The ITER full size plasma source device design Sonato, P.; Agostinetti, P.; Anaclerio, G. ...
Fusion engineering and design,
06/2009, Letnik:
84, Številka:
2
Journal Article, Conference Proceeding
Recenzirano
In the framework of the strategy for the development and the procurement of the NB systems for ITER, it has been decided to build in Padova a test facility, including two experimental devices: a full ...size plasma source with low voltage extraction and a full size NB injector at full beam power (1
MV). These two different devices will separately address the main scientific and technological issues of the 17
MW NB injector for ITER. In particular the full size plasma source of negative ions will address the ITER performance requirements in terms of current density and uniformity, limitation of the electron/ion ratio and stationary operation at full current with high reliability and constant performances for the whole operating time up to 1
h. The required negative ion current density to be extracted from the plasma source ranges from 290
A/m
2 in D
2 (D
−) and 350
A/m
2 in H
2 (H
−) and these values should be obtained at the lowest admissible neutral pressure in the plasma source volume, nominally at 0.3
Pa. The electron to ion ratio should be limited to less than 1 and the admissible ion inhomogeneity extracted from the grids should be better than 10% on the whole plasma cross-section having a surface exposed to the extraction grid of the order of 1
m
2.
The main design choices will be presented in the paper as well as an overview of the design of the main components and systems.
This paper proposes a control-oriented approach to the tokamak plasma current profile dynamics. It is established based on a consistent set of simplified relationships, in particular for the ...microwave current drive sources, rather than exact physical modelling. Assuming that a proper model for advanced control schemes can be established using the socalled cylindrical approximation and neglecting the diamagnetic effects, we propose a model that focuses on the flux diffusion (from which the current profile is inferred). Its inputs are some real-time measurements available on modern tokamaks and the effects of some major actuators, such as the magnetic coils, Lower Hybrid (LHCD), Electron and Ion Cyclotron Frequency (ECCD and ICRH) systems, are particularly taken into account. More precisely, the non-inductive current profile sources are modelled as 3-parameters functions of the control inputs derived either from approximate theoretical formulae for the ECCD and bootstrap terms or from experimental scaling laws specifically developed from Hard X-ray Tore Supra data for the LHCD influence. The use of scaling laws in this model reflects the fact that the operation of future reactors will certainly depend upon a great number of scaling laws and specific engineering parameters. The discretisation issues are also specifically addressed, to ensure the robustness with respect to discretisation errors and the efficiency (in terms of computation time) of the associated algorithm. This model is compared with experimental results and the CRONOS solver for Tore Supra Tokamak.
Disruptions remain one of the most hazardous events in the operation of a tokamak device, since they can cause damage to the vacuum vessel and surrounding structures. Their potential danger increases ...with the plasma volume and energy content and therefore they will constitute an even more serious issue for the next generation of machines. For these reasons, in the recent years a lot of attention has been devoted to devise predictors, capable of foreseeing the imminence of a disruption sufficiently in advance, to allow time for undertaking remedial actions. In this paper, the results of applying fuzzy logic and classification and regression trees (CART) to the problem of predicting disruptions at JET are reported. The conceptual tools of fuzzy logic, in addition to being well suited to accommodate the opinion of experts even if not formulated in mathematical but linguistic terms, are also fully transparent, since their governing rules are human defined. They can therefore help not only in forecasting disruptions but also in studying their behaviour. The analysis leading to the rules of the fuzzy predictor has been complemented with a systematic investigation of the correlation between the various experimental signals and the imminence of a disruption. This has been performed with an exhaustive, non-linear and unbiased method based on decision trees. This investigation has confirmed that the relative importance of various signals can change significantly depending on the plasma conditions. On the basis of the results provided by CART on the information content of the various quantities, the prototype of an adaptive fuzzy logic predictor was trained and tested on JET database. Its performance is significantly better than the previous static one, proving that more flexible prediction strategies, not uniform over the whole discharge but tuned to the operational region of the plasma at any given time, can be very competitive and should be investigated systematically.
To operate advanced plasma scenario (long pulse with high stored energy) in present and future tokamak devices under safe operation conditions, the control requirements of the plasma control system ...(PCS) leads to the development of advanced feedback control and real time handling exceptions.
To develop these controllers and these exceptions handling strategies, a project aiming at setting up a flight simulator has started at CEA in 2009. Now, the new WEST (W Environment in Steady-state Tokamak) project deals with modifying Tore Supra into an ITER-like divertor tokamak. This upgrade impacts a lot of systems including Tore Supra PCS and is the opportunity to improve the current PCS architecture to implement the previous works and to fulfill the needs of modern tokamak operation.
This paper is dealing with the description of the architecture of WEST PCS. Firstly, the requirements will be presented including the needs of new concepts (segments configuration, alternative (or backup) scenario, …). Then, the conceptual design of the PCS will be described including the main components and their functions.
The third part will be dedicated to the proposal RT framework and to the technologies that we have to implement to reach the requirements.
During a tokamak discharge, several control modes may have to be run in sequence in order to perform the control of the different discharge phases. The transitions between these control modes are not ...always easy to handle because in most cases the coupling between the controlled plasma quantities is not taken into account in each control mode design process. This paper presents a new Multi-Inputs/Multi-Outputs (MIMO) controller applied on Tore Supra to control both plasma current and flux variations through the central solenoid voltage and the lower hybrid current drive (LHCD) system power. It deals with the transition from a loop voltage floating mode to a loop voltage control mode. The controller, synthesized and tuned using a model-based approach, has been validated in simulation before its successful implementation on Tore Supra experiments.
This paper summarizes the main achievements of the RFX fusion science program in the period between the 2008 and 2010 IAEA Fusion Energy Conferences. RFX-mod is the largest reversed field pinch in ...the world, equipped with a system of 192 coils for active control of MHD stability. The discovery and understanding of helical states with electron internal transport barriers and core electron temperature >1.5 keV significantly advances the perspectives of the configuration. Optimized experiments with plasma current up to 1.8 MA have been realized, confirming positive scaling. The first evidence of edge transport barriers is presented. Progress has been made also in the control of first-wall properties and of density profiles, with initial first-wall lithization experiments. Micro-turbulence mechanisms such as ion temperature gradient and micro-tearing are discussed in the framework of understanding gradient-driven transport in low magnetic chaos helical regimes. Both tearing mode and resistive wall mode active control have been optimized and experimental data have been used to benchmark numerical codes. The RFX programme also provides important results for the fusion community and in particular for tokamaks and stellarators on feedback control of MHD stability and on three-dimensional physics. On the latter topic, the result of the application of stellarator codes to describe three-dimensional reversed field pinch physics will be presented.
With the exploration of the MA plasma current regime in up to 0.5 s long discharges, RFX-mod has opened new and very promising perspectives for the reversed field pinch (RFP) magnetic configuration, ...and has made significant progress in understanding and improving confinement and in controlling plasma stability. A big leap with respect to previous knowledge and expectations on RFP physics and performance has been made by RFX-mod since the last 2006 IAEA Fusion Energy Conference. A new self-organized helical equilibrium has been experimentally achieved (the Single Helical Axis—SHAx—state), which is the preferred state at high current. Strong core electron transport barriers characterize this regime, with electron temperature gradients comparable to those achieved in tokamaks, and by a factor of 4 improvement in confinement time with respect to the standard RFP. RFX-mod is also providing leading edge results on real-time feedback control of MHD instabilities, of general interest for the fusion community.
High current regimes in RFX-mod Valisa, M; Bolzonella, T; Buratti, P ...
Plasma physics and controlled fusion,
12/2008, Letnik:
50, Številka:
12
Journal Article
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