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Abstract: 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 p ^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 p of 15MA, and at full toroidal field BT of 5.3T. 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 17MA 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 15MA plasma in ITER is likely to be adequate in the range of li 0.7–0.9 whereas, for a 17MA 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. [Copyright &y& Elsevier]

• Preliminary results on the magnetic control of the plasma shape on DEMO. • Current, position and shape control performed at the same time. • Controller performance tested in the presence of an H-L transition. About 85% of the primary energy is currently obtained from fossil sources. In the next future, nuclear fusion can significantly contribute to energy production: the fuel is in principle unlimited and radioactive waste are short lived. Next steps in fusion research are represented by the two tokamaks ITER, which is under construction, and DEMO, which is in its conceptual design phase. In this paper we focus on a specific aspect of DEMO design, that is indeed crucial for tokamak safe operation: plasma vertical and shape control. It is well known that a plasma with elongated cross-section exhibits a vertical instability that needs to be feedback controlled. Typically on operating tokamaks, and in ITER, this task is accomplished using in-vessel actuator coils. Since in the present DEMO design these coils are not foreseen, hereafter we assumed that all the actuator coils, located outside the vessel, are used at the same time to guarantee both vertical stabilization and shape control, resorting to a suitable geometric decoupling. The performance of the controller is shown in simulation using a nonlinear evolution code during a plasma H-L back-transition. [ABSTRACT FROM AUTHOR]

This paper gives an overview of the scope and context of the CODAC high-level real-time applications (Supervision and Plasma Control) and presents the strategy and current state of design of the tools to support the implementation. A real-time framework, which is currently under development with strong support of the worldwide fusion community will not only support the implementation of plasma control strategies with the extensive exception handling and forecasting functionality foreseen for ITER, but also integrated commissioning, orchestration and supervision as well as the real-time needs of ITER plant system developers. A second cornerstone in the implementation strategy is the development of a powerful simulation environment (Plasma Control System Simulation Platform – PCSSP) to design and verify control strategies, event handling and orchestration and automation. The development of PCSSP is currently under contract and this paper will also give an overview of its current state of development. [ABSTRACT FROM AUTHOR]

Abstract: ITER is targeting Q =10 with 500MW 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 plasma boundary), and the strong component of eddy current at the start-up (resulting in a poor S/N ratio for plasma reconstruction for I p <2MA requiring a robust plasma control) make the ITER magnetic diagnostic system a demanding part of the magnetic control and investment protection systems. Finally the paper will illustrate the identified roles of magnetic control in the PCS (plasma control system) as formally defined in the recent first step of the design and development of the system. [Copyright &y& Elsevier]

Highlights: [•] Describes task to develop simulation tool to aid development/testing of ITER PCS. [•] Requirements and use cases and preliminary architecture have been delivered. [•] Detailed design is now being developed. [•] Provides overview of use cases and requirements. [•] Provides overview of architecture and status of development. [ABSTRACT FROM AUTHOR]

Abstract: The Fusion Advanced Studies Torus (FAST) conceptual study has been proposed [A. Pizzuto on behalf of the Italian Association, The Fusion Advanced Studies Torus (FAST): a proposal for an ITER Satellite facility in support of the development of fusion energy, in: Proceedings of 22nd IAEA Fusion Energy Conference, Geneva, Switzerland, October 13–18, 2008; Nucl. Fusion, submitted for publication] as possible European ITER Satellite facility with the aim of preparing ITER operation scenarios and helping DEMO design and R&D. Insights into ITER regimes of operation in deuterium plasmas can be obtained from investigations of non linear dynamics that are relevant for the understanding of alpha particle behaviours in burning plasmas by using fast ions accelerated by heating and current drive systems. FAST equilibrium configurations have been designed in order to reproduce those of ITER with scaled plasma current, but still suitable to fulfil plasma conditions for studying burning plasma physics issues in an integrated framework. In this paper we report the plasma scenarios that can be studied on FAST, with emphasis on the aspect of its flexibility in terms of both performance and physics that can be investigated. All plasma equilibria satisfy the following constraints: (a) minimum distance of 3 energy e-folding length (assumed to be 1cm on the equatorial plane) between plasma and first wall to avoid interaction between plasma and main chamber; (b) maximum current density in the poloidal field coils, transiently, up to around 30MA/m2. The discharge duration is always limited by the heating of the toroidal field coils that are inertially cooled by helium gas at 30K. The location of the poloidal field coils has been optimized in order to: minimize the magnetic energy; produce enough magnetic flux (up to 35Wb stored) for the formation and sustainment of each scenario; produce a good field null at the plasma break-down (B P/B T <2×10−4 at low field, i.e. B T =4T and E T =2V/m for at least 40ms). Plasma position and shape control studies will also be presented. The optimization of the passive shell position slows the vertical stability growth time down to 100ms. [Copyright &y& Elsevier]

Abstract: 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 15MA 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. [Copyright &y& Elsevier]