Physics of the conceptual design of the ITER plasma control system

The ITER plasma control system (PCS) will play a central role in enabling the experimental program to attempt to sustain DT plasmas with Q = 10 for several hundred seconds and also support research toward the development of steady-state operation in ITER. The PCS is now in the final phase of its conceptual design. The PCS relies on about 45 diagnostic systems to assess real-time plasma conditions and about 20 actuator systems for overall control of ITER plasmas. It will integrate algorithms required for active control of a wide range of plasma parameters with sophisticated event forecasting and handling functions, which will enable appropriate transitions to be implemented, in real-time, in response to plasma evolution or actuator constraints.In specifying the PCS conceptual design, it is essential to define requirements related to all phases of plasma operation, ranging from early (non-active) H/He plasmas through high fusion gain inductive plasmas to fully non-inductive steady-state operation, to ensure that the PCS control functionality and architecture will be capable of satisfying the demands of the ITER research plan. The scope of the control functionality required of the PCS includes plasma equilibrium and density control commonly utilized in existing experiments, control of the plasma heat exhaust, control of a range of MHD instabilities (including mitigation of disruptions), and aspects such as control of the non-inductive current and the current profile required to maintain stable plasmas in steady-state scenarios. Control areas are often strongly coupled and the integrated control of the plasma to reach and sustain high plasma performance must apply multiple control functions simultaneously with a limited number of actuators. A sophisticated shared actuator management system is being designed to prioritize the goals that need to be controlled or weigh the algorithms and actuators in real-time according to dynamic control needs. The underlying architecture will be event-based so that many possible plasma or plant system events or faults could trigger automatic changes in the control algorithms or operational scenario, depending on real-time operating limits and conditions.     

A simulation environment for ITER PCS development

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.     

Towards the conceptual design of the ITER real-time plasma control system

ITER will be the world's largest magnetic confinement tokamak fusion device and is currently under construction in southern France. The ITER Plasma Control System (PCS) is a fundamental component of the ITER Control, Data Access and Communication system (CODAC). It will control the evolution of all plasma parameters that are necessary to operate ITER throughout all phases of the discharge. The design and implementation of the PCS poses a number of unique challenges. The timescales of phenomena to be controlled spans three orders of magnitude, ranging from a few milliseconds to seconds. Novel control schemes, which have not been implemented at present-day machines need to be developed, and control schemes that are only done as demonstration experiments today will have to become routine. In addition, advances in computing technology and available physics models make the implementation of real-time or faster-than-real-time predictive calculations to forecast and subsequently to avoid disruptions or undesired plasma regimes feasible. This requires the PCS design to be adaptable in real-time to the results of these forecasting algorithms. A further novel feature is a sophisticated event handling system, which provides a means to deal with plasma related events (such as MHD instabilities or L-H transitions) or component failure. Finally, the schedule for design and implementation poses another challenge. The beginning of ITER operation will be in late 2020, but the conceptual design activity of the PCS has already commenced as required by the on-going development of diagnostics and actuators in the domestic agencies and the need for integration and testing. This activity is presently underway as a collaboration of international experts and the results will be published as a subsequent publication. In this paper, an overview about the main areas of intervention of the plasma control system will be given as well as a summary of the interfaces and the integration into ITER CODAC (networks, other applications, etc.). The limited amount of commissioning time foreseen for plasma control will make extensive testing and validation necessary. This should be done in an environment that is as close to the PCS version running the machine as possible. Furthermore, the integration with an Integrated Modeling Framework will lead to a versatile tool that can also be employed for pulse validation, control system development and testing as well as the development and validation of physics models. An overview of the requirements and possible structure of such an environment will also be presented.     

Event generation and simulation of exception handling with the ITER PCSSP

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.     

High voltage power supplies for ITER RF heating and current drive systems

The RF heating and current drive (H&CD) systems to be installed for the ITER fusion machine are the electron cyclotron (EC), ion cyclotron (IC) and, although not in the first phase of the project, lower hybrid (LH). These systems require high voltage, high current power supplies (HVPS) in CW operation.These HVPS should deliver around 50 MW electrical power to each of the RF H&CD systems with stringent requirements in terms of accuracy, voltage ripple, response time, turn off time and fault energy. The PSM (Pulse Step Modulation) technology has demonstrated over the past 20 years its ability to fulfill these requirements in many industrial facilities and other fusion reactors and has therefore been chosen as reference design for the IC and EC HVPS systems.This paper describes the technical specifications, including interfaces, the resulting constraints on the design, the conceptual design proposed for ITER EC and IC HVPS systems and the current status.     

Overview of magnetic control in ITER

ITER is targeting Q = 10 with 500 MW of fusion power. To meet this target, the plasma needs to be controlled and shaped for a period of hundreds of seconds, avoiding contact with internal components, and acting against instabilities that could result in the loss of control of the plasma and in its disruptive termination.Axisymmetric magnetic control is a well-understood area being the basic control for any tokamak device. ITER adds more stringent constraints to the control primarily due to machine protection and engineering limits. The limits on the actuators by means of the maximum current and voltage at the coils and the few hundred ms time response of the vacuum vessel requires optimization of the control strategies and the validation of the capabilities of the machine in controlling the designed scenarios.Scenarios have been optimized with realistic control strategies able to guarantee robust control against plasma behavior and engineering limits due to recent changes in the ITER design. Technological issues such as performance changes associated with the optimization of the final design of the central solenoid, control of fast transitions like H to L mode to avoid plasma-wall contact, and optimization of the plasma ramp-down have been modeled to demonstrate the successful operability of ITER and compatibility with the latest refinements in the magnetic system design.Validation and optimization of the scenarios refining the operational space available for ITER and associated control strategies will be proposed. The present capabilities of magnetic control will be assessed and the remaining critical aspects that still need to be refined will be presented. The paper will also demonstrate the capabilities of the diagnostic system for magnetic control as a basic element for control. In fact, the noisy environment (affecting primarily vertical stability), the non-axisymmetric elements in the machine structure (affecting the accuracy of the identification of the plasma boundary), and the strong component of eddy current at the start-up (resulting in a poor S/N ratio for plasma reconstruction for I_p < 2 MA requiring a robust plasma control) make the ITER magnetic diagnostic system a demanding part of the magnetic control and investment protection systems. Finally the paper will illustrate the identified roles of magnetic control in the PCS (plasma control system) as formally defined in the recent first step of the design and development of the system.     

Design and performance analysis of ITER ex-vessel magnetic diagnostics

Accurate magnetic diagnostics are essential to perform reliable operation of any tokamak. The ITER magnetic diagnostics include a wide variety of sensors located on the inner and outer surfaces of the vacuum vessel, in the divertor cassettes and in the casing of the toroidal field coils. As the measurement accuracy of the inner set of magnetic sensors might be compromised by various radiation effects and high heat loads, the complementary ex-vessel set is essential to provide backup information. This paper is an overview of the ex-vessel magnetic diagnostic which consists mainly of pick-up coils, steady state sensors, Rogowski coils in the toroidal field coil casing and fibre optic current sensors. The work presented aims at designing these sensors to meet the performance requirements in spite of the constraints due to the tokamak environment. The manufacturing constraints and the positioning requirements for all the ex-vessel magnetic sensors are described. The use and expected accuracy of the entire ex-vessel magnetic diagnostic is assessed in terms of magnetic equilibrium reconstruction and plasma current measurement precision.     

Progress and improvement of KSTAR plasma control using model-based control simulators

Superconducting tokamaks like KSTAR, EAST and ITER need elaborate magnetic controls mainly due to either the demanding experiment schedule or tighter hardware limitations caused by the superconducting coils. In order to reduce the operation runtime requirements, two types of plasma simulators for the KSTAR plasma control system (PCS) have been developed for improving axisymmetric magnetic controls. The first one is an open-loop type, which can reproduce the control done in an old shot by loading the corresponding diagnostics data and PCS setup. The other one, a closed-loop simulator based on a linear nonrigid plasma model, is designed to simulate dynamic responses of the plasma equilibrium and plasma current (I_p) due to changes of the axisymmetric poloidal field (PF) coil currents, poloidal beta, and internal inductance. The closed-loop simulator is the one that actually can test and enable alteration of the feedback control setup for the next shot. The simulators have been used routinely in 2012 plasma campaign, and the experimental performances of the axisymmetric shape control algorithm are enhanced. Quality of the real-time EFIT has been enhanced by utilizations of the open-loop type. Using the closed-loop type, the decoupling scheme of the plasma current control and axisymmetric shape controls are verified through both the simulations and experiments. By combining with the relay feedback tuning algorithm, the improved controls helped to maintain the shape suitable for longer H-mode (10–16 s) with the number of required commissioning shots largely reduced.     

Design status of the ITER ECRH upper launcher mm-wave system

The purpose of the ITER electron cyclotron resonance heating (ECRH) upper launcher (UL), or antennae will be to provide localised current drive by accurately directing mm-wave beams up to 2MW, out of the four allocated upper port plugs, at chosen rational magnetic flux surfaces in order to stabilise neoclassical tearing modes (NTMs). This paper will present an overview of the UL, with emphasis on the mm-wave components. The mm-wave layout includes corrugated waveguide sections and a quasi-optical path with both focusing mirrors and plane steering mirrors. One of the essential components of the UL is the Steering Mechanism Assembly (SMA), providing variable poloidal injection angles fulfilling high deposition accuracy requirements at the plasma location. The Actuator principle and rotor bearings are frictionless and backlash free, avoiding tribological difficulties such as stickslip and seizure. The underlying working principle is the use of mechanically compliant structures. Validation and proof testing of the steering principle is achieved with an uncooled first prototype demonstrator. A second prototype is currently being manufactured, comprising the functionalities needed for the ITER compatible system such as water cooling and high power mm-wave compatibility. In order to perform the fatigue tests of the actuator bellows, a test facility has been built, under ITER-like vacuum and temperature working conditions. Results of the cyclic fatigue tests are compared to the various manufacturer standards and codes, combining stress and strain controlled material fatigue properties.     

Development of the supervisory systems for the ITER diagnostic systems in JADA

In ITER, it is important how the CODAC system conducts many plant systems including diagnostic systems. In order to establish necessary communications between the diagnostics systems and the CODAC system, Japan domestic agency (JADA) has proposed the new concept of supervisory system for the diagnostic system based on our experiences in operating plasma diagnostic systems. The supervisory system manages operation sequences, current state and configuration parameters for the measurement. JADA designed the supervisory system satisfying the requirements from both CODAC system and diagnostic systems. In our design, the tool which converts operational steps described as flowcharts into the EPICS (experimental physics and industrial control system) records source codes is introduced. This tool will ensure reduction of the system designers’ efforts. We designed a communication protocol to configure measurement parameters and proposed configuration parameter validation function. We also analyzed the management of the central/local control mode for the diagnostic systems. The function which selects the adequate limit values and consistency check algorithms in accordance with the conditions of the diagnostics system is proposed. JADA will develop a prototype of the supervisory system and validate the design in 2013.     

Progress on the heating and current drive systems for ITER

The electron cyclotron (EC), ion cyclotron (IC), heating-neutral beam (H-NB) and, although not in the day 1 baseline, lower hybrid (LH) systems intended for ITER have been reviewed in 2007/2008 in light of progress of physics and technology in the field. Although the overall specifications are unchanged, notable changes have been approved. Firstly, it has been emphasized that the H&CD systems are vital for the ITER programme. Consequently, the full 73 MW should be commissioned and available on a routine basis before the D/T phase. Secondly, significant changes have been approved at system level, most notably: the possibility to operate the heating beams at full power during the hydrogen phase requiring new shine through protection; the possibility to operate IC with 2 antennas with increased robustness (no moving parts); the possible increase to 2 MW of key components of the EC transmission systems in order to provide an easier upgrading of the EC power as may be required by the project; the addition of a building dedicated to the RF power sources and to a testing facility for acceptance of diagnostics and heating port plugs. Thirdly, the need of a plan for developing, in time for the active phase, a CD system such as LH suitable for very long pulse operation of ITER was recognised. The review describes these changes and their rationale.     

Architectural concept for the ITER Plasma Control System

The plasma control system is a key instrument for successfully investigating the physics of burning plasma at ITER. It has the task to execute an experimental plan, known as pulse schedule, in the presence of complex relationships between plasma parameters like temperature, pressure, confinement and shape. The biggest challenge in the design of the control system is to find an adequate breakdown of this task in a hierarchy of feedback control functions. But it is also important to foresee structures that allow handling unplanned exceptional situations to protect the machine. Also the management of the limited number of actuator systems for multiple targets is an aspect with a strong impact on system architecture. Finally, the control system must be flexible and reconfigurable to cover the manifold facets of plasma behaviour and investigation goals.In order to prepare the development of a control system for ITER plasma operation, a conceptual design has been proposed by a group of worldwide experts and reviewed by an ITER panel in 2012. In this paper we describe the fundamental principles of the proposed control system architecture and how they were derived from a systematic collection and analysis of use cases and requirements. The experience and best practices from many fusion devices and research laboratories, augmented by the envisaged ITER specific tasks, build the foundation of this collection. In the next step control functions were distilled from this input. An analysis of the relationships between the functions allowed sequential and parallel structures, alternate branches and conflicting requirements to be identified. Finally, a concept of selectable control layers consisting of nested “compact controllers” was synthesised. Each control layer represents a cascaded scheme from high-level to elementary controllers and implements a control hierarchy. The compact controllers are used to resolve conflicts when several control functions would use the same command signals as their outputs. They consist of a collection of potentially conflicting control functions from which one at a time is exclusively activated by a mode selector signal.It can be shown that this architectural design is capable of implementing all of the presently known functional control requirements. Furthermore, this design takes already into account that the result of future experiments at ITER will create additional requirements on the functions or performance of ITER plasma control.     

Design and technical status of the EU contribution to ITER

Europe is involved in the procurement of most of the high-technology items for the ITER device (e.g. parts of the superconducting Toroidal (TF) and Poloidal Field (PF) coils, the vacuum vessel (VV), the in-vessel components, the remote handling, the additional heating systems, the tritium plant and cryoplant and finally parts of the diagnostics). In many cases the technologies required to manufacture these components are well established, in others there is still ongoing design and R&D work to select and optimise the final design solutions and to consolidate the underlying technologies as, for example, in the areas of heating and current drive, plasma diagnostics, shield blanket and first wall, remote handling, etc. A design review has recently been conducted by the ITER Organisation, with the support of the Domestic Agencies (DAs) established by the countries participating to ITER, to address the remaining outstanding technical issues and understand the associated implications for design, machine performance, schedule and cost.This paper provides an update of the design and technical status of EU contributions to ITER.     

Structural Design Study for ITER Upper ELM Coils

ITER ELM coils are important parts of in-vessel coils and they are mounted on the vacuum vessel and behind the blanket module. They consist of three sets of coils, referred to as the upper, mid, and lower coils. In order to verify the structural design feasibility and find the better structure for upper edge localized modes (ELM) coil, two different variants of coil support structures are studied under the electromagnetic load, thermal and other loads. Results show that besides the bracket of variant 2 does not meet the fatigue criteria, the conductor, jacket and bracket of the two structures can meet the static, fatigue and crack threshold criteria and both of them are valid and feasible. In addition, the better structure is chosen for upper ELM coil.     

HL-2A托卡马克装置的工程和实验概况

中国环流器二号A装置(HL-2A)是核工业西南物理研究院2002年投入实验运行的托卡马克,它是我国第1个具有偏滤器、等离子体截面具有一定垂直拉长的托卡马克.HL-2A的磁体使用铜导体,具有良好的灵活性和等离子体的可近性,其极向场线圈全部位于环向场线圈之内,位于真空室内的偏滤器的成形线圈可建立双零和单零的偏滤器位形.HL-2A已发展了30多套先进的等离子体诊断系统和总功率4 MW的辅助加热系统,加料技术得到持续发展.随着上述系统的建设和放电综合控制技术的提高,HL-2A装置已获得了高约束模式,这为开展先进托卡马克(AT)物理实验,ITER和聚变堆的科学、技术和工程问题等的研究奠定了基础.HL-2A也成为国际上最活跃的中型托卡马克,为国际托卡马克物理活动(ITPA)作出了积极贡献.     

Real-time mirror steering for improved closed loop neoclassical tearing mode suppression by electron cyclotron current drive in DIII-D

The development and operation of the neoclassical tearing mode (NTM) avoidance and control system for DIII-D, which uses six sets of real-time steerable mirrors in order to move the electron cyclotron current drive (ECCD) deposition location in plasma, is described. The real-time DIII-D NTM control algorithm residing in the Plasma Control System (PCS) automatically detects an NTM by analysis of the Mirnov diagnostics, employs motional Stark effect (MSE) EFIT MHD equilibrium reconstruction to locate the rational q-surface where the NTM island can be found, then calculates the appropriate mirror position for alignment of the ECCD with the island using ray tracing. The control commands from PCS are sent to the electron cyclotron system to switch on and off or modulate the gyrotrons and to the steerable mirror system to move the steerable mirrors to the requested positions. Successful NTM suppression has been achieved in DIII-D using this control system to rapidly align the NTM island and the ECCD deposition location, and to actively maintain the alignment as plasma conditions change.     

1-D thermal-hydraulic analysis of the high temperature superconducting current leads for the ITER magnet system from 5 K to 300 K

The magnet system of ITER includes high temperature superconducting (HTS) current leads with a maximum current of 68 kA for the toroidal field (TF) coils, 55 kA for the poloidal field (PF)/central solenoid (CS) coils and 10 kA for the control coils (CC), respectively. Although different in terms of size and operative conditions, the ITER HTS current leads have been all designed on the basis of an established concept, which was successfully developed for the LHC at CERN and proven by the so-called 70 kA “demonstrator” lead made by KIT and by the ITER pre-prototypes made by ASIPP in China. A broad R&D campaign has been undertaken by ASIPP and CERN in order to find optimized designs for each component of the leads. Nevertheless, a comprehensive picture of the performance of the entire HTS current leads is not yet available. In this paper, a steady state, full length, thermal-hydraulic 1-D modeling is applied to the study of the three types (TF, PF/CS, CC) of ITER HTS current leads. The results of this predictive analysis are then compared with relevant ITER requirements. It was found that the present design of the HTS current leads will fulfill these specifications.     

Concept design of fusion power shutdown system for ITER

ITER is a nuclear facility. It is essential to maintain operational safety or to bring this facility to a safe state in case of accidents or incidents. During plasma operation ITER plasma will generate significant heat loads on the plasma facing components. For a few reference accidents there is the need to stop plasma reliably within a certain time. Fusion power shutdown system is the safety system to implement this termination function for ITER. It is based on the concept of massive gas injection.This paper summarizes the safety requirements, logics and the physics requirements on this system for reliable termination of ITER plasma. With regard to the quantity of gas, transient behavior simulation is shown, subsequently providing guideline for laboratory bench-testing. Conceptual engineering design of the system is given together with instrumentation and control specifications.     

First tests of diagnostic mirrors in a tokamak divertor: An overview of experiments in DIII-D

Mirrors will be used in ITER in all optical diagnostic systems observing the plasma radiation in the ultraviolet, visible and infrared ranges. Diagnostic mirrors in ITER will suffer from electromagnetic radiation, energetic particles and neutron irradiation. Erosion due to impact of fast neutrals from plasma and deposition of plasma impurities may significantly degrade optical and polarization characteristics of mirrors influencing the overall performance of the respective diagnostics. Therefore, maintaining the best possible performance of mirrors is of the crucial importance for the ITER optical diagnostics. Mirrors in ITER divertor are expected to suffer from deposition of impurities. The dedicated experiment in a tokamak divertor was needed to address this issue. Investigations with molybdenum diagnostic mirrors were made in DIII-D divertor. Mirror samples were exposed at different temperatures in the private flux region to a series of ELMy H-mode discharges with partially detached divertor plasmas. An increase of temperature of mirrors during the exposure generally led to the mitigation of carbon deposition, primarily due to temperature-enhanced chemical erosion of carbon layers by D atoms. Finally, for the mirrors exposed at the temperature of ∼160 °C neither carbon deposition nor degradation of optical properties was detected.     

Distributed digital real-time control system for TCV tokamak

A new digital feedback control system (named the SCD “Système de Contrôle Distribué”) has been developed, integrated and used successfully to control TCV (Tokamak à Configuration Variable) plasmas. The system is designed to be modular, distributed, and scalable, accommodating hundreds of diagnostic inputs and actuator outputs. With many more inputs and outputs available than previously possible, it offers the possibility to design advanced control algorithms with better knowledge of the plasma state and to coherently control all TCV actuators, including poloidal field (PF) coils, gas valves, the gyrotron powers and launcher angles of the electron cyclotron heating and current drive system (ECRH/ECCD) together with diagnostic triggering signals. The system consists of multiple nodes; each is a customised Linux desktop or embedded PC which may have local ADC and DAC cards. Each node is also connected to a memory network (reflective memory) providing a reliable, deterministic method of sharing memory between all nodes. Control algorithms are programmed as block diagrams in Matlab-Simulink providing a powerful environment for modelling and control design. The C code is generated automatically from the Simulink block diagram and compiled, with the Simulink Embedded Coder (SEC, formerly Real-Time Workshop Embedded Coder), into a Linux shared library (“.so” file) and distributed to target nodes in the discharge preparation phase. During the TCV discharge, an application on each node is executed that dynamically loads the shared library at runtime. In order to obtain reliable and reproducible real time execution of the algorithm, all interrupts to the CPU on each node are suspended just before firing the shot and re-enabled afterwards. Since installation, the new digital control system has been used for a multitude of plasma control applications, ranging from basic experiments of coil current and density control to advanced experiments of MHD (magnetohydrodynamics) and plasma profile control, as well as real-time plasma transport simulations. Recently, a real-time version of a plasma equilibrium reconstruction code was developed and implemented, providing the future possibility to control the plasma shape and profiles directly during the discharge evolution. This paper presents the architecture of the new control system, its integration into the TCV plant and a sample of control applications used for TCV plasma discharges.