A. Portone

Principal physics developments evaluated in the ITER design review

As part of the ITER Design Review and in response to the issues identified by the Science and Technology Advisory Committee, the ITER physics requirements were reviewed and as appropriate updated. The focus of this paper will be on recent work affecting the ITER design with special emphasis on topics affecting near-term procurement arrangements. This paper will describe results on: design sensitivity studies, poloidal field coil requirements, vertical stability, effect of toroidal field ripple on thermal confinement, material choice and heat load requirements for plasma-facing components, edge localized modes control, resistive wall mode control, disruptions and disruption mitigation.

Experimental vertical stability studies for ITER performance and design guidance

United States Department of Energy (DE-FC02-04ER54698, DEAC52- 07NA27344, and DE-FG02-04ER54235)

Linearly perturbed MHD equilibria and 3D eddy current coupling via the control surface method

In this paper, a coupling strategy based on the control surface concept is used to self-consistently couple linear MHD solvers to 3D codes for the eddy current computation of eddy currents in the metallic structures surrounding the plasma. The coupling is performed by assuming that the plasma inertia (and, with it, all Alfven wave-like phenomena) can be neglected on the time scale of interest, which is dictated by the relevant electromagnetic time of the metallic structures. As is shown, plasma coupling with the metallic structures results in perturbations to the inductance matrix operator. In particular, by adopting the Fourier decomposition in poloidal and toroidal modes, it turns out that each toroidal mode can be associated with a matrix (additively) perturbing the inductance matrix that commonly describes the magnetic coupling of currents in vacuum. In this way, the treatment of resistive wall modes instabilities of various toroidal mode numbers and their possible cross-talk through the currents induced in the metallic structures can be easily studied.

Development of ITER 15 MA ELMy H-mode inductive scenario

The poloidal field (PF) coil system on ITER, which provides both feedforward and feedback control of plasma position, shape, and current, is a critical element for achieving mission performance. Analysis of PF capabilities has focused on the 15 MA Q = 10 scenario with a 300-500 s flattop burn phase. The operating space available for the 15 MA ELMy H-mode plasma discharges in ITER and upgrades to the PF coils or associated systems to establish confidence that ITER mission objectives can be reached have been identified. Time dependent self-consistent free-boundary calculations were performed to examine the impact of plasma variability, discharge programming, and plasma disturbances. Based on these calculations a new reference scenario was developed based upon a large bore initial plasma, early divertor transition, low level heating in L-mode, and a late H-mode onset. Equilibrium analyses for this scenario indicate that the original PF coil limitations do not allow low li (<0.8) operation or lower flux states, and the flattop burn durations were predicted to be less than the desired 400 s. This finding motivates the expansion of the operating space, considering several upgrade options to the PF coils. Analysis was also carried out to examine the feedback current reserve required in the CS and PF coils during a series of disturbances and a feasibility assessment of the 17 MA scenario was undertaken. Results of the studies show that the new scenario and modified PF system will allow a wide range of 15 MA 300-500 s operation and more limited but finite 17 MA operation.

Development of the EFDA Dipole High Field Conductor

Following the outcome of the conceptual design phase the EFDA dipole magnet will be made of rectangular cable-in-conduit conductors (CICC) jacketed in 316LN. In order to optimize the required amount of superconductor two different conductor types are used: a high-field (HF) conductor consisting of 144 strands and a low-field (LF) conductor with 108 strands. A high strand with a critical current density (at 4.2 K and 12 T) and an effective filament diameter of was selected. The first series of conductor prototype specimens was tested in summer 2006 but the conductor performances were lower than expected from the pre-prototype tests of 2005 and not fulfilling the design criteria. The conductor layouts were modified to increase the strand support inside the cable and the revised HF conductor design was qualified successfully end of 2006. A current sharing temperature 6 K was found at the dipole operating conditions (12.8 T, 17 kA) confirming the required temperature margin of more than 1 K. The HF conductor qualification process including the design modifications, analysis of the test results and comparison to the expectations are discussed.

Experimental studies of ITER demonstration discharges

Key parts of the ITER scenarios are determined by the capability of the proposed poloidal field (PF) coil set. They include the plasma breakdown at low loop voltage, the current rise phase, the performance during the flat top (FT) phase and a ramp down of the plasma. The ITER discharge evolution has been verified in dedicated experiments. New data are obtained from C-Mod, ASDEX Upgrade, DIII-D, JT-60U and JET. Results show that breakdown for Eaxis < 0.23–0.33 V m−1 is possible unassisted (ohmic) for large devices like JET and attainable in devices with a capability of using ECRH assist. For the current ramp up, good control of the plasma inductance is obtained using a full bore plasma shape with early X-point formation. This allows optimization of the flux usage from the PF set. Additional heating keeps li(3) < 0.85 during the ramp up to q95 = 3. A rise phase with an H-mode transition is capable of achieving li(3) < 0.7 at the start of the FT. Operation of the H-mode reference scenario at q95 ~ 3 and the hybrid scenario at q95 = 4–4.5 during the FT phase is documented, providing data for the li (3) evolution after the H-mode transition and the li (3) evolution after a back-transition to L-mode. During the ITER ramp down it is important to remain diverted and to reduce the elongation. The inductance could be kept ≤1.2 during the first half of the current decay, using a slow Ip ramp down, but still consuming flux from the transformer. Alternatively, the discharges can be kept in H-mode during most of the ramp down, requiring significant amounts of additional heating.

Design and Procurement of the European Dipole (EDIPO) Superconducting Magnet

A 12.5 T superconducting dipole magnet (European DIPOle, EDIPO) has been designed by EFDA and it is now being procured within the framework of the European Fusion Programme in order to be installed in CRPP-PSI. This saddle-shaped magnet is designed to reach 12.5 T in a 100 times 150 mm rectangular bore over a length of about 1.5 m in order to test full size conductor samples that shall be produced during the ITER magnets procurement. The magnet uses Cable In Conduit Conductor (CICC) technology and the cables are made of high Jc (about 2300 A/mm2 at 4.2 K, 12 T) superconducting strands. In this paper the main magnet parameters are given together with the key supporting electromagnetic, mechanical and thermal analyses. An update on the general status of the procurement of the strand, conductors, dipole magnet and facility is also given together with the key results of the on-going supporting R&D.

Plasma current and shape control in tokamaks using H/sub /spl infin// and /spl mu/-synthesis

This paper is focused on the plasma shape and current control in a large, next generation tokamak characterized by long-duration plasma discharges. The new features of these fusion devices are described and their impact on the feedback control requirements are investigated. Many of the controller design issues are addressed in the H/sub /spl infin////spl mu/ synthesis framework, including the model reduction, the power constraints, the voltage limits and plant uncertainties. The application to the case of the International Thermonuclear Experimental Reactor (ITER) shows the effectiveness of this approach for the design of plasma controllers.

A control scheme to deal with coil current saturation in a Tokamak

We present a new approach for the design of a plasma current and shape feedback control system in a Tokamak. This approach takes explicitly into account the problem of current saturation in the coils which are used for the control action. The proposed design technique is essentially aimed at enlarging the operating envelope in which the Tokamak can operate without current saturation. The proposed controller structure consists of two feedback loops: one is a standard linear time invariant controller designed with the H/sub /spl infin// technique, which operates in the absence of current saturation; the second gracefully degrades the system performance when the coil currents approach their saturation limits. Some stability considerations on the overall control system are given, and a case study is presented.

Plasma control in ITER

This paper describes the ITER tokamak from the standpoint of operation, with emphasis on the plasma control system. It discusses the plasma operation scenario sequence that is carried out in feedforward control mode, and presents the overall feedback control system, which can be roughly separated into two weakly coupled control loops. Details about each control loop, the magnetic and kinetic controller, are also given.