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ABSTRACT: Knowledge discovery consists of finding new knowledge from databases where dimension, complexity, or amount of data is prohibitively
large for human observation alone. The need for efficient data visualization and clustering is often faced, for instance,
in the analysis, monitoring, fault detection, or prediction of various engineering plants. In this paper, two clustering techniques,
K-means and Self-Organizing Maps, are used for the identification of characteristic regions for plasma scenario in nuclear
fusion experimental devices. The choice of the number of clusters, which heavily affects the performance of the mapping, is
firstly faced. Then, the ASDEX Upgrade Tokamak high-dimensional operational space is mapped into lower-dimensional maps, allowing
to detect the regions with high risk of disruption, and, finally, the current process state and its history in time are visualized
as a trajectory on the Self-Organizing Map, in order to predict the safe or disruptive state of the plasma.
KeywordsKnowledge discovery–Clustering–Self-Organizing Maps–Tokamak–Disruptions
Neural Computing and Applications 04/2012; 20(6):851-863. · 0.70 Impact Factor
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[show abstract]
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ABSTRACT: The mapping of the n-dimensional plasma parameter space of ASDEX Upgrade has been performed using a 2-dimensional Self Organizing Map, which reveals the map potentiality in data visualization. The proposed approach allows us the definition of simple displays capable of presenting meaningful information on the actual state of the plasma, but it also suggests to use the Self Organizing Map as a disruption predictor. In this paper, different criteria have been studied to associate the risk of disruption of each cluster in the map to a disruption alarm threshold. Data for this study comes from ASDEX Upgrade experiments executed between July 2002 and November 2009. The prediction performance of the proposed system has been evaluated on a set of discharges different from those used for the map training, obtaining a quite good prediction success rate. A deep analysis of the wrong predictions has been performed in order to identify possible common causes, and some criteria to increase prediction performance have been derived.
Fusion Engineering (SOFE), 2011 IEEE/NPSS 24th Symposium on; 07/2011
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ABSTRACT: In this paper, a neural predictor has been built using plasma discharges selected from two years of ASDEX Upgrade experiments, from July 2002 to July 2004. In order to test the real-time prediction capability of the system, its performance has been evaluated using discharges coming from different experimental campaigns, from June 2005 to July 2007. All disruptions that occurred in the chosen experimental campaigns were included with the exception of those occurring in the ramp-up phase, in the ramp-down phase (if the disruption does not happen in the first 100 ms), those caused by massive gas injection and disruptions following vertical displacement events. The large majority of selected disruptions are of the cooling edge type and typically preceded by the growth of tearing modes, degradation of the thermal confinement and enhanced plasma radiation. A very small percentage of them happen at large beta after a short precursor phase. For each discharge, seven plasma diagnostic signals have been selected from numerous signals available in real-time. During the training procedure, a self-organizing map has been used to reduce the database size in order to improve the training of the neural network. Moreover, an optimization procedure has been performed to discriminate between safe and pre-disruptive phases. The prediction success rate has been further improved, performing an adaptive training of the network whenever a missed alarm is triggered by the predictor.
Nuclear Fusion 06/2010; 50(7):075004. · 4.09 Impact Factor
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J. Adamek,
C. Angioni,
G. Antar,
C. V. Atanasiu,
M. Balden,
W Becker,
K. Behler,
K. Behringer,
A Bergmann,
R. Bilato, [......],
C. Wigger,
M. Wischmeier,
E. Wolfrum,
E. Würsching,
D. Yadikin,
Q Yu,
D. Zasche,
T. Zehetbauer,
M. Zilker,
H. Zohm
The Review of scientific instruments 03/2010; 81(3):039903. · 1.52 Impact Factor
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J Adamek,
C Angioni,
G Antar,
C V Atanasiu,
M Balden,
W Becker,
K Behler,
K Behringer,
A Bergmann,
R Bilato, [......],
C Wigger,
M Wischmeier,
E Wolfrum,
E Wuersching,
D Yadikin,
Q Yu,
D Zasche,
T Zehetbauer,
M Zilker,
H Zohm
Reviews of Scientific Instruments. 01/2010; 81(3):-.
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B. Esposito,
G. Granucci,
S. Nowak,
J.R. Martin-Solis,
L. Gabellieri,
E. Lazzaro,
P. Smeulders,
M. Maraschek, G. Pautasso,
J. Stober,
W. Treutterer,
L. Urso,
F. Volpe,
H. Zohm,
FTU,
ECRH,
ASDEX Upgrade Teams
[show abstract]
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ABSTRACT: The use of ECRH has been investigated as a promising technique to avoid or postpone disruptions in dedicated experiments in FTU and ASDEX Upgrade. Disruptions have been produced by injecting Mo through laser blow-off (FTU) or by puffing deuterium gas above the Greenwald limit (FTU and ASDEX Upgrade). The toroidal magnetic field is kept fixed and the ECRH launching mirrors have been steered before every discharge in order to change the deposition radius. The loop voltage signal is used as disruption precursor to trigger the ECRH power before the plasma current quench. In the FTU experiments (Ip = 0.35–0.5 MA, Bt = 5.3 T, PECRH = 0.4–1.2 MW) it is found that the application of ECRH modifies the current quench starting time depending on the power deposition location. A scan in deposition location has shown that the direct heating of one of the magnetic islands produced by magnetohydrodynamic (MHD) resistive instabilities (either m/n = 3/2, 2/1 or 3/1) prevents its further growth and also produces the stabilization of the other coupled modes and the delay of the current quench or its full avoidance. Disruption avoidance and complete discharge recovery are obtained when the ECRH power is applied on rational surfaces. The modes involved in the disruption are found to be tearing modes stabilized by a strong local ECRH heating. The Rutherford equation has been used to reproduce the evolution of the MHD modes. In the ASDEX Upgrade experiments L-mode plasmas (Ip = 0.6 MA, Bt = 2.5 T, PECRH = 0.6 MW ~ POHM) the injection of ECRH close to q = 2 significantly delays the 2/1 onset and prolongs the duration of the discharge: during this phase the density continues to increase. No delay in the onset of the 2/1 mode is observed when the injected power is reduced to 0.35 MW.
Nuclear Fusion 05/2009; 49(6):065014. · 4.09 Impact Factor
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L. Giannone,
W. Schneider,
P.J. McCarthy,
A.C.C. Sips,
W. Treutterer,
K.Behler,
T.Eich,
J.C. Fuchs,
N. Hicks,
A. Kallenbach, [......],
A. Mlynek,
G. Neu, G. Pautasso,
G. Raupp,
M. Reich,
K.H. Schuhbeck,
J. Stober,
F. Volpe,
T. Zehetbauer,
ASDEX Upgrade Team
Fusion Engineering and Design 01/2009; · 1.49 Impact Factor
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A. Kallenbach,
R. Dux,
T. Eich,
R. Fischer,
L. Giannone,
J. Harhausen,
A. Herrmann,
H.W. Müller, G. Pautasso,
M. Wischmeier,
the ASDEX Upgrade Team
[show abstract]
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ABSTRACT: Particle, electric charge and power fluxes for type-I ELMy H-modes are measured in the divertor of the ASDEX Upgrade tokamak by triple Langmuir probes, shunts, infrared (IR) thermography and spectroscopy. The discharges are in the medium to high density range, resulting in predominantly convective edge localized modes (ELMs) with moderate fractional stored energy losses of 2% or below. Time resolved data over ELM cycles are obtained by coherent averaging of typically one hundred similar ELMs, spatial profiles from the flush-mounted Langmuir probes are obtained by strike point sweeps. The application of simple physics models is used to compare different diagnostics and to make consistency checks, e.g. the standard sheath model applied to the Langmuir probes yields power fluxes which are compared with the thermographic measurements. In between ELMs, Langmuir probe and thermography power loads appear consistent in the outer divertor, taking into account additional load due to radiation and charge exchange neutrals measured by thermography. The inner divertor is completely detached and no significant power flow by charged particles is measured. During ELMs, quite similar power flux profiles are found in the outer divertor by thermography and probes, albeit larger uncertainties in Langmuir probe evaluation during ELMs have to be taken into account. In the inner divertor, ELM power fluxes from thermography are a factor 10 larger than those derived from probes using the standard sheath model. This deviation is too large to be caused by deficiencies of probe analysis. The total ELM energy deposition from IR is about a factor 2 higher in the inner divertor compared with the outer divertor. Spectroscopic measurements suggest a quite moderate contribution of radiation to the target power load. Shunt measurements reveal a significant positive charge flow into the inner target during ELMs. The net number of elementary charges correlates well with the total core particle loss obtained from highly resolved density profiles. As a consequence, the discrepancy between probe and IR measurements is attributed to the ion power channel via a high mean impact energy of the ions at the inner target. The dominant contributing mechanism is proposed to be the directed loss of ions from the pedestal region into the inner divertor.
Nuclear Fusion 07/2008; 48(8):085008. · 4.09 Impact Factor
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R Neu,
M Balden,
V Bobkov,
R Dux,
O Gruber,
A Herrmann,
A Kallenbach,
M Kaufmann,
C F Maggi,
H Maier, [......],
D Wagner,
M Wischmeier,
E Wolfrum,
E Würsching,
D Yadikin,
Q Yu,
D Zasche,
T Zehetbauer,
M Zilker,
H Zohm
[show abstract]
[hide abstract]
ABSTRACT: ASDEX Upgrade has recently finished its transition towards an all-W divertor tokamak, by the exchange of the last remaining graphite tiles to W-coated ones. The plasma start-up was performed without prior boronization. It was found that the large He content in the plasma, resulting from DC glow discharges for conditioning, leads to a confinement reduction. After the change to D glow for inter-shot conditioning, the He content quickly dropped and, in parallel, the usual H-Mode confinement with H factors close to one was achieved. After the initial conditioning phase, oxygen concentrations similar to that in previous campaigns with boronizations could be achieved. Despite the removal of all macroscopic carbon sources, no strong change in C influxes and C content could be observed so far. The W concentrations are similar to the ones measured previously in discharges with old boronization and only partial coverage of the surfaces with W. Concomitantly it is found that although the W erosion flux in the divertor is larger than the W sources in the main chamber in most of the scenarios, it plays only a minor role for the W content in the main plasma. For large antenna distances and strong gas puffing, ICRH power coupling could be optimized to reduce the W influxes. This allowed a similar increase of stored energy as yielded with comparable beam power. However, a strong increase of radiated power and a loss of H-Mode was observed for conditions with high temperature edge plasma close to the antennas. The use of ECRH allowed keeping the central peaking of the W concentration low and even phases of improved H-modes have already been achieved.
Plasma Physics and Controlled Fusion 11/2007; 49(12B):B59. · 2.42 Impact Factor
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B. Lipschultz,
X. Bonnin,
G. Counsell,
A. Kallenbach,
A. Kukushkin,
K. Krieger,
A. Leonard,
A. Loarte,
R. Neu,
R.A. Pitts, [......],
Y. Pan, G. Pautasso,
V. Philipps,
V. Rohde,
D. Rudakov,
P. Stangeby,
S. Takamura,
T. Tanabe,
Y. Yang,
S. Zhu
[show abstract]
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ABSTRACT: Recent research in scrape-off layer (SOL) and divertor physics is reviewed; new and existing data from a variety of experiments have been used to make cross-experiment comparisons with implications for further research and ITER. Studies of the region near the separatrix have addressed the relationship of profiles to turbulence as well as the scaling of the parallel power flow. Enhanced low-field side radial transport is implicated as driving parallel flows to the inboard side. The medium-n nature of edge localized modes (ELMs) has been elucidated and new measurements have determined that they carry ~10–20% of the ELM energy to the far SOL with implications for ITER limiters and the upper divertor. The predicted divertor power loads for ITER disruptions are reduced while those to main chamber plasma facing components (PFCs) increase. Disruption mitigation through massive gas puffing is successful at reducing PFC heat loads. New estimates of ITER tritium retention have shown tile sides to play a significant role; tritium cleanup may be necessary every few days to weeks. ITER's use of mixed materials gives rise to a reduction of surface melting temperatures and chemical sputtering. Advances in modelling of the ITER divertor and flows have enhanced the capability to match experimental data and predict ITER performance.
Nuclear Fusion 08/2007; 47(9):1189. · 4.09 Impact Factor
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G. Pautasso,
C.J. Fuchs,
O. Gruber,
C.F. Maggi,
M. Maraschek,
T. Pütterich,
V. Rohde,
C. Wittmann,
E. Wolfrum,
P. Cierpka,
M. Beck,
the ASDEX Upgrade Team
[show abstract]
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ABSTRACT: The massive injection of impurity gas into a plasma has been proved to reduce forces and localized thermal loads caused by disruptions in tokamaks. This mitigation system is routinely used on ASDEX Upgrade to shut down plasmas with a locked mode. The plasma response to impurity injection and the mechanism of reduction of the mechanical forces is discussed in the paper.
Nuclear Fusion 07/2007; 47(8):900. · 4.09 Impact Factor
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T.C. Hender,
J.C Wesley,
J. Bialek,
A. Bondeson,
A.H. Boozer,
R.J. Buttery,
A. Garofalo,
T.P Goodman,
R.S. Granetz,
Y. Gribov, [......],
M. Sugihara,
M. Takechi,
A.D. Turnbull,
E. Westerhof,
D.G. Whyte,
R. Yoshino,
H. Zohm,
the ITPA MHD,
Disruption,
Magnetic Control Topical Group
[show abstract]
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ABSTRACT: Progress in the area of MHD stability and disruptions, since the publication of the 1999 ITER Physics Basis document (1999 Nucl. Fusion 39 2137–2664), is reviewed. Recent theoretical and experimental research has made important advances in both understanding and control of MHD stability in tokamak plasmas. Sawteeth are anticipated in the ITER baseline ELMy H-mode scenario, but the tools exist to avoid or control them through localized current drive or fast ion generation. Active control of other MHD instabilities will most likely be also required in ITER. Extrapolation from existing experiments indicates that stabilization of neoclassical tearing modes by highly localized feedback-controlled current drive should be possible in ITER. Resistive wall modes are a key issue for advanced scenarios, but again, existing experiments indicate that these modes can be stabilized by a combination of plasma rotation and direct feedback control with non-axisymmetric coils. Reduction of error fields is a requirement for avoiding non-rotating magnetic island formation and for maintaining plasma rotation to help stabilize resistive wall modes. Recent experiments have shown the feasibility of reducing error fields to an acceptable level by means of non-axisymmetric coils, possibly controlled by feedback. The MHD stability limits associated with advanced scenarios are becoming well understood theoretically, and can be extended by tailoring of the pressure and current density profiles as well as by other techniques mentioned here. There have been significant advances also in the control of disruptions, most notably by injection of massive quantities of gas, leading to reduced halo current fractions and a larger fraction of the total thermal and magnetic energy dissipated by radiation. These advances in disruption control are supported by the development of means to predict impending disruption, most notably using neural networks. In addition to these advances in means to control or ameliorate the consequences of MHD instabilities, there has been significant progress in improving physics understanding and modelling. This progress has been in areas including the mechanisms governing NTM growth and seeding, in understanding the damping controlling RWM stability and in modelling RWM feedback schemes. For disruptions there has been continued progress on the instability mechanisms that underlie various classes of disruption, on the detailed modelling of halo currents and forces and in refining predictions of quench rates and disruption power loads. Overall the studies reviewed in this chapter demonstrate that MHD instabilities can be controlled, avoided or ameliorated to the extent that they should not compromise ITER operation, though they will necessarily impose a range of constraints.
Nuclear Fusion 05/2007; 47(6):S128. · 4.09 Impact Factor
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A Loarte,
G Saibene,
R Sartori,
V Riccardo,
P Andrew,
J Paley,
W Fundamenski,
T Eich,
A Herrmann, G Pautasso,
A Kirk,
G Counsell,
G Federici,
G Strohmayer,
D Whyte,
A Leonard,
R A Pitts,
I Landman,
B Bazylev,
S Pestchanyi
[show abstract]
[hide abstract]
ABSTRACT: New experimental results on transient loads during ELMs and disruptions in present divertor tokamaks are described and used to carry out a extrapolation to ITER reference conditions and to draw consequences for its operation. In particular, the achievement of low energy/convective type I edge localized modes (ELMs) in ITER-like plasma conditions seems the only way to obtain transient loads which may be compatible with an acceptable erosion lifetime of plasma facing components (PFCs) in ITER. Power loads during disruptions, on the contrary, seem to lead in most cases to an acceptable divertor lifetime because of the relatively small plasma thermal energy remaining at the thermal quench and the large broadening of the power flux footprint during this phase. These conclusions are reinforced by calculations of the expected erosion lifetime, under these load conditions, which take into account a realistic temporal dependence of the power fluxes on PFCs during ELMs and disruptions.
Physica Scripta 03/2007; 2007(T128):222. · 1.20 Impact Factor
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[show abstract]
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ABSTRACT: First results are reported on the prediction of disruptions in one tokamak, based on neural networks trained on another tokamak. The studies use data from the JET and ASDEX Upgrade devices, with a neural network trained on just seven normalized plasma parameters. In this way, a simple single layer perceptron network trained solely on JET correctly anticipated 67% of disruptions on ASDEX Upgrade in advance of 0.01 s before the disruption. The converse test led to a 69% success rate in advance of 0.04 s before the disruption in JET. Only one overall time scaling parameter is allowed between the devices, which can be introduced from theoretical arguments. Disruption prediction performance based on such networks trained and tested on the same device shows even higher success rates (JET, 86%; ASDEX Upgrade, 90%), despite the small number of inputs used and simplicity of the network. It is found that while performance for networks trained and tested on the same device can be improved with more complex networks and many adjustable weights, for cross-machine testing the best approach is a simple single layer perceptron. This offers the basis of a potentially useful technique for large future devices such as ITER, which with further development might help to reduce disruption frequency and minimize the need for a large disruption campaign to train disruption avoidance systems.
Nuclear Fusion 04/2005; 45(5):337. · 4.09 Impact Factor
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Fusion Eng. аnd Design. 01/2005; 75-79:589-593.
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Europhysics Conference Abstracts (CD-ROM, Proc. of the 31th EPS Conference on Controlled Fusion and Plasma Physics, London 2004). 01/2004;
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[show abstract]
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ABSTRACT: The presence of an extended region of open flux surfaces (halo), during the current quench phase of the disruption of elongated plasmas, is supported by measurements of halo currents and by numerical simulations. The halo, in addition to providing a poloidal current path between the plasma and the first-wall components, allows rapid conduction and convection of energy along field lines, and therefore a mechanism for the localized deposition of energy onto the wall. The heat load to the region of the plasma-first-wall interaction is higher than in the scenario in which the magnetic energy is mostly dissipated by radiative processes
Nuclear Fusion 10/2002; 34(3):455. · 4.09 Impact Factor
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[show abstract]
[hide abstract]
ABSTRACT: Injection of impurity pellets has been shown to be a successful method of reducing thermal and mechanical loads during disruptions. The evolution of the quenching plasma after pellet injection in ASDEX Upgrade is described and the requirements of such a method for mitigating disruptions in future devices are discussed
Nuclear Fusion 10/2002; 36(10):1291. · 4.09 Impact Factor
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O. Gruber,
H.-S. Bosch,
S. Günter,
A. Herrmann,
A. Kallenbach,
M. Kaufmann,
K. Krieger,
K. Lackner,
V. Mertens,
R. Neu, [......],
U. Wenzel,
F. Wesner,
R. Wunderlich,
N. Xantopoulos,
Q. Yu,
D. Zasche,
T. Zehetbauer,
H.-P. Zehrfeld,
H. Zohm,
M. Zouhar
[show abstract]
[hide abstract]
ABSTRACT: The closed ASDEX Upgrade Divertor II, `LYRA', is capable of handling heating powers of up to 20 MW or P/R of 12 MW/m, owing to a reduction of the maximum heat flux to the target plates by more than a factor of 2 compared with the open Divertor I. This reduction is caused by high radiative losses from carbon and hydrogen inside the divertor region and is in agreement with B2-EIRENE modelling predictions. At medium densities in the H mode, the type I ELM behaviour shows no dependence on the heating method (NBI, ICRH). ASDEX Upgrade-JET dimensionless identity experiments showed compatibility of the L-H transition with core physics constraints, while in the H mode confinement, inconsistencies with the invariance principle were established. At high densities close to the Greenwald density, the MHD limited edge pressures, the influence of divertor detachment on separatrix parameters and increasing edge transport lead to limited edge densities and finally to temperatures below the critical edge temperatures for H mode. This results in a drastic increase of the H mode threshold power and an upper H mode density limit with gas puff refuelling. The H mode confinement degradation approaching this density limit is caused by the ballooning mode limited edge pressures and `stiff' temperature profiles relating core and edge temperatures. Repetitive high field side pellet injection allows for H mode operation well above the Greenwald density; moreover, higher confinement than with gas fuelling is found up to the highest densities. Neoclassical tearing modes limit the achievable β depending on the collisionality at the resonant surface. In agreement with the polarization current model, the onset β is found to be proportional to the ion gyroradius in the collisionless regime, while higher collisionalities are stabilizing. The fractional energy loss connected with saturated modes at high pressures is about 25%. A reduction of neoclassical mode amplitude and an increase of β have been demonstrated by using phased ECRH and ECCD in the O point of islands. Advanced tokamak operation with internal transport barriers for both ions and electrons has been achieved with flat shear profiles and q0 > 1 or with reversed shear and qmin > 2. With flat shear a stationary H mode scenario was maintained for 40 confinement times and several internal skin times with βN = 2 and HITERL-89P = 2.4, where fishbones keep q0 at 1. βN is limited by either neoclassical tearing modes in the case of flat shear or kink modes with reversed shear.
Nuclear Fusion 05/2002; 39(9Y):1321. · 4.09 Impact Factor
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[show abstract]
[hide abstract]
ABSTRACT: Disruption is a sudden loss of magnetic confinement that can cause damage to the machine walls and support structures. For this reason, it is of practical interest to be able to detect the onset of such an event early. A novel technique is presented of early prediction of plasma disruption in tokamak reactors which uses neural networks and `fuzzy' inference. The studies carried out in the work make use of an experimental database of disruptive shots made available by the ASDEX Upgrade Team. The main result of the work is that, in the limit of the available database, it is possible to predict the onset of the disruptive event sufficiently in advance in order to put the control system into action. The proposed system is a modular scheme that exploits a decomposition of the original database carried out in a proper way.
Nuclear Fusion 05/2002; 41(11):1715. · 4.09 Impact Factor
