Integrated scenario in JET using real-time profile control

Joffrin, E.; Crisanti, F.; Felton, R.; Litaudon, X.; Mazon, D.; Moreau, D.; Zabeo, L.; Albanese, R.; Ariola, M.; Alves, D.; Barana, O.; Basiuk, V.; Bécoulet, A.; Bécoulet, M.; Blum, Jacques; Bolzonnella, T.; Bosak, K.; Chareau, J. M.; Baar, M. de; Luna, E. de la; Vries, P. de; Dumortier, P.; Elbeze, D.; Farthing, J.; Fernandes, H.; Fenzi, C.; Giannella, R.; Guenther, Karl H.; Harling, J.; Hawkes, N.; Hender, T. C.; Howell, D. F.; Heesterman, P.; Imbeaux, F.; Innocente, P.; Laborde, Laurent; Lloyd, G. Kenneth; Lomas, P. J.; McDonald, D. C.; Mailloux, J.; Mantsinen, M.; Messiaen, A.; Murari, A.; Ongena, J.; Orsitto, F.; Pericoli‐Ridolfini, V.; Riva, M.; Sánchez, J.; Sartori, F.; Sauter, O.; Sips, A. C. C.; Tala, T.; Tuccillo, A. A.; Ester, D. van; Zastrow, K.‐D.; Zerbini, M.

doi:10.1088/0741-3335/45/12a/024

Cited by 59 publications

(31 citation statements)

References 36 publications

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“…In this hybrid scenario [19][20][21][22][23] plasma on DIII-D, increasing the normalized beta (β N ) from 2.6 to 3.2 at 4250 ms, causes a m/n = 3/2 NTM to be destabilized ≈350 ms later. A direct measurement of the safety factor minimum using the MSE pitch angles [15] shows an abrupt increase from q min < 1 to q min ≈ 1.1 after the destabilization of the m/n = 3/2 NTM (figure 1(c)); this is consistent with the disappearance of the sawtooth oscillation after 4700 ms.…”

Section: Measurement Of Missing Bootstrap Currentmentioning

confidence: 99%

Local formulae for combinatorial Pontryagin classes

Gaifullin¹

2004

Izv. Math.

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Using direct analysis of the motional Stark effect (MSE) signals, an explicit measurement of the 'missing' bootstrap current density around the island location of a neoclassical tearing mode (NTM) is made for the first time. When the NTM is suppressed using co-electron cyclotron current drive, the measured changes in the current profile that restore the bootstrap current are also directly found from the MSE measurements. Additionally, direct analysis of helical perturbations in the MSE signals during slowly rotating 'quasi-stationary' modes shows the first explicit measurement of the deficit in the toroidal current density in the island O-point.

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Section: Measurement Of Missing Bootstrap Currentmentioning

confidence: 99%

Local formulae for combinatorial Pontryagin classes

Gaifullin¹

2004

Izv. Math.

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“…Substantial contributions to all ITER operational scenarios were made due to continuous development of JET Real Time Control capabilities. The implemented real time diagnostics and control techniques proved to enhance performance of the JET plasmas and minimise the internal disturbances [11]. In a recent campaign, pressure and safety factor profiles were simultaneously feedback controlled in ITB plasmas (I p = 1.7 MA, B T = 3 T) for up to 7 seconds for the first time.…”

Section: Overview Of Recent Jet Resultsmentioning

confidence: 99%

JET: Preparing the future in fusion

et al. 2004

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JET (Joint European Torus) is the largest tokamak in the world and the only fusion facility able to operate with Tritium, the fusion fuel, and Beryllium, the ITER first wall material. JET also features the most complete remote handling equipment for invessel maintenance. As a multinational research center, JET provides logistic experience in preparing for operation of the global facility, tokamak ITER.Experiments on JET are focused on ITER-relevant studies, in particular on detailing the operational scenarios (ELMy H-modes and advanced regimes), on enhancing the heating systems, on developing diagnostics for burning plasmas etc. Pioneering real-time control techniques have been implemented that maximize performance and minimize internal disturbances of JET plasmas. In helium plasmas, ion cyclotron heating (ICRH) created fast α-particles, mimicking their populations in future burning plasmas. The recent successful Trace Tritium campaign provided important new data on fuel transport. Current enhancements on JET include a new ITER-like ELM-resilient high power ICRH antenna (7 MW) and over twenty new diagnostics that will further extend the JET scientific capabilities and push the facility even closer to the ITER parameters.A special mention is given to the involvement of the fusion experts from Association EURATOM-IPP.CR, who have been actively participating in the collective use of JET facility for more than three years. PACS : 52.55.Fa 1 JET, the European center of fusion research JET (Joint European Torus, [1], see Fig. 1) with the plasma major and minor radii R = 2.96 m and a = 1.25 m, respectively, is the world biggest tokamak. Plasma with approx. volume 80 m 3 is confined by magnetic field with toroidal on-axis magnetic field B T max. 4 T, with plasma currents I p up to 4 MA. The central plasma ion temperature can reach up to 40 keV, plasma densities have the order

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“…Several tokamaks have now developed comprehensive real time measurement networks capable of issuing most of the data required for the control of steady state discharges, in particular JET [45], Tore Supra [71], JT-60U [72] and DIII-D [73]. Other plasma parameters relevant to steady state control are also calculated online by dedicated codes.…”

Section: Specific Control Issues For Steady State Operationmentioning

confidence: 99%

“…Initially real time control systems were used to control one plasma parameter with one actuator only. This enabled one to maintain the performance (and avoid major instabilities) of advanced scenarios in various experiments by using feedback control of the stored energy or neutron rate through the modulation of the neutral beam power [45,73,[77][78][79].…”

Section: Specific Control Issues For Steady State Operationmentioning

confidence: 99%

Chapter 8: Plasma operation and control

et al. 2007

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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 and control is similar in ITER and present tokamaks, there is a principal qualitative difference. To minimize its cost, ITER has been designed with small margins in many plasma and engineering parameters. These small margins result in a significantly narrower operational space compared with present tokamaks. Furthermore, ITER operation is expensive and component damage resulting from purely operational errors might lead to a high and avoidable repair cost. These factors make it judicious to use validated plasma diagnostics and employ simulators to 'pre-test' the combined ITER operation and control systems. Understanding of how to do this type of pre-test validation is now developed in present day experiments. This research push should provide us with fully functional simulators before the first ITER operation.

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Integrated scenario in JET using real-time profile control

Abstract: The recent development of real time measurements and control tools in JET has enhanced the reliability and reproducibility of the relevant ITER scenarios.

Cited by 59 publications

References 36 publications

Local formulae for combinatorial Pontryagin classes

Local formulae for combinatorial Pontryagin classes

JET: Preparing the future in fusion

Chapter 8: Plasma operation and control

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