate the device capability to detect erosions on ITER first wall and divertor and to estimate the amount of eroded material. In ENEA Frascati laboratories, an IVVS probe prototype was developed along with a method and a computational procedure applied to a reference erosion plate target simulating ITER vessel components and their possible erosions. Experimental tests were carried out by this system performing several scans of the reference target with different incidence angles, estimating the eroded volume and comparing this volume with its true value. A dedicated study has been also done by changing the power of the laser source; a discussion about the quality of the 3D laser images is reported. The main results obtained during laboratory tests and data processing are presented and discussed.

  4. Analysis of global warming stabilization scenarios. The Asian-Pacific Integrated Model

    International Nuclear Information System (INIS)

    Kainuma, Mikiko; Morita, Tsuneyuki; Masui, Toshihiko; Takahashi, Kiyoshi; Matsuoka, Yuzuru

    2004-01-01

    This paper analyzes the economic and climatic impacts of the EMF 19 emission scenarios. A reference scenario, three emission scenarios targeting 550 ppmv atmospheric concentration, and three tax scenarios are analyzed. The profiles of energy consumption and economic losses of each policy scenario are compared to the reference scenario. The model also estimates that global mean temperature will increase 1.7-2.9 C in 2100, and the sea level will rise 40-51 cm, compared to the 1990 levels under the EMF scenarios. Impacts on food productivity and malaria infection are estimated to be very severe in some countries in the Asian region

  5. 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)

  6. 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)

  7. Model-based iterative reconstruction technique for radiation dose reduction in chest CT: comparison with the adaptive statistical iterative reconstruction technique

    Energy Technology Data Exchange (ETDEWEB)

    Katsura, Masaki; Matsuda, Izuru; Akahane, Masaaki; Sato, Jiro; Akai, Hiroyuki; Yasaka, Koichiro; Kunimatsu, Akira; Ohtomo, Kuni [University of Tokyo, Department of Radiology, Graduate School of Medicine, Bunkyo-ku, Tokyo (Japan)

    2012-08-15

    To prospectively evaluate dose reduction and image quality characteristics of chest CT reconstructed with model-based iterative reconstruction (MBIR) compared with adaptive statistical iterative reconstruction (ASIR). One hundred patients underwent reference-dose and low-dose unenhanced chest CT with 64-row multidetector CT. Images were reconstructed with 50 % ASIR-filtered back projection blending (ASIR50) for reference-dose CT, and with ASIR50 and MBIR for low-dose CT. Two radiologists assessed the images in a blinded manner for subjective image noise, artefacts and diagnostic acceptability. Objective image noise was measured in the lung parenchyma. Data were analysed using the sign test and pair-wise Student's t-test. Compared with reference-dose CT, there was a 79.0 % decrease in dose-length product with low-dose CT. Low-dose MBIR images had significantly lower objective image noise (16.93 {+-} 3.00) than low-dose ASIR (49.24 {+-} 9.11, P < 0.01) and reference-dose ASIR images (24.93 {+-} 4.65, P < 0.01). Low-dose MBIR images were all diagnostically acceptable. Unique features of low-dose MBIR images included motion artefacts and pixellated blotchy appearances, which did not adversely affect diagnostic acceptability. Diagnostically acceptable chest CT images acquired with nearly 80 % less radiation can be obtained using MBIR. MBIR shows greater potential than ASIR for providing diagnostically acceptable low-dose CT images without severely compromising image quality. (orig.)

  8. Opinion of the Consultative Committee for the EURATOM Research and Training Programme in the Field of Nuclear Energy (Fusion) - CCE-FU - on the European domestic assessment of the ITER-FEAT outline design report. Brussels, 11 July 2000

    International Nuclear Information System (INIS)

    2001-01-01

    The CCE-FU endorses the findings of the FTC on the European Domestic Assessment of the ITER-FEAT Outline Design Review, and expresses the opinion that: The machine, ITER-FEAT, the design of which is presented in the Outline Design Report and accompanying document (Technical Basis for the ITER-FEAT Outline Design) successfully responds, in broad terms, to the requirements set by the Special Working Group, established by the ITER Council, in response to Task No. 1 and adopted by the Council; The parameters chosen represent convergence towards a coherent design, based upon preserving adequate margins within the cost target against the new objectives and yet retaining flexibility to exploit advances in physics understanding; ITER-FEAT can meet its objectives of extended burn in induction operation with power amplification Q>10 at the reference operating values of Ip=15MA, Paux=40MW, thus providing 400 MW of fusion power. The margins to achieve this objective are in the range of 15-25%. The design also has sufficient flexibility to explore hybrid scenarios with long pulse capability (> 2000 seconds), and scenarios aiming at demonstrating steady state operation with the ratio of fusion power to input power for current drive of at least 5, provided further confinement enhancement can be achieved; Although most of the components of ITER-FEAT are still in a preliminary design state, the new design appears suitable to be developed into the final design stage, the new design appears suitable to be developed into the final design of a machine capable of achieving the objectives set by the ITER Council; The target cost for the realisation of ITER-FEAT was set by the ITER Council at about half of the 1998 ITER cost estimates. The present preliminary analysis provides confidence that this target will be reached. By the end of 2000, the detailed cost estimates will be provided from the detailed design specifications and manufacturing studies in the industries of the three Parties

  9. Iterative normalization technique for reference sequence generation for zero-tail discrete fourier transform spread orthogonal frequency division multiplexing

    DEFF Research Database (Denmark)

    2017-01-01

    , and performing an iterative manipulation of the input sequence. The performing of the iterative manipulation of the input sequence may include, for example: computing frequency domain response of the sequence, normalizing elements of the computed frequency domain sequence to unitary power while maintaining phase...

  10. Preparing ITER ICRF: development and analysis of the load resilient matching systems based on antenna mock-up measurements

    International Nuclear Information System (INIS)

    Messiaen, A.; Vervier, M.; Dumortier, P.; Grine, D.; Lamalle, P.U.; Durodie, F.; Koch, R.; Louche, F.; Weynants, R.

    2009-01-01

    The reference design for the ICRF antenna of ITER is constituted by a tight array of 24 straps grouped in eight triplets. The matching network must be load resilient for operation in ELMy discharges and must have antenna spectrum control for heating or current drive operation. The load resilience is based on the use of either hybrid couplers or conjugate-T circuits. However, the mutual coupling between the triplets at the low expected loading strongly counteracts the load resilience and the spectrum control. Using a mock-up of the ITER antenna array with adjustable water load matching solutions are designed. These solutions are derived from transmission line modelling based on the measured scattering matrix and are finally tested. We show that the array current spectrum can be controlled by the anti-node voltage distribution and that suitable decoupler circuits can not only neutralize the adverse mutual coupling effects but also monitor this anti-node voltage distribution. A matching solution using four 3 dB hybrids and the antenna current spectrum feedback control by the decouplers provides outstanding performance if each pair of poloidal triplets undergoes a same load variation. Finally, it is verified by modelling that this matching scenario has the same antenna spectrum and load resilience performances as the antenna array loaded by plasma as described by the TOPICA simulation. This is true for any phasing and frequency in the ITER frequency band. The conjugate-T solution is presently considered as a back-up option.

  11. ITER-FEAT - outline design report. Report by the ITER Director. ITER meeting, Tokyo, January 2000

    International Nuclear Information System (INIS)

    2001-01-01

    It is now possible to define the key elements of ITER-FEAT. This report provides the results, to date, of the joint work of the Special Working Group in the form of an Outline Design Report on the ITER-FEAT design which, subject to the views of ITER Council and of the Parties, will be the focus of further detailed design work and analysis in order to provide to the Parties a complete and fully integrated engineering design within the framework of the ITER EDA extension

  12. Conceptual Design of the ITER ECE Diagnostic - An Update

    Science.gov (United States)

    Austin, M. E.; Pandya, H. K. B.; Beno, J.; Bryant, A. D.; Danani, S.; Ellis, R. F.; Feder, R.; Hubbard, A. E.; Kumar, S.; Ouroua, A.; Phillips, P. E.; Rowan, W. L.

    2012-09-01

    The ITER ECE diagnostic has recently been through a conceptual design review for the entire system including front end optics, transmission line, and back-end instruments. The basic design of two viewing lines, each with a single ellipsoidal mirror focussing into the plasma near the midplane of the typical operating scenarios is agreed upon. The location and design of the hot calibration source and the design of the shutter that directs its radiation to the transmission line are issues that need further investigation. In light of recent measurements and discussion, the design of the broadband transmission line is being revisited and new options contemplated. For the instruments, current systems for millimeter wave radiometers and broad-band spectrometers will be adequate for ITER, but the option for employing new state-of-the-art techniques will be left open.

  13. Tritium loading in ITER plasma-facing surfaces and its release under accident conditions

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Anderl, R.A.; Pawelko, R.J.

    1996-01-01

    Plasma-facing surfaces of the International Thermonuclear Experimental Reactor (ITER) will take up tritium from the plasma. These surfaces will probably consist of matures of Be, C, and possibly W together with other impurities. Recent experimental results have suggested mechanisms, not previously considered in analyses, by which tritium and other hydrogen isotopes are retained in Be. This warrants revised modeling and estimation of the amount of tritium that will be deposited in ITER beryllium plasma-facing surfaces and the rates at which it can be released under postulated accident scenarios. In this paper we describe improvements in modeling and experiments planned at the Idaho National Engineering Laboratory (INEL) to investigate the tritium uptake and thermal release behavior for mixed plasma- facing materials. TMAP4 calculations were made using recent data to estimate first-wall tritium inventories in ITER. 16 refs., 1 fig

  14. New methods of testing nonlinear hypothesis using iterative NLLS estimator

    Science.gov (United States)

    Mahaboob, B.; Venkateswarlu, B.; Mokeshrayalu, G.; Balasiddamuni, P.

    2017-11-01

    This research paper discusses the method of testing nonlinear hypothesis using iterative Nonlinear Least Squares (NLLS) estimator. Takeshi Amemiya [1] explained this method. However in the present research paper, a modified Wald test statistic due to Engle, Robert [6] is proposed to test the nonlinear hypothesis using iterative NLLS estimator. An alternative method for testing nonlinear hypothesis using iterative NLLS estimator based on nonlinear hypothesis using iterative NLLS estimator based on nonlinear studentized residuals has been proposed. In this research article an innovative method of testing nonlinear hypothesis using iterative restricted NLLS estimator is derived. Pesaran and Deaton [10] explained the methods of testing nonlinear hypothesis. This paper uses asymptotic properties of nonlinear least squares estimator proposed by Jenrich [8]. The main purpose of this paper is to provide very innovative methods of testing nonlinear hypothesis using iterative NLLS estimator, iterative NLLS estimator based on nonlinear studentized residuals and iterative restricted NLLS estimator. Eakambaram et al. [12] discussed least absolute deviation estimations versus nonlinear regression model with heteroscedastic errors and also they studied the problem of heteroscedasticity with reference to nonlinear regression models with suitable illustration. William Grene [13] examined the interaction effect in nonlinear models disused by Ai and Norton [14] and suggested ways to examine the effects that do not involve statistical testing. Peter [15] provided guidelines for identifying composite hypothesis and addressing the probability of false rejection for multiple hypotheses.

  15. Linear MHD stability analysis of post-disruption plasmas in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Aleynikova, K., E-mail: ksenia.aleynikova@gmail.com [EURATOM Association, Max-Planck-Institut für Plasmaphysik (Germany); Huijsmans, G. T. A. [ITER Organization (France); Aleynikov, P. [EURATOM Association, Max-Planck-Institut für Plasmaphysik (Germany)

    2016-05-15

    Most of the plasma current can be replaced by a runaway electron (RE) current during plasma disruptions in ITER. In this case the post-disruption plasma current profile is likely to be more peaked than the pre-disruption profile. The MHD activity of such plasma will affect the runaway electron generation and confinement and the dynamics of the plasma position evolution (Vertical Displacement Event), limiting the timeframe for runaway electrons and disruption mitigation. In the present paper, we evaluate the influence of the possible RE seed current parameters on the onset of the MHD instabilities. By varying the RE seed current profile, we search for subsequent plasma evolutions with the highest and the lowest MHD activity. This information can be applied to a development of desirable ITER disruption scenario.

  16. Acoustical source reconstruction from non-synchronous sequential measurements by Fast Iterative Shrinkage Thresholding Algorithm

    Science.gov (United States)

    Yu, Liang; Antoni, Jerome; Leclere, Quentin; Jiang, Weikang

    2017-11-01

    Acoustical source reconstruction is a typical inverse problem, whose minimum frequency of reconstruction hinges on the size of the array and maximum frequency depends on the spacing distance between the microphones. For the sake of enlarging the frequency of reconstruction and reducing the cost of an acquisition system, Cyclic Projection (CP), a method of sequential measurements without reference, was recently investigated (JSV,2016,372:31-49). In this paper, the Propagation based Fast Iterative Shrinkage Thresholding Algorithm (Propagation-FISTA) is introduced, which improves CP in two aspects: (1) the number of acoustic sources is no longer needed and the only making assumption is that of a "weakly sparse" eigenvalue spectrum; (2) the construction of the spatial basis is much easier and adaptive to practical scenarios of acoustical measurements benefiting from the introduction of propagation based spatial basis. The proposed Propagation-FISTA is first investigated with different simulations and experimental setups and is next illustrated with an industrial case.

  17. 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

  18. The European contribution to the ITER Remote Maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Damiani, C., E-mail: carlo.damiani@f4e.europa.eu [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Annino, C.; Balagué, S.; Bates, P.; Ceccanti, F.; Di Mascio, T.; Dubus, G.; Esqué, S.; Gonzalez, C.; Lewczanin, M.; Locke, D.; Mont, L.; Olajos, K.; Ranz, R.; Shuff, R.; Puiu, A.; Van Hille, C.; Van Uffelen, M. [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Choi, C.H.; Friconneau, J.P. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); and others

    2014-10-15

    Highlights: •The article introduces the needs for remote maintenance in ITER. •It also discusses some of the issues related to the cultural transition from tokamaks as plasma physics to nuclear reactors. •It highlights the related cultural change and the implications on plant topology and maintenance. •Then, it presents those remote handling systems that will be procured by Europe. •The article emphasises the need of a major involvement of industries from now on. -- Abstract: For a first-of-a-kind nuclear fusion reactor like ITER, remote maintainability of neutron-activated components is one of the key aspects of plant design and operations, and a fundamental ingredient for the demonstration of long-term viability of fusion as energy source. The European Domestic Agency (EU DA, i.e. Fusion for Energy, F4E) is providing important support to the ITER Organisation (IO) in specifying the functional requirements of the Remote Handling (RH) Procurement Packages (i.e. the subsystems allocated to EU DA belonging to the overall ITER Remote Maintenance Systems IRMS), and in performing design and R and D activities – with the support of national laboratories and industries – in order to define a sound concept for these packages. Furthermore, domestic industries are being involved in the subsequent detailed design, validation, manufacturing and installation activities, in order to actually fulfil our procurement-in-kind obligations. After an introduction to ITER Remote Maintenance, this paper will present status and next stages for the RH systems allocated to EU DA, and will also illustrate complementary aspects related to cross cutting technologies like radiation tolerant components and RH control systems. Finally, the way all these efforts are coordinated will be presented together with the overall implementation scenario and key milestones.

  19. Final report of the ITER EDA. Final report of the ITER Engineering Design Activities. Prepared by the ITER Council

    International Nuclear Information System (INIS)

    2001-01-01

    This is the Final Report by the ITER Council on work carried out by ITER participating countries on cooperation in the Engineering Design Activities (EDA) for the ITER. In this report the main ITER EDA technical objectives, the scope of ITER EDA, its organization and resources, engineering design of ITER tokamak and its main parameters are presented. This Report also includes safety and environmental assessments, site requirements and proposed schedule and estimates of manpower and cost as well as proposals on approaches to joint implementation of the project

  20. Roaming Reference: Reinvigorating Reference through Point of Need Service

    Directory of Open Access Journals (Sweden)

    Kealin M. McCabe

    2011-11-01

    Full Text Available Roaming reference service was pursued as a way to address declining reference statistics. The service was staffed by librarians armed with iPads over a period of six months during the 2010-2011 academic year. Transactional statistics were collected in relation to query type (Research, Facilitative or Technology, location and approach (librarian to patron, patron to librarian or via chat widget. Overall, roaming reference resulted in an additional 228 reference questions, 67% (n=153 of which were research related. Two iterations of the service were implemented, roaming reference as a standalone service (Fall 2010 and roaming reference integrated with traditional reference desk duties (Winter 2011. The results demonstrate that although the Weller Library’s reference transactions are declining annually, they are not disappearing. For a roaming reference service to succeed, it must be a standalone service provided in addition to traditional reference services. The integration of the two reference models (roaming reference and reference desk resulted in a 56% decline in the total number of roaming reference questions from the previous term. The simple act of roaming has the potential to reinvigorate reference services as a whole, forcing librarians outside their comfort zones, allowing them to reach patrons at their point of need.

  1. Structural analysis of ITER TBM Frame and Dummy TBM

    International Nuclear Information System (INIS)

    Marin, Anna; Kim, Byoung Yoon; Bertolini, Claudio; Lucca, Flavio; Komarov, Victor; Merola, Mario; Giancarli, Luciano; Gicquel, Stefan

    2013-01-01

    One of the main engineering performance goals of ITER is to test and validate design concepts of tritium breeding blankets. To accomplish these goals, three ITER equatorial ports are dedicated to the test of Test Blanket Modules (TBMs) that are mock-ups of tritium breeding blankets. These TBMs, associated with appropriate shield blocks, will also provide the same thermal and nuclear shielding as the main blanket. The main function of TBM Port Plug (PP) is to accommodate TBMs and provide a standardized interface with the vacuum vessel (VV)/port structure. The feasibility of the design concept of the Frame including two Dummy TBMs has been investigated by proposing design improvements of the reference design through an extensive set of thermal, electromagnetic (EM) and stress analyses. As well, the related static strength was evaluated in accordance with the structural design criteria for ITER in-vessel components (SDC-IC). This paper outlines the engineering aspects of the ITER TBM Frame and Dummy TBM and focuses on the feasibility of the present design by structural assessment

  2. 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

  3. Threshold-driven optimization for reference-based auto-planning

    Science.gov (United States)

    Long, Troy; Chen, Mingli; Jiang, Steve; Lu, Weiguo

    2018-02-01

    We study threshold-driven optimization methodology for automatically generating a treatment plan that is motivated by a reference DVH for IMRT treatment planning. We present a framework for threshold-driven optimization for reference-based auto-planning (TORA). Commonly used voxel-based quadratic penalties have two components for penalizing under- and over-dosing of voxels: a reference dose threshold and associated penalty weight. Conventional manual- and auto-planning using such a function involves iteratively updating the preference weights while keeping the thresholds constant, an unintuitive and often inconsistent method for planning toward some reference DVH. However, driving a dose distribution by threshold values instead of preference weights can achieve similar plans with less computational effort. The proposed methodology spatially assigns reference DVH information to threshold values, and iteratively improves the quality of that assignment. The methodology effectively handles both sub-optimal and infeasible DVHs. TORA was applied to a prostate case and a liver case as a proof-of-concept. Reference DVHs were generated using a conventional voxel-based objective, then altered to be either infeasible or easy-to-achieve. TORA was able to closely recreate reference DVHs in 5-15 iterations of solving a simple convex sub-problem. TORA has the potential to be effective for auto-planning based on reference DVHs. As dose prediction and knowledge-based planning becomes more prevalent in the clinical setting, incorporating such data into the treatment planning model in a clear, efficient way will be crucial for automated planning. A threshold-focused objective tuning should be explored over conventional methods of updating preference weights for DVH-guided treatment planning.

  4. Analysis of the ITER cryoplant operational modes

    International Nuclear Information System (INIS)

    Henry, D.; Journeaux, J.Y.; Roussel, P.; Michel, F.; Poncet, J.M.; Girard, A.; Kalinin, V.; Chesny, P.

    2007-01-01

    In the framework of an EFDA task, CEA is carrying out an analysis of the various ITER cryoplant operational modes. According to the project integration document, ITER is designed to be operated 365 days per year in order to optimize the available time of the Tokamak. It is anticipated that operation will be performed in long periods separated by maintenance periods (e.g. 10 days continuous operation and 1 week break) with annual or bi-annual major shutdown periods of a few months for maintenance, further installation and commissioning. For this operation schedule, auxiliary subsystems like the cryoplant and the cryodistribution have to cope with different heat loads which depend on the different ITER operating states. The cryoplant consists of four identical 4.5 K refrigerators and two 80 K helium loops coupled with two LN2 modules. All of these cryogenic subsystems have to operate in parallel to remove the heat loads from the magnet, 80 K shields, cryopumps and other small users. After a brief recall of the main particularities of a cryogenic system operating in a Tokamak environment, the first part of this study is dedicated to the assessment of the main ITER operation states. A new design of refrigeration loop for the HTS current leads, the updated layout of the cryodistribution system and revised strategy for operations of the cryopumps have been taken into consideration. The relevant normal operating scenarios of the cryoplant are checked for the typical ITER operating states like plasma operation state, short term stand by, short term maintenance, or test and conditioning state. The second part of the paper is dedicated to the abnormal operating modes coming from the magnets and from those generated by the cryoplant itself. The occurrence of a fast discharge or a quench of the magnets generates large heat loads disturbances and produces exceptional high mass flow rates which have to be managed by the cryoplant, while a failure of a cryogenic component induces

  5. Model-based iterative reconstruction technique for radiation dose reduction in chest CT: comparison with the adaptive statistical iterative reconstruction technique

    International Nuclear Information System (INIS)

    Katsura, Masaki; Matsuda, Izuru; Akahane, Masaaki; Sato, Jiro; Akai, Hiroyuki; Yasaka, Koichiro; Kunimatsu, Akira; Ohtomo, Kuni

    2012-01-01

    To prospectively evaluate dose reduction and image quality characteristics of chest CT reconstructed with model-based iterative reconstruction (MBIR) compared with adaptive statistical iterative reconstruction (ASIR). One hundred patients underwent reference-dose and low-dose unenhanced chest CT with 64-row multidetector CT. Images were reconstructed with 50 % ASIR-filtered back projection blending (ASIR50) for reference-dose CT, and with ASIR50 and MBIR for low-dose CT. Two radiologists assessed the images in a blinded manner for subjective image noise, artefacts and diagnostic acceptability. Objective image noise was measured in the lung parenchyma. Data were analysed using the sign test and pair-wise Student's t-test. Compared with reference-dose CT, there was a 79.0 % decrease in dose-length product with low-dose CT. Low-dose MBIR images had significantly lower objective image noise (16.93 ± 3.00) than low-dose ASIR (49.24 ± 9.11, P < 0.01) and reference-dose ASIR images (24.93 ± 4.65, P < 0.01). Low-dose MBIR images were all diagnostically acceptable. Unique features of low-dose MBIR images included motion artefacts and pixellated blotchy appearances, which did not adversely affect diagnostic acceptability. Diagnostically acceptable chest CT images acquired with nearly 80 % less radiation can be obtained using MBIR. MBIR shows greater potential than ASIR for providing diagnostically acceptable low-dose CT images without severely compromising image quality. (orig.)

  6. Fusion Power Measurement Using a Combined Neutron Spectrometer-Camera System at ITER

    International Nuclear Information System (INIS)

    Sjoestrand, Henrik; Sunden, E. Andersson; Conroy, S.; Ericsson, G.; Johnson, M. Gatu; Giacomelli, L.; Hellesen, C.; Hjalmarsson, A.; Ronchi, E.; Weiszflog, M.; Kaellne, J.

    2008-01-01

    A central task for fusion plasma diagnostics is to measure the 2.5 and 14 MeV neutron emission rate in order to determine the fusion power. A new method for determining the neutron yield has been developed at JET. It makes use of the magnetic proton recoil neutron spectrometer and a neutron camera and provides the neutron yield with small systematic errors. At ITER a similar system could operate if a high-resolution, high-performance neutron spectrometer similar to the MPR was installed. In this paper, we present how such system could be implemented and how well it would perform under different assumption of plasma scenarios and diagnostic capabilities. It is found that the systematic uncertainty for using such a system as an absolute calibration reference is as low as 3% and hence it would be an excellent candidate for the calibration of neutron monitors such as fission chambers. It is also shown that the system could provide a 1 ms time resolved estimation of the neutron rate with a total uncertainty of 5%

  7. Model-based iterative reconstruction and adaptive statistical iterative reconstruction: dose-reduced CT for detecting pancreatic calcification

    International Nuclear Information System (INIS)

    Yasaka, Koichiro; Katsura, Masaki; Akahane, Masaaki; Sato, Jiro; Matsuda, Izuru; Ohtomo, Kuni

    2016-01-01

    Iterative reconstruction methods have attracted attention for reducing radiation doses in computed tomography (CT). To investigate the detectability of pancreatic calcification using dose-reduced CT reconstructed with model-based iterative construction (MBIR) and adaptive statistical iterative reconstruction (ASIR). This prospective study approved by Institutional Review Board included 85 patients (57 men, 28 women; mean age, 69.9 years; mean body weight, 61.2 kg). Unenhanced CT was performed three times with different radiation doses (reference-dose CT [RDCT], low-dose CT [LDCT], ultralow-dose CT [ULDCT]). From RDCT, LDCT, and ULDCT, images were reconstructed with filtered-back projection (R-FBP, used for establishing reference standard), ASIR (L-ASIR), and MBIR and ASIR (UL-MBIR and UL-ASIR), respectively. A lesion (pancreatic calcification) detection test was performed by two blinded radiologists with a five-point certainty level scale. Dose-length products of RDCT, LDCT, and ULDCT were 410, 97, and 36 mGy-cm, respectively. Nine patients had pancreatic calcification. The sensitivity for detecting pancreatic calcification with UL-MBIR was high (0.67–0.89) compared to L-ASIR or UL-ASIR (0.11–0.44), and a significant difference was seen between UL-MBIR and UL-ASIR for one reader (P = 0.014). The area under the receiver-operating characteristic curve for UL-MBIR (0.818–0.860) was comparable to that for L-ASIR (0.696–0.844). The specificity was lower with UL-MBIR (0.79–0.92) than with L-ASIR or UL-ASIR (0.96–0.99), and a significant difference was seen for one reader (P < 0.01). In UL-MBIR, pancreatic calcification can be detected with high sensitivity, however, we should pay attention to the slightly lower specificity

  8. Model-based iterative reconstruction and adaptive statistical iterative reconstruction: dose-reduced CT for detecting pancreatic calcification.

    Science.gov (United States)

    Yasaka, Koichiro; Katsura, Masaki; Akahane, Masaaki; Sato, Jiro; Matsuda, Izuru; Ohtomo, Kuni

    2016-01-01

    Iterative reconstruction methods have attracted attention for reducing radiation doses in computed tomography (CT). To investigate the detectability of pancreatic calcification using dose-reduced CT reconstructed with model-based iterative construction (MBIR) and adaptive statistical iterative reconstruction (ASIR). This prospective study approved by Institutional Review Board included 85 patients (57 men, 28 women; mean age, 69.9 years; mean body weight, 61.2 kg). Unenhanced CT was performed three times with different radiation doses (reference-dose CT [RDCT], low-dose CT [LDCT], ultralow-dose CT [ULDCT]). From RDCT, LDCT, and ULDCT, images were reconstructed with filtered-back projection (R-FBP, used for establishing reference standard), ASIR (L-ASIR), and MBIR and ASIR (UL-MBIR and UL-ASIR), respectively. A lesion (pancreatic calcification) detection test was performed by two blinded radiologists with a five-point certainty level scale. Dose-length products of RDCT, LDCT, and ULDCT were 410, 97, and 36 mGy-cm, respectively. Nine patients had pancreatic calcification. The sensitivity for detecting pancreatic calcification with UL-MBIR was high (0.67-0.89) compared to L-ASIR or UL-ASIR (0.11-0.44), and a significant difference was seen between UL-MBIR and UL-ASIR for one reader (P = 0.014). The area under the receiver-operating characteristic curve for UL-MBIR (0.818-0.860) was comparable to that for L-ASIR (0.696-0.844). The specificity was lower with UL-MBIR (0.79-0.92) than with L-ASIR or UL-ASIR (0.96-0.99), and a significant difference was seen for one reader (P < 0.01). In UL-MBIR, pancreatic calcification can be detected with high sensitivity, however, we should pay attention to the slightly lower specificity.

  9. The ITER poloidal field system: control and power supplies

    International Nuclear Information System (INIS)

    Mondino, P.L.; Benfatto, I.; Gribov, Y.; Matsukawa, M.; Odajima, K.; Portone, A.; Roshal, A.; Bareyt, B.; Bertolini, E.; Bottereau, J.M.; Huart, M.; Maschio, A.; Bulgakov, S.; Kuchinski, V.

    1995-01-01

    The paper reports the preliminary scenario of the ITER Poloidal Field (PF) system operation, the method used to evaluate the installed power, the basic structure of the circuits and finally the concepts of the preliminary design of control and power supply. The superconducting coils are energized from the HV Grid with conventional AC/DC converters. R and D is required for circuit breakers, make switches and resistors, the basic components of both the switching networks and the discharge circuits. (orig.)

  10. 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

  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. COMPARISON OF HOLOGRAPHIC AND ITERATIVE METHODS FOR AMPLITUDE OBJECT RECONSTRUCTION

    Directory of Open Access Journals (Sweden)

    I. A. Shevkunov

    2015-01-01

    Full Text Available Experimental comparison of four methods for the wavefront reconstruction is presented. We considered two iterative and two holographic methods with different mathematical models and algorithms for recovery. The first two of these methods do not use a reference wave recording scheme that reduces requirements for stability of the installation. A major role in phase information reconstruction by such methods is played by a set of spatial intensity distributions, which are recorded as the recording matrix is being moved along the optical axis. The obtained data are used consistently for wavefront reconstruction using an iterative procedure. In the course of this procedure numerical distribution of the wavefront between the planes is performed. Thus, phase information of the wavefront is stored in every plane and calculated amplitude distributions are replaced for the measured ones in these planes. In the first of the compared methods, a two-dimensional Fresnel transform and iterative calculation in the object plane are used as a mathematical model. In the second approach, an angular spectrum method is used for numerical wavefront propagation, and the iterative calculation is carried out only between closely located planes of data registration. Two digital holography methods, based on the usage of the reference wave in the recording scheme and differing from each other by numerical reconstruction algorithm of digital holograms, are compared with the first two methods. The comparison proved that the iterative method based on 2D Fresnel transform gives results comparable with the result of common holographic method with the Fourier-filtering. It is shown that holographic method for reconstructing of the object complex amplitude in the process of the object amplitude reduction is the best among considered ones.

  14. Scenario-informed multiple criteria analysis for prioritizing investments in electricity capacity expansion

    International Nuclear Information System (INIS)

    Martinez, Lauro J.; Lambert, James H.; Karvetski, Christopher W.

    2011-01-01

    Planning the expansion and energy security of electricity capacity for a national electricity utility is a complex task in almost any economy. Planning is usually an iterative activity and can involve the use of large scale planning optimization systems accompanied by assessment of uncertain scenarios emerging from economic, technological, environmental, and regulatory developments. This paper applies a multiple criteria decision analysis to prioritize investment portfolios in capacity expansion and energy security while principally studying the robustness of the prioritization to multiple uncertain and emergent scenarios. The scenarios are identified through interaction with decision makers and stakeholders. The approach finds which scenarios most affect the prioritization of the portfolios and which portfolios have the greatest upside and downside potential across scenarios. The approach fosters innovation in the use of robust and efficient technologies, renewable energy sources, and cleaner energy fuels. A demonstration is provided for assessing the performance of technology portfolios constructed from investments in nine electricity generation technologies in Mexico.

  15. Conceptual Design of the ITER ECE Diagnostic – An Update

    Directory of Open Access Journals (Sweden)

    Ouroua A.

    2012-09-01

    Full Text Available The ITER ECE diagnostic has recently been through a conceptual design review for the entire system including front end optics, transmission line, and back-end instruments. The basic design of two viewing lines, each with a single ellipsoidal mirror focussing into the plasma near the midplane of the typical operating scenarios is agreed upon. The location and design of the hot calibration source and the design of the shutter that directs its radiation to the transmission line are issues that need further investigation. In light of recent measurements and discussion, the design of the broadband transmission line is being revisited and new options contemplated. For the instruments, current systems for millimeter wave radiometers and broad-band spectrometers will be adequate for ITER, but the option for employing new state-of-the-art techniques will be left open.

  16. Carbon fiber composites application in ITER plasma facing components

    Science.gov (United States)

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

    1998-10-01

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

  17. Carbon fiber composites application in ITER plasma facing components

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  18. Plasma position and shape control for ITER

    International Nuclear Information System (INIS)

    Portone, A.; Gribov, Y.; Huguet, M.

    1995-01-01

    Key features and main results about the control of the plasma shape in ITER are presented. A control algorithm is designed to control up to 6 gaps between the plasma separatrix and the plasma facing components during the reference burn phase. Nonlinear simulations show the performances of the controller in the presence of plasma vertical position offsets, beta drops and power supply voltage saturation

  19. Exploring the reversibility of marine climate change impacts in temperature overshoot scenarios

    Science.gov (United States)

    Zickfeld, K.; Li, X.; Tokarska, K.; Kohfeld, K. E.

    2017-12-01

    Artificial carbon dioxide removal (CDR) from the atmosphere has been proposed as a measure for mitigating climate change and restoring the climate system to a `safe' state after overshoot. Previous studies have demonstrated that the changes in surface air temperature due to anthropogenic CO2 emissions can be reversed through CDR, while some oceanic properties, for example thermosteric sea level rise, show a delay in their response to CDR. This research aims to investigate the reversibility of changes in ocean conditions after implementation of CDR with a focus on ocean biogeochemical properties. To achieve this, we analyze climate model simulations based on two sets of emission scenarios. We first use RCP2.6 and its extension until year 2300 as the reference scenario and design several temperature and cumulative CO2 emissions "overshoot" scenarios based on other RCPs, which represents cases with less ambitious mitigation policies in the near term that temporarily exceed the 2 °C target adopted by the Paris Agreement. In addition, we use a set of emission scenarios with a reference scenario limiting warming to 1.5°C in the long term and two overshoot scenarios. The University of Victoria Earth System Climate Model (UVic ESCM), a climate model of intermediate complexity, is forced with these emission scenarios. We compare the response of select ocean variables (seawater temperature, pH, dissolved oxygen) in the overshoot scenarios to that in the respective reference scenario at the time the same amount of cumulative emissions is achieved. Our results suggest that the overshoot and subsequent return to a reference CO2 cumulative emissions level would leave substantial impacts on the marine environment. Although the changes in global mean sea surface variables (temperature, pH and dissolved oxygen) are largely reversible, global mean ocean temperature, dissolved oxygen and pH differ significantly from those in the reference scenario. Large ocean areas exhibit

  20. Rotation and neoclassical ripple transport in ITER

    Science.gov (United States)

    Paul, E. J.; Landreman, M.; Poli, F. M.; Spong, D. A.; Smith, H. M.; Dorland, W.

    2017-11-01

    Neoclassical transport in the presence of non-axisymmetric magnetic fields causes a toroidal torque known as neoclassical toroidal viscosity (NTV). The toroidal symmetry of ITER will be broken by the finite number of toroidal field coils and by test blanket modules (TBMs). The addition of ferritic inserts (FIs) will decrease the magnitude of the toroidal field ripple. 3D magnetic equilibria in the presence of toroidal field ripple and ferromagnetic structures are calculated for an ITER steady-state scenario using the variational moments equilibrium code (VMEC). Neoclassical transport quantities in the presence of these error fields are calculated using the stellarator Fokker-Planck iterative neoclassical conservative solver (SFINCS). These calculations fully account for E r , flux surface shaping, multiple species, magnitude of ripple, and collisionality rather than applying approximate analytic NTV formulae. As NTV is a complicated nonlinear function of E r , we study its behavior over a plausible range of E r . We estimate the toroidal flow, and hence E r , using a semi-analytic turbulent intrinsic rotation model and NUBEAM calculations of neutral beam torque. The NTV from the \\vert{n}\\vert = 18 ripple dominates that from lower n perturbations of the TBMs. With the inclusion of FIs, the magnitude of NTV torque is reduced by about 75% near the edge. We present comparisons of several models of tangential magnetic drifts, finding appreciable differences only for superbanana-plateau transport at small E r . We find the scaling of calculated NTV torque with ripple magnitude to indicate that ripple-trapping may be a significant mechanism for NTV in ITER. The computed NTV torque without ferritic components is comparable in magnitude to the NBI and intrinsic turbulent torques and will likely damp rotation, but the NTV torque is significantly reduced by the planned ferritic inserts.

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

    International Nuclear Information System (INIS)

    Janeschitz, G.; Antipenkov, A.; Federici, G.; Ibbott, C.; Kukushkin, A.; Ladd, P.; Martin, E.; Tivey, R.

    2001-01-01

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

  2. Iterative learning control an optimization paradigm

    CERN Document Server

    Owens, David H

    2016-01-01

    This book develops a coherent theoretical approach to algorithm design for iterative learning control based on the use of optimization concepts. Concentrating initially on linear, discrete-time systems, the author gives the reader access to theories based on either signal or parameter optimization. Although the two approaches are shown to be related in a formal mathematical sense, the text presents them separately because their relevant algorithm design issues are distinct and give rise to different performance capabilities. Together with algorithm design, the text demonstrates that there are new algorithms that are capable of incorporating input and output constraints, enable the algorithm to reconfigure systematically in order to meet the requirements of different reference signals and also to support new algorithms for local convergence of nonlinear iterative control. Simulation and application studies are used to illustrate algorithm properties and performance in systems like gantry robots and other elect...

  3. 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

  4. 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.

  5. Iteration of ultrasound aberration correction methods

    Science.gov (United States)

    Maasoey, Svein-Erik; Angelsen, Bjoern; Varslot, Trond

    2004-05-01

    Aberration in ultrasound medical imaging is usually modeled by time-delay and amplitude variations concentrated on the transmitting/receiving array. This filter process is here denoted a TDA filter. The TDA filter is an approximation to the physical aberration process, which occurs over an extended part of the human body wall. Estimation of the TDA filter, and performing correction on transmit and receive, has proven difficult. It has yet to be shown that this method works adequately for severe aberration. Estimation of the TDA filter can be iterated by retransmitting a corrected signal and re-estimate until a convergence criterion is fulfilled (adaptive imaging). Two methods for estimating time-delay and amplitude variations in receive signals from random scatterers have been developed. One method correlates each element signal with a reference signal. The other method use eigenvalue decomposition of the receive cross-spectrum matrix, based upon a receive energy-maximizing criterion. Simulations of iterating aberration correction with a TDA filter have been investigated to study its convergence properties. A weak and strong human-body wall model generated aberration. Both emulated the human abdominal wall. Results after iteration improve aberration correction substantially, and both estimation methods converge, even for the case of strong aberration.

  6. Thermo-mechanical design of the SINGAP accelerator grids for ITER NB Injectors

    International Nuclear Information System (INIS)

    Agostinetti, P.; Dal Bello, S.; Palma, M.D.; Zaccaria, P.

    2006-01-01

    The SINGle Aperture - SINgle GAP (SINGAP) accelerator for ITER neutral beam injector foresees four grids for the extraction and acceleration of negative ions, instead of the seven grids of the Multi Aperture Multi Grid (MAMuG) reference configuration. Optimized geometry of the SINGAP grids (plasma, extraction, pre-acceleration, and grounded grid) was identified by CEA Association considering specific requirements for ions extraction and beam generation referring to experimental data and code simulations. This paper focuses on the thermo-hydraulic and thermo-mechanical design of the grids carried out by Consorzio RFX for the design of the first ITER NB Injector and the ITER NB Test Facility. The cooling circuit design (position and shape of the channels) and the cooling parameters (water coolant temperatures, pressure and velocity) were optimized with thermo-hydraulic and thermo-mechanical sensitivity analyses in order to satisfy the grid functional requirements (temperatures, in plane and out of plane deformations). A complete and detailed thermo-structural design assessment of the SINGAP grids was accomplished applying the structural design rules for ITER in-vessel components and considering both the reference load conditions and the maximum load provided by the power supplies. The design required a complete modelling of the grids and their support frames by means of 3D FE and CAD models. The grids were finally integrated with the support and cooling systems inside the beam source vessel. The main results of the thermo-hydraulic and thermo-mechanical analyses are presented. The open issues are then reported, mainly regarding the material properties characterization (static and fatigue tests) and the qualification of technologies for OFHC copper electro-deposition, brazing, and welding of heterogeneous materials. (author)

  7. 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)

  8. 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)

  9. 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

  10. 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.

  11. Manufacturing and testing of reference samples for the definition of acceptance criteria for the ITER divertor

    International Nuclear Information System (INIS)

    Visca, Eliseo; Cacciotti, E.; Libera, S.; Mancini, A.; Pizzuto, A.; Roccella, S.; Riccardi, B.; Escourbiac, F.; Sanguinetti, G.P.

    2010-01-01

    The most critical part of a high heat flux (HHF) plasma facing component (PFC) is the armour to heat sink joint. An experimental study was launched by EFDA in order to define the acceptance criteria to be used for the procurements of the ITER Divertor PFCs. ENEA is involved in the European International Thermonuclear Experimental Reactor (ITER) R and D activities and together with Ansaldo Ricerche S.p.A. has manufactured several PFCs mock-ups using the Hot Radial Pressing and Pre-Brazed Casting technologies. According to the technical specifications issued by EFDA, ENEA and Ansaldo have collaborated to manufacture half of the samples with calibrated artificial defects required for this experimental study. After manufacturing, the samples were examined by ultrasonic and SATIR non-destructive examination (NDE) methods in order to confirm the size and position of the artificial defects. In particular, it was concluded that defects are detectable with these NDE techniques and they finally gave indication about the threshold of propagation during high heat flux experiments relevant with heat fluxes expected in ITER Divertor. This paper reports the manufacturing procedure used to obtain the required calibrated artificial defects in the CFC and W armoured samples as well as the NDE results and the thermal high heat flux results.

  12. Manufacturing and testing of reference samples for the definition of acceptance criteria for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Visca, Eliseo, E-mail: visca@frascati.enea.i [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Cacciotti, E.; Libera, S.; Mancini, A.; Pizzuto, A.; Roccella, S. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Riccardi, B., E-mail: Bruno.Riccardi@f4e.europa.e [Fusion For Energy, Barcelona (Spain); Escourbiac, F., E-mail: frederic.escourbiac@iter.or [ITER Organization, Cadarache (France); Sanguinetti, G.P., E-mail: gianpaolo.sanguinetti@aen.ansaldo.i [Ansaldo Energia S.p.A., Genova (Italy)

    2010-12-15

    The most critical part of a high heat flux (HHF) plasma facing component (PFC) is the armour to heat sink joint. An experimental study was launched by EFDA in order to define the acceptance criteria to be used for the procurements of the ITER Divertor PFCs. ENEA is involved in the European International Thermonuclear Experimental Reactor (ITER) R and D activities and together with Ansaldo Ricerche S.p.A. has manufactured several PFCs mock-ups using the Hot Radial Pressing and Pre-Brazed Casting technologies. According to the technical specifications issued by EFDA, ENEA and Ansaldo have collaborated to manufacture half of the samples with calibrated artificial defects required for this experimental study. After manufacturing, the samples were examined by ultrasonic and SATIR non-destructive examination (NDE) methods in order to confirm the size and position of the artificial defects. In particular, it was concluded that defects are detectable with these NDE techniques and they finally gave indication about the threshold of propagation during high heat flux experiments relevant with heat fluxes expected in ITER Divertor. This paper reports the manufacturing procedure used to obtain the required calibrated artificial defects in the CFC and W armoured samples as well as the NDE results and the thermal high heat flux results.

  13. 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)

  14. ICP (ITER Collaborative Platform)

    Energy Technology Data Exchange (ETDEWEB)

    Capuano, C.; Carayon, F.; Patel, V. [ITER, 13 - St. Paul-Lez Durance (France)

    2009-07-01

    The ITER organization has the necessity to manage a massive amount of data and processes. Each team requires different process and databases often interconnected with those of others teams. ICP is the current central ITER repository of structured and unstructured data. All data in ICP is served and managed via a web interface that provides global accessibility with a common user friendly interface. This paper will explain the model used by ICP and how it serves the ITER project by providing a robust and agile platform. ICP is developed in ASP.NET using MSSQL Server for data storage. It currently houses 15 data driven applications, 150 different types of record, 500 k objects and 2.5 M references. During European working hours the system averages 150 concurrent users and 20 requests per second. ICP connects to external database applications to provide a single entry point to ITER data and a safe shared storage place to maintain this data long-term. The Core model provides an easy to extend framework to meet the future needs of the Organization. ICP follows a multi-tier architecture, providing logical separation of process. The standard three-tier architecture is expanded, with the data layer separated into data storage, data structure, and data access components. The business or applications logic layer is broken up into a common business functionality layer, a type specific logic layer, and a detached work-flow layer. Finally the presentation tier comprises a presentation adapter layer and an interface layer. Each layer is built up from small blocks which can be combined to create a wide range of more complex functionality. Each new object type developed gains access to a wealth of existing code functionality, while also free to adapt and extend this. The hardware structure is designed to provide complete redundancy, high availability and to handle high load. This document is composed of an abstract followed by the presentation transparencies. (authors)

  15. Iterative Equalization and Interference Alignment for Multiuser MIMO HetNets with Imperfect CSI

    Directory of Open Access Journals (Sweden)

    Daniel Castanheira

    2015-01-01

    Full Text Available In this paper we consider a scenario, where several small-cells work under the same coverage area and spectrum of a macrocell. The signals stemming from the small-cell (macrocell users if not carefully dealt with will generate harmful interference into the macrocell (small-cell. To tackle this problem interference alignment and iterative equalization techniques are considered. By using IA all interference generated by the small-cell (macrocell users is aligned along a low dimensional subspace, at the macrocell (small-cells. This reduces considerably the amount of resources allocated, to enable the coexistence of the two systems. However, perfect IA requires the availability of error-free channel state information (CSI at the transmitters. Due to CSI errors one can have substantial performance degradation due to imperfect alignments. Since in this work the IA precoders are based on imperfect CSI, an efficient iterative space-frequency equalization is designed at the receiver side to cope with the residual aligned interference. The results demonstrate that iterative equalization is robust to imperfect CSI and removes efficiently the interference generated by the poorly aligned interference. Close to matched filter bound performance is achieved, with a very few number of iterations.

  16. 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)

  17. The High Aspect Ratio Design (HARD): A candidate ITER concept with improved technology phase performance

    International Nuclear Information System (INIS)

    Nevins, W.M.; Perkins, L.J.; Wesley, J.C.

    1992-10-01

    The High Aspect Ratio Design (HARD) International Thermonuclear Experimental Reactor (ITER) concept developed by the US ITER team is an alternate to the low-aspect-ratio ITER design developed by the ITER participants during the Conceptual Design Activity (CDA). The CDA design, referred to hereafter as ITER CDA, has an aspect ratio, A, of 2.79, a toroidal magnetic field, B T , of 4.85 T, and a plasma current, I p , of 22 MA for operation with an ignited plasma. In contrast, HARD employs higher aspect ratio, A = 4.0, higher toroidal field, B T = 7.11 T, and lower plasma current, I p = 14.8 MA for ignition operation. The cross sections of the two designs are compared in. The parameters and performance of HARD and ITER CDA for inductively driven ignition operation are compared in. The HARD parameters provide the same ignition performance (ignition margin evaluated against ITER-89P confinement scaling) as ITER CDA in a device with comparable size and cost. However, the reason for advancing HARD rather than ITER CDA as the ITER design concept is not inductively driven ignition performance but HARD's significantly enhanced potential to achieve the technology testing and steady-state operation goals of the ITER objectives with non-inductive current drive

  18. Tungsten covered graphite and copper elements and ITER-like actively cooled tungsten divertor plasma facing units for the WEST project

    International Nuclear Information System (INIS)

    Guilhem, D; Bucalossi, J; Burles, S; Corre, Y; Ferlay, F; Firdaouss, M; Languille, P; Lipa, M; Martinez, A; Missirlian, M; Proust, M; Richou, M; Samaille, F; Tsitrone, E

    2016-01-01

    After a brief introduction giving some insight of the WEST project, we present the three types of plasma facing units (PFUs) developed for the WEST project taking into account the envisaged main scenarios: (1) high power short pulse scenario (a few seconds) where the objective is to maximize the power handling of the PFUs, up to 20 MW m −2 , (2) high fluence scenario (a few 100 s) on actively cooled ITER-like tungsten (W) PFUs, up to 10 MW m −2 during 1000 s. For the graphite PFUs, the high heat flux tests have been done at GLADIS (ion beam test facility), and for the CuCrZr PFUs on the JUDITH (electron beam test facility). The tests were successful, as no damage occurred for the different load cases. This confirms that the modelling done during the design phase is appropriate to describe these PFUs. Series productions are expected to be achieved by the end of 2015 for the graphite and CuCrZr PFUs, and few ITER-like W PFUs are expected at the beginning of 2016. The lower divertor will be complemented with ITER-like W PFUs as soon as available from our partners so that different fabrication procedures could be evaluated in a real industrial process and a real tokamak environment. (paper)

  19. The 2 oC scenario-A sustainable world energy perspective

    International Nuclear Information System (INIS)

    Krewitt, Wolfram; Simon, Sonja; Graus, Wina; Teske, Sven; Zervos, Arthouros; Schaefer, Oliver

    2007-01-01

    A target-oriented scenario of future energy demand and supply is developed in a backcasting process. The main target is to reduce global CO 2 emissions to around 10 Gt/a in 2050, thus limiting global average temperature increase to 2 o C and preventing dangerous anthropogenic interference with the climate system. A 10-region energy system model is used for simulating global energy supply strategies. A review of sector and region-specific energy efficiency measures resulted in the specification of a global energy demand scenario incorporating strong energy efficiency measures. The corresponding supply scenario has been developed in an iterative process in close cooperation with stakeholders and regional counterparts from academia, NGOs and the renewable energy industry. The 2 o C scenario shows that renewable energy could provide as much as half of the world's energy needs by 2050. Developing countries can virtually stabilise their CO 2 emissions, while at the same time increasing energy consumption through economic growth. OECD countries will be able to reduce their emissions by up to 80%

  20. Design of the ITER high-frequency magnetic diagnostic coils

    International Nuclear Information System (INIS)

    Toussaint, M.; Testa, D.; Baluc, N.; Chavan, R.; Fournier, Y.; Lister, J.B.; Maeder, T.; Marmillod, P.; Sanchez, F.; Stoeck, M.

    2011-01-01

    This paper is an overview of work carried out on the design of the ITER high-frequency magnetic diagnostic coil (HF sensor). In the first part, the ITER requirements for the HF sensor are presented. In the second part, the ITER reference design of the HF sensor has been assessed and showed some potential weaknesses, which led us to the conclusion that alternative designs could usefully be examined. Several options have been explored, and are presented in the third part: (a) direct laser cutting a metallic tube, (b) stacking of plane windings manufactured from a tungsten plate by electrical discharge machining, (c) coil using the conventional spring manufacture. In the fourth part, sensors using the low temperature co-fired ceramic technology (LTCC) are presented: (d) monolithic 1D magnetic flux sensors based on LTCC technology, and (e) monolithic 3D magnetic flux sensors based on the same LTCC technology. The solution which showed the best results is the monolithic 3D magnetic flux sensor based on LTCC.

  1. Scenarios for Benefits Analysis of Energy Research, Development,Demonstration and Deployment

    Energy Technology Data Exchange (ETDEWEB)

    Gumerman, Etan; Marnay, Chris

    2005-09-07

    For at least the last decade, evaluation of the benefits of research, development, demonstration, and deployment (RD3) by the U.S. Department of Energy has been conducted using deterministic forecasts that unrealistically presume we can precisely foresee our future 10, 25,or even 50 years hence. This effort tries, in a modest way, to begin a process of recognition that the reality of our energy future is rather one rife with uncertainty. The National Energy Modeling System (NEMS) is used by the Department of Energy's Office of Energy Efficiency and Renewable Energy (EE) and Fossil Energy (FE) for their RD3 benefits evaluation. In order to begin scoping out the uncertainty in these deterministic forecasts, EE and FE designed two futures that differ significantly from the basic NEMS forecast. A High Fuel Price Scenario and a Carbon Cap Scenario were envisioned to forecast alternative futures and the associated benefits. Ernest Orlando Lawrence Berkeley National Laboratory (LBNL) implemented these scenarios into its version of NEMS,NEMS-LBNL, in late 2004, and the Energy Information Agency created six scenarios for FE in early 2005. The creation and implementation of the EE-FE scenarios are explained in this report. Both a Carbon Cap Scenario and a High Fuel Price Scenarios were implemented into the NEMS-LBNL. EIA subsequently modeled similar scenarios using NEMS. While the EIA and LBNL implementations were in some ways rather different, their forecasts do not significantly diverge. Compared to the Reference Scenario, the High Fuel Price Scenario reduces energy consumption by 4 percent in 2025, while in the EIA fuel price scenario (known as Scenario 4) reduction from its corresponding reference scenario (known as Scenario 0) in 2025 is marginal. Nonetheless, the 4 percent demand reduction does not lead to other cascading effects that would significantly differentiate the two scenarios. The LBNL and EIA carbon scenarios were mostly identical. The only major

  2. Computational models for electromagnetic transients in ITER vacuum vessel, cryostat and thermal shield

    International Nuclear Information System (INIS)

    Alekseev, A.; Arslanova, D.; Belov, A.; Belyakov, V.; Gapionok, E.; Gornikel, I.; Gribov, Y.; Ioki, K.; Kukhtin, V.; Lamzin, E.; Sugihara, M.; Sychevsky, S.; Terasawa, A.; Utin, Y.

    2013-01-01

    A set of detailed computational models are reviewed that covers integrally the system “vacuum vessel (VV), cryostat, and thermal shields (TS)” to study transient electromagnetics (EMs) in the ITER machine. The models have been developed in the course of activities requested and supervised by the ITER Organization. EM analysis is enabled for all ITER operational scenarios. The input data are derived from results of DINA code simulations. The external EM fields are modeled accurate to the input data description. The known magnetic shell approach can be effectively applied to simulate thin-walled structures of the ITER machine. Using an integral–differential formulation, a single unknown is determined within the shells in terms of the vector electric potential taken only at the nodes of a finite-element (FE) mesh of the conducting structures. As a result, the FE mesh encompasses only the system “VV + Cryostat + TS”. The 3D model requires much higher computational resources as compared to a shell model based on the equivalent approximation. The shell models have been developed for all principal conducting structures in the system “VV + Cryostat + TS” including regular ports and neutral beam ports. The structures are described in details in accordance with the latest design. The models have also been applied for simulations of EM transients in components of diagnostic systems and cryopumps and estimation of the 3D effects of the ITER structures on the plasma performance. The developed models have been elaborated and applied for the last 15 years to support the ITER design activities. The finalization of the ITER VV design enables this set of models to be considered ready to use in plasma-physics computations and the development of ITER simulators

  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. Nuclear Analysis of an ITER Blanket Module

    Science.gov (United States)

    Chiovaro, P.; Di Maio, P. A.; Parrinello, V.

    2013-08-01

    ITER blanket system is the reactor's plasma-facing component, it is mainly devoted to provide the thermal and nuclear shielding of the Vacuum Vessel and external ITER components, being intended also to act as plasma limiter. It consists of 440 individual modules which are located in the inboard, upper and outboard regions of the reactor. In this paper attention has been focused on to a single outboard blanket module located in the equatorial zone, whose nuclear response under irradiation has been investigated following a numerical approach based on the Monte Carlo method and adopting the MCNP5 code. The main features of this blanket module nuclear behaviour have been determined, paying particular attention to energy and spatial distribution of the neutron flux and deposited nuclear power together with the spatial distribution of its volumetric density. Moreover, the neutronic damage of the structural material has also been investigated through the evaluation of displacement per atom and helium and hydrogen production rates. Finally, an activation analysis has been performed with FISPACT inventory code using, as input, the evaluated neutron spectrum to assess the module specific activity and contact dose rate after irradiation under a specific operating scenario.

  5. Integrated modeling of plasma ramp-up in DIII-D ITER-like and high bootstrap current scenario discharges

    Science.gov (United States)

    Wu, M. Q.; Pan, C. K.; Chan, V. S.; Li, G. Q.; Garofalo, A. M.; Jian, X.; Liu, L.; Ren, Q. L.; Chen, J. L.; Gao, X.; Gong, X. Z.; Ding, S. Y.; Qian, J. P.; Cfetr Physics Team

    2018-04-01

    Time-dependent integrated modeling of DIII-D ITER-like and high bootstrap current plasma ramp-up discharges has been performed with the equilibrium code EFIT, and the transport codes TGYRO and ONETWO. Electron and ion temperature profiles are simulated by TGYRO with the TGLF (SAT0 or VX model) turbulent and NEO neoclassical transport models. The VX model is a new empirical extension of the TGLF turbulent model [Jian et al., Nucl. Fusion 58, 016011 (2018)], which captures the physics of multi-scale interaction between low-k and high-k turbulence from nonlinear gyro-kinetic simulation. This model is demonstrated to accurately model low Ip discharges from the EAST tokamak. Time evolution of the plasma current density profile is simulated by ONETWO with the experimental current ramp-up rate. The general trend of the predicted evolution of the current density profile is consistent with that obtained from the equilibrium reconstruction with Motional Stark effect constraints. The predicted evolution of βN , li , and βP also agrees well with the experiments. For the ITER-like cases, the predicted electron and ion temperature profiles using TGLF_Sat0 agree closely with the experimental measured profiles, and are demonstrably better than other proposed transport models. For the high bootstrap current case, the predicted electron and ion temperature profiles perform better in the VX model. It is found that the SAT0 model works well at high IP (>0.76 MA) while the VX model covers a wider range of plasma current ( IP > 0.6 MA). The results reported in this paper suggest that the developed integrated modeling could be a candidate for ITER and CFETR ramp-up engineering design modeling.

  6. Mechanical design aspects of a "Long-Waveguide" version of the remote steering ECRH upper port launcher for ITER

    NARCIS (Netherlands)

    Ronden, D. M. S.; van den Berg, M.; Bongers, W. A.; Elzendoorn, B. S. Q.; M. F. Graswinckel,; Lamers, B.; Van Nigtevecht, K.; Verhoeven, A. G. A.; Henderson, M. A.

    2008-01-01

    The current status of the mechanical design of the remote steering electron cyclotron resonance heating upper port launching system for ITER is presented. Although an alternative front steering launcher has now been selected as the reference design for ITER, the development of a remote steering

  7. 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

  8. 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

  9. Selective Iterative Waterfilling for Digital Subscriber Lines

    Directory of Open Access Journals (Sweden)

    Xu Yang

    2007-01-01

    Full Text Available This paper presents a high-performance, low-complexity, quasi-distributed dynamic spectrum management (DSM algorithm suitable for DSL systems. We analytically demonstrate that the rate degradation of the distributed iterative waterfilling (IW algorithm in near-far scenarios is caused by the insufficient utilization of all available frequency and power resources due to its nature of noncooperative game theoretic formulation. Inspired by this observation, we propose the selective IW (SIW algorithm that can considerably alleviate the performance degradation of IW by applying IW selectively to different groups of users over different frequency bands so that all the available resources can be fully utilized. For users, the proposed SIW algorithm needs at most times the complexity of the IW algorithm, and is much simpler than the centralized optimal spectrum balancing (OSB, while it can offer a rate performance much better than that of the IW and close to the maximum possible rate region computed by the OSB in realistic near-far DSL scenarios. Furthermore, its predominantly distributed structure makes it suitable for DSL implementation.

  10. Transport simulation of ITER [International Thermonuclear Engineering Reactor] startup

    International Nuclear Information System (INIS)

    Attenberger, S.E.; Houlberg, W.A.

    1989-01-01

    The present International Thermonuclear Engineering Reactor (ITER) reference configurations are the ''Technology Phase,'' in which the plasma current is maintained noninductively at a subignition density, and the ''Physics Phase,'' which is ignited but requires inductive maintenance of the current. The WHIST 1.5-D transport code is used to evaluate the volt-second requirements of both configurations. A slow current ramp (60-80's) is required for fixed-radius startup in ITER to avoid hollow current density profiles. To reach the operating point requires about 203 V·s for the Technology Phase (18 MA) and about 270 V·s for the Physics Phase (22 MA). The resistive losses can be reduced with expanding-radius startup. 5 refs., 4 figs

  11. Analysis of advanced European nuclear fuel cycle scenarios including transmutation and economical estimates

    International Nuclear Information System (INIS)

    Merino Rodriguez, I.; Alvarez-Velarde, F.; Martin-Fuertes, F.

    2013-01-01

    Four European fuel cycle scenarios involving transmutation options have been addressed from a point of view of resources utilization and economics. Scenarios include the current fleet using Light Water Reactor (LWR) technology and open fuel cycle (as a reference scenario), a full replacement of the initial fleet with Fast Reactors (FR) burning U-Pu MOX fuel and two fuel cycles with Minor Actinide (MA) transmutation in a fraction of the FR fleet or in dedicated Accelerator Driven Systems (ADS).Results reveal that all scenarios are feasible according to nuclear resources demand. Regarding the economic analysis, the estimations show an increase of LCOE - averaged over the whole period - with respect to the reference scenario of 20% for Pu management scenario and around 35% for both transmutation scenarios respectively.

  12. Analysis of advanced European nuclear fuel cycle scenarios including transmutation and economical estimates

    Energy Technology Data Exchange (ETDEWEB)

    Merino Rodriguez, I.; Alvarez-Velarde, F.; Martin-Fuertes, F.

    2013-07-01

    Four European fuel cycle scenarios involving transmutation options have been addressed from a point of view of resources utilization and economics. Scenarios include the current fleet using Light Water Reactor (LWR) technology and open fuel cycle (as a reference scenario), a full replacement of the initial fleet with Fast Reactors (FR) burning U-Pu MOX fuel and two fuel cycles with Minor Actinide (MA) transmutation in a fraction of the FR fleet or in dedicated Accelerator Driven Systems (ADS).Results reveal that all scenarios are feasible according to nuclear resources demand. Regarding the economic analysis, the estimations show an increase of LCOE - averaged over the whole period - with respect to the reference scenario of 20% for Pu management scenario and around 35% for both transmutation scenarios respectively.

  13. 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

  14. 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.

  15. 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.

  16. Feedback controlled, reactor relevant, high-density, high-confinement scenarios at ASDEX Upgrade

    Science.gov (United States)

    Lang, P. T.; Blanken, T. C.; Dunne, M.; McDermott, R. M.; Wolfrum, E.; Bobkov, V.; Felici, F.; Fischer, R.; Janky, F.; Kallenbach, A.; Kardaun, O.; Kudlacek, O.; Mertens, V.; Mlynek, A.; Ploeckl, B.; Stober, J. K.; Treutterer, W.; Zohm, H.; ASDEX Upgrade Team

    2018-03-01

    One main programme topic at the ASDEX Upgrade all-metal-wall tokamak is development of a high-density regime with central densities at reactor grade level while retaining high-confinement properties. This required development of appropriate control techniques capable of coping with the pellet tool, a powerful means of fuelling but one which presented challenges to the control system for handling of related perturbations. Real-time density profile control was demonstrated, raising the core density well above the Greenwald density while retaining the edge density in order to avoid confinement losses. Recently, a new model-based approach was implemented that allows direct control of the central density. Investigations focussed first on the N-seeding scenario owing to its proven potential to yield confinement enhancements. Combining pellets and N seeding was found to improve the divertor buffering further and enhance the operational range accessible. For core densities up to about the Greenwald density, a clear improvement with respect to the non-seeding reference was achieved; however, at higher densities this benefit is reduced. This behaviour is attributed to recurrence of an outward shift of the edge density profile, resulting in a reduced peeling-ballooning stability. This is similar to the shift seen during strong gas puffing, which is required to prevent impurity influx in ASDEX Upgrade. First tests indicate that highly-shaped plasma configurations like the ITER base-line scenario, respond very well to pellet injection, showing efficient fuelling with no measurable impact on the edge density profile.

  17. Iterative learning-based decentralized adaptive tracker for large-scale systems: a digital redesign approach.

    Science.gov (United States)

    Tsai, Jason Sheng-Hong; Du, Yan-Yi; Huang, Pei-Hsiang; Guo, Shu-Mei; Shieh, Leang-San; Chen, Yuhua

    2011-07-01

    In this paper, a digital redesign methodology of the iterative learning-based decentralized adaptive tracker is proposed to improve the dynamic performance of sampled-data linear large-scale control systems consisting of N interconnected multi-input multi-output subsystems, so that the system output will follow any trajectory which may not be presented by the analytic reference model initially. To overcome the interference of each sub-system and simplify the controller design, the proposed model reference decentralized adaptive control scheme constructs a decoupled well-designed reference model first. Then, according to the well-designed model, this paper develops a digital decentralized adaptive tracker based on the optimal analog control and prediction-based digital redesign technique for the sampled-data large-scale coupling system. In order to enhance the tracking performance of the digital tracker at specified sampling instants, we apply the iterative learning control (ILC) to train the control input via continual learning. As a result, the proposed iterative learning-based decentralized adaptive tracker not only has robust closed-loop decoupled property but also possesses good tracking performance at both transient and steady state. Besides, evolutionary programming is applied to search for a good learning gain to speed up the learning process of ILC. Copyright © 2011 ISA. Published by Elsevier Ltd. All rights reserved.

  18. Structural analysis of the ITER Vacuum Vessel regarding 2012 ITER Project-Level Loads

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, J.-M., E-mail: jean-marc.martinez@live.fr [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul lez Durance (France); Jun, C.H.; Portafaix, C.; Choi, C.-H.; Ioki, K.; Sannazzaro, G.; Sborchia, C. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul lez Durance (France); Cambazar, M.; Corti, Ph.; Pinori, K.; Sfarni, S.; Tailhardat, O. [Assystem EOS, 117 rue Jacquard, L' Atrium, 84120 Pertuis (France); Borrelly, S. [Sogeti High Tech, RE2, 180 rue René Descartes, Le Millenium – Bat C, 13857 Aix en Provence (France); Albin, V.; Pelletier, N. [SOM Calcul – Groupe ORTEC, 121 ancien Chemin de Cassis – Immeuble Grand Pré, 13009 Marseille (France)

    2014-10-15

    Highlights: • ITER Vacuum Vessel is a part of the first barrier to confine the plasma. • ITER Vacuum Vessel as Nuclear Pressure Equipment (NPE) necessitates a third party organization authorized by the French nuclear regulator to assure design, fabrication, conformance testing and quality assurance, i.e. Agreed Notified Body (ANB). • A revision of the ITER Project-Level Load Specification was implemented in April 2012. • ITER Vacuum Vessel Loads (seismic, pressure, thermal and electromagnetic loads) were summarized. • ITER Vacuum Vessel Structural Margins with regards to RCC-MR code were summarized. - Abstract: A revision of the ITER Project-Level Load Specification (to be used for all systems of the ITER machine) was implemented in April 2012. This revision supports ITER's licensing by accommodating requests from the French regulator to maintain consistency with the plasma physics database and our present understanding of plasma transients and electro-magnetic (EM) loads, to investigate the possibility of removing unnecessary conservatism in the load requirements and to review the list and definition of incidental cases. The purpose of this paper is to present the impact of this 2012 revision of the ITER Project-Level Load Specification (LS) on the ITER Vacuum Vessel (VV) loads and the main structural margins required by the applicable French code, RCC-MR.

  19. 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

  20. MHD equilibrium methods for ITER [International Thermonuclear Experimental Reactor] PF [poloidal field] coil design and systems analysis

    International Nuclear Information System (INIS)

    Strickler, D.J.; Galambos, J.D.; Peng, Y.K.M.

    1989-03-01

    Two versions of the Fusion Engineering Design Center (FEDC) free-boundary equilibrium code designed to computer the poloidal field (PF) coil current distribution of elongated, magnetically limited tokamak plasmas are demonstrated and applied to the systems analysis of the impact of plasma elongation on the design point of the International Thermonuclear Experimental Reactor (ITER). These notes were presented at the ITER Specialists' Meeting on the PF Coil System and Operational Scenario, held at the Max Planck Institute for Plasma Physics in Garching, Federal Republic of Germany, May 24--27, 1988. 8 refs., 6 figs., 4 tabs

  1. ITER-FEAT outline design report

    International Nuclear Information System (INIS)

    2001-01-01

    In July 1998 the ITER Parties were unable, for financial reasons, to proceed with construction of the ITER design proposed at that time, to meet the detailed technical objectives and target cost set in 1992. It was therefore decided to investigate options for the design of ITER with reduced technical objectives and with possibly decreased technical margins, whose target construction cost was one half that of the 1998 ITER design, while maintaining the overall programmatic objective. To identify designs that might meet the revised objectives, task forces involving the JCT and Home Teams met during 1998 and 1999 to analyse and compare a range of options for the design of such a device. This led at the end of 1999 to a single configuration for the ITER design with parameters considered to be the most credible consistent with technical limitations and the financial target, yet meeting fully the objectives with appropriate margins. This new design of ITER, called ''ITER-FEAT'', was submitted to the ITER Director to the ITER Parties as the ''ITER-FEAT Outline Design Report'' (ODR) in January 2000, at their meeting in Tokyo. The Parties subsequently conducted their domestic assessments of this report and fed the resulting comments back into the progressing design. The progress on the developing design was reported to the ITER Technical Advisory Committee (TAC) in June 2000 in the report ''Progress in Resolving Open Design Issues from the ODR'' alongside a report on Progress in Technology R and D for ITER. In addition, the progress in the ITER-FEAT Design and Validating R and D was reported to the ITER Parties. The ITER-FEAT design was subsequently approved by the governing body of ITER in Moscow in June 2000 as the basis for the preparation of the Final Design Report, recognising it as a single mature design for ITER consistent with its revised objectives. This volume contains the documents pertinent to the process described above. More detailed technical information

  2. From scenarios to domain models: processes and representations

    Science.gov (United States)

    Haddock, Gail; Harbison, Karan

    1994-03-01

    The domain specific software architectures (DSSA) community has defined a philosophy for the development of complex systems. This philosophy improves productivity and efficiency by increasing the user's role in the definition of requirements, increasing the systems engineer's role in the reuse of components, and decreasing the software engineer's role to the development of new components and component modifications only. The scenario-based engineering process (SEP), the first instantiation of the DSSA philosophy, has been adopted by the next generation controller project. It is also the chosen methodology of the trauma care information management system project, and the surrogate semi-autonomous vehicle project. SEP uses scenarios from the user to create domain models and define the system's requirements. Domain knowledge is obtained from a variety of sources including experts, documents, and videos. This knowledge is analyzed using three techniques: scenario analysis, task analysis, and object-oriented analysis. Scenario analysis results in formal representations of selected scenarios. Task analysis of the scenario representations results in descriptions of tasks necessary for object-oriented analysis and also subtasks necessary for functional system analysis. Object-oriented analysis of task descriptions produces domain models and system requirements. This paper examines the representations that support the DSSA philosophy, including reference requirements, reference architectures, and domain models. The processes used to create and use the representations are explained through use of the scenario-based engineering process. Selected examples are taken from the next generation controller project.

  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. Influence of Ultra-Low-Dose and Iterative Reconstructions on the Visualization of Orbital Soft Tissues on Maxillofacial CT.

    Science.gov (United States)

    Widmann, G; Juranek, D; Waldenberger, F; Schullian, P; Dennhardt, A; Hoermann, R; Steurer, M; Gassner, E-M; Puelacher, W

    2017-08-01

    Dose reduction on CT scans for surgical planning and postoperative evaluation of midface and orbital fractures is an important concern. The purpose of this study was to evaluate the variability of various low-dose and iterative reconstruction techniques on the visualization of orbital soft tissues. Contrast-to-noise ratios of the optic nerve and inferior rectus muscle and subjective scores of a human cadaver were calculated from CT with a reference dose protocol (CT dose index volume = 36.69 mGy) and a subsequent series of low-dose protocols (LDPs I-4: CT dose index volume = 4.18, 2.64, 0.99, and 0.53 mGy) with filtered back-projection (FBP) and adaptive statistical iterative reconstruction (ASIR)-50, ASIR-100, and model-based iterative reconstruction. The Dunn Multiple Comparison Test was used to compare each combination of protocols (α = .05). Compared with the reference dose protocol with FBP, the following statistically significant differences in contrast-to-noise ratios were shown (all, P ≤ .012) for the following: 1) optic nerve: LDP-I with FBP; LDP-II with FBP and ASIR-50; LDP-III with FBP, ASIR-50, and ASIR-100; and LDP-IV with FBP, ASIR-50, and ASIR-100; and 2) inferior rectus muscle: LDP-II with FBP, LDP-III with FBP and ASIR-50, and LDP-IV with FBP, ASIR-50, and ASIR-100. Model-based iterative reconstruction showed the best contrast-to-noise ratio in all images and provided similar subjective scores for LDP-II. ASIR-50 had no remarkable effect, and ASIR-100, a small effect on subjective scores. Compared with a reference dose protocol with FBP, model-based iterative reconstruction may show similar diagnostic visibility of orbital soft tissues at a CT dose index volume of 2.64 mGy. Low-dose technology and iterative reconstruction technology may redefine current reference dose levels in maxillofacial CT. © 2017 by American Journal of Neuroradiology.

  5. Impacts of high energy prices on long-term energy-economic scenarios for Germany

    Energy Technology Data Exchange (ETDEWEB)

    Krey, V.; Markewitz, P. [Research Center Juelich, Inst. of Energy Res., Systems Analysis and Technology Evaluation, Juelich (Germany); Horn, M. [DIW Berlin, Berlin (Germany); Matthes, C.; Graichen, V.; Harthan, R.O.; Repenning, J. [Oeko-Institut, Berlin (Germany)

    2007-05-15

    Prices of oil and other fossil fuels on global markets have reached a high level in recent years. These levels were not able to be reproduced on the basis of scenarios and prognoses that were published in the past. New scenarios, based on higher energy price trajectories, have appeared only recently. The future role of various energy carriers and technologies in energy-economic scenarios will greatly depend on the level of energy prices. Therefore, an analysis of the impact of high energy prices on long-term scenarios for Germany was undertaken. Based on a reference scenario with moderate prices, a series of consistent high price scenarios for primary and secondary energy carriers were developed. Two scenarios with (i) continuously rising price trajectories and (ii) a price shock with a price peak during the period 2010-15 and a subsequent decline to the reference level are analysed. Two types of models have been applied in the analysis. The IKARUS energy systems optimisation model covers the whole of the German energy system from primary energy supply down to the end-use sectors. Key results in both high price scenarios include a replacement of natural gas by hard coal and renewable energy sources in electricity and heat generation. Backstop technologies like coal liquefaction begin to play a role under such conditions. Up to 10% of final energy consumption is saved in the end-use sectors, with the residential and transport sector being the greatest contributors. Even without additional restrictions, CO{sub 2} emissions significantly drop in comparison to the reference scenario. The ELIAS electricity investment analysis model focuses on the power sector. In the reference scenario with current allocation rules in the emissions trading scheme, the CO{sub 2} emissions decrease relatively steadily. The development is characterised by the phaseout of nuclear energy which is counterweighted by the increase of renewable. In the high price scenario, the CO{sub 2

  6. Safety studies: Review of loss of plasma control transients in ITER with AINA 3.0 code

    International Nuclear Information System (INIS)

    Rivas, J.C.; Dies, J.

    2013-01-01

    Highlights: ► We have examined the methodology employed in the analysis of the “Loss of plasma transients in ITER” safety reference events. ► We have developed a new methodology based on the study of the plasma operating window. ► We have concluded that the combined effect of different perturbations should be studied also to determine the most severe transients. -- Abstract: The loss of plasma control events in ITER are safety cases investigated to give an upper bound of the worse effects foreseeable from a total failure of the plasma control function. In the past, conservative analyses based on simple 0D models for plasma balance equations and 1D models for wall heat transfer showed that a hypothetical scenario of first wall coolant tubes melting and subsequent entering of water in the vacuum vessel could not be totally excluded. AINA (Analyses of IN vessel Accidents) code is a safety code developed at Fusion Energy Engineering Laboratory (FEEL) in Barcelona. It uses a 0D–1D architecture, similar to that used for previous analyses of ITER loss of plasma control events. The results of this study show the simultaneous effect of two perturbations (overfuelling and overheating) over a plasma transient, and compare it with the isolated effects of each perturbation. It is shown that the combined effect can be more severe, and a method is outlined to locate the most dangerous transients over a nT diagram

  7. 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

  8. 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].

  9. Japanese contribution of ITER PF systems design during CDA

    International Nuclear Information System (INIS)

    Azumi, Masafumi; Hasegawa, Mitsuru; Kameari, Akihisa

    1992-03-01

    This report presents the brief summary of major part of Japanese contributions to the poloidal field coil systems design during conceptual design activity (CDA). The followings are studied; layout of poloidal field coil system, effects of plasma triangularity and current profile, alternative configurations such as semi-double null and single null divertor plasmas apart from the reference double null plasma, magnetic line tracing study for the detailed examination of the divertor plasmas, repulsive force on the center solenoid for the volt-second saving operation scenario, general criteria of vertical stability by passive structure, stability performance of two types of passive structure, saddle loop type and twin loop type, in-vessel coil current and voltage evaluation required for active control of plasma verrical position, TOLFEX method for the identification of plasma shape and position for the control of ITER plasma, control of horizontal plasma position during breakdown, slow control algorithm for the plasma shape, means to reduce the effective divertor peak heat load, separatrix sweeping both by SC coils and in-vessel coils, feasibility to generate the ergodic magnetic layer by in-vessel coils for the vertical position control, simulation of plasma dynamic behavior after the thermal quench by TSC code. Many of the results are not fully complete, while should be sufficient o demonstrate the feasibility of the CDA design. (J.P.N.)

  10. The European contribution to the procurement of the ITER Remote Handling systems

    International Nuclear Information System (INIS)

    Damiani, Carlo; Irving, Mike; Semeraro, Luigi

    2009-01-01

    Fusion for Energy (F4E) will manage the European in-kind contribution of various remote handling (RH) systems for the maintenance of ITER components: (i) the divertor cassette movers, end effectors, manipulator arms and tooling; (ii) 50% of the transfer casks, in particular the air transfer systems and some in-cask devices; (iii) the in-vessel viewing and metrology system (IVVS); (iv) the Neutral Beam (NB) Cell crane, manipulator arms, tooling, Caesium Oven replacement tooling, NB source installation/removal trolley, auxiliary vehicles. A wide range of technologies is involved: special monorail crane, movers, manipulator arms, pipe cutting/welding tooling, special cameras, laser-based metrology devices, control systems, virtual reality. An important aspect to consider is the resistance to radiation levels that range from max ∼10 KGy/h for IVVS down to ∼1 Gy/h for the RH devices operating in the NB cell. Given the unprecedented complexity of the ITER maintenance scenario, a development strategy is being implemented that includes prototyping and testing of RH subsystems before proceeding with the final production for ITER. This paper presents an overview of the various procurement packages, the status of development for each of them, the validation and procurement strategy, including issues like radiation resistance and standardisation policy, and the organisational and managerial challenges in relation with the complex ITER Organisation (IO).

  11. SNS Superconducting RF cavity modeling-iterative learning control

    CERN Document Server

    Kwon, S I; Wang, Y M

    2002-01-01

    The Spallation Neutron Source (SNS) Superconducting RF (SRF) linear accelerator is operated with a pulsed beam. For the SRF control system to track the repetitive electromagnetic field reference trajectory, both feedback and feedforward controllers have been proposed. The feedback controller is utilized to guarantee the closed loop system stability and the feedforward controller is used to improve the tracking performance for the repetitive reference trajectory and to suppress repetitive disturbances. As the iteration number increases, the feedforward controller decreases the tracking error. Numerical simulations demonstrate that inclusion of the feedforward controller significantly improves the control system performance over its performance with just the feedback controller.

  12. SNS Superconducting RF cavity modeling-iterative learning control

    International Nuclear Information System (INIS)

    Kwon, S.-I.; Regan, Amy; Wang, Y.-M.

    2002-01-01

    The Spallation Neutron Source (SNS) Superconducting RF (SRF) linear accelerator is operated with a pulsed beam. For the SRF control system to track the repetitive electromagnetic field reference trajectory, both feedback and feedforward controllers have been proposed. The feedback controller is utilized to guarantee the closed loop system stability and the feedforward controller is used to improve the tracking performance for the repetitive reference trajectory and to suppress repetitive disturbances. As the iteration number increases, the feedforward controller decreases the tracking error. Numerical simulations demonstrate that inclusion of the feedforward controller significantly improves the control system performance over its performance with just the feedback controller

  13. 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

  14. Design and rescue scenario of common repair equipment for in-vessel components in ITER hot cell

    International Nuclear Information System (INIS)

    Kakudate, Satoshi; Takeda, Nobukazu; Nakahira, Masataka; Shibanuma, Kiyoshi

    2006-06-01

    Transportation of the in-vessel components to be repaired in the ITER hot cell is carried by two kinds of transporters, i.e., overhead cranes and floor vehicles. The access area for repair operations in the hot cell is duplicated by these transporters. Clear sharing of the respective roles of these transporters with the minimum duplication is therefore useful for rationalization. The overhead cranes, which are independently installed in the respective cells in the hot sell, cannot pass through the components to be repaired between cells, i.e., receiving cell and refurbishment cell as an example. If the floor vehicle with simple mechanisms can cover the inaccessible area for the overhead cranes, a global transporter system in the hot cell will be simplified and the reliability will be increased. Based on this strategy, the overhead crane and floor vehicle concepts are newly proposed. The overhead crane has an adapter for change of the end-effectors, which can be easily changed, to grasp many kinds of components to be repaired. The floor vehicle, which is equipped with wheel mechanisms for transportation, is just to pass through the components between cells with only straight (linear) motion on the floor. The simple wheel mechanism can solve the spread of the dust, which is the critical issue of the original air bearing mechanism for traveling in the 2001 FDR design. Rescue scenarios and procedures in the hot cell are also studied in this report. The proposed rescue crane has major two functions for rescue operations of the hot cell facility, i.e., one for the overhead crane and the other for refurbishment equipment such as workstation for divertor repair. The rescue of the faulty overhead crane is carried out using the rescue tool installed on the rescue crane or directly traveled by pushing/pulling by the rescue crane after docking on the faulty overhead crane. For the rescue of the workstation, the rescue crane consists of a telescopic manipulator (maximum length

  15. Nuclear systems and testing programs for ITER. Progress report for FY 1998

    International Nuclear Information System (INIS)

    1998-01-01

    entitled Three-Dimensional Calculations of ITER Building Dose Rate Profiles and Assessment of Accessibility Inside the Building During Operation and After Shutdown of ITER can be located through ITER Reference Number of ITER Task S 62 TD 12, ID No: D325 ITER/US/98/S62TD12-D325 UCLA-FNT-100 UCLA-ENG-98-190

  16. 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

  17. Design analysis of the ITER divertor

    International Nuclear Information System (INIS)

    Samuelli, G.; Marin, A.; Roccella, M.; Lucca, F.; Merola, M.; Riccardi, B.; Petrizzi, L.; Villari, R.

    2007-01-01

    The divertor is one of the most challenging components of the ITER machine. Its function is to reduce the impurity in the plasma and consists essentially of two parts: the plasma facing components (PFCs) and a massive support structure called the cassette body (CB). Considerable R and D effort (developed by EFDA CSU GARCHING and the ITER International Team together with the EU Associations and the EU Industries) has been spent in designing divertor components capable of withstanding the expected electromagnetic (EM) loads and to take into account the latest ITER design conditions. In support of such efforts extensive and very detailed Neutronic, Thermal, EM and Structural analyses have been performed. A summary of the analyses performed will be presented. One of the main result is a typical exercise of integration between the different kind of analyses and the importance of keeping the consistency between the different assumptions and simplifications. The models used for the numerical analyses include a detailed geometrical description of the CB, the inlet, outlet hydraulic manifolds, the CB to vacuum vessel locking system and three configurations of the PFU. The effect of electrical bridging, both in poloidal and toroidal direction, of the PFU castellation, due to a possible melting at the W mono-block or tiles, occurring during the plasma disruptions, has been analyzed. For all these configurations 2 VDE scenarios including the effect of the Toroidal Field Variation and the HaloCurrent with the related out of plane induced EM forces have been extensively analyzed and a detailed poloidal and radial distribution of the nuclear heating has been used for the neutronic flux on the divertor components. The aim of this activity is to produce a comprehensive design and assessment of the ITER divertor via: -The estimation of the neutronic heat deposition and shielding capability; -The calculation of the related thermal and mechanical effects and the comparison of the

  18. Design analysis of the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Samuelli, G.; Marin, A.; Roccella, M.; Lucca, F. [L.T. Calcoli SaS, Merate (Lecco) (Italy); Merola, M. [ITER Team, Cadarache (France); Riccardi, B. [EFDA CSU Garching (Germany); Petrizzi, L.; Villari, R. [CRE ENEA sulla Fusione Frascati, Roma (Italy)

    2007-07-01

    The divertor is one of the most challenging components of the ITER machine. Its function is to reduce the impurity in the plasma and consists essentially of two parts: the plasma facing components (PFCs) and a massive support structure called the cassette body (CB). Considerable R and D effort (developed by EFDA CSU GARCHING and the ITER International Team together with the EU Associations and the EU Industries) has been spent in designing divertor components capable of withstanding the expected electromagnetic (EM) loads and to take into account the latest ITER design conditions. In support of such efforts extensive and very detailed Neutronic, Thermal, EM and Structural analyses have been performed. A summary of the analyses performed will be presented. One of the main result is a typical exercise of integration between the different kind of analyses and the importance of keeping the consistency between the different assumptions and simplifications. The models used for the numerical analyses include a detailed geometrical description of the CB, the inlet, outlet hydraulic manifolds, the CB to vacuum vessel locking system and three configurations of the PFU. The effect of electrical bridging, both in poloidal and toroidal direction, of the PFU castellation, due to a possible melting at the W mono-block or tiles, occurring during the plasma disruptions, has been analyzed. For all these configurations 2 VDE scenarios including the effect of the Toroidal Field Variation and the HaloCurrent with the related out of plane induced EM forces have been extensively analyzed and a detailed poloidal and radial distribution of the nuclear heating has been used for the neutronic flux on the divertor components. The aim of this activity is to produce a comprehensive design and assessment of the ITER divertor via: -The estimation of the neutronic heat deposition and shielding capability; -The calculation of the related thermal and mechanical effects and the comparison of the

  19. Iterative and iterative-noniterative integral solutions in 3-loop massive QCD calculations

    International Nuclear Information System (INIS)

    Ablinger, J.; Radu, C.S.; Schneider, C.; Behring, A.; Imamoglu, E.; Van Hoeij, M.; Von Manteuffel, A.; Raab, C.G.

    2017-11-01

    Various of the single scale quantities in massless and massive QCD up to 3-loop order can be expressed by iterative integrals over certain classes of alphabets, from the harmonic polylogarithms to root-valued alphabets. Examples are the anomalous dimensions to 3-loop order, the massless Wilson coefficients and also different massive operator matrix elements. Starting at 3-loop order, however, also other letters appear in the case of massive operator matrix elements, the so called iterative non-iterative integrals, which are related to solutions based on complete elliptic integrals or any other special function with an integral representation that is definite but not a Volterra-type integral. After outlining the formalism leading to iterative non-iterative integrals,we present examples for both of these cases with the 3-loop anomalous dimension γ (2) qg and the structure of the principle solution in the iterative non-interative case of the 3-loop QCD corrections to the ρ-parameter.

  20. Iterative and iterative-noniterative integral solutions in 3-loop massive QCD calculations

    Energy Technology Data Exchange (ETDEWEB)

    Ablinger, J.; Radu, C.S.; Schneider, C. [Johannes Kepler Univ., Linz (Austria). Research Inst. for Symbolic Computation (RISC); Behring, A. [RWTH Aachen Univ. (Germany). Inst. fuer Theoretische Teilchenphysik und Kosmologie; Bluemlein, J.; Freitas, A. de [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany); Imamoglu, E.; Van Hoeij, M. [Florida State Univ., Tallahassee, FL (United States). Dept. of Mathematics; Von Manteuffel, A. [Michigan State Univ., East Lansing, MI (United States). Dept. of Physics and Astronomy; Raab, C.G. [Johannes Kepler Univ., Linz (Austria). Inst. for Algebra

    2017-11-15

    Various of the single scale quantities in massless and massive QCD up to 3-loop order can be expressed by iterative integrals over certain classes of alphabets, from the harmonic polylogarithms to root-valued alphabets. Examples are the anomalous dimensions to 3-loop order, the massless Wilson coefficients and also different massive operator matrix elements. Starting at 3-loop order, however, also other letters appear in the case of massive operator matrix elements, the so called iterative non-iterative integrals, which are related to solutions based on complete elliptic integrals or any other special function with an integral representation that is definite but not a Volterra-type integral. After outlining the formalism leading to iterative non-iterative integrals,we present examples for both of these cases with the 3-loop anomalous dimension γ{sup (2)}{sub qg} and the structure of the principle solution in the iterative non-interative case of the 3-loop QCD corrections to the ρ-parameter.

  1. Particle Reduction Strategies - PAREST. Emissions in PAREST reference scenario 2005-2020; Strategien zur Verminderung der Feinstaubbelastung - PAREST. Emissionen im PAREST-Referenzszenario 2005-2020

    Energy Technology Data Exchange (ETDEWEB)

    Joerss, Wolfram [Institut fuer Zukunftsstudien und Technologiebewertung gGmbH (IZT), Berlin (Germany); Kugler, Ulrike; Theloke, Jochen [Stuttgart Univ. (Germany). Inst. fuer Energiewirtschaft und Rationelle Energieanwendung (IER)

    2013-06-15

    In the research project PAREST emission scenarios to 2020 for particulate matter (PM{sub 10} und PM{sub 2,5}), and aerosol precursors SO{sub 2}, NO{sub x}, NH{sub 3} and NMVOC were prepared for Germany and Europe, mitigation measures evaluated and the results used to model the PM pollution load for Germany. In this sub-report of PAREST Germany's aggregate emissions totals of the considered pollutants are presented in the reference scenario. [German] Im Forschungsvorhaben PAREST wurden Emissionsszenarien bis 2020 fuer Feinstaub (PM{sub 10} und PM{sub 2,5}) und die Aerosol-Vorlaeufersubstanzen SO{sub 2}, NO{sub x}, NH{sub 3} und NMVOC fuer Deutschland und Europa erstellt, Minderungsmassnahmen evaluiert und daraus die PM-Immissionsbelastung fuer Deutschland modelliert. In diesem Teilbericht zu PAREST werden die deutschlandweit aggregierten Emissionssummen der betrachteten Schadstoffe im Referenzszenario vorgestellt.

  2. 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

  3. Numerical study of the ITER divertor plasma with the B2-EIRENE code package

    Energy Technology Data Exchange (ETDEWEB)

    Kotov, V.; Reiter, D. [Forschungszentrum Juelich (DE). Inst. fuer Energieforschung (IEF), Plasmaphysik (IEF-4); Kukushkin, A.S. [ITER International Team, Cadarache (France)

    2007-11-15

    The problem of plasma-wall interaction and impurity control is one of the remaining critical issues for development of an industrial energy source based on nuclear fusion of light isotopes. In this field sophisticated integrated numerical tools are widely used both for the analysis of current experiments and for predictions guiding future device design. The present work is dedicated to the numerical modelling of the edge plasma region in divertor configurations of large-scale tokamak fusion devices. A well established software tool for this kind of modelling is the B2-EIRENE code. It was originally developed for a relatively hot (>> 10 eV) ''high recycling divertor''. It did not take into account a number of physical effects which can be potentially important for ''detached conditions'' (cold, - several eV, - high density, - {approx} 10{sup 21} m{sup -3}, - plasma) typical for large tokamak devices. This is especially critical for the modelling of the divertor plasma of ITER: an international project of an experimental tokamak fusion reactor to be built in Cadarache, France by 2016. This present work is devoted to a major upgrade of the B2-EIRENE package, which is routinely used for ITER modelling, essentially with a significantly revised version of EIRENE: the Monte-Carlo neutral transport code. The main part of the thesis address three major groups of the new physical effects which have been added to the model in frame of this work: the neutral-neutral collisions, the up-to date hydrogen molecular reaction kinetics and the line radiation transport. The impact of the each stage of the upgrade on the self-consistent (between plasma, the neutral gas and the radiation field) solution for the reference ITER case is analysed. The strongest effect is found to be due to the revised molecular collision kinetics, in particular due to hitherto neglected elastic collisions of hydrogen molecules with ions. The newly added non

  4. Numerical study of the ITER divertor plasma with the B2-EIRENE code package

    International Nuclear Information System (INIS)

    Kotov, V.; Reiter, D.; Kukushkin, A.S.

    2007-11-01

    The problem of plasma-wall interaction and impurity control is one of the remaining critical issues for development of an industrial energy source based on nuclear fusion of light isotopes. In this field sophisticated integrated numerical tools are widely used both for the analysis of current experiments and for predictions guiding future device design. The present work is dedicated to the numerical modelling of the edge plasma region in divertor configurations of large-scale tokamak fusion devices. A well established software tool for this kind of modelling is the B2-EIRENE code. It was originally developed for a relatively hot (>> 10 eV) ''high recycling divertor''. It did not take into account a number of physical effects which can be potentially important for ''detached conditions'' (cold, - several eV, - high density, - ∼ 10 21 m -3 , - plasma) typical for large tokamak devices. This is especially critical for the modelling of the divertor plasma of ITER: an international project of an experimental tokamak fusion reactor to be built in Cadarache, France by 2016. This present work is devoted to a major upgrade of the B2-EIRENE package, which is routinely used for ITER modelling, essentially with a significantly revised version of EIRENE: the Monte-Carlo neutral transport code. The main part of the thesis address three major groups of the new physical effects which have been added to the model in frame of this work: the neutral-neutral collisions, the up-to date hydrogen molecular reaction kinetics and the line radiation transport. The impact of the each stage of the upgrade on the self-consistent (between plasma, the neutral gas and the radiation field) solution for the reference ITER case is analysed. The strongest effect is found to be due to the revised molecular collision kinetics, in particular due to hitherto neglected elastic collisions of hydrogen molecules with ions. The newly added non-linear effects (neutral-neutral collisions, radiation opacity

  5. Enhanced Confinement Scenarios Without Large Edge Localized Modes in Tokamaks: Control, Performance, and Extrapolability Issues for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Maingi, R [PPPL

    2014-07-01

    Large edge localized modes (ELMs) typically accompany good H-mode confinement in fusion devices, but can present problems for plasma facing components because of high transient heat loads. Here the range of techniques for ELM control deployed in fusion devices is reviewed. The two baseline strategies in the ITER baseline design are emphasized: rapid ELM triggering and peak heat flux control via pellet injection, and the use of magnetic perturbations to suppress or mitigate ELMs. While both of these techniques are moderately well developed, with reasonable physical bases for projecting to ITER, differing observations between multiple devices are also discussed to highlight the needed community R & D. In addition, recent progress in ELM-free regimes, namely Quiescent H-mode, I-mode, and Enhanced Pedestal H-mode is reviewed, and open questions for extrapolability are discussed. Finally progress and outstanding issues in alternate ELM control techniques are reviewed: supersonic molecular beam injection, edge electron cyclotron heating, lower hybrid heating and/or current drive, controlled periodic jogs of the vertical centroid position, ELM pace-making via periodic magnetic perturbations, ELM elimination with lithium wall conditioning, and naturally occurring small ELM regimes.

  6. AgMIP Climate Data and Scenarios for Integrated Assessment. Chapter 3

    Science.gov (United States)

    Ruane, Alexander C.; Winter, Jonathan M.; McDermid, Sonali P.; Hudson, Nicholas I.

    2015-01-01

    Climate change presents a great challenge to the agricultural sector as changes in precipitation, temperature, humidity, and circulation patterns alter the climatic conditions upon which many agricultural systems rely. Projections of future climate conditions are inherently uncertain owing to a lack of clarity on how society will develop, policies that may be implemented to reduce greenhouse-gas (GHG) emissions, and complexities in modeling the atmosphere, ocean, land, cryosphere, and biosphere components of the climate system. Global climate models (GCMs) are based on well-established physics of each climate component that enable the models to project climate responses to changing GHG concentration scenarios (Stocker et al., 2013).The most recent iteration of the Coupled Model Intercomparison Project (CMIP5; Taylor et al., 2012) utilized representative concentration pathways (RCPs) to cover the range of plausible GHG concentrations out past the year 2100, with RCP8.5 representing an extreme scenario and RCP4.5 representing a lower concentrations scenario (Moss et al., 2010).

  7. Iterative-Transform Phase Retrieval Using Adaptive Diversity

    Science.gov (United States)

    Dean, Bruce H.

    2007-01-01

    A phase-diverse iterative-transform phase-retrieval algorithm enables high spatial-frequency, high-dynamic-range, image-based wavefront sensing. [The terms phase-diverse, phase retrieval, image-based, and wavefront sensing are defined in the first of the two immediately preceding articles, Broadband Phase Retrieval for Image-Based Wavefront Sensing (GSC-14899-1).] As described below, no prior phase-retrieval algorithm has offered both high dynamic range and the capability to recover high spatial-frequency components. Each of the previously developed image-based phase-retrieval techniques can be classified into one of two categories: iterative transform or parametric. Among the modifications of the original iterative-transform approach has been the introduction of a defocus diversity function (also defined in the cited companion article). Modifications of the original parametric approach have included minimizing alternative objective functions as well as implementing a variety of nonlinear optimization methods. The iterative-transform approach offers the advantage of ability to recover low, middle, and high spatial frequencies, but has disadvantage of having a limited dynamic range to one wavelength or less. In contrast, parametric phase retrieval offers the advantage of high dynamic range, but is poorly suited for recovering higher spatial frequency aberrations. The present phase-diverse iterative transform phase-retrieval algorithm offers both the high-spatial-frequency capability of the iterative-transform approach and the high dynamic range of parametric phase-recovery techniques. In implementation, this is a focus-diverse iterative-transform phaseretrieval algorithm that incorporates an adaptive diversity function, which makes it possible to avoid phase unwrapping while preserving high-spatial-frequency recovery. The algorithm includes an inner and an outer loop (see figure). An initial estimate of phase is used to start the algorithm on the inner loop, wherein

  8. Solving large test-day models by iteration on data and preconditioned conjugate gradient.

    Science.gov (United States)

    Lidauer, M; Strandén, I; Mäntysaari, E A; Pösö, J; Kettunen, A

    1999-12-01

    A preconditioned conjugate gradient method was implemented into an iteration on a program for data estimation of breeding values, and its convergence characteristics were studied. An algorithm was used as a reference in which one fixed effect was solved by Gauss-Seidel method, and other effects were solved by a second-order Jacobi method. Implementation of the preconditioned conjugate gradient required storing four vectors (size equal to number of unknowns in the mixed model equations) in random access memory and reading the data at each round of iteration. The preconditioner comprised diagonal blocks of the coefficient matrix. Comparison of algorithms was based on solutions of mixed model equations obtained by a single-trait animal model and a single-trait, random regression test-day model. Data sets for both models used milk yield records of primiparous Finnish dairy cows. Animal model data comprised 665,629 lactation milk yields and random regression test-day model data of 6,732,765 test-day milk yields. Both models included pedigree information of 1,099,622 animals. The animal model ¿random regression test-day model¿ required 122 ¿305¿ rounds of iteration to converge with the reference algorithm, but only 88 ¿149¿ were required with the preconditioned conjugate gradient. To solve the random regression test-day model with the preconditioned conjugate gradient required 237 megabytes of random access memory and took 14% of the computation time needed by the reference algorithm.

  9. 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

  10. Development of in-vessel neutron flux monitor equipped with microfission chambers to withstand the extreme ITER environment

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, Masao, E-mail: ishikawa.masao@jaea.go.jp; Takeda, Keigo; Itami, Kiyoshi

    2016-11-01

    Highlights: • The in-vessel components of MFC system must withstand the extreme ITER environment. • To verify this, the thermal cycle test and the vibration tests were conducted. • Both tests were conducted under much severer conditions than ITER environment. • Soundness verification tests after the tests indicated that no problemswere found. • It is shown that the in-vessel component is sufficiently robust ITER environment. - Abstract: Via thermal cycling and vibration tests, this study aims to demonstrate that the in-vessel components of the microfission chamber (MFC) system can withstand the extreme International Thermonuclear Experimental Reactor (ITER) environment. In thermal cycle tests, the signal cable of the device was bent into a smaller radius and it was given more bends than those in its actual configuration within ITER. A faster rate of temperature change than that under the typical ITER baking scenario was then imposed on in-vessel components. For the vibration tests, strong 10 G vibrational accelerations with frequencies ranging from 30 Hz to 2000 Hz were imposed to the detector and the connector of the in-vessel components to simulate various types of electromagnetic events. Soundness verification tests of the in-vessel components conducted after thermal cycling and vibration testing indicated that problems related to the signal transmission cable functioning were not found. Thus, it was demonstrated that the in-vessel components of the MFC can withstand the extreme environment within ITER.

  11. Failure Mode and Effect Analysis for remote handling transfer systems of ITER

    International Nuclear Information System (INIS)

    Pinna, T.; Caporali, R.; Tesini, A.

    2008-01-01

    A Failure Mode and Effect Analysis (FMEA) at component level was done to study safety-relevant implications arising from possible failures in performing remote handling (RH) operations at ITER facility . Autonomous air cushion transporter, pallet, sealed casks and tractor movers needed for port plug mounting/dismantling operation were analysed. For each sub-system, the breakdown of significant components was outlined and, for each component, possible failure modes have been investigated pointing out possible causes, possible actions to prevent the causes, consequences and actions to prevent or mitigate consequences. Off-normal events which may result in hazardous consequences to the public and the environment have been defined as Postulated Initiating Events (PIEs). Two safety-relevant PIEs have been defined by assessing elementary failures related to the analysed system. Each PIE has been discussed in order to qualitatively identify accident sequences arising from each of them. As an output of this FMEA study, possible incidental scenarios, where the intervention of rescue RH equipments is required to overcome critical situations determined by fault of RH components, were defined as well. Being rescue scenarios of main concern for ITER remote handling activities, such families could be helpful in defining the design requirements of port handling systems in general and on RH transfer system in particular. Furthermore, they could be useful in defining casks and vehicles to be used for rescue activities

  12. ITER Construction--Plant System Integration

    International Nuclear Information System (INIS)

    Tada, E.; Matsuda, S.

    2009-01-01

    This brief paper introduces how the ITER will be built in the international collaboration. The ITER Organization plays a central role in constructing ITER and leading it into operation. Since most of the ITER components are to be provided in-kind from the member countries, integral project management should be scoped in advance of real work. Those include design, procurement, system assembly, testing, licensing and commissioning of ITER.

  13. 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

  14. 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

  15. Progress of IRSN R&D on ITER Safety Assessment

    Science.gov (United States)

    Van Dorsselaere, J. P.; Perrault, D.; Barrachin, M.; Bentaib, A.; Gensdarmes, F.; Haeck, W.; Pouvreau, S.; Salat, E.; Seropian, C.; Vendel, J.

    2012-08-01

    The French "Institut de Radioprotection et de Sûreté Nucléaire" (IRSN), in support to the French "Autorité de Sûreté Nucléaire", is analysing the safety of ITER fusion installation on the basis of the ITER operator's safety file. IRSN set up a multi-year R&D program in 2007 to support this safety assessment process. Priority has been given to four technical issues and the main outcomes of the work done in 2010 and 2011 are summarized in this paper: for simulation of accident scenarios in the vacuum vessel, adaptation of the ASTEC system code; for risk of explosion of gas-dust mixtures in the vacuum vessel, adaptation of the TONUS-CFD code for gas distribution, development of DUST code for dust transport, and preparation of IRSN experiments on gas inerting, dust mobilization, and hydrogen-dust mixtures explosion; for evaluation of the efficiency of the detritiation systems, thermo-chemical calculations of tritium speciation during transport in the gas phase and preparation of future experiments to evaluate the most influent factors on detritiation; for material neutron activation, adaptation of the VESTA Monte Carlo depletion code. The first results of these tasks have been used in 2011 for the analysis of the ITER safety file. In the near future, this R&D global programme may be reoriented to account for the feedback of the latter analysis or for new knowledge.

  16. 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

  17. 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)

  18. Towards Current Profile Control in ITER: Potential Approaches and Research Needs

    Science.gov (United States)

    Schuster, E.; Barton, J. E.; Wehner, W. P.

    2014-10-01

    Many challenging plasma control problems still need to be addressed in order for the ITER Plasma Control System (PCS) to be able to successfully achieve the ITER project goals. For instance, setting up a suitable toroidal current density profile is key for one possible advanced scenario characterized by noninductive sustainment of the plasma current and steady-state operation. The nonlinearity and high dimensionality exhibited by the plasma demand a model-based current-profile control synthesis procedure that can accommodate this complexity through embedding the known physics within the design. The development of a model capturing the dynamics of the plasma relevant for control design enables not only the design of feedback controllers for regulation or tracking but also the design of optimal feedforward controllers for a systematic model-based approach to scenario planning, the design of state estimators for a reliable real-time reconstruction of the plasma internal profiles based on limited and noisy diagnostics, and the development of a fast predictive simulation code for closed-loop performance evaluation before implementation. Progress towards control-oriented modeling of the current profile evolution and associated control design has been reported following both data-driven and first-principles-driven approaches. An overview of these two approaches will be provided, as well as a discussion on research needs associated with each one of the model applications described above. Supported by the US Department of Energy under DE-SC0001334 and DE-SC0010661.

  19. 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

  20. Articulated inspection arm for ITER, a demonstration in the Tore Supra tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Cordier, J.J.; Gargiulo, L.; Grisolia, C.; Samaille, F. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Friconneau, J.P.; Perrot, Y. [CEA Fontenay-aux-Roses, LIST Robotics and Interactive Systems Unit, 92 (France); Palmer, J.D. [Max-Planck-Institut fuer Plasmaphysik Boltzmannstr.2, Garching (Germany)

    2003-07-01

    The aim of this program is to demonstrate for ITER the feasibility of an in-vessel remote handling inspection using a long reach, limited payload carrier (1 to 10 kg) for penetration of the ITER chamber through the openings. This device is dedicated to close inspection of the Plasma Facing Components (PFC). An articulated demonstrator called articulated inspection arm (AIA) has been manufactured. A feasibility study of a full AIA operation in Tore Supra was performed, taking into account ITER reference requirements. A scale one demonstration of the AIA under ITER relevant condition is feasible on Tore Supra and would give significant improvement in research results for ITER remote Handling equipment. The test of the AIA demonstrator behaviour is foreseen in 2005 in real Tokamak conditions. The paper presents the full robot concept, the results of the first test campaign, the AIA new design and its integration on Tore Supra. Several potential uses of the AIA for the in vessel components inspection are being studied such as PFC visual inspection, water loop leak testing, laser ablation for wall detritiation and carbon dust and flakes removal are foreseen as utilities to be placed at the AIA head. These various systems are described in the paper.

  1. Articulated inspection arm for ITER, a demonstration in the Tore Supra tokamak

    International Nuclear Information System (INIS)

    Cordier, J.J.; Gargiulo, L.; Grisolia, C.; Samaille, F.; Palmer, J.D.

    2003-01-01

    The aim of this program is to demonstrate for ITER the feasibility of an in-vessel remote handling inspection using a long reach, limited payload carrier (1 to 10 kg) for penetration of the ITER chamber through the openings. This device is dedicated to close inspection of the Plasma Facing Components (PFC). An articulated demonstrator called articulated inspection arm (AIA) has been manufactured. A feasibility study of a full AIA operation in Tore Supra was performed, taking into account ITER reference requirements. A scale one demonstration of the AIA under ITER relevant condition is feasible on Tore Supra and would give significant improvement in research results for ITER remote Handling equipment. The test of the AIA demonstrator behaviour is foreseen in 2005 in real Tokamak conditions. The paper presents the full robot concept, the results of the first test campaign, the AIA new design and its integration on Tore Supra. Several potential uses of the AIA for the in vessel components inspection are being studied such as PFC visual inspection, water loop leak testing, laser ablation for wall detritiation and carbon dust and flakes removal are foreseen as utilities to be placed at the AIA head. These various systems are described in the paper

  2. Structural analysis of ITER multi-purpose deployer

    International Nuclear Information System (INIS)

    Manuelraj, Manoah Stephen; Dutta, Pramit; Gotewal, Krishan Kumar; Rastogi, Naveen; Tesini, Alessandro; Choi, Chang-Hwan

    2016-01-01

    Highlights: • System modelling for structural analysis of the Multi-Purpose Deployer (MPD). • Finite element modeling of the Multi-Purpose Deployer (MPD). • Static, modal and seismic response analysis of the Multi-Purpose Deployer (MPD). • Iterative structural analysis and design update to satisfy the structural criteria. • Modal analysis for various kinematic configurations. • Reaction force calculations on the interfacing systems. - Abstract: The Multi-Purpose Deployer (MPD) is a general purpose ITER in-vessel remote handling (RH) system. The main handling equipment, known as the MPD Transporter, consists of a series of linked bodies, which provide anchoring to the vacuum vessel port and an articulated multi-degree of freedom motion to perform various in-vessel maintenance tasks. During the in-vessel operations, the structural integrity of the system should be guaranteed against various operational and seismic loads. This paper presents the structural analysis results of the concept design of the MPD Transporter considering the seismic events. Static structural, modal and frequency response spectrum analyses have been performed to verify the structural integrity of the system, and to provide reaction forces to the interfacing systems such as vacuum vessel and cask. Iterative analyses and design updates are carried out based on the reference design of the system to improve the structural behavior of the system. The frequency responses of the system in various kinematics and payloads are assessed.

  3. Hydrogen/hydrocarbon explosions in the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Goranson, P.L.

    1992-01-01

    The consequences of H 2 /hydrocarbon detonations in the vacuum vessel (torus) of the International Thermonuclear Experimental Reactor (ITER) have been studied. The most likely scenario for such a detonation involves a water leak into the torus and a vent of the torus to atmosphere, permitting the formation of an explosive fuel-air mixture. The generation of fuel gases and possible sources of air or oxygen are reviewed, and the severity and effects of specific fuel-air mixture explosions are evaluated. Detonation or deflagration of an explosive mixture could result in pressures exceeding the maximum allowable torus pressure. Further studies to examine the design details and develop an event-tree study of events following a gas detonation are recommended

  4. Prospective performances in JT-60SA towards the ITER and DEMO relevant plasmas

    International Nuclear Information System (INIS)

    Tamai, H.; Fujita, T.; Kikuchi, M.

    2006-01-01

    JT-60SA, the former JT-60SC and NCT, a superconducting tokamak positioned as the satellite machine of ITER, collaborating with Japan and EU fusion community, aims at contribution to ITER and DEMO through the demonstration of advanced plasma operation scenario and the plasma applicability test with advanced materials. After the discussions in JA-EU Satellite Tokamak Working Group in 2005, the increased heating power, higher heat removal capacity for the plasma facing components, improvement of the radiation shielding, the remote handling system for the maintenance of in-vessel components, and related equipment are planed to be additionally installed. With such full equipment towards the increased heating power of 41 MW (34 MW-NBI and 7 MW-ECH) with 100 s, the prospective plasma performances, analysed by the equilibrium and transport analysis codes, are rather improved in the view point of the contribution to ITER and DEMO relevant research. Accessibility for higher heating power in a higher density region enables the lower normalized Larmor radius and normalized collision frequency close to the reactor relevant plasma with the ITER-similar configuration of single null divertor plasma with the aspect ratio of A = 3.1, elongation of k95 = 1.7, triangularity of d95 (q95) in the plasma current of I p = 3.5 MA, toroidal magnetic field of B T = 2.59 T and the major radius of Rp=3.16 m. The increases in the electron temperature, beam driven and bootstrap current fraction by the increase of the power of Negative ion based NBI (10 MW) reduce the loop voltage and contribute to increase the maximum plasma current of ITER similar shape. Full non-inductive current drive controllability is also extended into the high density and high plasma current operation towards high beta plasma. Flexibility in aspect ratio and shape parameter is kept the same as NCT, i.e. a double null divertor plasma with A = 2.6, k95 = 1.83, d95 = 0.57, I p = 5.5 MA, B T = 2.72 T, and R p = 3.01 m which

  5. Evidence of dose saving in routine CT practice using iterative reconstruction derived from a national diagnostic reference level survey.

    Science.gov (United States)

    Thomas, P; Hayton, A; Beveridge, T; Marks, P; Wallace, A

    2015-09-01

    To assess the influence and significance of the use of iterative reconstruction (IR) algorithms on patient dose in CT in Australia. We examined survey data submitted to the Australian Radiation Protection and Nuclear Safety Agency (ARPANSA) National Diagnostic Reference Level Service (NDRLS) during 2013 and 2014. We compared median survey dose metrics with categorization by scan region and use of IR. The use of IR results in a reduction in volume CT dose index of between 17% and 44% and a reduction in dose-length product of between 14% and 34% depending on the specific scan region. The reduction was highly significant (p sum test) for all six scan regions included in the NDRLS. Overall, 69% (806/1167) of surveys included in the analysis used IR. The use of IR in CT is achieving dose savings of 20-30% in routine practice in Australia. IR appears to be widely used by participants in the ARPANSA NDRLS with approximately 70% of surveys submitted employing this technique. This study examines the impact of the use of IR on patient dose in CT on a national scale.

  6. Energy scenarios for the 21. century

    International Nuclear Information System (INIS)

    Lauerman, V.

    2002-01-01

    The Canadian Energy Research Institute (CERI) has adopted a scenario approach in developing its energy outlook to 2025. These scenarios can be used in developing strategies to optimize opportunities and avoid the dangers that often accompany the more deterministic approach to forecasting. The scenarios are not predictions, but are instead feasible outcomes for the future. They encompass 5 major areas of uncertainty for the world energy market through 2025 that impact energy demand and fuel mix. These include the economy, technology, energy policy, resource availability, and market structure. The drivers for the energy scenarios are the environment, demography, economics, culture geopolitics and technology. The paper referred to the world energy market, the OECD energy markets, the non-OECD energy markets and the FSU energy markets under 3 energy scenarios including: (1) the material world in which the new economy drives the global economy, (2) a very dark Orwellian world in which there is a rise in terrorism and a breakdown in international cooperation, and (3) a utopia in which the major powers become increasingly serious about achieving sustainable development. 22 refs., 48 figs

  7. Reflective metallic coatings for first mirrors on ITER

    International Nuclear Information System (INIS)

    Eren, Baran; Marot, Laurent; Litnovsky, Andrey; Matveeva, Maria; Steiner, Roland; Emberger, Valentin; Wisse, Marco; Mathys, Daniel; Covarel, Gregory; Meyer, Ernst

    2011-01-01

    Metallic mirrors are foreseen to play a crucial role for all optical diagnostics in ITER. Therefore, the development of reliable techniques for the production of mirrors which are able to maintain their optical properties in the harsh ITER environment is highly important. By applying magnetron sputtering and evaporation techniques, rhodium and molybdenum films have been prepared for tokamak tests. The films were characterised in terms of chemical composition, surface roughness, crystallite structure, reflectivity and adhesion. No impurities were detected on the surface after deposition. The effects of deposition parameters and substrate temperature on the resulting crystallite structure, surface roughness and hence on the reflectivity, were investigated. The films are found to exhibit nanometric crystallites with a dense columnar structure. Open boundaries between the crystallite columns, which are sometimes present after evaporation, are found to reduce the reflectivity as compared to rhodium or molybdenum references.

  8. Reflective metallic coatings for first mirrors on ITER

    Energy Technology Data Exchange (ETDEWEB)

    Eren, Baran, E-mail: baran.eren@unibas.ch [Department of Physics, University of Basel, Klingelbergstrasse 82, CH-4056 Basel (Switzerland); Marot, Laurent [Department of Physics, University of Basel, Klingelbergstrasse 82, CH-4056 Basel (Switzerland); Litnovsky, Andrey; Matveeva, Maria [Institut fuer Energieforschung (Plasmaphysik), Forschungszentrum Juelich, Association EURATOM-FZJ, D 52425 Juelich (Germany); Steiner, Roland; Emberger, Valentin; Wisse, Marco [Department of Physics, University of Basel, Klingelbergstrasse 82, CH-4056 Basel (Switzerland); Mathys, Daniel [Centre of Microscopy, University of Basel, Klingelbergstrasse 50/70, CH-4056 Basel (Switzerland); Covarel, Gregory [Laboratoire de Physique et Mecanique Textile EA CNRS 7189, Universite de Haute Alsace, 61 rue Albert Camus, 68093 Mulhouse Cedex (France); Meyer, Ernst [Department of Physics, University of Basel, Klingelbergstrasse 82, CH-4056 Basel (Switzerland)

    2011-10-15

    Metallic mirrors are foreseen to play a crucial role for all optical diagnostics in ITER. Therefore, the development of reliable techniques for the production of mirrors which are able to maintain their optical properties in the harsh ITER environment is highly important. By applying magnetron sputtering and evaporation techniques, rhodium and molybdenum films have been prepared for tokamak tests. The films were characterised in terms of chemical composition, surface roughness, crystallite structure, reflectivity and adhesion. No impurities were detected on the surface after deposition. The effects of deposition parameters and substrate temperature on the resulting crystallite structure, surface roughness and hence on the reflectivity, were investigated. The films are found to exhibit nanometric crystallites with a dense columnar structure. Open boundaries between the crystallite columns, which are sometimes present after evaporation, are found to reduce the reflectivity as compared to rhodium or molybdenum references.

  9. Chapter 8: Plasma operation and control [Progress in the ITER Physics Basis (PIPB)

    International Nuclear Information System (INIS)

    Gribov, Y.; Humphreys, D.; Kajiwara, K.; Lazarus, E.A.; Lister, J.B.; Ozeki, T.; Portone, A.; Shimada, M.; Sips, A.C.C.; Wesley, J.C.

    2007-01-01

    The ITER plasma control system has the same functional scope as the control systems in present tokamaks. These are plasma operation scenario sequencing, plasma basic control (magnetic and kinetic), plasma advanced control (control of RWMs, NTMs, ELMs, error fields, etc) and plasma fast shutdown. This chapter considers only plasma initiation and plasma basic control. This chapter describes the progress achieved in these areas in the tokamak experiments since the ITER Physics Basis (1999 Nucl. Fusion 39 2577) was written and the results of assessment of ITER to provide the plasma initiation and basic control. This assessment was done for the present ITER design (15 MA machine) at a more detailed level than it was done for the ITER design 1998 (21 MA machine) described in the ITER Physics Basis (1999 Nucl. Fusion 39 2577). The experiments on plasma initiation performed in DIII-D and JT-60U, as well as the theoretical studies performed for ITER, have demonstrated that, within specified assumptions on the plasma confinement and the impurity influx, ITER can produce plasma initiation in a low toroidal electric field (0.3 V m -1 ), if it is assisted by about 2 MW of ECRF heating. The plasma basic control includes control of the plasma current, position and shape-the plasma magnetic control, as well as control of other plasma global parameters or their profiles-the plasma performance control. The magnetic control is based on more reliable and simpler models of the control objects than those available at present for the plasma kinetic control. Moreover the real time diagnostics used for the magnetic control in many cases are more precise than those used for the kinetic control. Because of these reasons, the plasma magnetic control was developed for modern tokamaks and assessed for ITER better than the kinetic control. However, significant progress has been achieved in the plasma performance control during the last few years. Although the physics basis of plasma operation

  10. Plasma regimes and research goals of JT-60SA towards ITER and DEMO

    International Nuclear Information System (INIS)

    Kamada, Y.; Ide, S.; Fujita, T.; Suzuki, T.; Matsunaga, G.; Yoshida, M.; Shinohara, K.; Urano, H.; Nakano, T.; Sakurai, S.; Kawashima, H.; Barabaschi, P.; Lackner, K.; Ishida, S.; Bolzonella, T.

    2011-01-01

    The JT-60SA device has been designed as a highly shaped large superconducting tokamak with a variety of plasma actuators (heating, current drive, momentum input, stability control coils, resonant magnetic perturbation coils, W-shaped divertor, fuelling, pumping, etc) in order to satisfy the central research needs for ITER and DEMO. In the ITER- and DEMO-relevant plasma parameter regimes and with DEMO-equivalent plasma shapes, JT-60SA quantifies the operation limits, plasma responses and operational margins in terms of MHD stability, plasma transport and confinement, high-energy particle behaviour, pedestal structures, scrape-off layer and divertor characteristics. By integrating advanced studies in these research fields, the project proceeds 'simultaneous and steady-state sustainment of the key performances required for DEMO' with integrated control scenario development applicable to the highly self-regulating burning high-β high bootstrap current fraction plasmas.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-08-21

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

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

    Science.gov (United States)

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

    2014-08-01

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

  13. 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

  14. SACFIR: SDN-Based Application-Aware Centralized Adaptive Flow Iterative Reconfiguring Routing Protocol for WSNs.

    Science.gov (United States)

    Aslam, Muhammad; Hu, Xiaopeng; Wang, Fan

    2017-12-13

    Smart reconfiguration of a dynamic networking environment is offered by the central control of Software-Defined Networking (SDN). Centralized SDN-based management architectures are capable of retrieving global topology intelligence and decoupling the forwarding plane from the control plane. Routing protocols developed for conventional Wireless Sensor Networks (WSNs) utilize limited iterative reconfiguration methods to optimize environmental reporting. However, the challenging networking scenarios of WSNs involve a performance overhead due to constant periodic iterative reconfigurations. In this paper, we propose the SDN-based Application-aware Centralized adaptive Flow Iterative Reconfiguring (SACFIR) routing protocol with the centralized SDN iterative solver controller to maintain the load-balancing between flow reconfigurations and flow allocation cost. The proposed SACFIR's routing protocol offers a unique iterative path-selection algorithm, which initially computes suitable clustering based on residual resources at the control layer and then implements application-aware threshold-based multi-hop report transmissions on the forwarding plane. The operation of the SACFIR algorithm is centrally supervised by the SDN controller residing at the Base Station (BS). This paper extends SACFIR to SDN-based Application-aware Main-value Centralized adaptive Flow Iterative Reconfiguring (SAMCFIR) to establish both proactive and reactive reporting. The SAMCFIR transmission phase enables sensor nodes to trigger direct transmissions for main-value reports, while in the case of SACFIR, all reports follow computed routes. Our SDN-enabled proposed models adjust the reconfiguration period according to the traffic burden on sensor nodes, which results in heterogeneity awareness, load-balancing and application-specific reconfigurations of WSNs. Extensive experimental simulation-based results show that SACFIR and SAMCFIR yield the maximum scalability, network lifetime and stability

  15. SACFIR: SDN-Based Application-Aware Centralized Adaptive Flow Iterative Reconfiguring Routing Protocol for WSNs

    Directory of Open Access Journals (Sweden)

    Muhammad Aslam

    2017-12-01

    Full Text Available Smart reconfiguration of a dynamic networking environment is offered by the central control of Software-Defined Networking (SDN. Centralized SDN-based management architectures are capable of retrieving global topology intelligence and decoupling the forwarding plane from the control plane. Routing protocols developed for conventional Wireless Sensor Networks (WSNs utilize limited iterative reconfiguration methods to optimize environmental reporting. However, the challenging networking scenarios of WSNs involve a performance overhead due to constant periodic iterative reconfigurations. In this paper, we propose the SDN-based Application-aware Centralized adaptive Flow Iterative Reconfiguring (SACFIR routing protocol with the centralized SDN iterative solver controller to maintain the load-balancing between flow reconfigurations and flow allocation cost. The proposed SACFIR’s routing protocol offers a unique iterative path-selection algorithm, which initially computes suitable clustering based on residual resources at the control layer and then implements application-aware threshold-based multi-hop report transmissions on the forwarding plane. The operation of the SACFIR algorithm is centrally supervised by the SDN controller residing at the Base Station (BS. This paper extends SACFIR to SDN-based Application-aware Main-value Centralized adaptive Flow Iterative Reconfiguring (SAMCFIR to establish both proactive and reactive reporting. The SAMCFIR transmission phase enables sensor nodes to trigger direct transmissions for main-value reports, while in the case of SACFIR, all reports follow computed routes. Our SDN-enabled proposed models adjust the reconfiguration period according to the traffic burden on sensor nodes, which results in heterogeneity awareness, load-balancing and application-specific reconfigurations of WSNs. Extensive experimental simulation-based results show that SACFIR and SAMCFIR yield the maximum scalability, network lifetime

  16. Low carbon and clean energy scenarios for India: Analysis of targets approach

    International Nuclear Information System (INIS)

    Shukla, Priyadarshi R.; Chaturvedi, Vaibhav

    2012-01-01

    Low carbon energy technologies are of increasing importance to India for reducing emissions and diversifying its energy supply mix. Using GCAM, an integrated assessment model, this paper analyzes a targets approach for pushing solar, wind, and nuclear technologies in the Indian electricity generation sector from 2005 to 2095. Targets for these technologies have been constructed on the basis of Indian government documents, policy announcements, and expert opinions. Different targets have been set for the reference scenario and the carbon price scenario. In the reference scenario, wind and nuclear technologies exceed respective targets in the long run without any subsidy push, while solar energy requires subsidy push throughout the century in order to meet its high targets. In the short run, nuclear energy also requires significant subsidy, including a much higher initial subsidy relative to solar power, which is a result of its higher targets. Under a carbon price scenario, the carbon price drives the penetration of these technologies. Still, subsidy is required — especially in the short run when the carbon price is low. We also found that pushing solar, wind, and nuclear technologies leads to a decrease in share of CCS under the carbon price scenario and biomass under both the reference and carbon price scenarios. This is because low carbon technologies compete among themselves and substitute each other, thereby enhancing the need for subsidy or carbon price, highlighting that proposed targets are not set at efficient levels. In light of contemporary debate on external costs of nuclear energy, we also assess the sensitivity of the results to nuclear technology cost. We find that higher cost significantly decreases the share of nuclear power under both the reference and carbon price scenarios.

  17. ITER interim design report package documents

    International Nuclear Information System (INIS)

    1996-01-01

    This publication contains the Excerpt from the ITER Council (IC-8), the ITER Interim Design Report, Cost Review and Safety Analysis, ITER Site Requirements and ITER Site Design Assumptions and the Excerpt from the ITER Council (IC-9). 8 figs, 2 tabs

  18. 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

  19. Finding the Optimum Scenario in Risk-benefit Assessment: An Example on Vitamin D

    DEFF Research Database (Denmark)

    Berjia, Firew Lemma; Hoekstra, J.; Verhagen, H.

    2014-01-01

    when changing from the reference to the optimum scenario. Conclusion: The method allowed us to find the optimum serum level in the vitamin D example. Additional case studies are needed to further validate the applicability of the approach to other nutrients or foods, especially with regards...... a method for finding the optimum scenario that provides maximum net health gains. Methods: A multiple scenario simulation. The method is presented using vitamin D intake in Denmark as an example. In addition to the reference scenario, several alternative scenarios are simulated to detect the scenario...... that provides maximum net health gains. As a common health metric, Disability Adjusted Life Years (DALY) has been used to project the net health effect by using the QALIBRA (Quality of Life for Benefit Risk Assessment) software. Results: The method used in the vitamin D example shows that it is feasible to find...

  20. In-vessel tritium retention and removal in ITER

    International Nuclear Information System (INIS)

    Federici, G.; Anderl, R.A.

    1998-01-01

    The International Thermonuclear Experimental Reactor (ITER) is envisioned to be the next major step in the world's fusion program from the present generation of tokamaks and is designed to study fusion plasmas with a reactor relevant range of plasma parameters. During normal operation, it is expected that a fraction of the unburned tritium, that is used to routinely fuel the discharge, will be retained together with deuterium on the surfaces and in the bulk of the plasma facing materials (PFMs) surrounding the core and divertor plasma. The understanding of he basic retention mechanisms (physical and chemical) involved and their dependence upon plasma parameters and other relevant operation conditions is necessary for the accurate prediction of the amount of tritium retained at any given time in the ITER torus. Accurate estimates are essential to assess the radiological hazards associated with routine operation and with potential accident scenarios which may lead to mobilization of tritium that is not tenaciously held. Estimates are needed to establish the detritiation requirements for coolant water, to determine the plasma fueling and tritium supply requirements, and to establish the needed frequency and the procedures for tritium recovery and clean-up. The organization of this paper is as follows. Section 2 provides an overview of the design and operating conditions of the main components which define the plasma boundary of ITER. Section 3 reviews the erosion database and the results of recent relevant experiments conducted both in laboratory facilities and in tokamaks. These data provide the experimental basis and serve as an important benchmark for both model development (discussed in Section 4) and calculations (discussed in Section 5) that are required to predict tritium inventory build-up in ITER. Section 6 emphasizes the need to develop and test methods to remove the tritium from the codeposited C-based films and reviews the status and the prospects of the

  1. In-vessel tritium retention and removal in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Federici, G. [ITER JWS Garching Co-Center (Germany); Anderl, R.A. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.; Andrew, P. [JET Joint Undertaking, Abingdon (United Kingdom)] [and others

    1998-06-01

    The International Thermonuclear Experimental Reactor (ITER) is envisioned to be the next major step in the world`s fusion program from the present generation of tokamaks and is designed to study fusion plasmas with a reactor relevant range of plasma parameters. During normal operation, it is expected that a fraction of the unburned tritium, that is used to routinely fuel the discharge, will be retained together with deuterium on the surfaces and in the bulk of the plasma facing materials (PFMs) surrounding the core and divertor plasma. The understanding of he basic retention mechanisms (physical and chemical) involved and their dependence upon plasma parameters and other relevant operation conditions is necessary for the accurate prediction of the amount of tritium retained at any given time in the ITER torus. Accurate estimates are essential to assess the radiological hazards associated with routine operation and with potential accident scenarios which may lead to mobilization of tritium that is not tenaciously held. Estimates are needed to establish the detritiation requirements for coolant water, to determine the plasma fueling and tritium supply requirements, and to establish the needed frequency and the procedures for tritium recovery and clean-up. The organization of this paper is as follows. Section 2 provides an overview of the design and operating conditions of the main components which define the plasma boundary of ITER. Section 3 reviews the erosion database and the results of recent relevant experiments conducted both in laboratory facilities and in tokamaks. These data provide the experimental basis and serve as an important benchmark for both model development (discussed in Section 4) and calculations (discussed in Section 5) that are required to predict tritium inventory build-up in ITER. Section 6 emphasizes the need to develop and test methods to remove the tritium from the codeposited C-based films and reviews the status and the prospects of the

  2. A change navigation-based, scenario planning process within a developing world context from an Afro-centric leadership perspective

    Directory of Open Access Journals (Sweden)

    Chris A. Geldenhuys

    2011-02-01

    Research purpose: This study aimed to investigate how scenario-based planning (a strictly cognitive management tool can be combined with organisational change navigation (a practice addressing the emotionality of change and how this integrated process should be aligned with the prerequisites imposed by a developing country context and an Afro-centric leadership perspective in order to make the process more context relevant and aligned. Motivation for the study: The integration of organisational change navigation with conventional scenario based planning, as well as the incorporation of the perquisites of a developing countries and an Afro-centric leadership perspective, will give organisations a more robust, holistic strategic management tool that will add significantly more value within a rapidly, radically and unpredictably changing world. Research design, approach and method: The adopted research approach comprised a combination of the sourcing of the latest thinking in the literature (the ‘theory’ as well as the views of seasoned practitioners of scenario planning (the ‘practice’ through an iterative research process, moving between theory and practice, back to practice and finally returning to theory in order to arrive at a validated expanded and enhanced scenario-based planning process which is both theory and practice ‘proof’. Main findings: A management tool incorporating the change navigation and the unique features of developing countries and Afro-centric leadership was formulated and empirically validated. This management tool is referred to as a change navigation based, scenario planning process (CNBSPP. Practical/managerial implications: CNBSPP is available for use by organisations wishing to apply a strategic planning tool that fits within a developing country context and an Afro-centric leadership approach. Contribution/value add: The research makes a unique contribution to the current level of knowledge by integrating two disciplines

  3. Accident scenarios triggered by lightning strike on atmospheric storage tanks

    International Nuclear Information System (INIS)

    Necci, Amos; Argenti, Francesca; Landucci, Gabriele; Cozzani, Valerio

    2014-01-01

    Severe Natech accidents may be triggered by lightning strike affecting storage tanks containing relevant inventories of hazardous materials. The present study focused on the identification of event sequences and accident scenarios following lightning impact on atmospheric tanks. Reference event trees, validated using past accident analysis, are provided to describe the specific accident chains identified, accounting for reference protection and mitigation safety barriers usually adopted in current industrial practice. An overall methodology was outlined to allow the calculation of the expected frequencies of final scenarios following lightning impact on atmospheric storage tanks, taking into account the expected performance of available safety barriers. The methodology was applied to a case study in order to better understand the data that may be obtained and their importance in the framework of quantitative risk assessment (QRA) and of the risk management of industrial facilities with respect to external hazards due to natural events. - Highlights: • Event sequences following lightning impact on atmospheric tanks were identified. • Reference event trees including standard safety barriers were obtained. • Safety barriers applied in industrial practice were assessed to quantify event trees. • Frequencies of final scenarios following lightning impact on tanks were calculated. • Natech scenarios caused by lightning have an important influence on risk profiles

  4. A Frictionless Steering Mechanism for the Front Steering ECCD ITER Upper Port Launcher

    International Nuclear Information System (INIS)

    Chavan, R; Henderson, M A; Sanchez, F

    2005-01-01

    A FS launcher is being designed for the ITER upper port, which offers enhanced physics performance over the RS launcher. A two mirror system is used to decouple the focusing and steering aspects of the launcher and provide a relatively small beam waist ( 1.6 m from the steering mirror). The resulting NTM stabilization efficiency (maximum CD density divided by the local bootstrap current >1.6) is above marginal for the q = 2 and 3/2 rational flux surfaces of the relevant ITER equilibria (scenarios 2, 3a and 5) and a factor of ∼3 relative to an equivalent RS launcher. The performance of the FS launcher strongly depends on the reliability of the steering mechanism, which is used to rotate the plasma facing steering mirror. CRPP has designed a frictionless steering mechanism assembled in a compact cartridge capable of up to ±10 deg. rotation (corresponding to a poloidal steering range of up to ±20 deg. for the microwave beam around a fixed axis of rotation) that offers a high operation reliability despite the close proximity to the thermal and neutron flux coming from the ITER plasma

  5. 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.

  6. Review of the International Thermonuclear Experimental Reactor (ITER) detailed design report

    International Nuclear Information System (INIS)

    1997-01-01

    Dr. Martha Krebs, Director, Office of Energy Research at the US Department of Energy (DOE), wrote to the Fusion Energy Sciences Advisory Committee (FESAC), in letters dated September 23 and November 6, 1996, requesting that FESAC review the International Thermonuclear Experimental Reactor (ITER) Detailed Design Report (DDR) and provide its view of the adequacy of the DDR as part of the basis for the United States decision to enter negotiations with the other interested Parties regarding the terms and conditions for an agreement for the construction, operations, exploitation and decommissioning of ITER. The letter from Dr. Krebs, referred to as the Charge Letter, provided context for the review and a set of questions of specific interest

  7. Review of the International Thermonuclear Experimental Reactor (ITER) detailed design report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-04-18

    Dr. Martha Krebs, Director, Office of Energy Research at the US Department of Energy (DOE), wrote to the Fusion Energy Sciences Advisory Committee (FESAC), in letters dated September 23 and November 6, 1996, requesting that FESAC review the International Thermonuclear Experimental Reactor (ITER) Detailed Design Report (DDR) and provide its view of the adequacy of the DDR as part of the basis for the United States decision to enter negotiations with the other interested Parties regarding the terms and conditions for an agreement for the construction, operations, exploitation and decommissioning of ITER. The letter from Dr. Krebs, referred to as the Charge Letter, provided context for the review and a set of questions of specific interest.

  8. 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.

  9. 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.

  10. 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

  11. 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

  12. 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.)

  13. Use of direct and iterative solvers for estimation of SNP effects in genome-wide selection

    Directory of Open Access Journals (Sweden)

    Eduardo da Cruz Gouveia Pimentel

    2010-01-01

    Full Text Available The aim of this study was to compare iterative and direct solvers for estimation of marker effects in genomic selection. One iterative and two direct methods were used: Gauss-Seidel with Residual Update, Cholesky Decomposition and Gentleman-Givens rotations. For resembling different scenarios with respect to number of markers and of genotyped animals, a simulated data set divided into 25 subsets was used. Number of markers ranged from 1,200 to 5,925 and number of animals ranged from 1,200 to 5,865. Methods were also applied to real data comprising 3081 individuals genotyped for 45181 SNPs. Results from simulated data showed that the iterative solver was substantially faster than direct methods for larger numbers of markers. Use of a direct solver may allow for computing (covariances of SNP effects. When applied to real data, performance of the iterative method varied substantially, depending on the level of ill-conditioning of the coefficient matrix. From results with real data, Gentleman-Givens rotations would be the method of choice in this particular application as it provided an exact solution within a fairly reasonable time frame (less than two hours. It would indeed be the preferred method whenever computer resources allow its use.

  14. Engineering Design of the ITER AC/DC Power Supplies

    International Nuclear Information System (INIS)

    Oh, B. H.; Lee, K. W.; Hwang, C. K.; Jin, J. T.; Chang, D. S.; Kim, T. S.

    2009-02-01

    To design high power pulse power supplies, especially in huge power supplies have not designed till now, it is necessary to analyze a system's characteristics and relations with another systems as well as to know high voltage, high current control technologies. Contents of this project are; - Study for the engineering designs changed recently by ITER Organization(IO) and writing specifications for the power supplies to reduce project risk. - Detailed analysis of the AC/DC Converters and writing subtask reports on the Task Agreement. - Study for thyristor numbers, DCR's specifications for Korea-China sharing meetings. - Study for the grounding systems of the ITER power supply system. The results may used as one of reference for practical designs of the high power coil power supplies and also may used in various field such as electroplating, plasma arc furnaces, electric furnaces

  15. EU contribution to ITER CTA fuel cycle design and R and D

    International Nuclear Information System (INIS)

    Murdoch, D.; Bekris, N.; Boissin, J.-C.; Day, C.; Gilroy, J.; Glugla, M.; Laesser, R.; Mack, A.

    2003-01-01

    Design and supporting R and D are being carried out in the EU in preparation for the construction of International Thermonuclear Experimental Reactor (ITER). As the Fuel Cycle systems are located in 'time-critical' buildings the detailed design of piping, ducting, cabling and services is being done to ensure compatibility with the building. R and D is focussed on improvements to the reference designs for ITER, extension of parametric studies to cover all operating modes and long term experiments to test component reliability. In addition emphasis is being placed on design and R and D for the Water Detritiation System, an important barrier against routine release of tritium

  16. 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

  17. Analysis of advanced european nuclear fuel cycle scenarios including transmutation and economical estimates

    International Nuclear Information System (INIS)

    Merino Rodriguez, I.; Alvarez-Velarde, F.; Martin-Fuertes, F.

    2013-01-01

    In this work the transition from the existing Light Water Reactors (LWR) to the advanced reactors is analyzed, including Generation III+ reactors in a European framework. Four European fuel cycle scenarios involving transmutation options have been addressed. The first scenario (i.e., reference) is the current fleet using LWR technology and open fuel cycle. The second scenario assumes a full replacement of the initial fleet with Fast Reactors (FR) burning U-Pu MOX fuel. The third scenario is a modification of the second one introducing Minor Actinide (MA) transmutation in a fraction of the FR fleet. Finally, in the fourth scenario, the LWR fleet is replaced using FR with MOX fuel as well as Accelerator Driven Systems (ADS) for MA transmutation. All scenarios consider an intermediate period of GEN-III+ LWR deployment and they extend for a period of 200 years looking for equilibrium mass flows. The simulations were made using the TR-EVOL code, a tool for fuel cycle studies developed by CIEMAT. The results reveal that all scenarios are feasible according to nuclear resources demand (U and Pu). Concerning to no transmutation cases, the second scenario reduces considerably the Pu inventory in repositories compared to the reference scenario, although the MA inventory increases. The transmutation scenarios show that elimination of the LWR MA legacy requires on one hand a maximum of 33% fraction (i.e., a peak value of 26 FR units) of the FR fleet dedicated to transmutation (MA in MOX fuel, homogeneous transmutation). On the other hand a maximum number of ADS plants accounting for 5% of electricity generation are predicted in the fourth scenario (i.e., 35 ADS units). Regarding the economic analysis, the estimations show an increase of LCOE (Levelized cost of electricity) - averaged over the whole period - with respect to the reference scenario of 21% and 29% for FR and FR with transmutation scenarios respectively, and 34% for the fourth scenario. (authors)

  18. Analysis of ITER NbTi and Nb3Sn CICCs experimental minimum quench energy with JackPot, MCM and THEA models

    NARCIS (Netherlands)

    Bagni, T.; Duchateau, J.L.; Breschi, M.; Devred, A.; Nijhuis, A.

    2017-01-01

    Cable-in-conduit conductors (CICCs) for ITER magnets are subjected to fast changing magnetic fields during the plasma-operating scenario. In order to anticipate the limitations of conductors under the foreseen operating conditions, it is essential to have a better understanding of the stability

  19. The ITER Plasma Control System Simulation Platform

    International Nuclear Information System (INIS)

    Walker, M.L.; Ambrosino, G.; De Tommasi, G.; Humphreys, D.A.; Mattei, M.; Neu, G.; Rapson, C.J.; Raupp, G.; Treutterer, W.; Welander, A.S.; Winter, A.

    2015-01-01

    Highlights: • A development and test environment called PCSSP has been constructed for the ITER PCS. • A description of requirements and use cases, a final design and software architecture design, users guide, and a prototype implementation have been delivered. • The prototype implementation was demonstrated at IO in December of 2013. • PCSSP will be deployed for alpha testing to the IO, the development group, and selected other ITER partners upon completion of the next development phase. - Abstract: The Plasma Control System Simulation Platform (PCSSP) is a highly flexible, modular, time-dependent simulation environment developed primarily to support development of the ITER Plasma Control System (PCS). It has been under development since 2011 and is scheduled for first release to users in the ITER Organization (IO) and at selected additional sites in 2015. Modules presently implemented in PCSSP enable exploration of axisymmetric evolution and control, basic kinetic control, and tearing mode suppression. A basic capability for generation of control-relevant events is included, enabling study of exception handling in the PCS, continuous controllers, and PCS architecture. While the control design focus of PCSSP applications tends to require only a moderate level of accuracy and complexity in modules, more complex codes can be embedded or connected to access higher accuracy if needed. This paper describes the background and motivation for PCSSP, provides an overview of the capabilities, architecture, and features of PCSSP, and discusses details of the PCSSP vision and its intended goals and application. Completed work, including architectural design, prototype implementation, reference documents, and IO demonstration of PCSSP, is summarized and example use of PCSSP is illustrated. Near-term high-level objectives are summarized and include preparation for release of an “alpha” version of PCSSP and preparation for the next development phase. High

  20. The ITER Plasma Control System Simulation Platform

    Energy Technology Data Exchange (ETDEWEB)

    Walker, M.L., E-mail: walker@fusion.gat.com [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Ambrosino, G.; De Tommasi, G. [CREATE/Università di Napoli Federico II, Napoli (Italy); Humphreys, D.A. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Mattei, M. [CREATE/Seconda Università di Napoli, Napoli (Italy); Neu, G.; Rapson, C.J.; Raupp, G.; Treutterer, W. [Max Planck Institute for Plasma Physics, Boltzmannstr. 2, 85748 Garching (Germany); Welander, A.S. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Winter, A. [ITER Organization, Route de Vinon-sur-Verdon, 13115 St. Paul-lez-Durance (France)

    2015-10-15

    Highlights: • A development and test environment called PCSSP has been constructed for the ITER PCS. • A description of requirements and use cases, a final design and software architecture design, users guide, and a prototype implementation have been delivered. • The prototype implementation was demonstrated at IO in December of 2013. • PCSSP will be deployed for alpha testing to the IO, the development group, and selected other ITER partners upon completion of the next development phase. - Abstract: The Plasma Control System Simulation Platform (PCSSP) is a highly flexible, modular, time-dependent simulation environment developed primarily to support development of the ITER Plasma Control System (PCS). It has been under development since 2011 and is scheduled for first release to users in the ITER Organization (IO) and at selected additional sites in 2015. Modules presently implemented in PCSSP enable exploration of axisymmetric evolution and control, basic kinetic control, and tearing mode suppression. A basic capability for generation of control-relevant events is included, enabling study of exception handling in the PCS, continuous controllers, and PCS architecture. While the control design focus of PCSSP applications tends to require only a moderate level of accuracy and complexity in modules, more complex codes can be embedded or connected to access higher accuracy if needed. This paper describes the background and motivation for PCSSP, provides an overview of the capabilities, architecture, and features of PCSSP, and discusses details of the PCSSP vision and its intended goals and application. Completed work, including architectural design, prototype implementation, reference documents, and IO demonstration of PCSSP, is summarized and example use of PCSSP is illustrated. Near-term high-level objectives are summarized and include preparation for release of an “alpha” version of PCSSP and preparation for the next development phase. High

  1. Charging of mobile services by mobile payment reference model

    OpenAIRE

    Pousttchi, Key; Wiedemann, Dietmar Georg

    2005-01-01

    The purpose of the paper is to analyze mobile payments in the mobile commerce scenario. Therefore, we first classify the mobile payment in the mobile commerce scenario by explaining general offer models, charging concepts, and intermediaries. Second, we describe the mobile payment reference model, especially, the mobile payment reference organization model and different mobile payment standard types. Finally, we conclude our findings.

  2. ITER physics design guidelines: 1989

    International Nuclear Information System (INIS)

    Uckan, N.A.

    1990-01-01

    The physics basis for ITER has been developed from an assessment of the results of the last twenty-five years of tokamak research and from detailed analysis of important physics issues specifically for the ITER design. This assessment has been carried out with direct participation of members of the experimental teams of each of the major tokamaks in the world fusion program through participation in ITER workshops, contributions to the ITER Physics R and D Program, and by direct contacts between the ITER team and the cognizant experimentalists. Extrapolations to the present data base, where needed, are made in the most cautious way consistent with engineering constraints and performance goals of the ITER. In cases where a working assumptions had to be introduced, which is insufficiently supported by the present data base, is explicitly stated. While a strong emphasis has been placed on the physics credibility of the design, the guidelines also take into account that ITER should be designed to be able to take advantage of potential improvements in tokamak physics that may occur before and during the operation of ITER. (author). 33 refs

  3. 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

  4. Negatep: A Scenario for Combating Global Warming; Le scenario Negatep. Un scenario de lutte contre le rechauffement climatique

    Energy Technology Data Exchange (ETDEWEB)

    Acket, C.; Bacher, P. [Sauvons Le Climat, 92 - Boulogne Billancourt (France)

    2011-07-15

    There have been an increasing number of foresight exercises in the field of energy and global warming in recent years, as we have seen from the articles devoted to these questions by Futuribles in 2011 (both in this special issue and in the April number). It is certainly the case that the goals for greenhouse-gas emission reduction are rather ambitious, particularly in France, it being the aim of the 2005 French framework law on energy to reduce carbon gas discharges by a factor of four. Among these scenarios, the Negatep scenario developed by Claude Acket and Pierre Bacher from the 'Sauvons le climat' [Let's save the climate] Association proposes to achieve this ('factor 4') goal in France by 2050 by reducing fossil fuel use by 75% and replacing this as quickly as possible with electricity produced from non-carbon-gas-emitting sources - chiefly, nuclear power and renewables. The authors lay out their goals here, backed up by figures, comparing these with the reference scenario. They also show the path that must be followed to arrive at these goals, particularly in the residential and tertiary sectors, and in transport and industry (through control of needs and recourse to alternative energy sources). They close by comparing the Negatep scenario with two other more recent scenarios aimed also at reducing greenhouse gas emissions, on the one hand in Europe, and on the other in Germany. The comparison confirms that they were right to rely on electricity as a substitute for oil, but gives them cause for concern in respect of the consequences (formidable in their view) that the replacement of nuclear power and coal energy by intermittent renewable energies might have in Europe, both with regard to costs and to the effects on the power network. (authors)

  5. Control of dust production in ITER

    International Nuclear Information System (INIS)

    Rodriguez-Rodrigo, L.; Ciattaglia, S.; Elbez-Uzan, J.

    2006-01-01

    In the last years dust has been observed in a number of fusion devices and is being studied more in detail for understanding in particular the physical phenomena related to its formation, its composition, physical and chemical characteristics, and the amount of produced dust. The extrapolation of dust formation to ITER predicts (with large error bars), a large mass of dust production with a scattered size distribution. To evaluate the impact of dust on safety, assumptions have also been made on radionuclide inventory, and mobility in off-normal events, as well as any postulated contributions the dust may make to effluents or accidental releases. Solid activation products in structures are generally not readily mobilisable in incidental and accidental situations, so that activated dust, tritium and activated corrosions products are the important in-vessel source terms in postulated scenarios that assume a mobilisation and release of some fraction of this inventory. Such a release would require the simultaneous leak or bypass of several robust confinement barriers. Further concerns for dust may be the potential for chemical reactions between dust and coolant in the event of an in-vessel leak, and the theoretical possibility of a dust explosion, either of which could in principle cause a pressure rise that challenges one or more of the confinement barriers. Although these hazards can - and will - be controlled by other measures in the ITER design, application of the principle of Defence in Depth dictates that the dust inventory should also be minimised and controlled to prevent the potential hazard. A well-coordinated R-and-D programme is required to support this dust production control. This document provides from the safety point of view, an overview of existing data given in '' Dossier d'Options de Surete '', the first safety report presented in 2001 to the French Safety Authorities, and ITER documents; it also gathers information on status of studies on activated

  6. Iterative solution of large sparse systems of equations

    CERN Document Server

    Hackbusch, Wolfgang

    2016-01-01

    In the second edition of this classic monograph, complete with four new chapters and updated references, readers will now have access to content describing and analysing classical and modern methods with emphasis on the algebraic structure of linear iteration, which is usually ignored in other literature. The necessary amount of work increases dramatically with the size of systems, so one has to search for algorithms that most efficiently and accurately solve systems of, e.g., several million equations. The choice of algorithms depends on the special properties the matrices in practice have. An important class of large systems arises from the discretization of partial differential equations. In this case, the matrices are sparse (i.e., they contain mostly zeroes) and well-suited to iterative algorithms. The first edition of this book grew out of a series of lectures given by the author at the Christian-Albrecht University of Kiel to students of mathematics. The second edition includes quite novel approaches.

  7. Geometric properties of Banach spaces and nonlinear iterations

    CERN Document Server

    Chidume, Charles

    2009-01-01

    Nonlinear functional analysis and applications is an area of study that has provided fascination for many mathematicians across the world. This monograph delves specifically into the topic of the geometric properties of Banach spaces and nonlinear iterations, a subject of extensive research over the past thirty years. Chapters 1 to 5 develop materials on convexity and smoothness of Banach spaces, associated moduli and connections with duality maps. Key results obtained are summarized at the end of each chapter for easy reference. Chapters 6 to 23 deal with an in-depth, comprehensive and up-to-date coverage of the main ideas, concepts and results on iterative algorithms for the approximation of fixed points of nonlinear nonexpansive and pseudo-contractive-type mappings. This includes detailed workings on solutions of variational inequality problems, solutions of Hammerstein integral equations, and common fixed points (and common zeros) of families of nonlinear mappings. Carefully referenced and full of recent,...

  8. ITER concept definition. V.2

    International Nuclear Information System (INIS)

    1989-01-01

    Volume II of the two volumes describing the concept definition of the International Thermonuclear Experimental Reactor deals with the ITER concept in technical depth, and covers all areas of design of the ITER tokamak. Included are an assessment of the current database for design, scoping studies, rationale for concepts selection, performance flexibility, the ITER concept, the operations and experimental/testing program, ITER parameters and design phase schedule, and research and development specific to ITER. This latter includes a definition of specific research and development tasks, a division of tasks among members, specific milestones, required results, and schedules. Figs and tabs

  9. ITER CTA newsletter. No. 10

    International Nuclear Information System (INIS)

    2002-07-01

    This ITER CTA newsletter issue comprises the ITER backgrounder, which was approved as an official document by the participants in the Negotiations on the ITER Implementation agreement at their fourth meeting, held in Cadarache from 4-6 June 2002, and information about two ITER meetings: one is the third meeting of the ITER parties' designated Safety Representatives, which took place in Cadarache, France from 6-7 June 2002, and the other is the second meeting of the International Tokamak Physics Activity (ITPA) topical group on diagnostics, which was held at General Atomics, San Diego, USA, from 4-8 March 2002

  10. 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)

  11. Operation and control strategies in pre-series testing of cold circulating pumps for ITER

    International Nuclear Information System (INIS)

    Bhattacharya, R.; Vaghela, H.; Sarkar, B.; Srinivas, M.; Choukekar, K.

    2013-01-01

    Cryo-distribution system of ITER is responsible for the distribution and control of forced-flow supercritical helium for cooling of the superconducting magnets and the cryo-pumps. The requirements of cold circulating pumps (CCP) for mass flow rates and performance are much higher than presently existing and commercially available one used at 4.0 K helium. Design up-scaling with pre-series test of CCP has been proposed including test infrastructure. Operation and control strategies for the test distribution box (TDB) of test infrastructure have been developed and analyzed using steady state and dynamic process simulation to cope with the functional requirements of CCPs. Off-normal scenario with CCP inlet pressure variation is an important concern, dynamic process responses during such scenario have been evaluated to verify the operability of CCP. The paper describes process simulation to cope with the functional requirements of CCPs along with evaluation of off-normal scenario to verify the operability of CCP. (author)

  12. A design study of hydrogen isotope separation system for ITER-FEAT

    International Nuclear Information System (INIS)

    Iwai, Yasunori; Yamanishi, Toshihiko; Nishi, Masataka

    2001-03-01

    Preliminary design study of the hydrogen isotope separation system (ISS) for the fuel cycle of the ITER-FEAT, a fusion experimental reactor, was carried out based on the substantial reduction of hydrogen flow to the ISS resulting from the design study for scale reduction of the formerly-designed ITER. Three feed streams (plasma exhaust gas stream, streams from the water detritiation system and that from the neutral beam injectors) are fed to the ISS, and three product streams (high purity tritium gas, high purity deuterium gas and hydrogen gas) are made in it by the method of cryogenic distillation. In this study, an original four-column cascade was proposed to the ISS cryogenic distillation column system considering simplification and the operation scenario of the ITER-FEAT. Substantial reduction of tritium inventory in the ISS was found to be possible in the progress of investigation concerning of the corresponding flow rate of tritium product stream (T>90 %) for pellet injector which depends upon the operation condition. And it was found that tritium concentration in the released hydrogen stream into environment from the ISS could easily fluctuate with current design of column arrangement due to the small disturbance in mass flow balance in the ISS. To solve this problem, two-column system for treatment of this flow was proposed. (author)

  13. 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

  14. Decommissioning of offshore oil and gas facilities: a comparative assessment of different scenarios.

    Science.gov (United States)

    Ekins, Paul; Vanner, Robin; Firebrace, James

    2006-06-01

    A material and energy flow analysis, with corresponding financial flows, was carried out for different decommissioning scenarios for the different elements of an offshore oil and gas structure. A comparative assessment was made of the non-financial (especially environmental) outcomes of the different scenarios, with the reference scenario being to leave all structures in situ, while other scenarios envisaged leaving them on the seabed or removing them to shore for recycling and disposal. The costs of each scenario, when compared with the reference scenario, give an implicit valuation of the non-financial outcomes (e.g. environmental improvements), should that scenario be adopted by society. The paper concludes that it is not clear that the removal of the topsides and jackets of large steel structures to shore, as currently required by regulations, is environmentally justified; that concrete structures should certainly be left in place; and that leaving footings, cuttings and pipelines in place, with subsequent monitoring, would also be justified unless very large values were placed by society on a clear seabed and trawling access.

  15. The Scenario Planning Paradox

    DEFF Research Database (Denmark)

    Spaniol, Matthew Jon; Rowland, Nicholas James

    2017-01-01

    planning paradox. Contributing fresh theory supposedly attends to the “dismal” state of theory, while contributing new typologies purportedly helps bring order to methodological chaos. Repeated over time, the contribution strategy breaks down. Effort to resolve the theoretical and methodological issue......, foundational theoretical perspective in futures studies. Perceived chaos gives way to typologies, which, as they mount, contribute to the chaos they were meant to resolve. The end result, intended by no one, is that theory remains dismal and methods remain chaotic. This direction for the field is indefensible......For more than a decade, futures studies scholars have prefaced scholarly contributions by repeating the claim that there is insufficient theory to support chaotic scenario methodology. The strategy is formulaic, and the net effect is a curious one, which the authors refer to as the scenario...

  16. Update of the HL-LHC operational scenarios for proton operation

    CERN Document Server

    Metral, Elias; Antoniou, Fanouria; Appleby, Robert Barrie; Arduini, Gianluigi; Barranco Garcia, Javier; Baudrenghien, Philippe; Biancacci, Nicolo; Bracco, Chiara; Bruce, Roderik; Buffat, Xavier; Calaga, Rama; Carver, Lee Robert; Chapochnikova, Elena; Crouch, Matthew Paul; De Maria, Riccardo; Fartoukh, Stephane; Gamba, Davide; Giovannozzi, Massimo; Goncalves Jorge, Patrik; Hofle, Wolfgang; Iadarola, Giovanni; Karastathis, Nikos; Lasheen, Alexandre; Mastoridis, Themistoklis; Medina Medrano, Luis Eduardo; Mereghetti, Alessio; Mirarchi, Daniele; Muratori, Bruno; Papadopoulou, Parthena Stefania; Papaphilippou, Yannis; Pellegrini, Dario; Pieloni, Tatiana; Redaelli, Stefano; Rumolo, Giovanni; Salvant, Benoit; Solfaroli Camillocci, Matteo; Tambasco, Claudia; Tomas Garcia, Rogelio; Valuch, Daniel; CERN. Geneva. ATS Department

    2018-01-01

    The main aim of this document is to have a clearly identified set of beam and machine parameters to be used for numerical simulations and performance assessment. Two scenarios (referring only to the operation at the nominal beam energy of 7 TeV) are discussed: i) Nominal scenario (levelling at a luminosity of 5E34 cm-2s-1) ii) Ultimate scenario (levelling at a luminosity of 7.5E34 cm-2s-1)

  17. The Scenario Approach to the Development of Regional Waste Management Systems (Implementation Experience in the Regions of Russia)

    Science.gov (United States)

    Fomin, Eugene P.; Alekseev, Audrey A.; Fomina, Natalia E.; Dorozhkin, Vladimir E.

    2016-01-01

    The article illustrates a theoretical approach to scenario modeling of economic indicators of regional waste management system. The method includes a three-iterative algorithm that allows the executive authorities and investors to take a decision on logistics, bulk, technological and economic parameters of the formation of the regional long-term…

  18. Thermo-mechanical design of the SINGAP accelerator grids for ITER NB injectors

    Energy Technology Data Exchange (ETDEWEB)

    Agostinetti, P. [Consorzio RFX, Euratom-ENEA Association, Corso Stati Uniti 4, I35127 Padova (Italy)], E-mail: piero.agostinetti@igi.cnr.it; Dal Bello, S.; Dalla Palma, M.; Zaccaria, P. [Consorzio RFX, Euratom-ENEA Association, Corso Stati Uniti 4, I35127 Padova (Italy)

    2007-10-15

    The SINGle Aperture-SINgle GAP (SINGAP) accelerator for ITER neutral beam injector foresees four grids for the extraction and acceleration of negative ions, instead of the seven grids of the Multi-Aperture Multi-Grid (MAMuG) reference configuration. The grids have to fulfil specific requirements coming from ion extraction, beam optics and thermo-mechanical issues. This paper focuses on the thermo-hydraulic and thermo-mechanical design of the grids carried out by Consorzio RFX for the design of the first ITER NB injector and the ITER NB Test Facility. The cooling circuit design (position and shape of the channels) and the cooling parameters (water coolant temperatures, pressure and velocity) were optimized with sensitivity analyses in order to satisfy the grid functional requirements (temperatures, stresses, in plane and out of plane deformations). The design required a complete modelling of the grids and their support frames by means of 3D FE and CAD models.

  19. Discovering the new RCP and SSP scenarios used by the IPCC

    International Nuclear Information System (INIS)

    2013-01-01

    This report presents the scenarios defined by a group of experts within the perspective of the 5. IPCC report. This four reference scenarios named RCP (Representative Concentration Pathways) have been designed to foresee the evolution of concentrations of greenhouse gases, of ozone, and of aerosol precursors for the 21. century and beyond. The report also evokes the evolution of simulations used by climatologists, and the introduction of a representation of social and economic evolutions (definition of five families of scenario-types: sustainability, middle of the road, fragmentation, inequality, and conventional development). The consistency of these RCP scenarios and social-economical scenarios is outlined

  20. Global Water Scarcity Assessment under Post-SRES Scenarios

    Science.gov (United States)

    Hanasaki, N.; Fujimori, S.

    2011-12-01

    A large number of future projections contributed to the fourth Assessment Report of IPCC were based on Special Report on Emission Scenarios (SRES). Processes toward the fifth Assessment Report are under way, and post-SRES scenarios, called Shared Socio-economic Pathways (SSP) are being prepared. One of the key challenges of SSP is provision of detailed socio-economic scenarios compared to SRES for impact, adaptation and vulnerability studies. In this study, a comprehensive global water scarcity assessment was conducted, using a state of the art global water resources model H08 (Hanasaki et al., 2008a, 2008b, 2010). We used a prototype of SSP developed by National Institute for Environmental Studies, Japan. Two sets of socio economic scenarios and two sets of climate scenarios were prepared to run H08 for the period 2001-2100. Socio-economic scenarios include Business As Usual and High Mitigation Capacity. Climate scenarios include Reference and Mitigation which stabilizes green house gas concentration at a certain level. We analyzed the simulation results of four combinations, particularly focusing on the sensitivity of socio-economic scenarios to major water resources indices.

  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. Large Ensemble Analytic Framework for Consequence-Driven Discovery of Climate Change Scenarios

    Science.gov (United States)

    Lamontagne, Jonathan R.; Reed, Patrick M.; Link, Robert; Calvin, Katherine V.; Clarke, Leon E.; Edmonds, James A.

    2018-03-01

    An analytic scenario generation framework is developed based on the idea that the same climate outcome can result from very different socioeconomic and policy drivers. The framework builds on the Scenario Matrix Framework's abstraction of "challenges to mitigation" and "challenges to adaptation" to facilitate the flexible discovery of diverse and consequential scenarios. We combine visual and statistical techniques for interrogating a large factorial data set of 33,750 scenarios generated using the Global Change Assessment Model. We demonstrate how the analytic framework can aid in identifying which scenario assumptions are most tied to user-specified measures for policy relevant outcomes of interest, specifically for our example high or low mitigation costs. We show that the current approach for selecting reference scenarios can miss policy relevant scenario narratives that often emerge as hybrids of optimistic and pessimistic scenario assumptions. We also show that the same scenario assumption can be associated with both high and low mitigation costs depending on the climate outcome of interest and the mitigation policy context. In the illustrative example, we show how agricultural productivity, population growth, and economic growth are most predictive of the level of mitigation costs. Formulating policy relevant scenarios of deeply and broadly uncertain futures benefits from large ensemble-based exploration of quantitative measures of consequences. To this end, we have contributed a large database of climate change futures that can support "bottom-up" scenario generation techniques that capture a broader array of consequences than those that emerge from limited sampling of a few reference scenarios.

  3. Conceptual design of the radial gamma ray spectrometers system for α particle and runaway electron measurements at ITER

    Science.gov (United States)

    Nocente, M.; Tardocchi, M.; Barnsley, R.; Bertalot, L.; Brichard, B.; Croci, G.; Brolatti, G.; Di Pace, L.; Fernandes, A.; Giacomelli, L.; Lengar, I.; Moszynski, M.; Krasilnikov, V.; Muraro, A.; Pereira, R. C.; Perelli Cippo, E.; Rigamonti, D.; Rebai, M.; Rzadkiewicz, J.; Salewski, M.; Santosh, P.; Sousa, J.; Zychor, I.; Gorini, G.

    2017-07-01

    We here present the principles and main physics capabilities behind the design of the radial gamma ray spectrometers (RGRS) system for alpha particle and runaway electron measurements at ITER. The diagnostic benefits from recent advances in gamma-ray spectrometry for tokamak plasmas and combines space and high energy resolution in a single device. The RGRS system as designed can provide information on α ~ particles on a time scale of 1/10 of the slowing down time for the ITER 500 MW full power DT scenario. Spectral observations of the 3.21 and 4.44 MeV peaks from the 9\\text{Be}≤ft(α,nγ \\right){{}12}\\text{C} reaction make the measurements sensitive to α ~ particles at characteristic resonant energies and to possible anisotropies of their slowing down distribution function. An independent assessment of the neutron rate by gamma-ray emission is also feasible. In case of runaway electrons born in disruptions with a typical duration of 100 ms, a time resolution of at least 10 ms for runaway electron studies can be achieved depending on the scenario and down to a current of 40 kA by use of external gas injection. We find that the bremsstrahlung spectrum in the MeV range from confined runaways is sensitive to the electron velocity space up to E≈ 30 -40 MeV, which allows for measurements of the energy distribution of the runaway electrons at ITER.

  4. 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.

  5. The European Union, regionalism, and world order: five scenarios

    Directory of Open Access Journals (Sweden)

    Mario Telò

    2011-06-01

    Full Text Available This paper addresses the changing interplay between regional and global governance, with reference to five possible scenarios in the future shift of the international system. A ‘new multilateralism’ represents the only possible global framework consistent with an expanding and pluridimensional regional governance system. Furthermore, the weaknesses and contradictions of the alternative scenarios make a multilayered, more robust and legitimate multilateral governance more realistic.

  6. Development of ITER in-vessel viewing and metrology systems

    Energy Technology Data Exchange (ETDEWEB)

    Obara, Kenjiro; Kakudate, Satoshi; Nakahira, Masataka; Ito, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    The ITER in-vessel viewing system is vital for detecting and locating damage to in-vessel components such as the blankets and divertors and in monitoring and assisting in-vessel maintenance. This system must be able to operate at high temperature (200degC) under intense gamma radiation ({approx}30 kGy/h) in a high vacuum or 1 bar inert gas. A periscope viewing system was chosen as a reference due to its clear, wide view and a fiberscope viewing system chosen as a backup for viewing in narrow confines. According to the ITER R and D program, both systems and a metrology system are being developed through the joint efforts of Japan, the U.S., and RF Home Teams. This paper outlines design and technology development mainly on periscope in-vessel viewing and laser metrology contributed by the Japan Home Team. (author)

  7. Development of ITER in-vessel viewing and metrology systems

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Kakudate, Satoshi; Nakahira, Masataka; Ito, Akira

    1998-01-01

    The ITER in-vessel viewing system is vital for detecting and locating damage to in-vessel components such as the blankets and divertors and in monitoring and assisting in-vessel maintenance. This system must be able to operate at high temperature (200degC) under intense gamma radiation (∼30 kGy/h) in a high vacuum or 1 bar inert gas. A periscope viewing system was chosen as a reference due to its clear, wide view and a fiberscope viewing system chosen as a backup for viewing in narrow confines. According to the ITER R and D program, both systems and a metrology system are being developed through the joint efforts of Japan, the U.S., and RF Home Teams. This paper outlines design and technology development mainly on periscope in-vessel viewing and laser metrology contributed by the Japan Home Team. (author)

  8. 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

  9. Beryllium assessment and recommendation for application in ITER plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Barabash, V.; Tanaka, S.; Matera, R. [ITER Joint Central Team, Muenchen (Germany)

    1998-01-01

    The design status of the ITER Plasma Facing Components (PFC) is presented. The operational conditions of the armour material for the different components are summarized. Beryllium is the reference armour material for the Primary Wall, Baffle and Limiter and the back-up material for the Divertor Dome. The activities on the selection of the Be grades and the joining technologies are reviewed. (author)

  10. 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

  11. ITER blanket module shield block design and analysis

    International Nuclear Information System (INIS)

    Mitin, D.; Khomyakov, S.; Razmerov, A.; Strebkov, Yu.

    2008-01-01

    This paper presents the alternative design of the shield block cooling path for a typical ITER blanket module with a predominantly sequential flow circuit. A number of serious disadvantages have been observed for the reference design, where the parallel flow circuit is used, which is inherent in the majority of blanket modules. The paper discusses these disadvantages and demonstrates the benefit of the alternative design based on the detailed design and the technological, hydraulic, thermal, structural and strength analyses, conducted for module no. 17

  12. IHadoop: Asynchronous iterations for MapReduce

    KAUST Repository

    Elnikety, Eslam Mohamed Ibrahim

    2011-11-01

    MapReduce is a distributed programming frame-work designed to ease the development of scalable data-intensive applications for large clusters of commodity machines. Most machine learning and data mining applications involve iterative computations over large datasets, such as the Web hyperlink structures and social network graphs. Yet, the MapReduce model does not efficiently support this important class of applications. The architecture of MapReduce, most critically its dataflow techniques and task scheduling, is completely unaware of the nature of iterative applications; tasks are scheduled according to a policy that optimizes the execution for a single iteration which wastes bandwidth, I/O, and CPU cycles when compared with an optimal execution for a consecutive set of iterations. This work presents iHadoop, a modified MapReduce model, and an associated implementation, optimized for iterative computations. The iHadoop model schedules iterations asynchronously. It connects the output of one iteration to the next, allowing both to process their data concurrently. iHadoop\\'s task scheduler exploits inter-iteration data locality by scheduling tasks that exhibit a producer/consumer relation on the same physical machine allowing a fast local data transfer. For those iterative applications that require satisfying certain criteria before termination, iHadoop runs the check concurrently during the execution of the subsequent iteration to further reduce the application\\'s latency. This paper also describes our implementation of the iHadoop model, and evaluates its performance against Hadoop, the widely used open source implementation of MapReduce. Experiments using different data analysis applications over real-world and synthetic datasets show that iHadoop performs better than Hadoop for iterative algorithms, reducing execution time of iterative applications by 25% on average. Furthermore, integrating iHadoop with HaLoop, a variant Hadoop implementation that caches

  13. IHadoop: Asynchronous iterations for MapReduce

    KAUST Repository

    Elnikety, Eslam Mohamed Ibrahim; El Sayed, Tamer S.; Ramadan, Hany E.

    2011-01-01

    MapReduce is a distributed programming frame-work designed to ease the development of scalable data-intensive applications for large clusters of commodity machines. Most machine learning and data mining applications involve iterative computations over large datasets, such as the Web hyperlink structures and social network graphs. Yet, the MapReduce model does not efficiently support this important class of applications. The architecture of MapReduce, most critically its dataflow techniques and task scheduling, is completely unaware of the nature of iterative applications; tasks are scheduled according to a policy that optimizes the execution for a single iteration which wastes bandwidth, I/O, and CPU cycles when compared with an optimal execution for a consecutive set of iterations. This work presents iHadoop, a modified MapReduce model, and an associated implementation, optimized for iterative computations. The iHadoop model schedules iterations asynchronously. It connects the output of one iteration to the next, allowing both to process their data concurrently. iHadoop's task scheduler exploits inter-iteration data locality by scheduling tasks that exhibit a producer/consumer relation on the same physical machine allowing a fast local data transfer. For those iterative applications that require satisfying certain criteria before termination, iHadoop runs the check concurrently during the execution of the subsequent iteration to further reduce the application's latency. This paper also describes our implementation of the iHadoop model, and evaluates its performance against Hadoop, the widely used open source implementation of MapReduce. Experiments using different data analysis applications over real-world and synthetic datasets show that iHadoop performs better than Hadoop for iterative algorithms, reducing execution time of iterative applications by 25% on average. Furthermore, integrating iHadoop with HaLoop, a variant Hadoop implementation that caches

  14. ITER ITA newsletter No. 31, June 2006

    International Nuclear Information System (INIS)

    2006-07-01

    This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about initialling the ITER Agreement and its related instruments by seven ITER parties, which too place in Brussels on 24 May 2006. The initialling constituted the final act of the ITER negotiations. It confirmed the Parties' common acceptance of the negotiated texts, ad referendum, and signalled their intentions to move forward towards the entry into force of the ITER Agreement as soon as possible. 'ITER - Uniting science today, global energy tomorrow' was the theme of a number of media events timed to accompany a remarkable day in the history of the ITER international venture, May 24th 2006, initialling of the ITER international agreement

  15. Status of the ITER EDA

    International Nuclear Information System (INIS)

    Aymar, R.

    2000-01-01

    This article summarizes progress made in the ITER Engineering Design Activities in the period between the ITER Meeting in Tokyo (January 2000) and June 2000. Topics: Termination of EDA, Joint Central Team and Support, Task Assignments, ITER Physics, Urgent and High Priority Physics Research Areas

  16. ITER EDA newsletter. V. 10, special issue

    International Nuclear Information System (INIS)

    2001-07-01

    This ITER EDA Newsletter includes summaries of the reports of ITER EDA JCT Physics unit about ITER physics R and D during the Engineering Design Activities (EDA), ITER EDA JCT Naka JWC ITER technology R and D during the EDA, and Safety, Environment and Health group of ITER EDA JCT, Garching JWS on EDA activities related to safety

  17. ITER CTA newsletter. No. 13, October 2002

    International Nuclear Information System (INIS)

    2002-11-01

    This ITER CTA newsletter issue comprises concise information about an ITER related meeting concerning the joint implementation of ITER - the fifth ITER Negotiations Meeting - which was held in Toronto, Canada, 19-20 September, 2002, and information about assessment of the possible ITER site in Clarington, Ontario, Canada, which was the subject of the first official stage of the Joint Assessment of Specific Sites (JASS) for the ITER Project. This assessment was completed just before the Fifth ITER Negotiations Meeting

  18. Analysis of the direction of plasma vertical movement during major disruptions in ITER

    International Nuclear Information System (INIS)

    Lukash, Victor; Sugihara, Masayoshi; Gribov, Yuri; Fujieda, Hirobumi

    2005-01-01

    The plasma movement in the upward direction (away from the X-point) after the thermal quench (TQ) of major disruptions in ITER is favourable for the machine design, since the downward movement causes larger electromagnetic (EM) load due to the induced eddy and halo currents. Vertical directions of plasma movement after the TQ in ITER are investigated using the predictive mode of the DINA code. Three dominant parameters in determining the direction of plasma movement are identified: (i) the rate of plasma current quench (plasma temperature after the TQ) (ii) the width of plasma current mixing area just after the TQ (change of the internal plasma inductance l i ) and (iii) the initial vertical position of plasma column before the TQ. It is shown that the reference ITER plasma moves upwards after the TQ, if the electron temperature after the TQ is less than 10 eV and the drop of l i does not exceed 0.2 for the present reference initial vertical position (55.5 cm above the centre of the machine). It is also shown that the operational domain leading to the upward movement is considerably large for disruptions with fast current quench, which could generate quite severe EM load due to the induced eddy current combined with the induced halo current if the movement is downwards

  19. The ITER remote maintenance system

    International Nuclear Information System (INIS)

    Tesini, A.; Palmer, J.

    2008-01-01

    The aim of this paper is to summarize the ITER approach to machine components maintenance. A major objective of the ITER project is to demonstrate that a future power producing fusion device can be maintained effectively and offer practical levels of plant availability. During its operational lifetime, many systems of the ITER machine will require maintenance and modification; this can be achieved using remote handling methods. The need for timely, safe and effective remote operations on a machine as complex as ITER and within one of the world's most hostile remote handling environments represents a major challenge at every level of the ITER Project organization, engineering and technology. The basic principles of fusion reactor maintenance are presented. An updated description of the ITER remote maintenance system is provided. This includes the maintenance equipment used inside the vacuum vessel, inside the hot cell and the hot cell itself. The correlation between the functions of the remote handling equipment, of the hot cell and of the radwaste processing system is also described. The paper concludes that ITER has equipped itself with a good platform to tackle the challenges presented by its own maintenance and upgrade needs

  20. Localization of cask and plug remote handling system in ITER using multiple video cameras

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, João, E-mail: jftferreira@ipfn.ist.utl.pt [Instituto de Plasmas e Fusão Nuclear - Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Vale, Alberto [Instituto de Plasmas e Fusão Nuclear - Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Ribeiro, Isabel [Laboratório de Robótica e Sistemas em Engenharia e Ciência - Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal)

    2013-10-15

    Highlights: ► Localization of cask and plug remote handling system with video cameras and markers. ► Video cameras already installed on the building for remote operators. ► Fiducial markers glued or painted on cask and plug remote handling system. ► Augmented reality contents on the video streaming as an aid for remote operators. ► Integration with other localization systems for enhanced robustness and precision. -- Abstract: The cask and plug remote handling system (CPRHS) provides the means for the remote transfer of in-vessel components and remote handling equipment between the Hot Cell building and the Tokamak building in ITER. Different CPRHS typologies will be autonomously guided following predefined trajectories. Therefore, the localization of any CPRHS in operation must be continuously known in real time to provide the feedback for the control system and also for the human supervision. This paper proposes a localization system that uses the video streaming captured by the multiple cameras already installed in the ITER scenario to estimate with precision the position and the orientation of any CPRHS. In addition, an augmented reality system can be implemented using the same video streaming and the libraries for the localization system. The proposed localization system was tested in a mock-up scenario with a scale 1:25 of the divertor level of Tokamak building.

  1. Localization of cask and plug remote handling system in ITER using multiple video cameras

    International Nuclear Information System (INIS)

    Ferreira, João; Vale, Alberto; Ribeiro, Isabel

    2013-01-01

    Highlights: ► Localization of cask and plug remote handling system with video cameras and markers. ► Video cameras already installed on the building for remote operators. ► Fiducial markers glued or painted on cask and plug remote handling system. ► Augmented reality contents on the video streaming as an aid for remote operators. ► Integration with other localization systems for enhanced robustness and precision. -- Abstract: The cask and plug remote handling system (CPRHS) provides the means for the remote transfer of in-vessel components and remote handling equipment between the Hot Cell building and the Tokamak building in ITER. Different CPRHS typologies will be autonomously guided following predefined trajectories. Therefore, the localization of any CPRHS in operation must be continuously known in real time to provide the feedback for the control system and also for the human supervision. This paper proposes a localization system that uses the video streaming captured by the multiple cameras already installed in the ITER scenario to estimate with precision the position and the orientation of any CPRHS. In addition, an augmented reality system can be implemented using the same video streaming and the libraries for the localization system. The proposed localization system was tested in a mock-up scenario with a scale 1:25 of the divertor level of Tokamak building

  2. Design and development of in-vessel viewing periscope for ITER (International Thermonuclear Experimental Reactor)

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Kakudate, Satoshi; Ito, Akira; Shibanuma, Kiyoshi; Tada, Eisuke

    1999-02-01

    An in-vessel viewing system is essential not only to detect and locate damage of components exposed to plasma, but also to monitor and assist in-vessel maintenance operation. In ITER, the in-vessel viewing system must be capable of operating at high temperature (200degC), under intense gamma radiation (30 kGy/h) and high vacuum or 1 bar inert gas. A periscope-type in-vessel viewing system has been chosen as a reference of the ITER in-vessel viewing system due to its wide viewing capability and durability for sever environments. According to the ITER research and development program, a full-scale radiation hard periscope with a length of 15 m has been successfully developed by the Japan Home Team. The performance tests have been shown sufficient capability at high temperature up to 250degC and radiation resistance over 100 MGy. This report describes the design and R and D results of the ITER in-vessel viewing periscope based on the development of 15-m-length radiation hard periscope. (author)

  3. ITER EDA Newsletter. Vol. 1, No. 1

    International Nuclear Information System (INIS)

    1992-11-01

    After the ITER Engineering Design Activities (EDA) Agreement and Protocol 1 had been signed by the four ITER parties on July 21, 1992 and had entered into force, the ITER Council suggested at its first meeting (Vienna, September 10-11, 1992) that the publication of the ITER Newsletter be continued during the EDA with assistance of the International Atomic Energy Agency. This suggestion was supported by the Agency and subsequently the ITER office in Vienna assumed its responsibilities for planning and executing activities related to the publication of the Newsletter. The ITER EDA Newsletter is planned to be a monthly publication aimed at disseminating broad information and understanding, including the description of the personal and institutional involvements in the ITER project in addition to technical facts about it. The responsibility for the Newsletter rests with the ITER council. In this first issue the signing of the ITER EDA Activities and Protocol 1 is reported. The EDA organizational structure is described. This issue also reports on the first ITER EDA council meeting, the opening of the ITER EDA NAKA Co-Centre, the first meeting of the ITER Technical Advisory Committee, activities of special working groups, an ITER Technical Meeting, as well as ''News in Brief'' and ''Coming Events''

  4. Report of the international symposium for ITER. 'Burning plasma science and technology on ITER'

    International Nuclear Information System (INIS)

    2002-10-01

    This report contains the presentations on the International Symposium for ITER, held on Jan. 24, 2002 on the occasion of the ITER Governmental Negotiations in Tokyo. This symposium is organized by Japan Atomic Energy Research Institute with the support of the Ministry of Education, Culture, Sports, Science and Technology (MEXT). The meaningful results were obtained through this symposium especially on new frontiers of science and technology brought by ITER, accelerated road maps towards realizing fusion energy, and portfolio of other fusion configurations from ITER. The 5 of the presented papers are indexed individually (J.P.N.)

  5. ITER Council tour of Clarington site

    International Nuclear Information System (INIS)

    Dautovich, D.

    2001-01-01

    The ITER Council meeting was recently held in Toronto on 27 and 28 February. ITER Canada provided local arrangements for the Council meeting on behalf of Europe as the Official host. Following the meeting, on 1 March, ITER Canada conducted a tour of the proposed ITER construction site at Charington, and the ITER Council members attended a luncheon followed by a speech by Dr. Peter Barnard, Chairman and CEO of ITER Canada, at the Empire Club of Canada. The official invitation to participate in these events came from Dr. Peter Harrison, Deputy Minister of Natural Resources Canada. This report provides a brief summary of the events on 1 March

  6. 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)

  7. 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)

  8. Model-based iterative reconstruction for reduction of radiation dose in abdominopelvic CT: comparison to adaptive statistical iterative reconstruction.

    Science.gov (United States)

    Yasaka, Koichiro; Katsura, Masaki; Akahane, Masaaki; Sato, Jiro; Matsuda, Izuru; Ohtomo, Kuni

    2013-12-01

    To evaluate dose reduction and image quality of abdominopelvic computed tomography (CT) reconstructed with model-based iterative reconstruction (MBIR) compared to adaptive statistical iterative reconstruction (ASIR). In this prospective study, 85 patients underwent referential-, low-, and ultralow-dose unenhanced abdominopelvic CT. Images were reconstructed with ASIR for low-dose (L-ASIR) and ultralow-dose CT (UL-ASIR), and with MBIR for ultralow-dose CT (UL-MBIR). Image noise was measured in the abdominal aorta and iliopsoas muscle. Subjective image analyses and a lesion detection study (adrenal nodules) were conducted by two blinded radiologists. A reference standard was established by a consensus panel of two different radiologists using referential-dose CT reconstructed with filtered back projection. Compared to low-dose CT, there was a 63% decrease in dose-length product with ultralow-dose CT. UL-MBIR had significantly lower image noise than L-ASIR and UL-ASIR (all pASIR and UL-ASIR (all pASIR in diagnostic acceptability (p>0.65), or diagnostic performance for adrenal nodules (p>0.87). MBIR significantly improves image noise and streak artifacts compared to ASIR, and can achieve radiation dose reduction without severely compromising image quality.

  9. 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)

  10. Multiple Revolution Solutions for the Perturbed Lambert Problem using the Method of Particular Solutions and Picard Iteration

    Science.gov (United States)

    Woollands, Robyn M.; Read, Julie L.; Probe, Austin B.; Junkins, John L.

    2017-12-01

    We present a new method for solving the multiple revolution perturbed Lambert problem using the method of particular solutions and modified Chebyshev-Picard iteration. The method of particular solutions differs from the well-known Newton-shooting method in that integration of the state transition matrix (36 additional differential equations) is not required, and instead it makes use of a reference trajectory and a set of n particular solutions. Any numerical integrator can be used for solving two-point boundary problems with the method of particular solutions, however we show that using modified Chebyshev-Picard iteration affords an avenue for increased efficiency that is not available with other step-by-step integrators. We take advantage of the path approximation nature of modified Chebyshev-Picard iteration (nodes iteratively converge to fixed points in space) and utilize a variable fidelity force model for propagating the reference trajectory. Remarkably, we demonstrate that computing the particular solutions with only low fidelity function evaluations greatly increases the efficiency of the algorithm while maintaining machine precision accuracy. Our study reveals that solving the perturbed Lambert's problem using the method of particular solutions with modified Chebyshev-Picard iteration is about an order of magnitude faster compared with the classical shooting method and a tenth-twelfth order Runge-Kutta integrator. It is well known that the solution to Lambert's problem over multiple revolutions is not unique and to ensure that all possible solutions are considered we make use of a reliable preexisting Keplerian Lambert solver to warm start our perturbed algorithm.

  11. Integrated modelling of economic-energy-environment scenarios - The impact of China and India's economic growth on energy use and CO2 emissions

    International Nuclear Information System (INIS)

    Roques, F.; Sassi, O.; Guivarch, C.; Waisman, H.; Crassous, R.; Hourcade, J.Ch.

    2009-03-01

    A hybrid framework coupling the bottom-up energy sector WEM model with the top-down general equilibrium model IMACLIM-R is implemented to capture the macro-economic feedbacks of Chinese and Indian economic growth on energy and emissions scenarios. The iterative coupling procedure captures the detailed representation of energy use and supply while ensuring the micro-economic and macro-economic consistency of the different scenarios studied. The dual representation of the hybrid model facilitates the incorporation of energy sector expertise in internally consistent scenarios. The paper describes how the hybrid model was used to assess the effect of uncertainty on economic growth in China and India in the energy and emissions scenarios of the International Energy Agency. (authors)

  12. 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.

  13. Parametric studh on coupling loss in subsize ITER Nb3Sn cabled specimens

    NARCIS (Netherlands)

    Nijhuis, Arend; ten Kate, Herman H.J.; Bruzzone, Pierluigi; Bottura, Luca

    1996-01-01

    The cable in conduit conductors for the various ITER coils are required to function under pulse conditions and fields up to 13 T. A parametric study, restricted to a limited variation of the reference cable lay out, is carried out to clarify the quantitative impact of various cable parameters on the

  14. Structural materials for ITER in-vessel component design

    Energy Technology Data Exchange (ETDEWEB)

    Kalinin, G. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Gauster, W. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Matera, R. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Tavassoli, A.-A.F. [CEA Centre d`Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France); Rowcliffe, A. [Oak Ridge National Lab., TN (United States); Fabritsiev, S. [Research Inst. of Electrophysical Apparatus, St. Petersburg (Russian Federation); Kawamura, H. [JAERI, IMTR Project, Ibaraki (Japan). Blanket Irradiation Lab.

    1996-10-01

    The materials proposed for ITER in-vessel components have to exhibit adequate performance for the operating lifetime of the reactor or for specified replacement intervals. Estimates show that maximum irradiation dose to be up to 5-7 dpa (for 1 MWa/m{sup 2} in the basic performance phase (BPP)) within a temperature range from 20 to 300 C. Austenitic SS 316LN-ITER Grade was defined as a reference option for the vacuum vessel, blanket, primary wall, pipe lines and divertor body. Conventional technologies and mill products are proposed for blanket, back plate and manifold manufacturing. HIPing is proposed as a reference manufacturing method for the primary wall and blanket and as an option for the divertor body. The existing data show that mechanical properties of HIPed SS are no worse than those of forged 316LN SS. Irradiation will result in property changes. Minimum ductility has been observed after irradiation in an approximate temperature range between 250 and 350 C, for doses of 5-10 dpa. In spite of radiation-induced changes in tensile deformation behavior, the fracture remains ductile. Irradiation assisted corrosion cracking is a concern for high doses of irradiation and at high temperatures. Re-welding is one of the critical issues because of the need to replace failed components. It is also being considered for the replacement of shielding blanket modules by breeding modules after the BPP. (orig.).

  15. Light Rotor: The 10-MW reference wind turbine

    DEFF Research Database (Denmark)

    Bak, Christian; Bitsche, Robert; Yde, Anders

    2012-01-01

    design show a rather well performing wind turbine both in terms of power and loads, but in the further work towards the final design the challenges in the control needs to be solved and the balance between power performance and loads and between structural performance and mass will be investigated......This paper describes the design of a rotor and a wind turbine for an artificial 10-MW wind turbine carried out in the Light Rotor project. The turbine called the Light Rotor 10-MW Reference Wind Turbine (LR10-MW RWT), is designed with existing methods and techniques and serves as a reference...... like the determination of the specific power and upscaling of the turbine. The design of Iteration #2 of the LR10-MW RWT is carried out in a sequence between aerodynamic rotor design, structural design and aero-servo-elastic design. Each of these topics is described. The results from the Iteration #2...

  16. Virtual fringe projection system with nonparallel illumination based on iteration

    International Nuclear Information System (INIS)

    Zhou, Duo; Wang, Zhangying; Gao, Nan; Zhang, Zonghua; Jiang, Xiangqian

    2017-01-01

    Fringe projection profilometry has been widely applied in many fields. To set up an ideal measuring system, a virtual fringe projection technique has been studied to assist in the design of hardware configurations. However, existing virtual fringe projection systems use parallel illumination and have a fixed optical framework. This paper presents a virtual fringe projection system with nonparallel illumination. Using an iterative method to calculate intersection points between rays and reference planes or object surfaces, the proposed system can simulate projected fringe patterns and captured images. A new explicit calibration method has been presented to validate the precision of the system. Simulated results indicate that the proposed iterative method outperforms previous systems. Our virtual system can be applied to error analysis, algorithm optimization, and help operators to find ideal system parameter settings for actual measurements. (paper)

  17. Scenarios constructed for nominal flow in the presence of a repository at Yucca Mountain and vicinity

    International Nuclear Information System (INIS)

    Barr, G.E.; Hunter, R.L.; Dunn, E.; Flint, A.

    1995-03-01

    Scenario development for the system performance assessment of the Yucca Mountain Site Characterization Project defines a scenario as a well-posed problem connecting an initiating event with radionuclide release to the accessible environment by a logical and physically possible combination or sequence of features, events, and processes. Drawing on the advice and assistance of the Project's principal investigators (PIs), a collection of release scenarios initiated by the nominal ground-water flow occurring in the vicinity of the potential Yucca Mountain high-level-waste repository is developed and described in pictorial form. This collection of scenarios is intended to provide a framework to assist PIs in recognizing essential field and calculational analyses, to assist performance assessment in providing guidance to site characterization, and to continue the effort to exhaustively identify all features, events, and processes important to releases. It represents a step in the iterative process of identifying what details of the potential site are important for safe disposal. 67 refs

  18. Control of MHD instabilities by ECCD: ASDEX Upgrade results and implications for ITER

    International Nuclear Information System (INIS)

    Zohm, H.; Gantenbein, G.; Leuterer, F.; Manini, A.; Maraschek, M.; Yu, Q.

    2007-01-01

    The requirements for control of MHD instabilities by electron cyclotron current drive (ECCD) are reviewed. It is shown that a localized current drive is needed for control of both sawteeth and neoclassical tearing modes (NTMs). In the case of NTMs, the deposition width should be smaller than the island width for efficient control. At island widths smaller than the deposition width, as is predicted to occur in ITER, theory suggests that efficient control is possible only by modulating the ECCD power in phase with the island. These predictions are experimentally confirmed in ASDEX Upgrade for NTM control. Narrow deposition has also been used to extend the operational range of NTM stabilization in ASDEX Upgrade to lower q 95 and in the improved H-mode scenario. Our results suggest that, for the ITER ECCD system, good localization of the driven current profile as well as the capability to modulate the ECCD in phase with rotating modes will be needed for efficient MHD control by ECCD

  19. ANCRE scenarios for energy transition. Report 2013

    International Nuclear Information System (INIS)

    Alazard-Toux, Nathalie; Des Courtils, Nicolas; Hache, Emmanuel; Liegeard, Alban; Lorne, Daphne; Duplan, Jean-Luc; Kalaydjian, Francois; Heintze, Eric; Tilagone, Richard; Henriot, Stephane; Forti, Laurent; Barthelemy, Pascal; Merlen, Elisabeth; Criqui, Patrick; Mathy, Sandrine; Menanteau, Philippe; Devezeaux De Lavergne, Jean-Guy; Avril, Sophie; Cavata, Christian; Le Duigou, Alain; Le Net, Elisabeth; Marcucci-Demeure, Jeanne; Safa, Henri; Topper, Benjamin; Touboul, Francoise; Carre, Franck; Joly, Jean-Pierre; Charbit, Francoise; Mermilliod, Nicole; Mermilliod, Nicole; Le Net, Elisabeth; Teissier, Olivier; Charrue, Herve; Colonna, Paul; Legrand, Jack; Vidal, Olivier; Goffe, Bruno; Mueller, Alex; Flamant, Gilles; Allard, Francis; Most, Jean-Michel; Matarasso, Pierre; Brault, Pascal; Lemoine, Lionel; Achard, Jean-Luc; Uster, Guillaume; Delsey, Jean; Lucchese, Paul; Tadrist, Lounes; Hadjsaid, Nouredine

    2014-01-01

    This report first gives an overview of the energy system by presenting the determining factors of energy demand and of CO 2 emissions per sector (housing and office building, transport, industry, agriculture, forestry and biomass), by analysing energy systems and CO 2 emissions (energy sources, energy vectors, networks and storage, energy and CO 2 assessment for France), and by describing the guidelines of the scenarios proposed by ANCRE. The three main scenarios are characterized by a stronger sobriety, an electricity-based de-carbonation (with a variant based on nuclear and renewable energies), and diversified vectors. They are notably compared to a trend-based reference scenario. Results are discussed in terms of energy consumption (primary and final energy, consumption by the different sectors), of energy production and CO 2 emissions. Scenarios are assessed in terms of economic, environmental and societal, political and strategic criteria. Some consequences for research topics and funding are identified

  20. Neutronic performance of Indian LLCB TBM set conceptual design in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Swami, H.L., E-mail: hswami@ipr.res.in; Shaw, A.K.; Mistry, A.N.; Danani, C.

    2016-12-15

    Highlights: • Neutronic analyses of conceptual design of LLCB test blanket module in ITER have been performed. • The estimated total tritium production rate in the LLCB TBM is 1.66E + 17 tritons/s. • Total heat deposited in the LLCB TBM is 0.46 MW and highest power density at TBM first wall is 5.2 Watt/cc. • The estimation shows the maximum DPA 2.72 at TBM FW. - Abstract: Tritium breeding blanket testing program in ITER is an important milestone towards the development of the fusion reactors. ITER organization is providing an opportunity to the partner countries to test their breeding blanket concepts. A mock-up of Indian Lead Lithium Ceramic Breeder (LLCB) tritium breeding blanket known as LLCB Test Blanket Module (TBM) will be tested in ITER equatorial port no. 2. LLCB blanket consists of lead lithium (PbLi) as a neutron multiplier & tritium breeder, ceramic breeder (Li{sub 2}TiO{sub 3}) as a tritium breeder and India specific Reduced Activation Ferretic Martinic Steel (IN-RAFMS) as a structural material. A stainless steel block which is cooled by water, called as shield block, is attached with TBM to provide neutron shield to ITER TBM port. A comprehensive neutronic performance evaluation is required for the design of the LLCB TBM set (TBM + shield block) and associated ancillary systems in ITER. The neutronic performance of the conceptual design of TBM set in ITER has been carried out and reported here. In order to carry out the neutronic performance evaluation, the neutronic models of the LLCB TBM set along with TBM frame have been constructed and inserted in the equatorial port of ITER reference neutronic model C-lite. Neutronic responses such as tritium production rate, nuclear heating, neutron flux & spectra, gas production & DPA in the LLCB TBM set are calculated considering 500 MW fusion power & fluence level of 0.3 MWa/m{sup 2}. Radiation transport code MCNP6 and FENDL 2.1 nuclear cross-section data library are used to perform the neutronic

  1. Particle Reduction Strategies - PAREST. Traffic emission modelling. Model comparision and alternative scenarios. Sub-report

    International Nuclear Information System (INIS)

    Kugler, Ulrike; Theloke, Jochen; Joerss, Wolfram

    2013-01-01

    The modeling of the reference scenario and the various reduction scenarios in PAREST was based on the Central System of Emissions (CSE) (CSE, 2007). Emissions from road traffic were calculated by using the traffic emission model TREMOD (Knoerr et al., 2005) and fed into the CSE. The version TREMOD 4.17 has been used. The resulting emission levels in PAREST reference scenario were supplemented by the emission-reducing effect of the implementation of the future Euro 5 and 6 emission standards for cars and light commercial vehicles and Euro VI for heavy commercial vehicles in combination with the truck toll extension. [de

  2. Iterative analysis of cerebrovascular reactivity dynamic response by temporal decomposition.

    Science.gov (United States)

    van Niftrik, Christiaan Hendrik Bas; Piccirelli, Marco; Bozinov, Oliver; Pangalu, Athina; Fisher, Joseph A; Valavanis, Antonios; Luft, Andreas R; Weller, Michael; Regli, Luca; Fierstra, Jorn

    2017-09-01

    To improve quantitative cerebrovascular reactivity (CVR) measurements and CO 2 arrival times, we present an iterative analysis capable of decomposing different temporal components of the dynamic carbon dioxide- Blood Oxygen-Level Dependent (CO 2 -BOLD) relationship. Decomposition of the dynamic parameters included a redefinition of the voxel-wise CO 2 arrival time, and a separation from the vascular response to a stepwise increase in CO 2 (Delay to signal Plateau - DTP) and a decrease in CO 2 (Delay to signal Baseline -DTB). Twenty-five (normal) datasets, obtained from BOLD MRI combined with a standardized pseudo-square wave CO 2 change, were co-registered to generate reference atlases for the aforementioned dynamic processes to score the voxel-by-voxel deviation probability from normal range. This analysis is further illustrated in two subjects with unilateral carotid artery occlusion using these reference atlases. We have found that our redefined CO 2 arrival time resulted in the best data fit. Additionally, excluding both dynamic BOLD phases (DTP and DTB) resulted in a static CVR, that is maximal response, defined as CVR calculated only over a normocapnic and hypercapnic calibrated plateau. Decomposition and novel iterative modeling of different temporal components of the dynamic CO 2 -BOLD relationship improves quantitative CVR measurements.

  3. 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

  4. ITER EDA Newsletter. V. 3, no. 8

    International Nuclear Information System (INIS)

    1994-08-01

    This ITER EDA (Engineering Design Activities) Newsletter issue reports on the sixth ITER council meeting; introduces the newly appointed ITER director and reports on his address to the ITER council. The vacuum tank for the ITER model coil testing, installed at JAERI, Naka, Japan is also briefly described

  5. EU Developments of the ITER ECRH System

    International Nuclear Information System (INIS)

    Henderson, M.

    2006-01-01

    The electron cyclotron (EC) heating and current drive (H (and) CD) system of ITER will deliver 20 MW/CW in the plasma at 170 GHz for H (and) CD in addition to 2.5 MW/3 s at 120 GHz for plasma start-up. The EC system is composed of power supplies (PS), up to 24 H (and) CD gyrotrons (1 to 2 MW tubes), 3 start-up gyrotrons (1 MW tubes), 24 transmission lines and two sets of launching antennas: equatorial (EL) and upper (UL) launchers. Under the present ITER procurement package the EU is responsible for one third of the H (and) CD 170 GHz gyrotrons, all PSs associated with the H (and) CD system, and the whole set (4) of upper launchers. In all areas of participation, the EU EC partnership (coordinated by the European Fusion Development Association - EFDA) aims toward advancing the technology of each of these subsystems. For example, procurement of Pulse Step Modulator (PSM) HVPS is under consideration, which might have equivalent costs to the present ITER design (thyristor HVPS and HV series switch), but with an increased flexibility in operation and variation in the EC power waveform. The EU is at the forefront in gyrotron research and is developing a 2 MW CW 170 GHz coaxial cavity gyrotron offering an increase in output power while maintaining moderate power densities in the gyrotron cavity and collector. THALES R in collaboration with its EFDA partners (FZK, CRPP, TEKES) is manufacturing a series of prototype tubes in three phases of typically 1 s, 100 s and then CW pulse capacity (∼ 20 10 ). A 2 MW, CW gyrotron test facility is being built at CRPP that will be used to develop the 2 MW coaxial tube, in addition to testing various components required by the EC system. EFDA has undertaken a parallel development of two launcher options: front (FS) and remote (RS) steering, with the aim of providing an optimum launcher for ITER weighing EC physics aspects and operation reliability. The FS launcher (ITER reference design) offers a significant enhancement in physics

  6. Plasma cleaning of ITER edge Thomson scattering mock-up mirror in the EAST tokamak

    Science.gov (United States)

    Yan, Rong; Moser, Lucas; Wang, Baoguo; Peng, Jiao; Vorpahl, Christian; Leipold, Frank; Reichle, Roger; Ding, Rui; Chen, Junling; Mu, Lei; Steiner, Roland; Meyer, Ernst; Zhao, Mingzhong; Wu, Jinhua; Marot, Laurent

    2018-02-01

    First mirrors are the key element of all optical and laser diagnostics in ITER. Facing the plasma directly, the surface of the first mirrors could be sputtered by energetic particles or deposited with contaminants eroded from the first wall (tungsten and beryllium), which would result in the degradation of the reflectivity. The impurity deposits emphasize the necessity of the first mirror in situ cleaning for ITER. The mock-up first mirror system for ITER edge Thomson scattering diagnostics has been cleaned in EAST for the first time in a tokamak using radio frequency capacitively coupled plasma. The cleaning properties, namely the removal of contaminants and homogeneity of cleaning were investigated with molybdenum mirror insets (25 mm diameter) located at five positions over the mock-up plate (center to edge) on which 10 nm of aluminum oxide, used as beryllium proxy, were deposited. The cleaning efficiency was evaluated using energy dispersive x-ray spectroscopy, reflectivity measurements and x-ray photoelectron spectroscopy. Using argon or neon plasma without magnetic field in the laboratory and with a 1.7 T magnetic field in the EAST tokamak, the aluminum oxide films were homogeneously removed. The full recovery of the mirrors’ reflectivity was attained after cleaning in EAST with the magnetic field, and the cleaning efficiency was about 40 times higher than that without the magnetic field. All these results are promising for the plasma cleaning baseline scenario of ITER.

  7. Thermal-structural analysis for ITER in-wall shielding block

    International Nuclear Information System (INIS)

    Hao Junchuan; Song Yuntao; Wu Weiyue; Du Shuangsong; Wang, X.; Ioki, K.

    2012-01-01

    Highlights: ► IWS blocks shall withstand various types of mechanical loads including EM loads, inertial loads and thermal loads. ► Due to the complicated geometry, the finite element method is the suitable tool to solve the problem. ► Contact element has been selected to simulate the friction between the different components. ► At baking phase, secondary stresses due to preloading and temperature difference predominate in the total stress. ► At plasma operation phase, secondary stresses due to preloading and thermal loads were deducted from the total stresses. - Abstract: In order to verify the design strength of the in-wall shielding (IWS) blocks of the ITER, thermal-structural analyses of one IWS block under vacuum vessel (VV) baking and plasma operation conditions have been respectively performed with finite element (FE) method. Among the complicated operation scenarios of the ITER, two critical types of combined loads required by the load specification of IWS were applied on the shielding block. The stress of the block is judged by American Society of Mechanical Engineers (ASME) criterion. Results show that the structure of this block has enough safety margin, and it also supplies detailed information of the stress distribution in concerned region under certain loads.

  8. IVVS actuating system compatibility test to ITER gamma radiation conditions

    Energy Technology Data Exchange (ETDEWEB)

    Rossi, Paolo, E-mail: paolo.rossi@enea.it [Associazione EURATOM-ENEA sulla Fusione, 45 Via Enrico Fermi, 00044 Frascati, Rome (Italy); Collibus, M. Ferri de; Florean, M.; Monti, C.; Mugnaini, G.; Neri, C.; Pillon, M.; Pollastrone, F. [Associazione EURATOM-ENEA sulla Fusione, 45 Via Enrico Fermi, 00044 Frascati, Rome (Italy); Baccaro, S.; Piegari, A. [ENEA CR Casaccia, 301 Via Anguillarese, 00123 Santa Maria di Galeria, Rome (Italy); Damiani, C.; Dubus, G. [Fusion For Energy c/Josep Pla, n° 2 Torres Diagonal Litoral, 08019 Barcelona (Spain)

    2013-10-15

    Highlights: • ENEA developed and tested a prototype of a laser In Vessel Viewing and ranging System (IVVS) for ITER. • One piezo-motor prototype has been tested on the ENEA Calliope gamma irradiation facility to verify its compatibility to ITER gamma radiation conditions. • After a total dose of more than 4 MGy the piezo-motor maintained almost the same working parameters monitored before test without any evident and significant degradation of functionality. • After the full gamma irradiation test, the same piezo-motor assembly will be tested with 14 MeV neutrons irradiation using ENEA FNG facility. -- Abstract: The In Vessel Viewing System (IVVS) is a fundamental remote handling equipment, which will be used to make a survey of the status of the blanket first wall and divertor plasma facing components. A design and testing activity is ongoing, in the framework of a Fusion for Energy (F4E) grant agreement, to make the IVVS probe design compatible with ITER operating conditions and in particular, but not only, with attention to neutrons and gammas fluxes and both space constraints and interfaces. The paper describes the testing activity performed on the customized piezoelectric motors and the main components of the actuating system of the IVVS probe with reference to ITER gamma radiation conditions. In particular the test is performed on the piezoelectric motor, optical encoder and small scale optical samples .The test is carried out on the ENEA Calliope gamma irradiation facility at ITER relevant gamma fields at rate of about 2.5 kGy/h and doses of 4 MGy. The paper reports in detail the setup arrangement of the test campaign in order to verify significant working capability of the IVVS actuating components and the results are shown in terms of functional performances and parameters. The overall test campaign on IVVS actuating system will be completed on other ENEA testing facilities in order to verify compatibility to Magnetic field, neutrons and thermal

  9. IVVS actuating system compatibility test to ITER gamma radiation conditions

    International Nuclear Information System (INIS)

    Rossi, Paolo; Collibus, M. Ferri de; Florean, M.; Monti, C.; Mugnaini, G.; Neri, C.; Pillon, M.; Pollastrone, F.; Baccaro, S.; Piegari, A.; Damiani, C.; Dubus, G.

    2013-01-01

    Highlights: • ENEA developed and tested a prototype of a laser In Vessel Viewing and ranging System (IVVS) for ITER. • One piezo-motor prototype has been tested on the ENEA Calliope gamma irradiation facility to verify its compatibility to ITER gamma radiation conditions. • After a total dose of more than 4 MGy the piezo-motor maintained almost the same working parameters monitored before test without any evident and significant degradation of functionality. • After the full gamma irradiation test, the same piezo-motor assembly will be tested with 14 MeV neutrons irradiation using ENEA FNG facility. -- Abstract: The In Vessel Viewing System (IVVS) is a fundamental remote handling equipment, which will be used to make a survey of the status of the blanket first wall and divertor plasma facing components. A design and testing activity is ongoing, in the framework of a Fusion for Energy (F4E) grant agreement, to make the IVVS probe design compatible with ITER operating conditions and in particular, but not only, with attention to neutrons and gammas fluxes and both space constraints and interfaces. The paper describes the testing activity performed on the customized piezoelectric motors and the main components of the actuating system of the IVVS probe with reference to ITER gamma radiation conditions. In particular the test is performed on the piezoelectric motor, optical encoder and small scale optical samples .The test is carried out on the ENEA Calliope gamma irradiation facility at ITER relevant gamma fields at rate of about 2.5 kGy/h and doses of 4 MGy. The paper reports in detail the setup arrangement of the test campaign in order to verify significant working capability of the IVVS actuating components and the results are shown in terms of functional performances and parameters. The overall test campaign on IVVS actuating system will be completed on other ENEA testing facilities in order to verify compatibility to Magnetic field, neutrons and thermal

  10. ITER technical advisory committee meeting

    International Nuclear Information System (INIS)

    Fujiwara, M.

    2001-01-01

    The 17th Meeting of the ITER Technical Advisory Committee (TAC-17) was held on February 19-22, the ITER Garching Work Site in Germany. The objective of the meeting was to review the Draft Final Design Report of ITER-FEAT and assess the ability of the self-consistent overall design both to satisfy the technical objectives previously defined and to meet the cost limitations. TAC-17 was also organized to confirm that the design and critical elements, with emphasis on the key recommendations made at previous TAC meetings, are such as to extend the confidence in starting ITER construction. It was also intended to provide the ITER Council, scheduled to meet on 27 and 28 February in Toronto, with a technical assessment and key recommendations of the above mentioned report

  11. Temperature-based modeling of reference evapotranspiration using several artificial intelligence models: application of different modeling scenarios

    Science.gov (United States)

    Sanikhani, Hadi; Kisi, Ozgur; Maroufpoor, Eisa; Yaseen, Zaher Mundher

    2018-02-01

    The establishment of an accurate computational model for predicting reference evapotranspiration (ET0) process is highly essential for several agricultural and hydrological applications, especially for the rural water resource systems, water use allocations, utilization and demand assessments, and the management of irrigation systems. In this research, six artificial intelligence (AI) models were investigated for modeling ET0 using a small number of climatic data generated from the minimum and maximum temperatures of the air and extraterrestrial radiation. The investigated models were multilayer perceptron (MLP), generalized regression neural networks (GRNN), radial basis neural networks (RBNN), integrated adaptive neuro-fuzzy inference systems with grid partitioning and subtractive clustering (ANFIS-GP and ANFIS-SC), and gene expression programming (GEP). The implemented monthly time scale data set was collected at the Antalya and Isparta stations which are located in the Mediterranean Region of Turkey. The Hargreaves-Samani (HS) equation and its calibrated version (CHS) were used to perform a verification analysis of the established AI models. The accuracy of validation was focused on multiple quantitative metrics, including root mean squared error (RMSE), mean absolute error (MAE), correlation coefficient (R 2), coefficient of residual mass (CRM), and Nash-Sutcliffe efficiency coefficient (NS). The results of the conducted models were highly practical and reliable for the investigated case studies. At the Antalya station, the performance of the GEP and GRNN models was better than the other investigated models, while the performance of the RBNN and ANFIS-SC models was best compared to the other models at the Isparta station. Except for the MLP model, all the other investigated models presented a better performance accuracy compared to the HS and CHS empirical models when applied in a cross-station scenario. A cross-station scenario examination implies the

  12. ITER ITA newsletter. No. 8, September 2003

    International Nuclear Information System (INIS)

    2003-10-01

    This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about ITER related activities including Robert Aymar's leaving ITER for CERN, ITER related issues at the IAEA General Conference and status and prospects of thermonuclear power and activity during the ITA on materials foe vessel and in-vessel components

  13. ITER CTA newsletter. No. 9

    International Nuclear Information System (INIS)

    2002-06-01

    This ITER CTA newsletter contains information about the Fourth Negotiations Meeting on the Joint Implementation of ITER held in Cadarache, France on 4-6 June 2002 and about the meeting of the ITER CTA Project Board which took place on the occasion of the N4 Meeting at Cadarache on 3-4 June 2002

  14. ITER CTA newsletter. No. 1

    International Nuclear Information System (INIS)

    2001-01-01

    This ITER CTA newsletter comprises reports on ITER co-ordinated technical activities, information about the Meeting of the ITER CTA project board which took place in Vienna on 16 July 2001, and the Meeting of the expert group on MHD, disruptions and plasma control which was held on 25-26 June 2001 in Funchal, Madeira

  15. The JET ITER-like wall experiment: First results and lessons for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Horton, Lorne, E-mail: Lorne.Horton@jet.efda.org [EFDA-CSU Culham, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); European Commission, B-1049 Brussels (Belgium)

    2013-10-15

    Highlights: ► JET has recently completed the installation of an ITER-like wall. ► Important operational aspects have changed with the new wall. ► Initial experiments have confirmed the expected low fuel retention. ► Disruption dynamics have change dramatically. ► Development of wall-compatible, ITER-relevant regimes of operation has begun. -- Abstract: The JET programme is strongly focused on preparations for ITER construction and exploitation. To this end, a major programme of machine enhancements has recently been completed, including a new ITER-like wall, in which the plasma-facing armour in the main vacuum chamber is beryllium while that in the divertor is tungsten—the same combination of plasma-facing materials foreseen for ITER. The goal of the initial experimental campaigns is to fully characterise operation with the new wall, concentrating in particular on plasma-material interactions, and to make direct comparisons of plasma performance with the previous, carbon wall. This is being done in a progressive manner, with the input power and plasma performance being increased in combination with the commissioning of a comprehensive new real-time protection system. Progress achieved during the first set of experimental campaigns with the new wall, which took place from September 2011 to July 2012, is reported.

  16. ITER concept definition. V.1

    International Nuclear Information System (INIS)

    1989-01-01

    Under the auspices of the International Atomic Energy Agency (IAEA), an agreement among the four parties representing the world's major fusion programs resulted in a program for conceptual design of the next logical step in the fusion program, the International Thermonuclear Experimental Reactor (ITER). The definition phase, which ended in November, 1989, is summarized in two reports: a brief summary is contained in the ITER Definition Phase Report (IAEA/ITER/DS/2); the extended technical summary and technical details of ITER are contained in this two-volume report. The first volume of this report contains the Introduction and Summary, and the remainder will appear in Volume II. In the Conceptual Design Activities phase, ITER has been defined as being a tokamak device. The basic performance parameters of ITER are given in Volume I of this report. In addition, the rationale for selection of this concept, the performance flexibility, technical issues, operations, safety, reliability, cost, and research and development needed to proceed with the design are discussed. Figs and tabs

  17. ITER primary cryopump test facility

    International Nuclear Information System (INIS)

    Petersohn, N.; Mack, A.; Boissin, J.C.; Murdoc, D.

    1998-01-01

    A cryopump as ITER primary vacuum pump is being developed at FZK under the European fusion technology programme. The ITER vacuum system comprises of 16 cryopumps operating in a cyclic mode which fulfills the vacuum requirements in all ITER operation modes. Prior to the construction of a prototype cryopump, the concept is tested on a reduced scale model pump. To test the model pump, the TIMO facility is being built at FZK in which the model pump operation under ITER environmental conditions, except for tritium exposure, neutron irradiation and magnetic fields, can be simulated. The TIMO facility mainly consists of a test vessel for ITER divertor duct simulation, a 600 W refrigerator system supplying helium in the 5 K stage and a 30 kW helium supply system for the 80 K stage. The model pump test programme will be performed with regard to the pumping performance and cryogenic operation of the pump. The results of the model pump testing will lead to the design of the full scale ITER cryopump. (orig.)

  18. ITER ITA newsletter No. 32, July 2006

    International Nuclear Information System (INIS)

    2006-07-01

    This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about ITER related activities. The ITER Parties, at their Ministerial Meeting in May 2006 in Brussels, initialled the draft text of the prospective Agreement on the Establishment of the ITER International Fusion Energy Organization for the Joint Implementation of the ITER Project as well as the draft text of the Agreement on the Privileges and Immunities of the ITER International Fusion Energy Organisation for the Joint Implementation of the ITER Project. The Parties have requested that the IAEA Director General serve as Depositary of the two aforementioned Agreements and that the IAEA establish a Trust Fund to Support Common Expenditures under the ITER Transitional Arrangements, pending entry into force of the prospective Agreement on the Establishment of the ITER International Fusion Energy Organization for the Joint Implementation of the ITER Project. At its June Meeting in Vienna, the IAEA Board of Governors approved these requests. There is also information about the Tenth Meeting of the International Tokamak Physics Activity (ITPA) Topical Group (TG) on Diagnostics was held at the Kurchatov Institute, Moscow, from 10-14 April 2006

  19. Quench detection and behaviour in case of quench in the ITER magnet systems

    International Nuclear Information System (INIS)

    Coatanea-Gouachet, M.

    2012-02-01

    The quench of one of the ITER magnet system is an irreversible transition from superconducting to normal resistive state, of a conductor. This normal zone propagates along the cable in conduit conductor dissipating a large power. The detection has to be fast enough to dump out the magnetic energy and avoid irreversible damage of the systems. The primary quench detection in ITER is based on voltage detection, which is the most rapid detection. The very magnetically disturbed environment during the plasma scenario makes the voltage detection particularly difficult, inducing large inductive components in the coils and voltage compensations have to be designed to discriminate the resistive voltage associated with the quench. A conceptual design of the quench detection based on voltage measurements is proposed for the three majors magnet systems of ITER. For this, a clear methodology was developed. It includes the classical hot spot criterion, the quench propagation study using the commercial code Gandalf and the careful estimation of the inductive disturbances by developing the TrapsAV code. Specific solutions have been proposed for the compensation in the three ITER magnet systems and for the quench detection parameters, which are the voltage threshold (in the range of 0.1 V - 0.55 V) and the holding time (in the range of 1-1.4 s). The selected values, in particular the holding time, are sufficiently high to ensure the reliability of the system and avoid fast safety discharges not induced by a quench, which is a classical problem. (author)

  20. ITER EDA Newsletter. V. 10, no. 7

    International Nuclear Information System (INIS)

    2001-07-01

    This ITER EDA Newsletter presents an overview of meetings held at IAEA Headquarters in Vienna during the week 16-20 July 2001 related to the successful completion of the ITER Engineering Design Activities (EDA). Among them were the final meeting of the ITER Council, the closing ceremony to commemorate the EDA completion, the final meeting of the ITER Management Advisory Committee, a briefing of issues related to ITER developments, and discussions on the possible joint implementation of ITER

  1. Energy efficiency improvement potentials and a low energy demand scenario for the global industrial sector

    NARCIS (Netherlands)

    Kermeli, Katerina; Graus, Wina H J; Worrell, Ernst

    2014-01-01

    The adoption of energy efficiency measures can significantly reduce industrial energy use. This study estimates the future industrial energy consumption under two energy demand scenarios: (1) a reference scenario that follows business as usual trends and (2) a low energy demand scenario that takes

  2. IAEA activities related to ITER

    International Nuclear Information System (INIS)

    Dolan, T.J.; Schneider, U.

    2001-01-01

    As agreed between the IAEA and the ITER Parties, special sessions are dedicated to ITER at the IAEA Fusion Energy Conferences. At the 18th IAEA Fusion Energy Conference, held on 4-10 October 2000 in Sorrento, Italy, in the Artsimovich-Kadomtsev Memorial opening session there were special lectures by Carlo Rubbia (President, ENEA, Italy), A. Arima (Japan), and E.P. Velikhov (Russia); an overview talk on ITER by R. Aymar (ITER Director); and a talk on the FTU experiment by F. Romanelli. In total, 573 participants from 34 countries presented 389 papers (including 11 post-deadline papers and the 4 summaries)

  3. 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.

  4. 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

  5. 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.)

  6. ITER ITA Newsletter. No. 29, March 2006

    International Nuclear Information System (INIS)

    2006-05-01

    This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about ITER related activities and meetings, namely, the ITER Director-General Nominee, Dr. Kaname Ikeda, took up his position as ITER Project Leader in Cadarache on 13 March, the consolidation of information technology infrastructure for ITER and about he Thirty-Fifth Meeting of the Fusion Power Co-ordinating Committee (FPCC), which was held on 28 February-1 March 2006 at the headquarters of the International Energy Agency (IEA) in Paris

  7. ITER safety challenges and opportunities

    International Nuclear Information System (INIS)

    Piet, S.J.

    1992-01-01

    This paper reports on 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. Much research and development (R ampersand D) and design analysis is needed to establish that ITER meets regulatory requirements. There is a further oportunity 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, three programmatic challenges and three technical challenges must be overcome. The first step 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 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

  8. 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

  9. Reference Scenario Forecasting: A New Approach to Transport Project Assessment

    DEFF Research Database (Denmark)

    Salling, Kim Bang; Leleur, Steen; Skougaard, Britt Zoëga

    2010-01-01

    This paper presents a new approach to transport project assessment in terms of feasibility risk assessment and reference class forecasting. Normally, transport project assessment is based upon a cost-benefit approach where evaluation criteria such as net present values are obtained. Recent research...... construction cost estimates. Hereafter, a quantitative risk analysis is provided making use of Monte Carlo simulation. This stochastic approach facilitates random input parameters based upon reference class forecasting, hence, a parameter data fit has been performed in order to obtain validated probability...... forecasting (RSF) frame. The RSF is anchored in the cost-benefit analysis (CBA), thus, it provides decision-makers with a quantitative mean of assessing the transport infrastructure project. First, the RSF method introduces uncertainties within the CBA by applying Optimism Bias uplifts on the preliminary...

  10. NbTi Strands Verification for ITER PF CICC Process Qualification of CNDA

    Science.gov (United States)

    Liu, F.; Liu, H.; Liu, S.; Liu, B.; Lei, L.; Wu, Y.

    2014-05-01

    China is in charge of most of Poloidal Field (PF) conductors production for the International Thermonuclear Experimental Reactor (ITER). The execution for PF conductors shall be in three main phases. According to ITER Procurement Arrangement (PA), the Domestic Agency (DA) shall be required to verify the room and low temperature acceptance tests carried out by the strand suppliers. As the reference laboratory of Chinese DA (CNDA), the superconducting strands test laboratory of Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) was undertaking the task of strands verification for ITER conductors. The verification test includes: diameter, Nickel plating thickness, copper-to-non-copper volume ratio, twist pitch direction and length, standard critical current (IC) and resistive transition index (n), residual resistance ration (RRR), and hysteresis loss. 48 NbTi strands with 7 billets were supplied for the PF Cable-In-Conduit Conductor (CICC) process qualification. In total, 54 samples were measured. The verification level for PF CICC process qualification was 100%. The test method, facility and results of each item are described in detail in this publication.

  11. Impact of the plasma response in three-dimensional edge plasma transport modelling for RMP ELM control scenarios at ITER

    Science.gov (United States)

    Schmitz, Oliver

    2014-10-01

    The constrains used in magneto-hydrodynamic (MHD) modeling of the plasma response to external resonant magnetic perturbation (RMP) fields have a profound impact on the three-dimensional (3-D) shape of the plasma boundary induced by RMP fields. In this contribution, the consequences of the plasma response on the actual 3D boundary structure and transport during RMP application at ITER are investigated. The 3D fluid plasma and kinetic neutral transport code EMC3-Eirene is used for edge transport modeling. Plasma response modeling is conducted with the M3D-C1 code using a single fluid, non-linear and a two fluid, linear MHD constrain. These approaches are compared to results with an ideal MHD like plasma response. A 3D plasma boundary is formed for all cases consisting of magnetic finger structures at the X-point intersecting the divertor surface in a helical footprint pattern. The width of the helical footprint pattern is largely reduced compared to vacuum magnetic fields when using the ideal MHD like screening model. This yields increasing peak heat fluxes in contrast to a beneficial heat flux spreading seen with vacuum fields. The particle pump out as well as loss of thermal energy is reduced by a factor of two compared to vacuum fields. In contrast, the impact of the plasma response obtained from both MHD constrains in M3D-C1 is nearly negligible at the plasma boundary and only a small modification of the magnetic footprint topology is detected. Accordingly, heat and particle fluxes on the target plates as well as the edge transport characteristics are comparable to the vacuum solution. This span of modeling results with different plasma response models highlights the importance of thoroughly validating both, plasma response and 3D edge transport models for a robust extrapolation towards ITER. Supported by ITER Grant IO/CT/11/4300000497 and F4E Grant GRT-055 (PMS-PE) and by Start-Up Funds of the University of Wisconsin - Madison.

  12. ITER CTA newsletter. No. 4

    International Nuclear Information System (INIS)

    2001-12-01

    This ITER CTA Newsletter contains information about the organization of the ITER Co-ordinated Technical Activities (CTA) International Team as the follow-up of the ITER CTA project board meeting in Toronto on 7 November 2001. It also includes a summary on the start of the international tokamak physics activity by Dr. D. Campbell, Chair of the ITPA Co-ordinating Committee

  13. ITER management advisory committee meeting

    International Nuclear Information System (INIS)

    Yoshikawa, M.

    2001-01-01

    The ITER Management Advisory Committee (MAC) Meeting was held on 23 February in Garching, Germany. The main topics were: the consideration of the report by the Director on the ITER EDA Status, the review of the Work Programme, the review of the Joint Fund, the review of a schedule of ITER meetings, and the arrangements for termination and wind-up of the EDA

  14. Analysis of JT-60SA operational scenarios

    Science.gov (United States)

    Garzotti, L.; Barbato, E.; Garcia, J.; Hayashi, N.; Voitsekhovitch, I.; Giruzzi, G.; Maget, P.; Romanelli, M.; Saarelma, S.; Stankiewitz, R.; Yoshida, M.; Zagórski, R.

    2018-02-01

    Reference scenarios for the JT-60SA tokamak have been simulated with one-dimensional transport codes to assess the stationary state of the flat-top phase and provide a profile database for further physics studies (e.g. MHD stability, gyrokinetic analysis) and diagnostics design. The types of scenario considered vary from pulsed standard H-mode to advanced non-inductive steady-state plasmas. In this paper we present the results obtained with the ASTRA, CRONOS, JINTRAC and TOPICS codes equipped with the Bohm/gyro-Bohm, CDBM and GLF23 transport models. The scenarios analysed here are: a standard ELMy H-mode, a hybrid scenario and a non-inductive steady state plasma, with operational parameters from the JT-60SA research plan. Several simulations of the scenarios under consideration have been performed with the above mentioned codes and transport models. The results from the different codes are in broad agreement and the main plasma parameters generally agree well with the zero dimensional estimates reported previously. The sensitivity of the results to different transport models and, in some cases, to the ELM/pedestal model has been investigated.

  15. ITER ITA newsletter. No. 6, July 2003

    International Nuclear Information System (INIS)

    2003-09-01

    This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about ITER related activities. One of them was the farewell party for for Annick Lyraud and Robert Aymar, who will take up his position as Director-General of CERN in January 2004, another is information about Dr. Yasuo Shimomura, ITER interim project leader, and ITER technical work during the transitional arrangements

  16. ITER EDA newsletter. V. 7, no. 7

    International Nuclear Information System (INIS)

    1998-07-01

    This newsletter contains the articles: 'Extraordinary ITER council meeting', 'ITER EDA final safety meeting' and 'Summary report of the 3rd combined workshop of the ITER confinement and transport and ITER confinement database and modeling expert groups'

  17. Progress on advanced tokamak and steady-state scenario development on DIII-D and NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Doyle, E J [Department of Electrical Engineering and PSTI, University of California, Los Angeles, California 90095 (United States); Garofalo, A M [Columbia University, New York, New York 10027 (United States); Greenfield, C M [General Atomics, San Diego, California 92186-5608 (United States); Kaye, S M [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Menard, J E [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Murakami, M [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Sabbagh, S A [Columbia University, New York, New York 10027 (United States); Austin, M E [University of Texas-Austin, Austin, Texas 78712 (United States); Bell, R E [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Burrell, K H [General Atomics, San Diego, California 92186-5608 (United States); Ferron, J R [General Atomics, San Diego, California 92186-5608 (United States); Gates, D A [Princeton Plasma Physics Lab., Princeton, New Jersey 08543-0451 (United States); Groebner, R J; Hyatt, A W; Luce, T C; Petty, C C; Wade, M R; Waltz, R E [General Atomics, San Diego, California 92186-5608 (United States); Jayakumar, R J [Lawrence Livermore National Lab., Livermore, California 94550 (United States); Kinsey, J E [Lehigh Univ., Bethlehem, Pennsylvania 18015 (United States); LeBlanc, B P [Princeton Plasma Physics Lab., Princeton, New Jersey 08543-0451 (United States); McKee, G R [Univ. of Wisconsin-Madison, Madison, Wisconsin 53706 (United States); Okabayashi, M [Princeton Plasma Physics Lab., Princeton, New Jersey 08543-0451 (United States); Peng, Y-K M [Oak Ridge National Lab., Oak Ridge, Tennessee 37831 (United States); Politzer, P A [General Atomics, San Diego, California 92186-5608 (United States); Rhodes, T L [Dept. of Electrical Engineering and PSTI, Univ. of California, Los Angeles, California 90095 (United States)

    2006-12-15

    Advanced tokamak (AT) research seeks to develop steady-state operating scenarios for ITER and other future devices from a demonstrated scientific basis. Normalized target parameters for steady-state operation on ITER are 100% non-inductive current operation with a bootstrap current fraction f{sub BS} {>=} 60%, q{sub 95} {approx} 4-5 and G {identical_to}{beta}{sub N}H{sub scaling}/q{sub 95}{sup 2} {>=}0.3. Progress in realizing such plasmas is considered in terms of the development of plasma control capabilities and scientific understanding, leading to improved AT performance. NSTX has demonstrated active resistive wall mode stabilization with low, ITER-relevant, rotation rates below the critical value required for passive stabilization. On DIII-D, experimental observations and GYRO simulations indicate that ion internal transport barrier (ITB) formation at rational-q surfaces is due to equilibrium zonal flows generating high local E ? B shear levels. In addition, stability modelling for DIII-D indicates a path to operation at {beta}{sub N} {>=} 4 with q{sub min} {>=} 2, using broad, hollow current profiles to increase the ideal wall stability limit. Both NSTX and DIII-D have optimized plasma performance and expanded AT operational limits. NSTX now has long-pulse, high performance discharges meeting the normalized targets for an spherical torus-based component test facility. DIII-D has developed sustained discharges combining high beta and ITBs, with performance approaching levels required for AT reactor concepts, e.g. {beta}{sub N} = 4, H{sub 89} = 2.5, with f{sub BS} > 60%. Most importantly, DIII-D has developed ITER steady-state demonstration discharges, simultaneously meeting the targets for steady-state Q {>=} 5 operation on ITER set out above, substantially increasing confidence in ITER meeting its steady-state performance objective.

  18. ITER Central Solenoid Module Fabrication

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-23

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

  19. Scenarios for the assessment of urban exposures after radioactive contamination

    International Nuclear Information System (INIS)

    Silva, D.N.G.; Guimarães, J.R.D.; Rochedo, E.R.R.

    2015-01-01

    Accidents involving radioactive material are not frequent but may include releases of radionuclides to the air, land or waterways. These releases are usually uncontrollable and may lead to doses in the public in excess of the reference levels established by regulations defined by the national regulatory agencies of each country. Although they had occurred sporadically since the last century, it was observed that, after the emergency phase, the public concern is enhanced when they feel that there is an unpreparedness of authorities responsible for remediation actions, due to the lack of definition of strategies to be adopted in the long term after such events. The aim of this work is to describe reference urban scenarios, considering the characteristics observed in residential and free access areas of urban centers. These scenarios were developed based on the counties surrounding the Brazilian nuclear power plant. Considering the counties within 50 km from the nuclear power plant, nine belong to the state of Rio de Janeiro and seven belong to São Paulo state; the highest population densities were observed in five counties of Rio de Janeiro. Based on the different types of residences and outdoor areas observed in these 16 counties, six reference scenarios for urban areas were developed including areas comprised by four types of residential houses (with low, medium and high shielding building material and houses in a row), apartments in buildings, and park areas with lawn and trees. The characteristics of each of these scenarios were raised through Google Earth images considering 1 km 2 of different locations comprised by each type of area defined. In a next step, the information obtained in each scenario shall be used in computer simulations to characterize the effects and consequences on public exposure of the application of decontamination procedures. (authors)

  20. ITER EDA newsletter. V. 1, no. 2

    International Nuclear Information System (INIS)

    1992-12-01

    This second issue of the ITER Newsletter during the EDA (Engineering Design Activities) reports on (i) the second ITER Council Meeting held in the Russian Research Centre (RRC) ''Kurchatov Institute'', Moscow, Russia, December 15-16, 1992, (ii) the opening ceremony of the ITER Council Office at the RRC, (iii) the first meeting of the ITER Management Advisory Committee (MAC), (iv) the start-up of the ITER EDA at Garching, Germany, (v) descriptions of the ITER Co-Centres at Naka, Japan, and (vi) San Diego, USA, (vii) contact persons activities, (viii) the adoption by the ITER Council of the recommendations by the Special Working Group 1 (SWG-1), (ix) news in brief, and (x) coming events

  1. Fast ion power loads on ITER first wall structures in the presence of NTMs and microturbulence

    International Nuclear Information System (INIS)

    Kurki-Suonio, T.; Asunta, O.; Hirvijoki, E.; Koskela, T.; Snicker, A.; Sipilae, S.; Hauff, T.; Jenko, F.; Poli, E.

    2011-01-01

    The level and distribution of the wall power flux of energetic ions in ITER have to be known accurately in order to ensure the integrity of the first wall. Until now, most quantitative estimates have been based on the assumption that fast ion transport is dictated by neoclassical effects only. However, in ITER, the fast ion distribution is likely to be affected by various MHD effects and probably also by microturbulence. We have now upgraded our orbit-following Monte Carlo code ASCOT so that it has simple, theory-based models for neoclassical tearing mode (NTM)-type islands as well as for turbulent diffusion. ASCOT also allows for full-orbit following, which is important close to the material surfaces and, possibly, also when strong toroidal inhomogeneities are present in the magnetic field. Here we introduce the new models, preliminary results obtained with them, and how these models could be made more realistic in the future. The simulations are carried out for thermonuclear alpha particles in ITER scenario 2 plasma, because we consider this combination to be most critical for the successful operation of ITER. Neither the turbulent transport nor NTM-type islands are found to introduce alarming changes in the wall loads. However, at this stage it was not possible to combine the island structures with the non-axisymmetric magnetic field of ITER, and it remains to be seen what the combined effect of drift islands together with the toroidal ripple and local field aberrations, such as those due to test blanket modules and resonant magnetic perturbations will be.

  2. ITER ITA newsletter. No. 11, December 2003

    International Nuclear Information System (INIS)

    2003-12-01

    This issue of the ITER ITA (ITER transitional Arrangements) newsletter contains concise information about ITER including information from the editor about ITER update, about progress in ITER magnet design and preparation of procurement packages and about 25th anniversary of the First Steering Committee Meeting of the International Tokamak Reactor (INTOR) Workshop, organized under the auspices of the IAEA, took place at the IAEA Headquarters in Vienna

  3. Rokkasho: Japanese site for ITER

    International Nuclear Information System (INIS)

    Ohtake, S.; Yamaguchi, V.; Matsuda, S.; Kishimoto, H.

    2003-01-01

    The Atomic Energy Commission of Japan authorized ITER as the core machine of the Third Phase Basic Program of Fusion Energy Development. After a series of discussions in the Atomic Energy Commission and the Council of Science and Technology Policy, Japanese Government concluded formally with the Cabinet Agreement on 31 May 2002 that Japan should participate in the ITER Project and offer the Rokkasho-Mura site for construction of ITER to the Negotiations among Canada (CA), the European Union (EU), Japan (JA), and the Russian Federation (RF). The JA site proposal is now under the international assessment in the framework of the ITER Negotiations. (author)

  4. Revisiting the analysis of passive plasma shutdown during an ex-vessel loss of coolant accident in ITER blanket

    International Nuclear Information System (INIS)

    Rivas, J.C.; Dies, J.; Fajarnés, X.

    2015-01-01

    Highlights: • We have repeated the safety analysis for the hypothesis of passive plasma shutdown for beryllium evaporation during an ex-vessel LOCA of ITER first wall, with AINA code. • We have performed a sensitivity analysis over some key parameters that represents uncertainties in physics and engineering, to identify cliff edge effects. • The obtained results for the 500 MW inductive scenario, with an ex-vessel LOCA affecting a third of first wall surface are similar to those of previous studies and point to the possibility of a passive plasma shutdown during this safety case, before a serious damage is inflicted to the ITER wall. • The sensitivity analysis revealed a new scenario potentially damaging for the first wall if we increase fusion power and time delay for impurity transport, and decrease fraction of affected first wall area and initial beryllium fraction in plasma. • After studying the 700 MW inductive scenario, with an ex-vessel LOCA affecting 10% of first wall surface, with 0.5% of Be in plasma and a time delay twice the energy confinement time, it was found that affected area of first wall would melt before a passive plasma shutdown occurs. - Abstract: In this contribution, the analysis of passive safety during an ex-vessel loss of coolant accident (LOCA) in the first wall/shield blanket of ITER has been studied with AINA safety code. In the past, this case has been studied using robust safety arguments, based on simple 0D models for plasma balance equations and 1D models for wall heat transfer. The conclusion was that, after first wall heating up due to the loss of all coolant, the beryllium evaporation in the wall surface would induce a growing impurity flux into core plasma that finally would end in a passive shut down of the discharge. The analysis of plasma-wall transients in this work is based in results from AINA code simulations. AINA (Analyses of IN vessel Accidents) code is a safety code developed at Fusion Energy Engineering

  5. ITER EDA newsletter. V. 8, no. 12

    International Nuclear Information System (INIS)

    1999-12-01

    This ITER EDA Newsletter reports about the ITER Management Advisory Committee Meeting in Naka, the ITER Technical Advisory Committee Meeting in Naka and the meeting of the ITER SWG-P2 in Vienna. A separate abstract is prepared for each meeting

  6. Simulation of MGI efficiency for plasma energy conversion into Ar radiation in JET and implications for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Pestchanyi, Serguei, E-mail: serguei.pestchanyi@kit.edu [Association EURATOM-KIT, Karlsruhe (Germany); Koslowski, Rudi; Reux, Cedric [JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Lehnen, Michael [Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2015-10-15

    Highlights: • We simulated disruption mitigation using massive gas injection with the TOKES code. • Cross-reference analysis of JET experiments on MGI and their simulations have been done. • The analysis allows suggesting the mechanism for saturation of radiated energy fraction at 70–80%. • Rough extrapolation of the result on ITER conditions has been done. - Abstract: Effectiveness of massive gas injection (MGI) for mitigation of disruptive wall damage has been investigated. Cross-reference analysis of the available JET experiments on MGI and their simulations with the TOKES code allow suggesting that in JET conditions one can convert into radiation the electron thermal energy and the plasma current energy, but the ion thermal energy does not convert into radiation because of very ineffective excitation of injected noble gas (NG) ions by D ions and long equipartition time between D ions and electrons. The model assumes rather high electron temperature during current quench (CQ), which contradicts with its time duration. Rough extrapolation of the result on ITER conditions shows that one can expect irradiation of total plasma energy if CQ duration in ITER is not shorter as in JET.

  7. ITER EDA newsletter. V. 10, no. 1

    International Nuclear Information System (INIS)

    2001-01-01

    This article provides a summary of results of the ITER Physics Committee Meeting, which was held on 14 October 2000 at the ITER Garching Joint Work Site, Germany. The ITER Physics Committee is the body responsible for overseeing, through the seven specialized Expert Groups, the R and D activities contributed voluntarily by the ITER Parties. The Parties' Physics Designated Persons, the Chairs and Co-Chairs of ITER Physics Expert Groups and the JCT members involved attended the Meeting. As usual, the meeting was chaired by the ITER Director, Dr. R. Aymar, who reported on the status of the ITER EDA. Dr. Aymar described the steps being taken in preparing the ITER-FEAT Final Design Report (FDR), and further stated that the Report would be available in time to be of benefit to the Negotiations on the ITER Joint Implementation, expected to start around May 2001. All Parties recognize that the ITER Physics Expert Group structure has been useful in focusing the tokamak physics activity on the ITER-relevant issues and provides an efficient worldwide collaboration on confirming innovative solutions. The concept of an international workshop to be organized as a pre-meeting of each Expert Group meeting, in order to involve U.S. scientists in the discussion of generic tokamak physics issues, was introduced in 2000, with some success, and its goal should be pursued

  8. Existence test for asynchronous interval iterations

    DEFF Research Database (Denmark)

    Madsen, Kaj; Caprani, O.; Stauning, Ole

    1997-01-01

    In the search for regions that contain fixed points ofa real function of several variables, tests based on interval calculationscan be used to establish existence ornon-existence of fixed points in regions that are examined in the course ofthe search. The search can e.g. be performed...... as a synchronous (sequential) interval iteration:In each iteration step all components of the iterate are calculatedbased on the previous iterate. In this case it is straight forward to base simple interval existence and non-existencetests on the calculations done in each step of the iteration. The search can also...... on thecomponentwise calculations done in the course of the iteration. These componentwisetests are useful for parallel implementation of the search, sincethe tests can then be performed local to each processor and only when a test issuccessful do a processor communicate this result to other processors....

  9. ITER EDA Newsletter. V.3, no.3

    International Nuclear Information System (INIS)

    1994-03-01

    This ITER EDA Newsletter issue contains reports on (i) the completion of the ITER EDA Protocol 1, (ii) the signing of ITER EDA Protocol 2, (iii) a technical meeting on pumping and fuelling and (iv) a technical meeting on the ITER Tritium Plant

  10. Implementing the Gaia Astrometric Global Iterative Solution (AGIS) in Java

    OpenAIRE

    O'Mullane, William; Lammers, Uwe; Lindegren, Lennart; Hernandez, Jose; Hobbs, David

    2011-01-01

    This paper provides a description of the Java software framework which has been constructed to run the Astrometric Global Iterative Solution for the Gaia mission. This is the mathematical framework to provide the rigid reference frame for Gaia observations from the Gaia data itself. This process makes Gaia a self calibrated, and input catalogue independent, mission. The framework is highly distributed typically running on a cluster of machines with a database back end. All code is written in ...

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

    International Nuclear Information System (INIS)

    Eriksson, John; Sjoeberg, A.; Sponton, L.L.

    2001-05-01

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

  12. Magnetic Configuration Control of ITER Plasmas

    International Nuclear Information System (INIS)

    Albanese, R.; Artaserse, G.; Mattei, M.; Ambrosino, G.; Crisanti, F.; Tommasi, G. de; Fresa, R.; Portone, A.; Sartori, F.; Villone, F.

    2006-01-01

    The aim of this paper is to review the capability of the ITER Poloidal Field (PF) system of controlling the broad range of plasma configurations presently forecasted during ITER operation. The attention is focused on the axi-symmetric aspects of plasma magnetic configuration control since they pose the greatest challenges in terms of control power and they have the largest impact on machine capital cost. The paper is broadly divided in two main sections devoted, respectively, to open loop (feed-forward) and closed loop (feedback) control. In the first part of the study the PF system is assessed with respect to the initiation, ramp-up, sustained burn, ramp-down phases of the main plasma inductive scenario. The limiter-to-divertor configuration transition phase is considered in detail with the aim of assessing the PF capability to form an X-point at the lowest possible current and, therefore, to relax the thermal load on the limiter surfaces. Moreover, during the sustained burn it is important to control plasmas with a broad range of current density profiles. In the second part of the study the plasma vertical feedback control requirements are assessed in details, in particular for the high elongation configurations achievable during the early limiter-to-X point transition phase. Non-rigid plasma displacement models are used to assess the control system voltage and current requirements of different radial field control circuits obtained, for example, by connecting the outermost PF coils, some CS coils, coils sub-sections etc. At last, the main 3D effects of the vessel ports are modeled and their impact of vertical stabilization evaluated. (author)

  13. ITER EDA newsletter. V. 7, no. 1

    International Nuclear Information System (INIS)

    1998-01-01

    This issue of the ITER Newsletter contains a summary report on the Thirteenth meeting of the ITER Management Advisory Committee (MAC), a report on ITER at the International Conference on Fusion Reactor Materials and a report of a Russian scientist working at ITER Garching JWS

  14. Plasma control concepts for ITER

    International Nuclear Information System (INIS)

    Lister, J.B.; Nieswand, C.

    1997-01-01

    This overview paper skims over a wide range of issues related to the control of ITER plasmas. Although operation of the ITER project will require extensive developmental work to achieve the degree of control required, there is no indication that any of the identified problems will present overwhelming difficulties compared with the operation of present tokamaks. However, the precision of control required and the degree of automation of the final ITER plasma control system will present a challenge which is somewhat greater than for present tokamaks. In order to operate ITER optimally, integrated use of a large amount of diagnostic information will be necessary, evaluated and interpreted automatically. This will challenge both the diagnostics themselves and their supporting interpretation codes. The intervening years will provide us with the opportunity to implement and evaluate most of the new features required for ITER on existing tokamaks, with the exception of the control of an ignited plasma. (author) 7 figs., 7 refs

  15. Intensity earthquake scenario (scenario event - a damaging earthquake with higher probability of occurrence) for the city of Sofia

    Science.gov (United States)

    Aleksandrova, Irena; Simeonova, Stela; Solakov, Dimcho; Popova, Maria

    2014-05-01

    . The usable and realistic ground motion maps for urban areas are generated: - either from the assumption of a "reference earthquake" - or directly, showing values of macroseimic intensity generated by a damaging, real earthquake. In the study, applying deterministic approach, earthquake scenario in macroseismic intensity ("model" earthquake scenario) for the city of Sofia is generated. The deterministic "model" intensity scenario based on assumption of a "reference earthquake" is compared with a scenario based on observed macroseimic effects caused by the damaging 2012 earthquake (MW5.6). The difference between observed (Io) and predicted (Ip) intensities values is analyzed.

  16. Industrial engineering studies for the manufacture of the ITER PF coils

    International Nuclear Information System (INIS)

    Libeyre, P.; Decool, P.; Guerin, O.; Perrella, M.; Bourquard, A.

    2007-01-01

    Industrial studies have been carried out in Europe to prepare the manufacture of the five poloidal field (PF) coils, which will be manufactured on the ITER site. A first study, carried out by Ansaldo Superconduttori, addressed the manufacturing sequence, assuming the manufacture of the PF coils inside the two buildings which will further host the cryogenic system. A second study, carried out by Alstom investigated how to achieve the manufacture of some crucial points. A new layout of the manufacturing line was proposed, aiming at manufacture of the PF2-6 coils within 36 months. A recent study performed by Alstom, assumes the manufacture of the PF coils in a single dedicated building, releasing so the constraint of meeting the deadline fixed in ITER reference scheme by the starting point of installation of the cryogenic components

  17. 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

  18. Iterative Multiview Side Information for Enhanced Reconstruction in Distributed Video Coding

    Directory of Open Access Journals (Sweden)

    2009-03-01

    Full Text Available Distributed video coding (DVC is a new paradigm for video compression based on the information theoretical results of Slepian and Wolf (SW and Wyner and Ziv (WZ. DVC entails low-complexity encoders as well as separate encoding of correlated video sources. This is particularly attractive for multiview camera systems in video surveillance and camera sensor network applications, where low complexity is required at the encoder. In addition, the separate encoding of the sources implies no communication between the cameras in a practical scenario. This is an advantage since communication is time and power consuming and requires complex networking. In this work, different intercamera estimation techniques for side information (SI generation are explored and compared in terms of estimating quality, complexity, and rate distortion (RD performance. Further, a technique called iterative multiview side information (IMSI is introduced, where the final SI is used in an iterative reconstruction process. The simulation results show that IMSI significantly improves the RD performance for video with significant motion and activity. Furthermore, DVC outperforms AVC/H.264 Intra for video with average and low motion but it is still inferior to the Inter No Motion and Inter Motion modes.

  19. Selection of a quench detection system for the ITER CS magnet

    International Nuclear Information System (INIS)

    Coatanea, Marc; Duchateau, Jean-Luc; Lacroix, Benoit; Nicollet, Sylvie; Rodriguez-Mateos, Felix; Topin, Frederic

    2011-01-01

    At variance with most of the existing superconducting systems operating in the world, the ITER central solenoid (CS) magnet is a fast pulsed system. This peculiarity creates a specific situation regarding the quench detection system, as a small resistive signal associated with a quench has to be discriminated from the high inductive signals imposed by the plasma scenario. The quench detection is based on an inductive compensation built from three adjacent double pancakes. The ITER protection rules for a superconducting magnet impose to respect the so-called maximum hot spot temperature criterion of 250 K in the quenched cable at the end of the fast discharge. A careful analysis of the residual inductive signals in the detection voltage shows that a blanking of the quench detection cannot be avoided during the early times of the plasma discharge (i.e. during 3.5 s). It is demonstrated that this blanking is, however, acceptable while fulfilling the hot spot criterion because the plasma initiation phase (PIP) is very similar to a fast safety discharge and corresponds to a fast decrease of the modules currents, which is favourable for the magnet protection.

  20. The ITER reduced cost design

    International Nuclear Information System (INIS)

    Aymar, R.

    2000-01-01

    Six years of joint work under the international thermonuclear experimental reactor (ITER) EDA agreement yielded a mature design for ITER which met the objectives set for it (ITER final design report (FDR)), together with a corpus of scientific and technological data, large/full scale models or prototypes of key components/systems and progress in understanding which both validated the specific design and are generally applicable to a next step, reactor-oriented tokamak on the road to the development of fusion as an energy source. In response to requests from the parties to explore the scope for addressing ITER's programmatic objective at reduced cost, the study of options for cost reduction has been the main feature of ITER work since summer 1998, using the advances in physics and technology databases, understandings, and tools arising out of the ITER collaboration to date. A joint concept improvement task force drawn from the joint central team and home teams has overseen and co-ordinated studies of the key issues in physics and technology which control the possibility of reducing the overall investment and simultaneously achieving the required objectives. The aim of this task force is to achieve common understandings of these issues and their consequences so as to inform and to influence the best cost-benefit choice, which will attract consensus between the ITER partners. A report to be submitted to the parties by the end of 1999 will present key elements of a specific design of minimum capital investment, with a target cost saving of about 50% the cost of the ITER FDR design, and a restricted number of design variants. Outline conclusions from the work of the task force are presented in terms of physics, operations, and design of the main tokamak systems. Possible implications for the way forward are discussed

  1. Analysis of the steady state hydraulic behaviour of the ITER blanket cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A., E-mail: pietroalessandro.dimaio@unipa.it [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Dell’Orco, G.; Furmanek, A. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Garitta, S. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Merola, M.; Mitteau, R.; Raffray, R. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Spagnuolo, G.A.; Vallone, E. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy)

    2015-10-15

    Highlights: • Nominal steady state hydraulic behaviour of ITER blanket standard sector cooling system has been investigated. • Numerical simulations have been run adopting a qualified thermal-hydraulic system code. • Hydraulic characteristic functions and coolant mass flow rates, velocities and pressure drops have been assessed. • Most of the considered circuits are able to effectively cool blanket modules, meeting ITER requirements. - Abstract: The blanket system is the ITER reactor component devoted to providing a physical boundary for plasma transients and contributing to thermal and nuclear shielding of vacuum vessel, magnets and external components. It is expected to be subjected to significant heat loads under nominal conditions and its cooling system has to ensure an adequate cooling, preventing any risk of critical heat flux occurrence while complying with pressure drop limits. At the University of Palermo a study has been performed, in cooperation with the ITER Organization, to investigate the steady state hydraulic behaviour of the ITER blanket standard sector cooling system. A theoretical–computational approach based on the finite volume method has been followed, adopting the RELAP5 system code. Finite volume models of the most critical blanket cooling circuits have been set-up, realistically simulating the coolant flow domain. The steady state hydraulic behaviour of each cooling circuit has been investigated, determining its hydraulic characteristic function and assessing the spatial distribution of coolant mass flow rates, velocities and pressure drops under reference nominal conditions. Results obtained have indicated that the investigated cooling circuits are able to provide an effective cooling to blanket modules, generally meeting ITER requirements in term of pressure drop and velocity distribution, except for a couple of circuits that are being revised.

  2. On One-Point Iterations and DIIS

    DEFF Research Database (Denmark)

    Østerby, Ole; Sørensen, Hans Henrik Brandenborg

    2009-01-01

    We analyze various iteration procedures in many dimensions inspired by the SCF iteration used in first principles electronic structure calculations. We show that the simple mixing of densities can turn a divergent (or slowly convergent) iteration into a (faster) convergent process provided all...

  3. 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

  4. ITER ITA newsletter. No. 1, February 2003

    International Nuclear Information System (INIS)

    2003-04-01

    This first issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about ITER related meetings including eighth ITER Negotiations meeting, held on 18-19 February, 2003 in St. Petersburg, Russia, first meeting of the ITER preparatory committee, held on 17 February, 2003 in St. Petersburg, Russia and the third meeting of the ITPA (International Tokamak Physics Activity) coordinating committee, held on 24-25 October 2002 at the Max-Planck-Institut fuer Plasmaphysik, Garching

  5. EMF 9 scenario design (EMF WP 9.4)

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    This paper describes the specifications of the scenarios selected by the EMF 9 Working Group, which is focusing on North American natural gas markets. The four first-round scenarios include: (1) an upper oil price path, (2) a lower oil price path, (3) a lower resource base, and (4) a higher gas demand due to policies and technologies affecting the electric utility sector. Each scenario represents a combination of market and regulatory environments. For example, the upper oil price scenario combines an upper oil price trend with a reference set of resource and demand conditions and with a pro-competitive regulatory environment. This scenario also serves as a control case for comparing the other scenarios, which specify a change in one of the environments for oil prices, resources, or demand. The regulatory environment has been maintained constant across these first-round scenarios but may be changed depending upon the later recommendations of the regulatory policy and market structure study group. The next section describes the detailed specifications for modelers in simulating the upper oil price scenario. Guidance is offered for energy prices, economic activity, and resource base estimates. The next three sections describe the other three scenarios that involve changes in these inputs from their values in the upper oil price scenario. Special wellhead price assumptions for stand-alone supply models and plans for developing inputs for Canadian models are then discussed. The final section describes the output variables to be reported to the EMF staff for comparing model results

  6. Behaviour of molten beryllium with ITER reference CFC SEPCARB NB31 under moisture

    Energy Technology Data Exchange (ETDEWEB)

    Lipa, M. [CEA/Cadarache-DSM/DRFC, Euratom-Association, Saint Paul Lez Durance F-13108 (France)], E-mail: manfred.lipa@cea.fr; Linke, J. [Forschungszentrum Juelich, Euratom-Association, Juelich D-52425 (Germany); Martin, G. [CEA/Cadarache-DSM/DRFC, Euratom-Association, Saint Paul Lez Durance F-13108 (France); Wessel, E. [Forschungszentrum Juelich, Euratom-Association, Juelich D-52425 (Germany)

    2007-10-15

    A dramatic exothermic reaction has been observed in Tore Supra with molten aluminium enclosed in a thermally loaded 2D carbon fibre reinforced carbon composite (CFC) tile, exposed after component removal to atmospheric moisture. This led to an important local destruction of the graphite matrix and fibre texture. Beryllium, as aluminium, a carbide forming metal, is foreseen for the ITER first wall armour. Since it can be expected that sputtered, molten and/or evaporated beryllium material will be transported to the hot CFC based divertor surface during normal and off-normal plasma operation, a preliminary reaction experiment has been performed in order to investigate to which extent the Be contaminated graphite matrix and fibre texture of 3D-CFC SEPCARB NB31 could be locally damaged ('erosion') following a subsequent long lasting air venting or a supposed in-vessel ingress of coolant (steam) event (ICE)

  7. ITER Conceptual design: Interim report

    International Nuclear Information System (INIS)

    1990-01-01

    This interim report describes the results of the International Thermonuclear Experimental Reactor (ITER) Conceptual Design Activities after the first year of design following the selection of the ITER concept in the autumn of 1988. Using the concept definition as the basis for conceptual design, the Design Phase has been underway since October 1988, and will be completed at the end of 1990, at which time a final report will be issued. This interim report includes an executive summary of ITER activities, a description of the ITER device and facility, an operation and research program summary, and a description of the physics and engineering design bases. Included are preliminary cost estimates and schedule for completion of the project

  8. ITER ITA newsletter. No. 20, February-March 2005

    International Nuclear Information System (INIS)

    2005-03-01

    This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about ITER related activities including interview on the occasion of Academician E.P. Velikhov' 70th birthday conducted by Dr. Lev Golubbchikov, former ITER Contact Person of the Russian Federation and a new document management system of ITER called IDM (ITER Document Management), which supersedes the old IDoMS

  9. The ITER Remote Maintenance Management System

    International Nuclear Information System (INIS)

    Tesini, Alessandro; Rolfe, A.C.

    2009-01-01

    A major challenge for the ITER project is to develop and implement a Remote Maintenance System, which can deliver high Tokamak availability within the constraints of the overall ITER programme objectives. Much of the maintenance of ITER will be performed using remote handling methods and some with combined manual and remote activities working together. The organization and management of the ITER remote handling facilities will be of a scale unlike any other remote handling application in the world. The ITER remote handling design and procurement activities will require co-ordination and management across many different sites throughout the world. It will be a major challenge for the ITER project to ensure a consistent quality and technical approach in all of the contributing parties. To address this issue the IO remote handling team are implementing the ITER Maintenance Management Plan (IMMP) comprising an overarching document defining the policies and methodologies (ITER Remote Maintenance Management System or IMMS) and an associated ITER remote handling code of practise (IRHCOP). The IMMS will be in document form available as a pdf file or similar. The IRHCOP will be implemented as a web based application and will provide access to the central resource of the entire code of practise from any location in the world. The IRHCOP data library will be centrally controlled in order that users can be assured of the data relevance and authenticity. This paper will describe the overall approach being taken to deal with this challenge and go on to detail the structure and content of both the IMMS and the IRHCOP.

  10. The ITER remote maintenance system

    International Nuclear Information System (INIS)

    Tesini, A.; Palmer, J.

    2007-01-01

    ITER is a joint international research and development project that aims to demonstrate the scientific and technological feasibility of fusion power. As soon as the plasma operation begins using tritium, the replacement of the vacuum vessel internal components will need to be done with remote handling techniques. To accomplish these operations ITER has equipped itself with a Remote Maintenance System; this includes the Remote Handling equipment set and the Hot Cell facility. Both need to work in a cooperative way, with the aim of minimizing the machine shutdown periods and to maximize the machine availability. The ITER Remote Handling equipment set is required to be available, robust, reliable and retrievable. The machine components, to be remotely handle-able, are required to be designed simply so as to ease their maintenance. The baseline ITER Remote Handling equipment is described. The ITER Hot Cell Facility is required to provide a controlled and shielded area for the execution of repair operations (carried out using dedicated remote handling equipment) on those activated components which need to be returned to service, inside the vacuum vessel. The Hot Cell provides also the equipment and space for the processing and temporary storage of the operational and decommissioning radwaste. A conceptual ITER Hot Cell Facility is described. (orig.)

  11. Tritium inventory and permeation in the ITER breeding blanket

    International Nuclear Information System (INIS)

    Violante, V.; Tosti, S.; Sibilia, C.; Felli, F.; Casadio, S.; Alvani, C.

    2000-01-01

    A model has allowed us to perform the analysis of the tritium inventory and permeation in the international thermonuclear experimental reactor (ITER) breeding blanket under the hypothesis of steady state conditions. Li 2 ZrO 3 (reference) and Li 2 TiO 3 (alternative) have been studied as breeding materials. The total breeder inventory assessed is 7.64 g for the Li 2 ZrO 3 at reference temperature. The model has also been used for a parametric analysis of the tritium permeation. At reference temperature and purge helium velocity of 0.01 m/s, the HT partial pressure is ranging from 10 to 30 Pa in the breeder and 1.5x10 -3 Pa in the beryllium. At 0.1 m/s of purge helium velocity, the HT partial pressure is reduced of one order by magnitude in the breeder and becomes 5x10 -5 Pa in the beryllium. The tritium permeation into the coolant for the whole blanket is ranging from 100 to 250 mCi per day for purge helium velocity of 0.01 m/s. The analysis of the tritium inventory and permeation for the alternative Li 2 TiO 3 breeding material has been carried out too. The tritium inventory in the breeder is in the range from 6 to 375 g larger than in Li 2 ZrO 3 by about a factor 5; the tritium permeation into coolant is comparable to the Li 2 ZrO 3 one. This analysis provides indications on the influence of the operating parameters on the tritium control in the ITER breeding blanket; particularly the control of the tritium inventory by the temperature and the tritium permeation by the purge gas velocity

  12. ITER EDA newsletter. V. 4, no. 9

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    This issue of the ITER EDA (Engineering Design Activities) Newsletter contains reports on the first meeting of the ITER Test Blanket Working Group held 19-21 July 1995 at the ITER Garching Joint Work Site, and on the second workshop of the ITER Expert Group on Confinement and Transport.

  13. ITER EDA newsletter. V. 4, no. 9

    International Nuclear Information System (INIS)

    1995-09-01

    This issue of the ITER EDA (Engineering Design Activities) Newsletter contains reports on the first meeting of the ITER Test Blanket Working Group held 19-21 July 1995 at the ITER Garching Joint Work Site, and on the second workshop of the ITER Expert Group on Confinement and Transport

  14. ITER EDA newsletter. V. 10, no. 6

    International Nuclear Information System (INIS)

    2001-06-01

    This ITER EDA Newsletter issue includes information about the ITER Management Advisory Committee Meeting held in Vienna on 16 July 2001 and also a summary of the ninth ITER Technical Meeting on safety and environment held at the ITER Garching Joint Work site, 8 to 10 May, 2001

  15. ITER EDA newsletter. V. 4, no.12

    International Nuclear Information System (INIS)

    1995-12-01

    This issue of the ITER EDA (Engineering Design Activities) Newsletter contains a report on the ninth ITER council meeting held December 12 - 13, 1995 in Garching near Munich, Germany (by Dr. E. Canobbio), a report on the status of the ITER EDA (by Dr. R. Aymar, ITER Director) and a report on the ninth meeting of the ITER Technical Advisory Committee (by Professor P. Rutherford, TAC Chair) held 27 - 29 November 1995, in Garching near Munich, Germany

  16. ITER ITA newsletter. No. 4, May 2003

    International Nuclear Information System (INIS)

    2003-07-01

    This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about ITER related meetings, one of them the eighth meeting of the ITER negotiators' standing sub-group (NSSG-8) and a number of related meetings from 14 to 22 May 2003 at Garching, Germany, another was bilateral blanket meeting between ITER International Team (IT) and the Research and Development Institute of Power Engineering (ENTEK), which was held in Moscow, Russian Federation on 22 and 23 May, 2003

  17. Reference Projections Energy and Emissions 2005-2020

    International Nuclear Information System (INIS)

    Van Dril, A.W.N.; Elzenga, H.E.

    2005-10-01

    The Reference Projection 2005-2020 covers the future development of Dutch energy use, greenhouse gas emissions and air pollution up to 2020. The Reference projection is based on assumptions regarding economic, structural, technological and policy developments. Two scenarios have been used. The Strong Europe (SE) scenario is characterized by moderate economic growth and strong public responsibility. The Global Economy (GE) scenario assumes high economic growth and has a strong orientation towards private responsibility. Energy consumption continues to grow in both scenarios and energy intensity is declining in the GE-scenario. Gradual rise of temperature is now included in the estimates for space heating and air conditioning. Energy prices for end users will rise, due to increased imports of natural gas and rising costs of electricity generation. The share of renewables in electricity consumption increases considerably due to subsidies for wind at sea and biomass, up to the target of 9% in 2010. Emissions of non-CO2 greenhouse gases are reduced and stabilise after 2010. The Dutch Kyoto target is probably met in both scenarios, assuming considerable emission reduction ef-forts abroad. Acidifying emissions of NOx and SO2 stabilise after reductions, but at levels that exceed their national emission ceiling (NEC). Emissions of volatile organic compounds are projected to fall with approximately 25% between 2002 and 2010 below their NEC. Emissions of ammonia are projected to meet their NEC. The emission of particulate matter (PM10) will stabilise at present levels

  18. ITER EDA Newsletter. V. 4, no. 7

    International Nuclear Information System (INIS)

    1995-07-01

    This ITER EDA (Engineering Design Activities) Newsletter issue contains reports on (i) the 8th meeting of the ITER Technical Advisory Committee (TAC-8) held on June 29 - July 7, 1995 at the ITER San Diego Work Site, (ii) the 8th meeting of the ITER Management Advisory Committee (MAC-8) held at the ITER San Diego Work Site on July 9-10, 1995, (iii) the 33rd meeting of the International Fusion Research Council (FRC), held July 11, 1995 at the IAEA Headquarters in Vienna, Austria, and (iv) the ITER participation in the fifth topical meeting on Tritium Technology in Fission, Fusion and Isotopic Applications

  19. ITER ITA newsletter. No. 27, January 2006

    International Nuclear Information System (INIS)

    2006-02-01

    This issue of ITER ITA (ITER transitional arrangements) newsletter contains concise information about two ITER related meetings including the twelfth ITER Negotiations Meeting and The Ninth Meeting of the ITPA Topical Group (TG) on Diagnostics was held at the National Fusion Research Centre (NFRC), Daejeon, Korea, from 10-14 October 2005

  20. ITER EDA newsletter. V. 5, no. 9

    International Nuclear Information System (INIS)

    1996-09-01

    This issue of the Newsletter on the Engineering Design Activities (EDA) for the ITER project contains an overview of one of the seven large ITER Research and Development Projects identified by the ITER Director, namely the Vacuum Vessel Sector, as well as an account of computer animation created for ITER