D. Valcarcel

A protection system for the JET ITER-like wall based on imaging diagnostics

The new JET ITER-like wall (made of beryllium and tungsten) is more fragile than the former carbon fiber composite wall and requires active protection to prevent excessive heat loads on the plasma facing components (PFC). Analog CCD cameras operating in the near infrared wavelength are used to measure surface temperature of the PFCs. Region of interest (ROI) analysis is performed in real time and the maximum temperature measured in each ROI is sent to the vessel thermal map. The protection of the ITER-like wall system started in October 2011 and has already successfully led to a safe landing of the plasma when hot spots were observed on the Be main chamber PFCs. Divertor protection is more of a challenge due to dust deposits that often generate false hot spots. In this contribution we describe the camera, data capture and real time processing systems. We discuss the calibration strategy for the temperature measurements with cross validation with thermal IR cameras and bi-color pyrometers. Most importantly, we demonstrate that a protection system based on CCD cameras can work and show examples of hot spot detections that stop the plasma pulse. The limits of such a design and the associated constraints on the operations are also presented.

Engineering design of ITER prototype Fast Plant System Controller

The ITER control, data access and communication (CODAC) design team identified the need for two types of plant systems. A slow control plant system is based on industrial automation technology with maximum sampling rates below 100 Hz, and a fast control plant system is based on embedded technology with higher sampling rates and more stringent real-time requirements than that required for slow controllers. The latter is applicable to diagnostics and plant systems in closed-control loops whose cycle times are below 1 ms. Fast controllers will be dedicated industrial controllers with the ability to supervise other fast and/or slow controllers, interface to actuators and sensors and, if necessary, high performance networks. Two prototypes of a fast plant system controller specialized for data acquisition and constrained by ITER technological choices are being built using two different form factors. This prototyping activity contributes to the Plant Control Design Handbook effort of standardization, specifically regarding fast controller characteristics. Envisaging a general purpose fast controller design, diagnostic use cases with specific requirements were analyzed and will be presented along with the interface with CODAC and sensors. The requirements and constraints that real-time plasma control imposes on the design were also taken into consideration. Functional specifications and technology neutral architecture, together with its implications on the engineering design, were considered. The detailed engineering design compliant with ITER standards was performed and will be discussed in detail. Emphasis will be given to the integration of the controller in the standard CODAC environment. Requirements for the EPICS IOC providing the interface to the outside world, the prototype decisions on form factor, real-time operating system, and high-performance networks will also be discussed, as well as the requirements for data streaming to CODAC for visualization and archiving.

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.

Fast digital link for a tokamak current source control

The need for real-time plasma position control on the ISTTOK tokamak led to the development of two similar 240 A:500 kA/s fast power supplies, for the vertical and horizontal magnetic equilibrium fields. This paper presents the design, the digital link, the modeling and operation of the vertical B-field power supply. It consists of a digital hard real-time control embedded board, based on a microcontroller programmed in assembly to ensure a fault free operation. In order to avoid the magnetic noise present in tokamak operation, the communication with the power supply is optic. A robust and fast communication protocol was developed to reduce processing time. An optimized control for the vertical power supply was also developed and tested successfully. Simulation and experimental results are presented showing the efficiency of the proposed power supply.

A real-time architecture for the identification of faulty magnetic sensors in the JET tokamak

In a tokamak, the accurate estimation of the plasma boundary is essential to maximise the fusion performance and is also the first line of defence for the physical integrity of the device. In particular, the first wall components might get severely damaged if over-exposed to a high plasma thermal load. The most common approach to calculate the plasma geometry and related parameters is based in a large set of different types of magnetic sensors. Using this information, real-time plasma equilibrium codes infer a flux map and calculate the shape and geometry of the plasma boundary and its distance to a known reference (e.g. first wall). These are inputs to one or more controllers capable of acting on the shape and trajectory based in pre-defined requests. Depending on the device, the error of the estimated boundary distance must usually be less than 1 centimetre, which translates into very small errors on the magnetic measurement itself. Moreover, asymmetries in the plasma generated and surrounding magnetic fields can produce local shape deformations potentially leading to an unstable control of the plasma geometry. The JET tokamak was recently upgraded to a new and less thermally robust all-metal wall, also known as the ITER-like wall. Currently the shape controller system uses the output of a single reconstruction algorithm to drive the plasma geometry and the protection systems have no input from the plasma boundary reconstruction. These choices are historical and were due to architectural, hardware and processing power limitations. Taking advantage of new multi-core systems and of the already proved robustness of the JET real-time network, this paper proposes a distributed architecture for the real-time identification of faults in the magnetic measurements of the JET tokamak. Besides detecting simple faults, such as short-circuits and open-loops, the system compares the expected measurement at the coil location and the real measurement, producing a confidence value. Several magnetic reconstructions, using sensors from multiple toroidally distributed locations, can run in parallel, allowing for a voting or averaging scheme selection. Finally, any fault warnings can be directly fed to the real-time protection sequencer system, whose main function is to coordinate the protection of the JETs first wall.

ITER prototype fast plant system controller based on ATCA platform

The ITER Fast Plant System Controllers (FPSC) are based on embedded technologies and will be devoted to both data acquisition tasks (sampling rates >1 kSPS) and control purposes in closed-control loops whose cycle times are below 1 ms. Fast Controllers will be dedicated industrial controllers with the ability to: i) supervise other fast and/or slow controllers; ii) interface to actuators and sensors and high performance networks. This contribution presents an FPSC prototype, specialized for data acquisition, based on the ATCA (Advanced Telecommunications Computing Architecture) standard. This prototyping activity contributes to the ITER Plant Control Design Handbook (PCDH) effort of standardization, specifically regarding fast controller characteristics. For the prototype, IPFN is developing a new family of ATCA modules targeting ITER requirements. The modules comprise an AMC carrier/data hub/timing hub compliant with the upcoming ATCA extensions for Physics and a multi-channel with galvanic isolation hot-swappable digitizer designed for serviceability. The design and test of a peer-to-peer communications layer for the implementation of a reflective memory over PCI Express and the design and test of an IEEE-1588 transport layer over a high performance serial link was also performed. In this work, a complete description of the solution is presented as well as the integration of the controller into the standard CODAC environment. The most relevant results of real tests will be addressed, focusing in the benefits and limitations of the applied technologies.

Tomographic Visualization for Plasma Position Control in ISTTOK

Plasma positioning on the ISTTOK tokamak has usually been done, resorting to magnetic pickup coils. However, during alternating current (ac)-type discharges with plasma current reversal, this method has been found to be inadequate. The recently installed tomography diagnostic can be an alternative for determining the plasma position. The presented image shows tomographic reconstructions of the ISTTOK plasma during ac discharges, showing the position of maximum and average emissivity positions.

A Flexible System for the Control of External Magnetic Perturbations in the JET Tokamak

External magnetic perturbations are typically utilized in tokamak devices with two operational or experimental purposes: 1) correction of intrinsic 3-D error fields and 2) mitigation or suppression of edge localized modes (ELMs). At Joint European Torus (JET), dedicated coils are used for the generation of these toroidally asymmetric perturbations. While error fields exist even in the absence of plasma, in ELM mitigation experiments, the external fields are meant to slightly ergodize the magnetic topology in the plasma periphery hence reducing the drive for the destabilization of these instabilities. The control of the magnetic field produced by these coils is achieved by controlling the current flowing in them. The real-time system responsible for this control recently underwent a number of functional improvements since its original implementation utilizing the present voltage-controlled voltage sources. This paper describes the overall system, built-in functionality, and control algorithms and presents preliminary experimental results along with performance assessment studies. In particular, the main improvements are: 1) the possibility of automatically reducing the current references in case the plasma amplifies the applied perturbation; 2) a real-time limitation of dI/dt to reduce the electromotive force in machine protection diagnostic systems; 3) implementation of a model predictive controller as an alternative to the proportional integral derivative; and 4) the possibility of adapting the current references, in real time, using an external system. The result is a flexible control system contributing toward state-of-the-art physics research at JETs international and dynamic scientific environment.