Controlling extremely shaped plasmas in the JET tokamak
published in 42nd IEEE Conference on Decision and Control (CDC'03), Maui, Hawaii, December 2003
This paper deals with the design of a new plasma current and shape controller for the JET tokamak. This new controller... more This paper deals with the design of a new plasma current and shape controller for the JET tokamak. This new controller is aimed at improving the performance of the present controller so as to allow the control of extremely shaped plasmas with higher values of elongation and triangularity. The design approach makes use of a linearized mathematical model linking the voltages applied to the active coils with a number of geometrical descriptors characterizing the plasma shape. To describe accurately the plasma shape a number of geometrical descriptors, greater than the number of available active circuits, is needed. Hence the control problem under investigation is characterized by the fact that the number of parameters to be controlled is larger than the number of control inputs. To overcome this problem a singular value analysis is carried out to identify the principal directions of the algebraic mapping between coil currents and geometrical descriptors. These principal directions are then assumed as controller outputs so as the original multivariable (rectangular) control problem is transformed in a standard square problem. Moreover the singular value decomposition approach, makes it possible to solve this modified problem by means of a number of SISO problems. Each of this SISO control problem has been solved designing a suitable PID controller.
Download (.pdf) ITER vertical stabilization system
published in 'Fusion Engineering and Design', 2009
As part of the ITER design review, a reassessment of the specifications underlying the design of the vertical... more
As part of the ITER design review, a reassessment of the specifications underlying the design of the vertical stabilization system (VS) was performed.
Recent results from experiments, aimed at the evaluation of the feasibility of the ITER reference scenarios, have raised several concerns regarding mostly the ramp-up and ramp-down phases of the pulse. The main issue is the value of the internal inductance li which may reach values outside the range 0.7–1, considered as reference for the ITER control system design. Similar concerns apply to the low current L-mode plasmas, needed to the exploitation of the machine towards the development of the 15 MA pulse. The performance of the reference vertical stabilization system, under the revised conditions may be marginal, in particular if the effect of plasma generated noise on the velocity measurement is considered.
A reliable and robust VS is mandatory to guarantee the operation of ITER at the reference elongation and plasma current values. To avoid de-scoping of the machine mission, several solutions have been proposed to improve the VS performances, ranging from an upgrade of the maximum voltage available to the present external coils system, to the introduction of in-vessel passive and/or active conductors.
The paper presents an overview of the modelling and experimental effort aimed at the assessment of the baseline ITER VS and analyses the proposed solutions to improve the system performance.
Design, implementation and test of the XSC extreme shape controller in JET
published in 'Fusion Engineering and Design', 2005
A new model-based plasma current and shape controller has been set up and tested on the JET Tokamak with the existing... more A new model-based plasma current and shape controller has been set up and tested on the JET Tokamak with the existing active circuits and control. The installation has been carried out without causing any interference to the plasma operation and without requiring a long commissioning time. Eventually, the new controller was used on really extremely shaped internal transport barrier experiments at high poloidal beta and in the presence of quite large variations of the plasma current density profile (variation range Δβpol up to 1.5 and Δli up to 0.5). The extreme shape controller (XSC) controller architecture and philosophy also offer new interesting opportunities, e.g., the separatrix sweeping on the divertor plates without significantly affecting the overall plasma shape, and the possibility of improving the overall tokamak performance via combined control of plasma shape, current and profile. The adopted methodology constitutes also an important test bed for feedback control strategies of ITER relevance.
Modeling Tools for the ITER Central Interlock System
published in 'Fusion Engineering and Design', 2011
The ITER Central Interlock System (CIS) executes manual and automatic interlocks in order to protect tokamak’s... more The ITER Central Interlock System (CIS) executes manual and automatic interlocks in order to protect tokamak’s investments. When a new control system, such as the CIS, has to be deployed, it is strongly recommended to use modeling and simulation tools, in order to test and validate the control logics. This paper presents the software tools that have been recently developed to perform the modeling of the ITER CIS. In particular, the architectures of both a CIS oriented plant simulator and of a CIS logical prototype are presented, together with the simulation of two test cases.
Performance assessment of a dynamic current allocator for the JET eXtreme Shape Controller
published in 'Fusion Enginering and Design', 2011
This paper reports on a recently proposed dynamic allocation technique that can be effectively adopted to handle the... more This paper reports on a recently proposed dynamic allocation technique that can be effectively adopted to handle the current saturations of the Poloidal Field coils with the eXtreme Shape Controller. The proposed approach allows to automatically relax the plasma shape regulation when the reference shape requires current levels out of the available ranges, finding in real-time an optimal trade-off between shape control precision and currents saturation avoidance. In this paper the results attained during preliminary analysis are presented, showing the advantage arising from the use of the dynamic allocator, versus the bare use of the eXtreme Shape Controller.
Nonlinear dynamic allocator for optimal input/output performance trade-off: application to the JET Tokamak shape controller
published in 'Automatica', 2011
In this paper, we revisit the dynamic allocation scheme for input redundant plants proposed in Zaccarian (2009).... more In this paper, we revisit the dynamic allocation scheme for input redundant plants proposed in Zaccarian (2009). within a set-point regulation setting and propose generalizations that apply to cases where the plant under consideration is not input redundant but the control specifications allow us to modify the set-point regulation within certain bounds. In this case, the input allocator is extended to the nonredundant case by allowing it to improve the input allocation selection at the price of some output regulation error. We also establish here desirable convergence properties of the allocator, which were only addressed for the linear case in Zaccarian (2009). In particular, we design the allocator dynamics as the gradient of a cost function and establish its convergence to the minimum of the cost function under some mild conditions. The allocation scheme is applied to the JET tokamak shape controller by illustrating its capabilities to enforce coil currents selections that tend to move away from the saturation limits within the allowable degrees of freedom.
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Current, position, and shape control in tokamaks
published in 'Fusion Science and Technology', 2011
The need to achieve increasingly better performance in present and future tokamak devices has made plasma control... more The need to achieve increasingly better performance in present and future tokamak devices has made plasma control increasingly important in tokamak engineering. When high performance and robustness are required, it is essential to adopt a model-based approach to design a control system. We introduce the basics of plasma current, position, and shape control in tokamaks. As an example, the plasma magnetic control systems of the JET tokamak is presented, together with an approach proposed for plasma axisymmetric magnetic control at the ITER tokamak.
Plasma Vertical Stabilization in the ITER Tokamak via Constrained Static Output Feedback
published in 'IEEE Transactions on Control Systems Technology', 2011
The international thermonuclear experimental reactor (ITER) is the next step toward the realization of... more The international thermonuclear experimental reactor (ITER) is the next step toward the realization of electricity-producing fusion power plants. ITER has been designed so as to reach the plasma burning condition, and to operate with high elongated unstable plasmas. However, due to the constraints which affect the machine realization, these open-loop unstable high performance plasmas can be hardly stabilized using the Poloidal Field coils placed outside the tokamak vessel. In this paper, it is proposed to use the in-vessel coils to vertically stabilize the plasma. In particular a constraint on the current in the in-vessel coils is explicitly taken into account in the design procedure, so as to yield a controller which can operate safely. Furthermore, the proposed controller is designed as a static-output feedback, in order to obtain a simple structure which permits to envisage effective adaptive algorithms, as it is usually required in operating tokamaks. The sufficient condition provided to design the controller is expressed in terms of a Bilinear Matrix Inequalities feasibility problem. The effectiveness of the approach is shown in a study case.
Rapid Prototyping of Safety System for Nuclear Risks of the ITER Tokamak
published in 'IEEE Transactions on Plasma Science', 2010
The ITER tokamak (Latin for “the way”) is the next step toward the realization of electricity-producing fusion power... more The ITER tokamak (Latin for “the way”) is the next step toward the realization of electricity-producing fusion power plants, since it has been designed to reach the plasma burning condition. The Central Safety System for Nuclear Risk (CSS) is the control system in charge to assure nuclear safety for the ITER plant, the personnel, and the environment. Since the CSS is a critical safety system, its validation and commissioning play an important role, and the required level of reliability must be demonstrated. In such a scenario, it is strongly recommended to use modeling and simulation tools since the early design phase. Indeed, mathematical models will help in the definition of the control system requirements. These models can be used for the rapid prototyping of the safety system, and hardware-in-the-loop (HIL) simulations can be performed to assess the performance of the control hardware against a plant simulator. This paper introduces the methodology and the software/hardware architecture used to develop both a CSS prototype and a ITER plant model suitable for the test and validation of this prototype.
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