I. S. Carvalho

The COMPASS tokamak plasma control software performance

The COMPASS tokamak has began operation at the IPP Prague in December 2008. A new control system has been built using an ATCA-based real-time system developed at IST Lisbon. The control software is implemented on top of the MARTe real-time framework attaining control cycles as short as 50 μs, with a jitter of less than 1 μs. The controlled parameters, important for the plasma performance, are the plasma current, position of the plasma current center, boundary shape and horizontal and vertical velocities. These are divided in two control cycles: slow at 500 μs and fast at 50 μs. The project has two phases. First, the software implements a digital controller, similar to the analog one used during the COMPASS-D operation in Culham. In the slow cycle, the plasma current and position are measured and controlled with PID and feedforward controllers, respectively, the shaping magnetic field is preprogrammed. The vertical instability and horizontal equilibrium are controlled with the faster 50-μs cycle PID controllers. The second phase will implement a plasma-shape reconstruction algorithm and controller, aiming at optimized plasma performance. The system was designed to be as modular as possible by breaking the functional requirements of the control system into several independent and specialized modules. This splitting enabled tuning the execution of each system part and to use the modules in a variety of applications with different time constraints. This paper presents the design and overall performance of the COMPASS control software.

Fast feedback control for plasma positioning with a PCI hybrid DSP/FPGA board

The need to control in real-time the plasma parameters in fusion devices leads to the development of algorithms requiring intensive computation and providing results on a few hundred microseconds. The present works objective was the implementation of the current filaments method (CF) to model in real-time the ISTTOK plasma shape and position. The hardware used was the on site developed PCI-TR-256 hardware configurable module, which includes the latest technology in DSP and FPGA. The algorithm estimates the position of the plasma column and generates the control signals for the vertical magnetic field actuators. The main advantage of this system is to provide a digital approach to feedback plasma position control with similar cycle times to those of analog systems but allowing flexible, user defined, algorithms.

Sawtooth control in JET with ITER relevant low field side resonance ion cyclotron resonance heating and ITER-like wall

New experiments at JET with the ITER-like wall show for the first time that ITER-relevant low field side resonance first harmonic ion cyclotron resonance heating (ICRH) can be used to control sawteeth that have been initially lengthened by fast particles. In contrast to previous (Graves et al 2012 Nat. Commun. 3 624) high field side resonance sawtooth control experiments undertaken at JET, it is found that the sawteeth of L-mode plasmas can be controlled with less accurate alignment between the resonance layer and the sawtooth inversion radius. This advantage, as well as the discovery that sawteeth can be shortened with various antenna phasings, including dipole, indicates that ICRH is a particularly effective and versatile tool that can be used in future fusion machines for controlling sawteeth. Without sawtooth control, neoclassical tearing modes (NTMs) and locked modes were triggered at very low normalised beta. High power H-mode experiments show the extent to which ICRH can be tuned to control sawteeth and NTMs while simultaneously providing effective electron heating with improved flushing of high Z core impurities. Dedicated ICRH simulations using SELFO, SCENIC and EVE, including wide drift orbit effects, explain why sawtooth control is effective with various antenna phasings and show that the sawtooth control mechanism cannot be explained by enhancement of the magnetic shear. Hybrid kinetic-magnetohydrodynamic stability calculations using MISHKA and HAGIS unravel the optimal sawtooth control regimes in these ITER relevant plasma conditions.

ELM frequency feedback control on JET

This paper describes the first development and implementation of a closed loop edge localized mode (ELM) frequency controller using gas injection as the actuator. The controller has been extensively used in recent experiments on JET and it has proved to work well at ELM frequencies in the 15–40xa0Hz range. The controller responds effectively to a variety of disturbances, generally recovering the requested ELM frequency within approximately 500xa0ms. Controlling the ELM frequency has become of prime importance in the new JET configuration with all metal walls, where insufficient ELM frequency is associated with excessive tungsten influx. The controller has allowed successful operation near the minimum acceptable ELM frequency where the best plasma confinement can be achieved. Use of the ELM frequency controller in conjunction with pellet injection has enabled investigations of ELM triggering by pellets while maintaining the desired ELM frequency even when pellets fail to trigger ELMs.

Real-time control of ELM and sawtooth frequencies: similarities and differences

ELMs and Sawteeth, located in different parts of the plasma, are similar from a control engineering point of view. Both manifest themselves through quiescent periods interrupted by periodic collapses. For both, large collapses, following long quiescent periods, have detrimental effects while short periods are associated with decreased confinement. Following the installation of the all metal ITER like wall on JET, sawteeth and ELMs also play an important role by expelling tungsten from the core and edge of the plasma respectively. Control of tungsten has therefore been added to divertor heat load reduction, NTM avoidance and helium ash removal as reasons for requiring ELM and sawtooth control. It is therefore of interest to implement control systems to maintain the sawtooth and ELM frequencies in the desired ranges. On JET, ELM frequency control uses radial field kicks and pellet and gas injection as actuators, while sawtooth control uses ion cyclotron resonance heating (ICRH). JET experiments have, for the first time, established feedback control of the ELM frequency, via real time variation of the injected gas flow [1]. Using this controller in conjunction with pellet injection allows the ELM frequency to be kept as required despite variations in pellet ELM triggering efficiency. JET Sawtooth control experiments have, for the first time, demonstrated that low field side ICRH, as foreseen for ITER, can shorten sawteeth lengthened by central fast ions [2]. The development of ELM and sawtooth control could be key to achieve stable high performance JET discharges with minimal tungsten content. Integrating such schemes into an overall control strategy will be required in future tokamaks and gaining experience on current tokamaks is essential.

Plasma isotopic changeover experiments in JET under carbon and ITER-like wall conditions

In JET-ILW isotopic plasma wall changeover experiments have been carried out to determine the amount of particles accessible by changing the plasma from H to D and from D to H. The gas balance analysis integrated over the experimental sessions show that the total amount of H or D removed from the wall is in the range of (1–3) × 1022D. For both changeover experiments, the respective plasma isotopic ratio behaviour is exactly the same as a function of the pulse number. After only 80 s of plasma (4 pulses), the plasma isotopic ratio is lower than 10%, below 4.5% after 13 pulses and then saturates around ~2–3%. In these conditions, the removal efficiency through plasma operation becomes very poor. The saturation of the plasma isotopic ratio in the range of 10% is also observed for the JET-C configuration although the amount of tritium retained in the vessel after the DT pulses was more than one order of magnitude compared to the retention observed with the JET-ILW. This demonstrates that the amount of particle recovery through plasma changeover is independent from the long term retention. Since this long term reservoir results from codeposition, these experiments suggest that there is a limited access to these codeposited particles by plasma isotopic changeover. Finally, in ITER, change over from D/T to H at the end of the discharge for possibly reducing the long term retention does not appear as a good strategy.

ITER-Like Tokamak Exhaust Gases in JET Active Gas Handling System: Process Optioneering

Abstract The Active Gas Handling System (AGHS) collects Tokamak exhaust gases from the JET machine and recovers, purifies and recycles the deuterium and tritium for fuelling the plasma. With the increasing fusion power and all-metal first wall and diverter, the forthcoming DTE2 experiment will see a change in the Tokamak exhaust composition compared to the DTE1 first series of experiments. A range of gases additional to those for fuelling the plasma will be added for critical applications such as plasma detachment and disruption mitigation. One of these candidate gases is neon, which is anticipated to have a negative influence on storing the recovered hydrogen at an early stage of AGHS reprocessing. This paper will outline the trials and plant modifications which are in hand to provide a solution and enable downstream processes to operate as during DTE1. This will comprise a scrubbing circuit to mitigate the blanketing action of the neon in sorbing the hydrogen isotopes onto the existing depleted-uranium metal-hydride storage beds. A second approach is also under study that has the potential to bring the tritium process circuit into closer alignment to that intended by ITER, thus providing process data for supporting the ITER fuel cycle design and increasing further the value of the JET experiments to the ITER project. JET’s torus gases recovered onto AGHS’s ITER Prototype Cryosorption Pump (PCP) will, on regeneration, be representative of the gas compositions received into ITER’s Tokamak Exhaust Processing (TEP) system front end. The following adaptions of AGHS are being considered to make processing more representative of ITER: 1. Helium, neon and hydrogen isotopes from 80 K regenerations of the PCP could be routed directly to a Pd-Ag permeator that will replicate the first stage of TEP hydrogen processing. 2. Other gases liberated from regenerating the PCP at 130 K (so-called “warm” regenerations) could be cycled around a nickel bed and permeator train that will approximate to the operation of a palladium membrane reactor which is a second processing route within the TEP system.

JET diagnostic enhancements testing and commissioning in preparation for DT scientific campaigns

In order to optimize the scientific exploitation of JET (Joint European Torus) during the upcoming deuterium-tritium experiments, a set of diagnostic systems is being enhanced. These upgrades focus mainly on the experimental and operational conditions expected during tritium campaigns. It should be stressed that measurements relevant for burning plasmas are specifically targeted. Previously non-available capabilities, such as a current measurement system fully covering all poloidal field circuits, are described in detail. Instrument descriptions, performance prediction, testing, and initial commissioning results of these systems are presented.

Real-Time Vertical Plasma Position Control Using the Heavy Ion Beam Diagnostic

A future thermonuclear fusion reactor relies on the real-time control of, among others, instabilities and plasma position. At the ISTTOK tokamak, a novel approach of a real-time diagnostic system to control the vertical plasma position (in a 100-

An ATCA Embedded Data Acquisition and Control System for the Compass tokamak

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