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.     

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.     

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.     

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.     

ASDEX Upgrade Discharge Control System—A real-time plasma control framework

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 data-driven 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.     

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.     

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.     

Simulation-based design process for the verification of ITER remote handling systems

The work behind this paper takes place in the EFDA's European Goal Oriented Training programme on Remote Handling (RH) “GOT-RH”. The programme aims to train engineers for activities supporting the ITER project and the long-term fusion programme. One of the projects of this programme focuses on the verification and validation (V&V) of ITER RH system requirements using digital mock-ups (DMU). The purpose of this project is to study and develop efficient approach of using DMUs in the V&V process of ITER RH system design utilizing a System Engineering (SE) framework. Complex engineering systems such as ITER facilities lead to substantial rise of cost while manufacturing the full-scale prototype. In the V&V process for ITER RH equipment, physical tests are a requirement to ensure the compliance of the system according to the required operation. Therefore it is essential to virtually verify the developed system before starting the prototype manufacturing phase. This paper gives an overview of the current trends in using digital mock-up within product design processes. It suggests a simulation-based process design centralized around a simulation lifecycle management system. The purpose of this paper is to describe possible improvements in the formalization of the ITER RH design process and V&V processes, in order to increase their cost efficiency and reliability.     

Development of Powerful RF Plasma Sources for Present and Future NBI Systems

This paper deals with the topic of RF plasma sources and their application in high-power neutral beam heating systems for nuclear fusion devices. RF sources represent an interesting alternative to the conventional arc discharge sources. Due to the absence of hot filaments they exhibit an inherent simplicity both in mechanical and electrical aspects and consequently offer advantages in terms of cost savings, gain in availability and reliability and reduced maintenance. This renders the RF plasma source attractive for any long pulse (> 10 sec) NBI system and in particular for the ITER NBI system. The latter, however, requires that the RF plasma source is also capable of delivering negative rather than positive hydrogen ions.In the first part of the paper the types, characteristics and operation experience of RF plasma sources for positive ions in operation are described. The second part is devoted to the development for ITER NBI: the basic requirements, physics and technology issues and the present status     

Iter in vessel viewing system design and assessment activities

The In Vessel Viewing System (IVVS) is fundamental remote handling equipment, which will be used to make a survey of the status of the blanket first wall and divertor plasma facing components. A prototype of a laser In Vessel Viewing and ranging System was developed and tested at ENEA laboratories in Frascati under EFDA task agreements, it is able to perform sub-millimetric bi-dimensional and three-dimensional images inside ITER during maintenance procedure allowing the evaluation of the state and damages of the in-vessel surface. The present prototype has been designed to operate under room conditions and starting from springtime 2009 a Grant with F4E is in progress for the design and the assessment of the IVVS system for ITER, keeping in account all the environmental conditions and constraints.     

ITER CODAC interface for the visible and infra-red wide angle viewing cameras

The amount of data generated by the infra-red and visible cameras at ITER is expected to be considerably larger than most diagnostics. ITER will have 12 infra-red cameras plus 12 visible cameras in four different equatorial port plugs. Each of the ports will have a Plant System Host (PSH) that will provide a standard image of the plant system to the ITER's Control and Data Access and Communication (CODAC) system.The two key functions of these cameras will be the scientific exploitation with the detection of interesting physics events and the operational protection of the machine, namely the first wall. Already assuming high bandwidth requirements for both audio and video, ITER will provide a separate network for this kind of communication, which will be used to transmit both the experimental and informational data provided by the cameras.This paper presents the current camera plant system design and its interaction with ITER CODAC system and networks. Starting from the camera specifications several alternatives for data collection and compression are discussed. The required inputs from CODAC and a first specification for the internal finite state machine are also presented. Finally, a possible hardware straw man design solution for the plant system hardware is proposed.     

ITERDB—The Data Archiving System for ITER

For ITER, acquiring, managing and archiving its data is an essential task. ITER is foreseen to produce up to one terabyte of data per pulse and several petabytes of data per year. All the produced data needs to be stored and managed. The stored data is expected to serve the data access needs of ITER researchers located both on the ITER premises as well as worldwide during ITER's lifetime and beyond.ITERDB is a data management system being designed for centralized ITER data archival and data access. It is designed to manage and serve both unprocessed and processed data from the ITER plant systems and data analysis workflows.In this paper, we report the ITER Data Archiving System software requirements and priorities that have been identified by working with ITER staff and a large number of stakeholders. We will describe the design challenges and the proposed solutions. We will also present the current state of the ITERDB software architecture design.     

Architecture of SPIDER control and data acquisition system

The ITER Heating Neutral Beam injectors will be implemented in three steps: development of the ion source prototype, development of the full injector prototype, and, finally, construction of up to three ITER injectors. The first two steps will be carried out in the ITER neutral beam test facility under construction in Italy. The ion source prototype, referred to as SPIDER, which is currently in the development phase, is a complex experiment involving more than 20 plant units and operating with beam-on pulses lasting up to 1 h. As for control and data acquisition it requires fast and slow control (cycle time around 0.1 ms and 10 ms, respectively), synchronization (10 ns resolution), and data acquisition for about 1000 channels (analogue and images) with sampling frequencies up to tens of MS/s, data throughput up to 200 MB/s, and data storage volume of up to tens of TB/year. The paper describes the architecture of the SPIDER control and data acquisition system, discussing the SPIDER requirements and the ITER CODAC interfaces and specifications for plant system instrumentation and control.     

Plan of ITER remote experimentation center

Plan of ITER remote experimentation center (REC) based on the broader approach (BA) activity of the joint program of Japan and Europe (EU) is described. Objectives of REC activity are (1) to identify the functions and solve the technical issues for the construction of the REC for ITER at Rokkasho, (2) to develop the remote experiment system and verify the functions required for the remote experiment by using the Satellite Tokamak (JT-60SA) facilities in order to make the future experiments of ITER and JT-60SA effectively and efficiently implemented, and (3) to test the functions of REC and demonstrate the total system by using JT-60SA and existing other facilities in EU. Preliminary identified items to be developed are (1) Functions of the remote experiment system, such as setting of experiment parameters, shot scheduling, real time data streaming, communication by video-conference between the remote-site and on-site, (2) Effective data transfer system that is capable of fast transfer of the huge amount of data between on-site and off-site and the network connecting the REC system, (3) Storage system that can store/access the huge amount of data, including database management, (4) Data analysis software for the data viewing of the diagnostic data on the storage system, (5) Numerical simulation for preparation and estimation of the shot performance and the analysis of the plasma shot. Detailed specifications of the above items will be discussed and the system will be made in these four years in collaboration with tokamak facilities of JT-60SA and EU tokamak, experts of informatics, activities of plasma simulation and ITER. Finally, the function of REC will be tested and the total system will be demonstrated by the middle of 2017.     

The JET programme in support of ITER

As part of its mission to prepare the operation of ITER, a major programme of enhancements has just been completed on the JET tokamak. These enhancements include a complete replacement of the plasma-facing components in JET, from carbon-based to the combination of beryllium and tungsten foreseen for ITER, an upgrade of the neutral beam heating available on JET from 20 MW/short pulse to 30 MW/long pulse operation, the installation of a high frequency pellet injection system for plasma fuelling and ELM control studies, an upgrade to the JET vertical stability system and a suite of new diagnostics.The future JET programme is foreseen to proceed progressively from a test of fuel retention in the standard regimes of ITER operation towards more aggressive, high performance experiments that will demonstrate the operating space limits with the new wall. Depending on the results of the earlier experiments, the exploitation of the enhancements is foreseen to be completed with a deuterium-tritium experiment. This would represent the most integrated test of ITER operational scenarios possible before ITER itself.JET is a cooperative programme funded and exploited in collaboration by all of the European fusion laboratories. As such, JET is a test bed for multi-national use of a single fusion facility, as is foreseen for ITER. Opportunities for broadening the participation in JET to other ITER Parties are presently being explored. If these opportunities can be implemented, JET would provide not only an integrated test of ITER regimes of operation but also a demonstration of how ITER will be operated, even to the extent of including significant numbers of the same team who will eventually operate ITER.     

Actuator and diagnostic requirements of the ITER Plasma Control System

The ITER Plasma Control System (PCS) requires an extensive set of about 50 diagnostic systems to measure the plasma response and about 20 actuators to act on the plasma to carry out its control functions. The specifications and real limitations of the actuators and diagnostics are being assessed as part of the ongoing conceptual design of the PCS to understand the potential impact on plasma control. The actuators include magnetic coils (central solenoid (CS), poloidal field (PF), vertical stability (VS), edge localized mode (ELM), correction coils (CC)), heating and current drive (electron cyclotron (EC), ion cyclotron (IC), neutral beam injection (NBI), and possibly lower hybrid (LH)), glow discharge cleaning, fueling and impurity gas and pellet injection, vacuum pumping, and disruption mitigation systems. Diagnostic systems are prioritized according to their role in machine protection (MP), basic control (BC), advanced control (AC), and physics studies (PS). At the conceptual design phase, detailed control algorithms do not yet need to be specified, but conceptual solutions must be chosen that satisfy the PCS requirements for control within the real constraints of the diagnostics and actuators. The feasibility of the chosen solutions must be proven either through established control schemes on existing machines or through an R&D program to develop them before they will be required on ITER. The diagnostic and actuator requirements of the PCS will evolve from first plasma through the high performance DT phase. A comparison is made of the expected requirements to control vertical stability, sawteeth, neoclassical tearing modes (NTMs), edge localized modes (ELMs), error fields, resistive wall modes (RWMs), Alfvén eigenmodes, and disruptions with the ITER baseline actuator and diagnostic specifications.     

RADI—A RF source size-scaling experiment towards the ITER neutral beam negative ion source

IPP Garching is currently developing a negative hydrogen ion RF source for the ITER neutral beam system. The source demonstrated already current densities in excess of the ITER requirements (>200 A/m² D⁻) at the required source pressure and electron/ion ratio, but with only small extraction area and limited pulse length. A new test facility (RADI) went recently into operation for the demonstration of the required (plasma) homogeneity of a large RF source and the modular driver concept.The source with the dimension of 0.8 m × 0.76 m has roughly the width and half the height of the ITER source; its modular driver concept will allow an easy extrapolation in only one direction to the full size ITER source. The RF power supply consists of two 180 kW, 1 MHz RF generators capable of 30 s pulses. A dummy grid matches the conductance of the ITER source. Full size extraction is presently not possible due to the lack of an insulator, a large size extraction system and a beam dump.The main parameters determining the performance of this “half-size” source are the negative ion and electron density in front of the grid as well as the homogeneity of their profiles across the grid. Those will be measured by optical emission and cavity ring down spectroscopy, by Langmuir probes and laser detachment. These methods have been calibrated to the extracted current densities achieved at the smaller source test facilities at IPP for similar source parameters. However, in order to get some information about the possible ion and electron currents, local single aperture extraction with a Faraday cup system is planned.     

Study on the Flexible Support Legs for ITER Transfer Cask System

In fields of remote handing i.e. robot technology for fusion engineering reactor, such as ITER or the China fusion engineering test reactor, the flexible support legs are key components for their transfer cask system to adjust its position, joining to hot cell or tokamak ports for maintaining the fusion device. For ITER machine, each support leg should withstand maximum 50 tons load and adjust its height in 150 mm. Defect in original ITER design was presented. A new concept for the support legs was configured and its feasibility was proven. Detailed design and simulation was done for the new support leg with virtual prototype technology. Simulation results show that new support leg could not only meet all required function but also has merits of constant load during the tuning process with linear relation of control variable parameters, which is intended to be used for Tokamak reactors.