An overview of recent results obtained at the tokamak ASDEX Upgrade (AUG) is given. A work flow for predictive profile modelling of AUG discharges was established which is able to reproduce experimental H-mode plasma profiles based on engineering parameters only. In the plasma center, theoretical predictions on plasma current redistribution by a dynamo effect were confirmed experimentally. For core transport, the stabilizing effect of fast ion distributions on turbulent transport is shown to be important to explain the core isotope effect and improves the description of hollow low-Z impurity profiles. The L–H power threshold of hydrogen plasmas is not affected by small helium admixtures and it increases continuously from the deuterium to the hydrogen level when the hydrogen concentration is raised from 0 to 100%. One focus of recent campaigns was the search for a fusion relevant integrated plasma scenario without large edge localised modes (ELMs). Results from six different ELM-free confinement regimes are compared with respect to reactor relevance: ELM suppression by magnetic perturbation coils could be attributed to toroidally asymmetric turbulent fluctuations in the vicinity of the separatrix. Stable improved confinement mode plasma phases with a detached inner divertor were obtained using a feedback control of the plasma β. The enhanced D α H-mode regime was extended to higher heating power by feedback controlled radiative cooling with argon. The quasi-coherent exhaust regime was developed into an integrated scenario at high heating power and energy confinement, with a detached divertor and without large ELMs. Small ELMs close to the separatrix lead to peeling-ballooning stability and quasi continuous power exhaust. Helium beam density fluctuation measurements confirm that transport close to the separatrix is important to achieve the different ELM-free regimes. Based on separatrix plasma parameters and interchange-drift-Alfvén turbulence, an analytic model was derived that reproduces the experimentally found important operational boundaries of the density limit and between L- and H-mode confinement. Feedback control for the X-point radiator (XPR) position was established as an important element for divertor detachment control. Stable and detached ELM-free phases with H-mode confinement quality were obtained when the XPR was moved 10 cm above the X-point. Investigations of the plasma in the future flexible snow-flake divertor of AUG by means of first SOLPS-ITER simulations with drifts activated predict beneficial detachment properties and the activation of an additional strike point by the drifts.

The aim of this paper is to present a signal processing algorithm that, applied to the raw Locked Mode signal, allows us to obtain a disruption indicator in principle exploitable on different tokamaks. A common definition of such an indicator for different machines would facilitate the development of portable systems for disruption prediction, which is becoming of increasingly importance for the next tokamak generations. Moreover, the indicator allows us to overcome some intrinsic problems in the diagnostic system such as drift and offset. The behavior of the proposed indicator as disruption predictor, based on crossing optimized thresholds of the signal amplitude, has been analyzed using data of both JET and ASDEX Upgrade experiments. A thorough analysis of the disruption prediction performance shows how the indicator is able to recover some missed and tardy detections of the raw signal. Moreover, it intervenes and corrects premature or even wrong alarms due to, e.g., drifts and/or offsets.

In this paper we validate the finite element code RAPLICASOL, which models radiofrequency wave propagation in edge plasmas near ICRF antennas, against calculations with the TOPICA code. We compare the output of both codes for the ASDEX Upgrade 2-strap antenna, and for a 4-strap WEST-like antenna. Although RAPLICASOL requires considerably fewer computational resources than TOPICA, we find that the predicted quantities of experimental interest (including reflection coefficients, coupling resistances, - and -matrix entries, optimal matching settings, and even radiofrequency electric fields) are in good agreement provided we are careful to use the same geometry in both codes.

The ASDEX Upgrade (AUG) programme, jointly run with the EUROfusion MST1 task force, continues to significantly enhance the physics base of ITER and DEMO. Here, the full tungsten wall is a key asset for extrapolating to future devices. The high overall heating power, flexible heating mix and comprehensive diagnostic set allows studies ranging from mimicking the scrape-off-layer and divertor conditions of ITER and DEMO at high density to fully non-inductive operation (q(95) = 5.5, beta(N) <= 2.8) at low density. Higher installed electron cyclotron resonance heating power <= 6 MW, new diagnostics and improved analysis techniques have further enhanced the capabilities of AUG. Stable high-density H-modes with P-sep/R <= 11 MW m(-1) with fully detached strike-points have been demonstrated. The ballooning instability close to the separatrix has been identified as a potential cause leading to the H-mode density limit and is also found to play an important role for the access to small edge-localized modes (ELMs). Density limit disruptions have been successfully avoided using a path-oriented approach to disruption handling and progress has been made in understanding the dissipation and avoidance of runaway electron beams. ELM suppression with resonant magnetic perturbations is now routinely achieved reaching transiently H-H98(y,H-2) <= 1.1. This gives new insight into the field penetration physics, in particular with respect to plasma flows. Modelling agrees well with plasma response measurements and a helically localised ballooning structure observed prior to the ELM is evidence for the changed edge stability due to the magnetic perturbations. The impact of 3D perturbations on heat load patterns and fast-ion losses have been further elaborated. Progress has also been made in understanding the ELM cycle itself. Here, new fast measurements of T-i and E-r allow for inter ELM transport analysis confirming that E-r is dominated by the diamagnetic term even for fast timescales. New analysis techniques allow detailed comparison of the ELM crash and are in good agreement with nonlinear MHD modelling. The observation of accelerated ions during the ELM crash can be seen as evidence for the reconnection during the ELM. As type-I ELMs (even mitigated) are likely not a viable operational regime in DEMO studies of 'natural' no ELM regimes have been extended. Stable I-modes up to n/n(GW) <= 0.7 have been characterised using beta-feedback. Core physics has been advanced by more detailed characterisation of the turbulence with new measurements such as the eddy tilt angle-measured for the first time-or the cross-phase angle of T-e and n(e) fluctuations. These new data put strong constraints on gyro-kinetic turbulence modelling. In addition, carefully executed studies in different main species (H, D and He) and with different heating mixes highlight the importance of the collisional energy exchange for interpreting energy confinement. A new regime with a hollow T-e profile now gives access to regimes mimicking aspects of burning plasma conditions and lead to nonlinear interactions of energetic particle modes despite the sub-Alfvenic beam energy. This will help to validate the fast-ion codes for predicting ITER and DEMO.

Ammonia formation was studied in a series of dedicated nitrogen seeded H-mode discharges at ASDEX Upgrade.The evolution of ammonia formation was investigated with a reference phase before the seeding, and a long, stable flat-top nitrogen-seeded phase.It was monitored with divertor spectroscopy and analysis of the exhaust gas.The amount of the detected ammonia increased continuously over the course of five discharges with the same nitrogen seeding rate.The same trend was observed in the nitrogen density in the core plasma, as measured by charge exchange recombination spectroscopy and other signals, linked to the effects of nitrogen seeding.The results show that the rate of ammonia formation exhibited the same trend as the nitrogen density in the plasma.This density, in turn, was strongly influenced by the nitrogen wall inventory.The spatial distribution of the detected ammonia suggests that a significant contribution to the net ammonia formation is made in plasma-shaded areas, through surface reactions of neutral species.

The ASDEX Upgrade programme is directed towards physics input to critical elements of the ITER design and the preparation of ITER operation, as well as addressing physics issues for a future DEMO design. After the finalization of the tungsten coating of the plasma facing components, the re-availability of all flywheel-generators allowed high-power operation with up to 20 MW heating power at I p up to 1.2 MA. Implementation of alternative ECRH schemes (140 GHz O2- and X3-mode) facilitated central heating above n e = 1.2 × 1020 m−3 and low q 95 operation at B t = 1.8 T. Central O2-mode heating was successfully used in high P/R discharges with 20 MW total heating power and divertor load control with nitrogen seeding. Improved energy confinement is obtained with nitrogen seeding both for type-I and type-III ELMy conditions. The main contributor is increased plasma temperature, no significant changes in the density profile have been observed. This behaviour may be explained by higher pedestal temperatures caused by ion dilution in combination with a pressure limited pedestal and hollow nitrogen profiles. Core particle transport simulations with gyrokinetic calculations have been benchmarked by dedicated discharges using variations of the ECRH deposition location. The reaction of normalized electron density gradients to variations of temperature gradients and the T e/T i ratio could be well reproduced. Doppler reflectometry studies at the L–H transition allowed the disentanglement of the interplay between the oscillatory geodesic acoustic modes, turbulent fluctuations and the mean equilibrium E × B flow in the edge negative E r well region just inside the separatrix. Improved pedestal diagnostics revealed also a refined picture of the pedestal transport in the fully developed H-mode type-I ELM cycle. Impurity ion transport turned out to be neoclassical in between ELMs. Electron and energy transport remain anomalous, but exhibit different recovery time scales after an ELM. After recovery of the pre-ELM profiles, strong fluctuations develop in the gradients of n e and T e. The occurrence of the next ELM cannot be explained by the local current diffusion time scale, since this turns out to be too short. Fast ion losses induced by shear Alfvén eigenmodes have been investigated by time-resolved energy and pitch angle measurements. This allowed the separation of the convective and diffusive loss mechanisms.

An overview of the preliminary design of the ITER plasma control system (PCS) is described here, which focusses on the needs for 1st plasma and early plasma operation in hydrogen/helium (H/He) up to a plasma current of 15 MA with moderate auxiliary heating power in low confinement mode (L-mode). Candidate control schemes for basic magnetic control, including divertor operation and kinetic control of the electron density with gas puffing and pellet injection, were developed. Commissioning of the auxiliary heating systems is included as well as support functions for stray field topology and real-time plasma boundary reconstruction. Initial exception handling schemes for faults of essential plant systems and for disruption protection were developed. The PCS architecture was also developed to be capable of handling basic control for early commissioning and the advanced control functions that will be needed for future high performance operation. A plasma control simulator is also being developed to test and validate control schemes. To handle the complexity of the ITER PCS, a systems engineering approach has been adopted with the development of a plasma control database to keep track of all control requirements.

Visualization of measurements together with experimental settings is a general subject in experiments analysis. The complex engineering design, 3D geometry, and manifold of diagnostics in larger fusion research experiments justify the development of special analysis and visualization programs. Novel ASDEX Upgrade (AUG) software tools bring together virtual navigation through 3D device models and advanced play-back and interpretation of video streams from plasma discharges. A third little tool allows the web-based platform independent observation of real-time diagnostic signals. While all three tools stem from spontaneous development ideas and are not considered mission critical for the operation of a fusion device, they with time and growing completeness shaped up as valuable helpers to visualize acquired data in fusion research. A short overview on the goals, the features, and the design as well as the operation of these tools is given in this paper. (C) 2016 Elsevier B.V. All rights reserved.

The ASDEX Upgrade Discharge Control System (DCS) is a distributed real-time control system executing complex control and monitoring tasks. Up to now, DCS control algorithms have been implemented by coding dedicated application processes with the C++ programming language. Algorithm changes required code modification, compilation and commissioning which only experienced programmers could perform. This was a significant constraint of flexibility for both control system operation and design.The new approach extends DCS with the capability of configuration-defined control algorithms. These are composed of chains of small, configurable standard function blocks providing general purpose functions like algebraic operations, filters, feedback controllers, output limiters and decision logic. In a later phase a graphical editor could help to compose and modify such configuration in a Simulink-like fashion.Building algorithms from standard functions can result in a high number of elements. In order to achieve a similar performance as with C++ coding, it is essential to avoid administrative bottlenecks by design. As a consequence, DCS executes a function block chain in the context of a single real-time thread of an application process. No concurrency issues as in a multi-threaded context need to be considered resulting in strongly simplified signal handling and zero performance overhead for inter-block communication. Instead of signal-driven synchronization, a block scheduler derives the execution sequence automatically from the block dependencies as defined in the configuration. All blocks and connecting signals are instantiated dynamically, based on definitions in a configuration file. Algorithms thus are not defined in the code but only in the configuration.The concept has been developed in view of Simulink block libraries and MARTe General Application Modules (GAM) but extends these with the DCS virtues of distributed computing and multi-threading.With growing diversity of general-purpose blocks the DCS framework will reach an unprecedented degree of universality and flexibility. Configuration-defined algorithms will gradually replace many existing DCS applications. Finally, the concept might also become of interest for the upcoming ITER plasma control system. (C) 2016 Elsevier B.V. All rights reserved.

Recent experiments on the ASDEX Upgrade tokamak aim at improving the physics base for ITER and DEMO to aid the machine design and prepare efficient operation. Type I edge localized mode (ELM) mitigation using resonant magnetic perturbations (RMPs) has been shown at low pedestal collisionality . In contrast to the previous high ν* regime, suppression only occurs in a narrow RMP spectral window, indicating a resonant process, and a concomitant confinement drop is observed due to a reduction of pedestal top density and electron temperature. Strong evidence is found for the ion heat flux to be the decisive element for the L–H power threshold. A physics based scaling of the density at which the minimum PLH occurs indicates that ITER could take advantage of it to initiate H-mode at lower density than that of the final Q = 10 operational point. Core density fluctuation measurements resolved in radius and wave number show that an increase of R/LTe introduced by off-axis electron cyclotron resonance heating (ECRH) mainly increases the large scale fluctuations. The radial variation of the fluctuation level is in agreement with simulations using the GENE code. Fast particles are shown to undergo classical slowing down in the absence of large scale magnetohydrodynamic (MHD) events and for low heating power, but show signs of anomalous radial redistribution at large heating power, consistent with a broadened off-axis neutral beam current drive current profile under these conditions. Neoclassical tearing mode (NTM) suppression experiments using electron cyclotron current drive (ECCD) with feedback controlled deposition have allowed to test several control strategies for ITER, including automated control of (3,2) and (2,1) NTMs during a single discharge. Disruption mitigation studies using massive gas injection (MGI) can show an increased fuelling efficiency with high field side injection, but a saturation of the fuelling efficiency is observed at high injected mass as needed for runaway electron suppression. Large locked modes can significantly decrease the fuelling efficiency and increase the asymmetry of radiated power during MGI mitigation. Concerning power exhaust, the partially detached ITER divertor scenario has been demonstrated at Psep/R = 10 MW m−1 in ASDEX Upgrade, with a peak time averaged target load around 5 MW m−2, well consistent with the component limits for ITER. Developing this towards DEMO, full detachment was achieved at Psep/R = 7 MW m−1 and stationary discharges with core radiation fraction of the order of DEMO requirements (70% instead of the 30% needed for ITER) were demonstrated. Finally, it remains difficult to establish the standard ITER Q = 10 scenario at low q95 = 3 in the all-tungsten (all-W) ASDEX Upgrade due to the observed poor confinement at low βN. This is mainly due to a degraded pedestal performance and hence investigations at shifting the operational point to higher βN by lowering the current have been started. At higher q95, pedestal performance can be recovered by seeding N2 as well as CD4, which is interpreted as improved pedestal stability due to the decrease of bootstrap current with increasing Zeff. Concerning advanced scenarios, the upgrade of ECRH power has allowed experiments with central ctr-ECCD to modify the q-profile in improved H-mode scenarios, showing an increase in confinement at still good MHD stability with flat elevated q-profiles at values between 1.5 and 2.

We describe a software system known as the Plasma Control System Simulation Platform (PCSSP) that is being constructed to support development of the ITER plasma control system. When mature, PCSSP will provide support for industry standard practices such as model-based controller design, simulation testing of controllers, auto-generation of controller code, hardware-and software-in-the-loop testing, use of source code management tools, and open-source methods in development of the software and control algorithms for the ITER Plasma Control System (PCS). It will also contribute to fusion specific objectives such as validation of ITER pulse schedules prior to their use in experimental operation. We also describe a more mature but less sophisticated software suite known as TokSys, which was developed at General Atomics to support plasma control development and is expected to eventually merge with PCSSP. Plasma control development activities that use or have used these software systems are also described. We will discuss the ongoing ITER PCS control algorithm development for initial plasma operation, which takes advantage of the existing PCSSP infrastructure and library of modules. Plasma control development and initial plasma operation of the EAST and KSTAR tokamaks relied heavily on the TokSys suite, which is now in routine use at EAST, KSTAR, DIII-D, and NSTX-U.

The ASDEX Upgrade Discharge Control System DCS is a modern and mature product, originally designed to regulate and supervise ASDEX Upgrade Tokamak plasma operation. In its core DCS is based on a generic, versatile real-time software framework with a plugin architecture that allows to easily combine, modify and extend control function modules in order to tailor the system to required features and let it continuously evolve with the progress of an experimental fusion device. Due to these properties other fusion experiments like the WEST project have expressed interest in adopting DCS.For this purpose, essential parts of DCS must be unpinned from the ASDEX Upgrade environment by exposure or introduction of generalised interfaces. Re-organisation of DCS modules allows distinguishing between intrinsic framework core functions and device-specific applications. In particular, DCS must be prepared for deployment in different system environments with their own realisations for user interface, pulse schedule preparation, parameter server, time and event distribution, diagnostic and actuator systems, network communication and data archiving.The article explains the principles of the revised DCS structure, derives the necessary interface definitions and describes major steps to achieve the separation between general-purpose framework and fusion device specific components. (C) 2015 Elsevier B.V. All rights reserved.

The Plasma Control System Simulation Platform (PCSSP) is a highly flexible, modular, time-dependent simulation environment developed primarily to support development of the ITER Plasma Control System (PCS). It has been under development since 2011 and is scheduled for first release to users in the ITER Organization (IC) and at selected additional sites in 2015. Modules presently implemented in PCSSP enable exploration of axisymmetric evolution and control, basic kinetic control, and tearing mode suppression. A basic capability for generation of control-relevant events is included, enabling study of exception handling in the PCS, continuous controllers, and PCS architecture. While the control design focus of PCSSP applications tends to require only a moderate level of accuracy and complexity in modules, more complex codes can be embedded or connected to access higher accuracy if needed. This paper describes the background and motivation for PCSSP, provides an overview of the capabilities, architecture, and features of PCSSP, and discusses details of the PCSSP vision and its intended goals and application. Completed work, including architectural design, prototype implementation, reference documents, and 10 demonstration of PCSSP, is summarized and example use of PCSSP is illustrated. Near-term high-level objectives are summarized and include preparation for release of an "alpha" version of PCSSP and preparation for the next development phase. High-level objectives for future work are also discussed. (C) 2015 Elsevier BAT. All rights reserved.

Concepts for the configuration of plant systems and plasma control of modern devices such as ITER and W7-X are based on global data structures, or "pulse schedules" or "experiment programs", which specify all physics characteristics (waveforms for controlled actuators and plasma quantities) and all technical characteristics of the plant systems (diagnostics and actuators operation settings) for a planned pulse.At ASDEX Upgrade we use different approach. We observed that the physics characteristics driving the discharge control system (DCS) are frequently modified on a pulse-to-pulse basis. Plant system operation, however, relies on technical standard settings, or "basic configurations" to provide guaranteed resources or services, which evolve according to longer term session or campaign operation schedules. This is why AUG manages technical configuration items separately from physics items.Consistent computation of the DCS configuration requires access to all this physics and technical data, which include the discharge programme (DP), settings of actuator systems and real-time diagnostics, the current system state and a database of static parameters. A Parameter Server provides a unified view on all these parameter sets and acts as the central point of access.We describe the functionality and architecture of the Parameter Server and its embedding into the control environment. (C) 2015 Elsevier B.V. All rights reserved.

ASDEX Upgrade is a fusion experiment with a size and complexity to allow extrapolation of technical and physical conditions and requirements to devices like ITER and even beyond. In addressing advanced physics topics it makes extensive use of sophisticated real-time control methods. It comprises real-time diagnostic integration, dynamically adaptable multivariable feedback schemes, actuator management including load distribution schemes and a powerful monitoring and pulse supervision concept based on segment scheduling and exception handling. The Discharge Control System (DCS) supplies all this functionality on base of a modular software framework architecture designed for real-time operation. It provides system-wide services like workflow management, logging and archiving, self-monitoring and inter-process communication on Linux, VxWorks and Solaris operating systems. By default DCS supports distributed computing, and a communication layer allows multi-directional signal transfer and datadriven process synchronisation over shared memory as well as over a number of real-time networks. The entire system is built following the same common design concept combining a rich set of re-usable generic but highly customisable components with a configuration-driven component deployment method.We will give an overview on the architectural concepts as well as on the outstanding capabilities of DCS in the domains of inter-process communication, generic feedback control and pulse supervision. In each of these domains, DCS has contributed important ideas and methods to the on-going design of the ITER plasma control system. We will identify and describe these essential features and illustrate them with examples from ASDEX Upgrade operation. (C) 2014 Elsevier B.V. All rights reserved.

The plasma control system simulation platform (PCSSP) for ITER shall support the analysis and development of methods to be used by the ITER plasma control system (PCS) for handling exceptions to optimize pulses and assist in machine protection. PCSSP will permit to investigate physical and technical events, such as component failures, control degradation, operation domain excess, plasma state bifurcation or instabilities, and interlock activity. Serving that purpose, the plasma, actuator, diagnostics and PCS simulation modules in PCSSP will be enhanced to compute nominal and off-normal data. Configured by an event schedule, an event generator will orchestrate the activation and manipulate the characteristics of such off-normal computation. In the simulated PCS exceptions will be handled in a pulse supervision layer operating on top of the pulse continuous control (PCC) feedback loops. It will monitor events, decide on which exceptions to respond, and compute new control references to modify PCC behavior. We discuss basic concepts for the event generation in PCSSP, and a preliminary architecture for exception handling in PCS, and show how these will be configured with event and pulse schedules.

A simulation environment known as the Plasma Control System Simulation Platform (PCSSP), specifically designed to support development of the ITER Plasma Control System (PCS), is currently under construction by an international team encompassing a cross-section of expertise in simulation and exception handling for plasma control. The proposed design addresses the challenging requirements of supporting the PCS design. This paper provides an overview of the PCSSP project and a discussion of some of the major features of its design. Plasma control for the ITER tokamak will be significantly more challenging than for existing fusion devices. An order of magnitude greater performance (e.g. [1,2]) is needed for some types of control, which together with limited actuator authority, implies that optimized individual controllers and nonlinear saturation logic are required. At the same time, consequences of control failure are significantly more severe, which implies a conflicting requirement for robust control. It also implies a requirement for comprehensive and robust exception handling. Coordinated control of multiple competing objectives with significant interactions, together with many shared uses of actuators to control multiple variables, implies that highly integrated control logic and shared actuator management will be required. It remains a challenge for the integrated technologies to simultaneously address these multiple and often competing requirements to be demonstrated on existing fusion devices and adapted for ITER in time to support its operational schedule. We describe ways in which the PCSSP will help address these challenges to support design of both the ITER PCS itself and the algorithms that will be implemented therein, and at the same time greatly reduce the cost of that development. We summarize the current status of the PCSSP design task, including system requirements and preliminary design documents already delivered as well as features of the ongoing detailed architectural design. The methods being incorporated in the detailed design are based on prior experience with control simulation environments in fusion and on standard practices prevalent in development of control-intensive industrial product designs. (C) 2014 Elsevier B.V. All rights reserved.

This paper describes the design, implementation, and operation of the Video Real-Time (VRT) diagnostic system of the ASDEX Upgrade plasma experiment and its integration with the ASDEX Upgrade Discharge Control System (DCS). Hot spots produced by heating systems erroneously or accidentally hitting the vessel walls, or from objects in the vessel reaching into the plasma outer border, show up as bright areas in the videos during and after the reaction. A system to prevent damage to the machine by allowing for intervention in a running discharge of the experiment was proposed and implemented. The VRT was implemented on a multi-core real-time Linux system. Up to 16 analog video channels (color and b/w) are acquired and multiple regions of interest (ROI) are processed on each video frame. Detected critical states can be used to initiate appropriate reactions - e.g. gracefully terminate the discharge. The system has been in routine operation since 2007. (C) 2013 Published by Elsevier B.V.

Thermonuclear plasmas are governed by nonlinear characteristics: plasma operation can be classified into scenarios with pronounced features like L and H-mode, ELMs or MHD activity. Transitions between them may be treated as events. Similarly, technical systems are also subject to events such as failure of measurement sensors, actuator saturation or violation of machine and plant operation limits.Such situations often are handled with a mixture of pulse abortion and iteratively improved pulse schedule reference programming. In case of protection-relevant events, however, the complexity of even a medium-sized device as ASDEX Upgrade requires a sophisticated and coordinated shutdown procedure rather than a simple stop of the pulse. The detection of events and their intelligent handling by the control system has been shown to be valuable also in terms of saving experiment time and cost.This paper outlines how ASDEX Upgrade's discharge control system (DCS) detects events and handles exceptions in two stages: locally and centrally. The goal of local exception handling is to limit the effect of an unexpected or asynchronous event to a minimal part of the controlled system. Thus, local exception handling facilitates robustness to failures but keeps the decision structures lean. A central state machine deals with exceptions requiring coordinated action of multiple control components. DCS implements the state machine by means of pulse schedule segments containing pre-programmed waveforms to define discharge goal and control methods within a time window. Segments also include conditions to define exceptions and associated segment branching instructions. Thus, the state machine logic can automatically adapt to plasma and plant state. (C) 2013 Elsevier B.V. All rights reserved.

Establishing adequate technical and physical boundary conditions for a sustained nuclear fusion reaction is a challenging task. Phased feedback control and monitoring for heating, fuelling and magnetic shaping is mandatory, especially for fusion devices aiming at high performance plasmas. Technical and physical interrelations require close collaboration of many components in sequential as well as in parallel processing flows. Moreover, handling of asynchronous, off-normal events has become a key element of modern plasma performance optimisation and machine protection recipes.The manifoldness of plasma states and events, the variety of plant system operation states and the diversity in diagnostic data sampling rates can hardly be mastered with a rigid control scheme. Rather, an adaptive system topology in combination with sophisticated synchronisation and process scheduling mechanisms is suited for such an environment. Moreover, the system is subject to real-time control constraints: response times must be deterministic and adequately short.Therefore, the experimental tokamak device ASDEX Upgrade employs a discharge control system DCS, whose core has been designed to meet these requirements. In the paper we will compare the scheduling schemes for the parallelised realisation of a control workflow and show the advantage of a data-driven workflow over a managed workflow. The data-driven workflow as used in DCS is based on signals connecting process outputs and inputs. These are implemented as real-time streams of data samples. Consequently, real-time signal management forms the foundation of DCS. The paper explains the principal features such as tagged samples, signal groups, algorithmic blocks and processes as well as scheduling schemes which allow DCS control applications to be defined as self-contained modular building blocks glued together by a software framework.By virtue of this sound foundation, DCS is a mature but still evolving system for reliable, distributed control of an entire tokamak device coordinating and monitoring 20 diagnostic systems, 14 magnetic power supplies, 5 heating systems with a total power of more than 25 MW, 8 gas fuelling channels, a pellet injector and a killer gas gun. (c) 2012 Elsevier B.V. All rights reserved.