Search

This paper presents the breakdown studies carried out in the framework of JET Enhancement Projects for Plasma Control Upgrade (PCU) and Enhanced Radial Field Amplifier (ERFA), to obtain plasma formation with different sets of coil turns in the radial field circuit. The electromagnetic conditions to reach the plasma breakdown in the JET machine are strongly dependent on the properties of JET iron core and the effects of the eddy currents driven by the transient electric field on the present passive structures. The study has been carried out by using a linearized dynamic model of JET provided by 2D axisymmetric finite element code CREATE-L [R. Albanese, et al., Nucl. Fusion 38 (1998) 723–738]. The dynamic simulations have been compared with the experimental data. A new fast visible camera has been installed and has been used for the first time at JET for studies of plasma breakdown. The images show, coherently with the model, that the avalanche evolves dynamically towards a region where the stray field is perpendicular to the first wall.

The Current Limit Avoidance System (CLA) has been recently deployed at the JET tokamak to avoid current saturations in the poloidal field (PF) coils when the eXtreme Shape Controller is used to control the plasma shape. In order to cope with the current saturation limits, the CLA exploits the redundancy of the PF coils system to automatically obtain almost the same plasma shape using a different combination of currents in the PF coils. In the presence of disturbances it tries to avoid the current saturations by relaxing the constraints on the plasma shape control. The CLA system has been successfully implemented on the JET tokamak and fully commissioned in 2011. This paper presents the first experimental results achieved in 2011–2012 during the restart and the ITER-like wall campaigns at JET.

The recent installation of a full metal, ITER-like, first wall provided the opportunity to study the impact of the plasma-facing materials on plasma initiation or breakdown. This study for the first time presents a full experimental characterisation of tokamak breakdown at JET, using all discharges since 2008, covering both operations with a main chamber carbon and a beryllium ITER-like main chamber wall. It was found that the avalanche phase was unaffected by the change in wall material. However, changes in out-gassing by the wall and lower carbon levels resulted in better controlled density and significantly lower radiation during the burn-through phase with the ITER-like wall. Breakdown failures, that usually developed with a carbon wall during the burn-through phase (especially after disruptions) were absent with the ITER-like wall. These observations match with the results obtained from a new model of plasma burn-through that includes plasma–surface interactions (Kim et al 2012 Nucl. Fusion 52 [http://dx.doi.org/10.1088/0029-5515/52/10/103016] 103016 ). This shows that chemical sputtering of carbon is the determining factor for the impurity content, and hence also radiation, during the burn-through phase for operations with a carbon wall. As seen experimentally, with a beryllium main wall, the plasma surface effects predicted by the model do not raise the radiation levels much above those expected for pure deuterium plasmas. With the ITER-like wall, operation with higher pre-fill pressures, and thus higher breakdown densities, was possible, which helped maintaining the density after breakdown.

The vertical stabilization system of the JET tokamak has been recently upgraded. This new system enables a more sensitive control of the plasma geometry and can withstand larger perturbations, enabling to push the plasma performance to its limits without risking a severe control loss, which might endanger the machine integrity. The project was successfully delivered in the course of 2009. This paper introduces the architecture of the new JET vertical stabilization system, discussing how its modularity enabled the system to provide different experimental features in several operational environments. Furthermore, some of the major achievements of the commissioning activity and of the regular operation during the 2008/2009 experimental campaigns are presented.

Experiments and modelling have been carried out to optimize the magnetic field null during breakdown at JET. Such optimization may prove to be essential for reliable plasma initiation at low voltages, e.g. in ITER where the value of the electric field available will be limited to 0.33 V m−1. A two-dimensional FEM electromagnetic model has been employed to predict the stray field configuration during JET breakdown. This model includes the active poloidal field circuits, a description of the passive structure and the JET magnetic circuit. In particular, the model includes the gap at the top of the iron circuit (but not at the bottom), which introduces a perturbing field, with radial and vertical components, not previously considered. A number of experiments were run using the optimized magnetic null configuration, allowing one to achieve a more robust breakdown at a low electric field. The model calculation to adjust the magnetic null position was validated using the recordings from a fast visible camera. The optimized position and dynamics of the plasma start lead to a smoother behaviour of the JET radial field control system, far from the amplifier limits. Finally, an important indication was obtained on the precision needed for the active current measurements during the low electric field breakdown relevant from the perspective of the ITER real-time acquisition system.

Recently a new recipe for developing and deploying real-time systems has become increasingly adopted in the JET tokamak. Powered by the advent of x86 multi-core technology and the reliability of the JET’s well established Real-Time Data Network (RTDN) to handle all real-time I/O, an official Linux vanilla kernel has been demonstrated to be able to provide realtime performance to user-space applications that are required to meet stringent timing constraints. In particular, a careful rearrangement of the Interrupt ReQuests’ (IRQs) affinities together with the kernel’s CPU isolation mechanism allows to obtain either soft or hard real-time behavior depending on the synchronization mechanism adopted. Finally, the Multithreaded Application Real-Time executor (MARTe) framework is used for building applications particularly optimised for exploring multicore architectures. In the past year, four new systems based on this philosophy have been installed and are now part of the JET’s routine operation. The focus of the present work is on the configuration and interconnection of the ingredients that enable these new systems’ real-time capability and on the impact that JET’s distributed real-time architecture has on system engineering requirements, such as algorithm testing and plant commissioning. Details are given about the common real-time configuration and development path of these systems, followed by a brief description of each system together with results regarding their real-time performance. A cycle time jitter analysis of a user-space MARTe based application synchronising over a network is also presented. The goal is to compare its deterministic performance while running on a vanilla and on a Messaging Real time Grid (MRG) Linux kernel.

The JET enhancement project Plasma Control Upgrade (PCU) aimed at increasing the capabilities of the plasma vertical stabilization (VS) system. One of the activities of this project was devoted to the development of simple but sufficiently accurate models of the VS system so as to address the main design choices, use the simulation tools as reliable test-beds, and provide an adequate support to the engineering design and commissioning of the new Enhanced Radial Field Amplifier (ERFA). This paper illustrates some of the main achievements of the modelling activity, which gave rise to a closed loop model of the VS system, including plasma, PF coils and passive structures. In particular the paper deals with the selection of the set of turns to be used in the control coils and with the estimation of the eddy current effects on the VS system. The latter analysis addressed an upgrade of the converter units of ERFA, successfully implemented during its commissioning on plasma in August 2009.

A project dedicated to the enhancement of the JET vertical stabilization system was launched in 2006, including an upgrade of the Power Supply of the Radial Field Amplifier, of hardware and software of the vertical stabilization control system. The main aim was to double the JET capability in stabilising high current plasmas when subject to perturbations, in particular large Edge Localised Modes. We present here the results of first plasma operation with the new Enhanced Radial Field Amplifier and its data acquisition and control system, focussing on the benefits of an approach based on phased commissioning, modelling and offline algorithm validation. (C) 2011 EURATOM. Published by Elsevier B.V. All rights reserved.

In order to prepare adequate current ramp-up and ramp-down scenarios for ITER, present experiments from various tokamaks have been analysed by means of integrated modelling in view of determining relevant heat transport models for these operation phases. A set of empirical heat transport models for L-mode (namely, the Bohm-gyroBohm model and scaling based models with a specific fixed radial shape and energy confinement time factors of H96-L = 0.6 or HIPB98 = 0.4) has been validated on a multi-machine experimental dataset for predicting the li dynamics within ±0.15 accuracy during current ramp-up and ramp-down phases. Simulations using the Coppi-Tang or GLF23 models (applied up to the LCFS) overestimate or underestimate the internal inductance beyond this accuracy (more than ±0.2 discrepancy in some cases). The most accurate heat transport models are then applied to projections to ITER current ramp-up, focusing on the baseline inductive scenario (main heating plateau current of Ip = 15 MA). These projections include a sensitivity study to various assumptions of the simulation. While the heat transport model is at the heart of such simulations (because of the intrinsic dependence of the plasma resistivity on electron temperature, among other parameters), more comprehensive simulations are required to test all operational aspects of the current ramp-up and ramp-down phases of ITER scenarios. Recent examples of such simulations, involving coupled core transport codes, free-boundary equilibrium solvers and a poloidal field (PF) systems controller are also described, focusing on ITER current ramp-down. © 2011 IAEA, Vienna.

Tokamaks are the most promising approach for nuclear fusion on earth. They are toroidal machines where the plasma is heated in a ring-shaped vessel and kept away from the vessel by applied magnetic fields. To achieve high performance in tokamaks, plasmas with elongated poloidal cross-section are needed. Such elongated plasmas are vertically unstable, hence position control on a fast time scale is clearly an essential feature for all tokamak devices. In this context the Plasma Control Upgrade project was aimed at increasing the capabilities of the Vertical Stabilization (VS) of the JET tokamak. This paper introduces the new JET VS system and focuses on how the flexibility of this real-time system has been exploited to enlarge its operational limits in terms of maximum controllable disturbance. Eventually, some experimental results achieved during the last experimental campaigns are presented.

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 Ip^4. As a consequence, the achievement of the ITER objective of Q = 10 requires the machine to be routinely operated at a nominal current Ip of 15 MA, and at full toroidal field BT of 5.3 T. 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 17 MA 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 15 MA plasma in ITER is likely to be adequate in the range of li 0.7–0.9 whereas, for a 17 MA 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.

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 E axis < 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 l i(3) < 0.85 during the ramp up to q 95 = 3. A rise phase with an H-mode transition is capable of achieving l i(3) < 0.7 at the start of the FT. Operation of the H-mode reference scenario at q 95 ∼ 3 and the hybrid scenario at q 95 = 4–4.5 during the FT phase is documented, providing data for the l i (3) evolution after the H-mode transition and the l i (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 I p 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.