Commissioning and first results of the ITER-relevant negative ion beam test facility ELISE

The test facility ELISE which was constructed in the last three years at the Max-Planck-Institut für Plasmaphysik (IPP), Garching, is an important intermediate step of the development of the neutral beam system for ITER. ELISE allows gaining an early experience of the performance and operation of large RF driven sources for negative hydrogen ions and will give an important input for the commissioning and the design of the SPIDER and MITICA test facilities at Padua and the ITER neutral beam system. ELISE has gone recently into operation with first plasma and beam pulses. The experiments aim at the demonstration of an ion beam at the required parameters within 2 years of operation until end of 2014, the end of the service contract with F4E for the establishment and exploitation of ELISE.     

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.     

Optimization of the electrostatic and magnetic field configuration in the MITICA accelerator

MITICA (Megavolt ITER Injector Concept Advancement) is a test facility for the development of a full-size heating and current drive neutral beam injectors for the ITER Tokamak reactor. The optimized electrostatic and magnetic configuration has been defined by means of an iterative optimization involving all the physics and the engineering aspects. The acceleration grids have been designed considering optical performances and mechanical constraints related to embedded magnets, to cooling channels, to the grid stiffness and manufacturability. A combination of “local” vertical field and horizontal “long range” field has been found to be the most effective set-up for ion extraction, beam focusing and minimization and equalization of thermo-mechanical loads and minimal number of electrons exiting the accelerator.     

Characterization,test and interpretative simulations of one-dimensional Carbon Fiber Composite prototype for SPIDER experiment

For ITER operations, additional heating systems are required. One of these systems is the neutral beam injector (NBI). The SPIDER experiment, a small-scale NBI, is going to be built with the aim to optimize the beam source. For this reason it is provided with several diagnostics, among which the Short-Time Retractable Instrumented Kalorimeter Experiment (STRIKE). In this contribution, a characterization of the Carbon Fiber Composite (CFC) tiles, which are the main component of the diagnostic, is presented. Such analyses include tests with a power laser, exposure to particle beams and thermal stress tests. The results are discussed, which will drive the definition of the acceptance tests of the final supply of CFC tiles.     

Structural analyses and integrated design of the MITICA Injector assembly

In the framework of the activities foreseen for PRIMA (Padova Research on Injector Megavolt Accelerated) the MITICA neutral beam injector plays the role of main experiment, aiming to build, operate, test and optimize a full power and full scale prototype of the ITER Heating Neutral Beam Injector [1], [2], [3].The entire MITICA system will be housed in special buildings, suitably designed to provide all the necessary supports, interfaces and shielding walls for nuclear radiation safety. Therefore an integrated design of the MITICA system and relevant buildings shall be developed and verified carefully, considering all the different configurations, operational modes and load combinations.This paper presents the numerical models and the results of MITICA assembly integrated analyses. The model takes into account properly constraints to ground and surrounding buildings, to study and verify the static and seismic response of the whole assembly.The load cases are defined and the numerical analyses described. Load definition and analyses have been performed considering the requirements of both the ASME [4] and the National Standard NTC2008 [5] for the seismic verification of structures subject to design response spectra.The obtained results are finally shown in detail and discussed, also comparing some different design options for design optimization.     

An Indian test facility to characterise diagnostic neutral beam for ITER

The diagnostic neutral beam (DNB) line shall be used to diagnose the He ash content in the D–T phase of the ITER machine using the charge exchange recombination spectroscopy (CXRS). Implementation of a successful DNB at ITER requires several challenges related to the production, neutralization and transport of the neutral beam over path lengths of 20.665 m, to be overcome. The delivery is aided if the above effects are tested prior to onsite commissioning. As DNB is a procurement package for INDIA, an ITER approved Indian test facility, INTF, is under construction at Institute for Plasma Research (IPR), India and is envisaged to be operational in 2015. The timeline for this facility is synchronized with the RADI, ELISE (IPP, Garching), SPIDER (RFX, Padova) in a manner that best utilization of configurational inputs available from them are incorporated in the design. This paper describes the facility in detail and discusses the experiments planned to optimise the beam transmission and testing of the beam line components using various diagnostics.     

Progress in control and data acquisition for the ITER neutral beam test facility

SPIDER, the ion source test bed in the ITER neutral beam test facility, is under construction and its operation is expected to start in 2014. Control and data acquisition for SPIDER are undergoing final design. SPIDER CODAS, as the control and data acquisition system is referred to, is requested to manage 25 plant units, to acquire 1000 analogue signals with sampling rates ranging from a few S/s to 10 MS/s, to acquire images with up to 100 frames per second, to operate with long pulses lasting up to 1 h, and to sustain 200 MB/s data throughput into the data archive with an annual data storage amount of up to 50 TB. SPIDER CODAS software architecture integrates three open-source software frameworks each addressing specific system requirements. Slow control exploits the synergy among EPICS and Siemens S7 programmable controllers. Data handling is by MDSplus a data-centric framework that is geared towards the collection and organization of scientific data. Diagnostics based on imaging drive the design of data throughput and archive size. Fast control is implemented by using MARTe, a data-driven, object-oriented, real-time environment. The paper will describe in detail the progress of the system hardware and software architecture and will show how the software frameworks interact to provide the functions requested by SPIDER CODAS. The paper will focus on how the performance requirements can be met with the described SPIDER CODAS architecture, describing the progress achieved by carrying out prototyping activities.     

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.     

Realization of a first series of single channel prototypes for the SPIDER grids

The SPIDER and MITICA experiments, planned to be built at Consorzio RFX for the development and optimization of the ITER Heating and Current Drive Neutral Beam Injectors, feature a number of components intercepted by high heat fluxes, that must be cooled during operations by means of suitable high performance cooling systems. As the design for such cooling systems presents some technological and heat transfer issues, a specific R&D program has been carried out, particularly referred to the accelerator grids that are among the most critical components.These components are foreseen to be manufactured by electrodeposition of pure copper onto a copper base plate to realize cooling channels and magnets slots, while the connection with the feeding manifolds is foreseen to be realized by friction welding or electron beam welding.Suitable manufacturing parameters and production methodologies have been identified by constructing and testing a first series of prototypes. A set of thermocouples have been embedded in some prototypes, to allow the evaluation of local heat transfer coefficients and the identification of local spots where dry-out phenomena might occur. To this purpose, a dedicated electrodeposition process has been developed.     

基于EPICS的J-TEXT CODAC系统

J-TEXT装置是华中科技大学恢复建造的中型托卡马克装置,已于2007年放电运行,其控制系统采用分布式结构,由多个子系统组成。为提高子系统集成、维护和更新的效率,并有效地管理各子系统、控制装置的运行状态及保障设备和人员安全,J-TEXT装置参考ITER CODAC的设计思路,结合J-TEXT装置的需求设计了J-TEXT CODAC系统。J-TEXT CODAC系统为装置各子系统提供统一的设计模型和相关设计标准,使用EPICS软件作为通讯中间层,设计了全局控制系统、时序和同步控制系统、联锁保护系统,并将原有控制系统改造、集成到J-TEXT CODAC系统中。目前该系统已部署在J-TEXT装置上,在2012年春季以来的多轮实验中运行良好。     

Gas Flow in the Neutralization Region of EAST Neutral Beam Injector

The gas flow in the neutralization region affects the neutralization efficiency as well as the beam transition efficiency of whole neutral beam injector, which will be applied as a high power auxiliary heating and non-inductive current driver system for the Experimental Advanced Superconducting Tokamak (EAST). The neutralization region of EAST neutral beam injector includes not only the gas cell neutralizer in traditional sense, but also part of the ion source downstream the electrode grid, the gate valve, and the transitional piping and fitting. The gas flow in this neutralization region has been modeled and researched using an adjusted Direct Simulation Monte Carlo code to understand the formation mechanism of gas target thickness, which determines the neutralization efficiency. This paper presents the steady-state, viscous and transition region flowfields, the gas density distribution and the various centerline profiles including Knudsen number, temperature, pressure and axial velocity by injected the deuterium gas from the arc chamber and neutralizer, respectively. The target thickness in the neutralization region as a function of gas inlet quantity is also given in the absence of beam for future operation of EAST neutral beam injector.     

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.     

The J-TEXT CODAC system design and implementation

Inspired by the ITER COntrol, Data Access and Communication (CODAC) and ITER instrumentation and control system, J-TEXT tokamak has upgraded its control system with J-TEXT CODAC system. The J-TEXT CODAC system is based on Experimental Physics and Industrial Control System (EPICS). The J-TEXT CODAC system covers everything in the J-TEXT control system including both central and plant control systems, similar to the ITER I&C system. J-TEXT CODAC system is built around a single central control system called Central CODAC system. All the control functions including conventional control, interlock, safety and other common services are supervised by CCS. The J-TEXT CODAC system has been implemented and tested on J-TEXT. It not only tests some of the ideas in ITER CODAC in real life, but also explores the feasibility of new approaches that is unique in J-TEXT CODAC system.     

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.     

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.     

FTU toroidal magnet power supply slow control using ITER CODAC Core System

FTU (Frascati Tokamak Upgrade) three-level slow control system has undergone several enhancements during its lifetime, involving essentially the supervisory and medium level, while the lower level is still mainly based on old Westinghouse Numalogic PLCs (Programmable Logic Controller). The legacy PLC controlling the toroidal magnet flywheel generator, named MFG1, is now being replaced with a more modern Siemens Simatic S7 PLC, because of its versatility an the ability to be integrated via standard networking protocol.The upgrade to this family of Siemens PLCs, which in the meantime has been selected as standard by ITER CODAC, has made MFG1 slow control an ideal candidate to deploy ITER CODAC software technologies and architecture to a running plant in an operating tokamak environment. A project has thus been started to port MFG1 control to ITER CODAC I&C architecture using the software package CODAC Core System to interface the PLC with the ITER standard systems for instrumentation and control, Plant System Host (PSH) and Mini-CODAC, developing dedicated HMI (Human–Machine Interface) and realizing the communication layer between MFG1 plant system and FTU supervisor.This paper will give a full account of the project and will report the results that have been obtained up to now, focusing also on the definite advantages provided by a distributed control architecture compared to the supervisor-dependent one still running at FTU, in view of future fusion devices.     

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.     

Recent and future upgrades to the DIII-D tokamak

Research on the DIII-D tokamak focuses on support for next-generation devices such as ITER by providing physics solutions to key issues and advancing the fundamental understanding of fusion plasmas. To support this goal, the DIII-D facility is planning a number of upgrades that will allow improved plasma heating, control, and diagnostic measurement capabilities. The neutral beam system has recently added an eighth ion source and one of the beamlines is currently being rebuilt to allow injection of 5 MW of off-axis power at an angle of up to 16.5° from the horizontal. The electron cyclotron heating (ECH) system is adding two additional gyrotrons and is using new launchers that can be aimed poloidally in real-time by an improved plasma control system. The fast wave heating system is being upgraded to allow two of the three launchers to inject up to 2 MW each in future experiments. Several diagnostics are being added or upgraded to more thoroughly study fluctuations, fast ions, heat flux to the walls, plasma flows, rotation, and details of the plasma density and temperature profiles.