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The ITER EC system includes both an equatorial port launcher and upper port launchers as RF power injection devices, The waveguide switch in the transmission line (TL) is used to select the operating launcher. Its operation is required even during plasma operation, primarily for the mid-pulse switch operation. Changing the waveguide switch requires that the gyrotron stop RF power during the switch operation since the direction is changed by mechanical movement of the mirror position which takes a few seconds. Since the ITER EC system is based on a multi-subsystem concept, each subsystem has its own subsystem control unit (SCU). The EC main controller supervises all subsystem controllers. Hence, cooperative operation requires the sharing of the information of both RF power status and switch status between the gyrotron SCU and TL SCU via the main controller. Since the design of the inter-subsystem control scheme is a key issue for ITER EC system control, its evaluation is required. At QST, the gyrotron and ITER-relevant TL test stand were utilized for demonstration of mid-pulse switch operation. For this purpose, SCUs for each subsystem and the main controller were developed using the ITER-relevant control system. The operation of the mechanical switch during gyrotron pulse was demonstrated. During the 150 s operation of the high power gyrotron at 400 kW level, the waveguide switch in the TL was operated to change the direction of RF power. The time duration for the switch operation with inter-subsystem control scheme took 1.5 s in total. The synchronizing of RF power suspend and resume with switch motion has therefore been realized, and RF power direction control during the gyrotron operation was successfully demonstrated.

ITER requires extensive diagnostics to meet the demands for machine operation, protection, plasma control and physics studies. The interfaces between plant instrumentation and control (I&C) and the central control system follow mandatory rules described in the Plant Control Design Handbook (PCDH) [1] , while the design strategy for PCDH compliant plant I&C is covered in its guidelines and supported by hardware catalogues. During preliminary and final design review as well as factory and site acceptance testing it is therefore important to determine the maturity of the I&C design and its implementation to ensure its compliance with the PCDH and the diagnostics performance requirements.

A prototype of the ITER pulse scheduling system that prepares and manages parameters for ITER plasma operations has been designed. Based on the analyzed requirements on the system, structure of the parameters and necessary functions were determined. Segment and module structures were tuned to the ITER requirements. Three types of validations assure sanity of the parameters. The design limits check and the operation window check verify whether the values of the parameters do not exceed the limits. The consistency check calculates dependency among parameters in accordance with logics described in a scripting language. The ITER pulse scheduling system provides interface with a physics model and simulator. Some abstract physics parameters are converted to engineering parameters with the physics simulation. The results of simulation such as plasma characteristics of specified parameters are also shown to the researchers. The tool to specify the parameters is data-driven. Therefore, it is flexible for changes of number of the parameters. A prototype is being implemented in 2011. Using the prototype, this design will be verified and refined. The evaluation of the prototype will be a basis of the final production of the ITER pulse scheduling system.

ITER project's long time span and the nature of the instrumentation and control (I&C) procurement procedures for the Plant Systems require that the ITER Organization defines and follows well recognized standards which are used both by the industry and in physics experiments. The ITER I&C standards are defined in the Plant Control Design Handbook (PCDH) [1] . The ITER Organization has selected PCI Express and Ethernet for IO intercommunication to be used for plant system instrumentation for fast controllers. The decision on the usage of serialized I/O bus protocols is based on the impressive performance and the commercial availability. The form factors that will be supported by CODAC include PXIe, MicroTCA, and AdvancedTCA platforms. While the PXIe form factor is already well established for instrumentation purposes through the PXI Systems Alliance ( www.pxisa.org ), the AdvancedTCA and MicroTCA platforms which were originally targeted for the telecommunications market ( www.picmg.org ) are currently optimized and specified for instrumentation use through the xTCA extensions for physics [2] . The objective of this study is the evaluation of an integrated ATCA controller design using only commercial components.

In the ITER project, Reliability, Availability, Maintainability and Inspectability (RAMI) approach has been adopted for technical risk control to guide the design of components and the preparation for operation and maintenance. RAMI analysis of the ITER CODAC System, the central plant control system, was performed in the current design available in the conceptual design phase. A functional breakdown was prepared, resulting in the system being divided into 5 main functions themselves divided into sub-functions. The Failure Modes, Effects and Critical Analysis were performed to initiate risk mitigation actions. Criticality matrices highlight the risks of the different failure modes with regard to their probability of occurrence and the impact on the availability. Identified risks exist on the effect on data storage hardware and for computer software. Reliability block diagrams were prepared to estimate the reliability and availability of each function under stipulated operating conditions. The inherent availability of the mandatory functions for the control of plasma experiments with mitigations was calculated to be 99.2% that is higher than the project required value of 98.8%. Compensating provisions are advocated in terms of design, testing, operation procedure and maintenance requirements in order to reduce risk levels.

The ITER instrumentation and control (I&C) system is the term encompassing all hardware and software required to operate ITER. It has two levels of hierarchy: the central I&C systems and the plant systems I&C. The plant systems I&C consists of thousands of computers processing hundreds of thousands of signals. The plant systems I&C, being the hardware interfacing layer for operating the ITER machine, shall integrate, control and coordinate all these computers and signals and allow a limited number of staff to operate the ITER machine from a central location with minimum human intervention during experimental pulses and 24 h/7 day continuous operation. The procurement model, together with the current estimate of 161 plant systems I&C, poses a major challenge for the realization and integration of the ITER I&C systems. standardization of plant systems I&C is of primary importance and has been one of the highest priority tasks of CODAC by issuing the Plant Control Design Handbook (PCDH) to ensure ITER I&C systems will contribute and be instrumental in making ITER project a success. In this paper we will elaborate from a technical perspective on what the CODAC team defined in the current version of PCDH (v5) for this standardization process.

ITER requires extensive diagnostic systems in order to meet the requirements for machine operation, protection, plasma control and physics studies. The realization of these systems is a major challenge not only because of the harsh environment and the nuclear requirements but also with respect to Instrumentation and Control (I&C) of all the 59 diagnostics plants. The Plant Systems I&C are mostly “in-kind”, i.e. procured by the seven ITER Domestic Agencies (DAs), while the Central I&C Systems are “in-fund”, i.e. procured by ITER Organization (IO). Standardization of Plant Systems I&C is of primary importance and has been one of the highest priority tasks of CODAC. The standards are published in the Plant Control Design Handbook (PCDH) [1] which will be followed to ensure a homogeneous design, guarantee high availability and simplify maintenance and support future upgrades. Most important for a successful commissioning and operation of the ITER facility are the concepts of interfacing the diagnostics plant systems with CODAC and the standards for instrumentation and control which must be followed all contributing parties. In this paper, we will elaborate on the concepts of interfacing the diagnostics plant systems with CODAC and the standards that must be followed for the design.

The control system of ITER consists of thousands of computers processing hundreds of thousands of signals. The control system, being the primary tool for operating the machine, shall integrate, control and coordinate all these computers and signals and allow a limited number of staff to operate the machine from a central location with minimum human intervention. The primary functions of the ITER control system are plant control, supervision and coordination, both during experimental pulses and 24/7 continuous operation. The former can be split in three phases; preparation of the experiment by defining all parameters; executing the experiment including distributed feed-back control and finally collecting, archiving, analyzing and presenting all data produced by the experiment. We define the control system as a set of hardware and software components with well defined characteristics. The architecture addresses the organization of these components and their relationship to each other. We distinguish between physical and functional architecture, where the former defines the physical connections and the latter the data flow between components. In this paper, we identify the ITER control system based on the plant breakdown structure. Then, the control system is partitioned into a workable set of bounded subsystems. This partition considers at the same time the completeness and the integration of the subsystems. The components making up subsystems are identified and defined, a naming convention is introduced and the physical networks defined. Special attention is given to timing and real-time communication for distributed control. Finally we discuss baseline technologies for implementing the proposed architecture based on analysis, market surveys, prototyping and benchmarking carried out during the last year.

Up to six mock-ups of different tritium breeding blanket systems, referred to as Test Blanket Systems (TBS), will be tested in three equatorial ports of ITER. The paper describes the recent studies performed by the IO for the preparation of the most urgent interfaces required for the integration of the TBSs into the ITER device and Tokamak complex. Main addressed items concern the impact of the TBM ferromagnetic structural material on plasma performance, the preliminary design of the TBM frame, the main dimensions and location for the TBM cooling systems and for the Tritium circuits, and preliminary requirement for measurement systems. A large effort has also been devoted to the proposed maintenance/refurbishment procedure and to the space and equipment requirements in the hot cell facility, including provisions for the shipping of TBM sub-components outside the ITER site for post-irradiation examinations.

The ITER instrumentation and control (I&C) system is the term encompassing all hardware and software required to operate ITER. It has two levels of hierarchy: the central I&C systems and the plant systems I&C. The central I&C systems comprise CODAC (Control, Data Access and Communication), the central interlock system (CIS) and the central safety systems (CSS). The central I&C systems are “in-fund”, i.e. procured by ITER Organization (IO), while plant systems I&C are “in-kind”, i.e. procured by the seven ITER domestic agencies. This procurement model, together with the current estimate of 161 plant systems I&C, poses a major challenge for the realization and integration of the ITER I&C system. To address this challenge a main strategic focus of the CODAC group, formed in 2008, has been to establish good relations with the domestic agencies. By distributing the required R&D tasks and contracts fairly between the domestic agencies we build collaborations for the future at the same time as technical work proceed. The primary goal of ITER I&C system is to provide a fully integrated and automated control system for ITER. Standardization of plant systems I&C is of primary importance and has been the highest priority task during the last year. The target of associated R&D activities is to survey, benchmark and prototype main stream technologies, in order to choose the best and most widely used technology standards for plant systems I&C. In this paper we elaborate on our approach, both from a technical and a non-technical perspective, explain technology evaluation and decisions and finally present the way forward to ensure ITER I&C system will contribute and be instrumental in making ITER a success.

The operation of a complex experimental machine like ITER will involve a number of potential hazards to personnel, the environment, and to the machine itself. While some protections are usually embedded within the overall control system, when it comes to the protection of people, the environment or the safe operation of the machine, dedicated systems are required. At ITER, the safety control systems are dedicated to the protection of people and the environment. These systems represents one of the three independent tiers on which the ITER Instrumentation and Control is based. They have to respect stringent requirements in terms of reliability, availability, safety and maintainability for operation, security and national and/or international safety regulations. This paper describes the current status and plans of the safety control systems, the functions to be performed, the envisaged architecture and the main design options including the principles of separation and independence between the three tiers.

CODAC orchestrates the activity of 60–90 Plant Systems in normal ITER operation. Interlock Systems protect ITER from potentially damaging operating off-normal conditions. Safety Systems protect the personnel and the environment and will be subject to licensing. The principal challenges to be met in the design and implementation of CODAC include: complexity, reliability, transparent access respecting security, a high experiment data rate and data volume since ITER is an experimental reactor, scientific exploitation from multiple Participant Team Experiment Sites and the long 35-year period for construction and operation. Complexity is addressed by prescribing the communication interfaces to the Plant Systems and prescribing the technical implementation within the Plant Systems. Plant Systems export to CODAC all the information on their construction and operation as “self-description”. Complexity is also addressed by automating the operation of ITER and of the plasma, using a structured data description of “Operation Schedules” which encompass all non-manual control, including Plasma Control. Reliability is addressed by maximising code reuse and maximising the use of existing products thereby minimising in-house development. The design is hierarchical and modular in both hardware and software. The latter facilitates evolution of methods during the project lifetime. Guaranteeing security while maximising access is addressed by flow separation. Out-flowing data, including experimental signals and the status of ITER plant is risk-free. In-flowing commands and data originate from Experiment Sites. The Cadarache Experiment Site is equated with the Remote Experiment Sites and a rigorous “Operation Request Gatekeeper” is provided. The high data rates and data volumes are handled with high performance networks. Global Area Networks allow Participant Teams to access all CODAC data and applications. Scientific exploitation of ITER will remain a human as well as technical challenge, to find methods of effectively combining the experience and observations of a geographically distributed research team.

The large tokamak fusion device JT-60 is expected to explore more advanced tokamak discharge scenarios toward the ITER and a future power reactor. Since various experimental issues are to be adequately discussed, and possibly to be solved in JT-60, the plasma real-time control system has been drastically improved with remodelling in hardware as well as in software. To satisfy the requirements, a “multiple networks” structure is employed as a basic principle of the remodelling. Distributed processors for diagnostics, actuators, and supervisory controllers are linked through a reflective memory (RM) network for fast, real-time communication. Similarly, advanced and complex calculations to reproduce plasma shape and profiles are performed by several processors connected to the same RM network. In this report, we discuss the developments to improve the JT-60 plasma control and data processing systems. In addition, a future plasma control system leading to a standard design for a power reactor is outlined, which is based on the 20-year plasma operation experience.

The JT-60U control system, composed of a UNIX-based workstation and a VME-bus system, has been successfully performing advanced plasma control for the JT-60U plasma discharges. Various hardware modifications and advanced plasma control algorithms have been continuously integrated into the JT-60U control system since its upgrade in 1991. The ZENKEI, the supervisory control system, has been entirely converted from its original 16-bit minicomputers and CAMAC system to the combination of a workstation and VME-bus system linked through an Ethernet network. The plasma equilibrium control system utilizes a full VME-bus system composed of three DEC alpha CPU modules that are currently running at control cycle times of 250 μs. The parameters for plasma equilibrium control are calculated by a newly developed real-time plasma shape calculation and reconstruction system. These parameters are fed to the plasma equilibrium control system synchronized by 250-μs clock pulses. To progress toward the advanced plasma control concept, which includes adoption of a global plasma parameter control algorithm and suppression of MHD instabilities, improvements were extensively consolidated into the plasma control system (PCS). Currently, additional modifications of the JT-60U control system have extended the pulse duration of the discharge without a major change in JT-60 equipment.

The present CAMAC based timing system has been used for synchronizing sequential events of the discharge and the data collection of the interesting JT-60U experiment plasma phenomena. However, a more flexible and sophisticated state-of-the-art timing system now is required to realize advanced plasma control with minimal maintenance costs. In this context, the versa module Europe (VME-bus) system with a high-speed data communication network using reflective memory (RM) modules and user-friendly application software based on MATLAB™ tools has been selected to develop the new prototype timing system. In the ZENKEI, the supervisory control system of the JT-60, the supervisory timing system provides the 50-μs master clock pulses, the various timing signal preparation logic, which is built into the digital signal processing (DSP) module in conjunction with the discharge sequence event signals, and the 6.2 MB/s high-speed communication data link provided by the RM module. Except the clock pulse generator (CPG) module, no other special timing module is necessary for this new timing system. The timing signal is prepared by software logic in conjunction with sequential events and the preset timer, is transferred to the subsystems through the RM module, where it is synchronized to the 50-μs clock pulses. The timing system of the subsystems also consists of hardware similar in structure to the ZENKEI timing system. The fundamental timing system configuration, the necessary functions, and the preliminary test results of the prototype system are reported in this presentation.

The old discharge control system (DCS) with HIDIC-80E mini-computer has been replaced by a control system with a unix workstation and a VME-bus system. Functionally, the workstation performs message communication with VME controller on the network, data processing using plant data and data acquisition. The commands and receiving the status data in real time. The new DCS is flexible for system extension and able to quickly respond to various requirements of software modifications. The data acquisition speed has been improved by almost a factor of two over the old one. Troubles and countermeasures experienced in the replacement work are also presented.

Remodeling of the JT-60 discharge control system (DCS) is initiated to minimize the cost and time for hardware maintenance and software development. Three principles have been adopted for remodeling. (1) The functions of the DCS must remain intact. (2) The shutdown period must be minimized to maximize JT-60 experiment time. To realize the new DCS, (3) a combination of the host workstation (WS) and target Versa Module Europe (VME)-bus system with real time OS was chosen as the fundamental structure. The prototype DCS was designed and tested to evaluate the performance of the converted discharge sequential control software modules and the message communication program. The new configuration of the DCS and procedure of smooth remodeling without shutdown is described in detail.

Features incorporated in the design of the International Thermonuclear Experimental Reactor (ITER) tokamak and its ancillary and plasma diagnostic systems that facilitate operation and control of ignited and/or high?Q DT plasmas are presented. Control methods based upon straightforward extrapolation of techniques employed in the present generation of tokamaks are found to be adequate and effective for ITER plasma control with fusion powers of up to 1.5?GW and burn durations of ? 1000?s. Examples of simulations of key plasma control functions, including plasma magnetic configuration control and fusion burn (power) control, are given. The prospects for the creation and control of steady state plasmas sustained by non-inductive current drive and bootstrap current are also discussed.

Plasma control requirements for the International thermonuclear Experimental Reactor (ITER) are identified, and an overview of proposed ITER plasma control concepts is presented. ITER will operate with a burning deuterium-tritium plasma to produce 1.5 GW of fusion power for durations of 1000 s or more. Key plasma control requirements to achieve these objectives encompass (a) plasma scenario and sequencing: plasma initiation, current rampup, divertor formation, auxiliary heating, ignition and burn, deignition (fusion power shutdown), and current rampdown and termination; (b) plasma magnetics control: plasma current and shape (R 0 , a, κ, δ) versus time, plus control of critical plasma-to-first-wall clearance gaps, including ion-cyclotron coupling gap and divertor magnetic configuration, during the diverted heating/ignition/burn/deignition phase of the plasma scenario; (c) plasma kinetics and divertor control: core plasma density and/or fusion power, core impurity content and/or radiated power fraction; core profile control (auxiliary heating and/or current drive), and divertor control (pumping, in-divertor gas and/or impurity injection and magnetic configuration optimization for divertor performance); and (d) fast plasma shutdown: fusion power and current shutdown by means of impurity injection. Physics and hardware concepts are presented as to how these plasma control junctions will be implemented. Diagnostic measurements needed for plasma control are summarized. The relationship of plasma control to machine protection and public safety is also addressed.

ITER is a large-scale nuclear fusion scientific experiment; the first to produce net energy. Its control and instrumentation will be complex, unique and challenging. Integrated and automated operation will require coordination of around 220 semi-autonomous plant system instrumentation & controls (IC the central layer that ensures coordination and the local layer dedicated to the individual control of each plant system. Specifically, the I&C of the ITER coil power supply system is described, this system will be the main actuators to control the plasma shape and position. Unlike past and present fusion experiments operating in tens of seconds pulse mode, ITER will have plant systems operating in steady state mode with plasma pulse lengths running from a few hundred seconds to ultimately some tens of minutes. The paper highlights the challenges, techniques and strategies of the distributed control systems. (6 pages)

Improvement of an enhanced confinement state in a high poloidal beta (βp) regime without sawtooth activity has been achieved in JT-60U. A confinement mode has been demonstrated where both the edge and the core confinement are improved. The attainable βp was also extended to higher values in this improved mode, because of its broader pressure profile. As a result of the improvement in confinement and in attainable βp, the highest value of the fusion triple product has been extended by a factor of 2.5 over that achieved in the 1992 experiments; it has reached (1.1 ± 0.3) × 1021 m-3.s.keV with a central ion temperature of about 37 keV. The D-D neutron emission rate has also been doubled in these experiments and has reached (5.6 ± 0.6) × 1016 s-1

CODAC (Control, Data Access and Communication) is the central control system responsible for operating the ITER device. CODAC interfaces to more than 160 plant systems containing actuators, sensors and local control. CODAC is responsible for coordinating and orchestrating the operation of these plant systems including plasma feedback control. CODAC is developed by ITER Organization, while plant systems are developed by the seven ITER parties (China, Europe, India, Japan, Korea, Russia and United States). This procurement model poses enormous challenges, has a big impact on architecture design and requires a strong standardization for better integration and future maintenance. In this paper we briefly describe the CODAC conceptual design, elaborate on the actions taken by the CODAC team to move from conceptual to engineering design during the last year and outline the plans ahead.

Recent lower hybrid current drive (LHCD), and heating (LHH) experiments on JT-60 are reported. The current drive product of neRpIRF approximately 12.5*1019 m-2 MA was achieved at the LH power of approximately 4.5 MW, and the CD efficiency, the energy confinement, the global power balance and the heat load on divertor plates were investigated in high power LHCD plasmas. Nearly steady state H-mode discharges were found during LHCD with two different frequency injections. Sawtooth suppression in NB heated plasmas by LHCD have shown an improvement in confinement near the plasma center. Parametric instabilities in LH heating experiments were significantly reduced by increasing the plasma current, and the stored energy increased linearly with heating power of up to approximately 9 MW at ne approximately 7*1019 m-3 and Ip=2.75 MA. Parametric instabilities near the plasma edge in the ion heating regime were also reduced in peaked density plasmas produced by pellet injection and LH waves increased the central plasma pressure at ne(0) > 1.4*1020 m-3.

The major challenges facing the COntrol, Data Access and Communication system (CODAC) for ITER have been identified. Workable approaches to many of these challenges have been developed over the last two years and no specific technical problem appears to be missing an adequate technical solution. The major challenge not demonstrated is for the facility availability not to be significantly reduced by CODAC. The conceptual CODAC design is in the process of being reviewed in 2007, to open the way for a preliminary design.

A new digital integrator with high input voltage and a long-time low drift speed has been developed in JT-60. This integrator uses an integrating method with a pair of a voltage-to-frequency converter (VFC) and an up and down counter (UDC) (VFC-UDC unit). However, this method has a saturation of the VFC for large input, leading to integration error. To improve it, the new digital integrator is composed of three sets of VFC-UDC units in parallel and a digital signal processor (DSP). Three VFC-UDC units with different input ranges integrate an identical input signal respectively, and the DSP selects a suitable integrated signal among three integrated outputs at a sampling frequency of 10 kHz and makes a chain of integrated signals. The performance of the integrator has been tested using a disruptive discharge in JT-60. A good integration result has been obtained though large signal is input. Also, fundamental characteristics of the VFC, linearity, thermal drift, dead band, which are causes of integration errors, are described.

The authors describe the design and initial operation of a new JT-60 (JAERI Tokamak-60) man-machine interface system built in workstations. Development of the system was stimulated by the requirements of making of JT-60 operator interface more friendly on the basis of the past five years of operational experience. Eleven Sun/3 workstations and their supervisory minicomputer HIDIC V90/45 are connected through the standard network, Ethernet. The network is also connected to the existing ZENKEI minicomputer system through the shared memory on the HIDIC V90/45 minicomputer. Improved software, such as automatic setting of the discharge conditions, consistency check among the related parameters, and easy operation for discharge result data display, provides the user-friendly environments. This man-machine interface system leads to the efficient JT-60 upgrade operation. >

ITER is a long-pulse tokamak with elongated plasma. The nominal inductive operation produces a D-T fusion power of 500 MW for a burn length of 300-500 s, with the injection of 50 MW of auxiliary power. With non-inductive current drive from the HC conventional control, interlock system and safety system, and two horizontal layers; central I&C systems and plant system I&C. CODAC (Control, Data Access and Communication) system forms the upper level of the hierarchy, and is the conventional central control system of ITER architecture. CODAC system is responsible for integrating all plant system I&C and enable operation of ITER as a single integrated plant. CODAC system provides overall plant systems coordination, supervision, plant status monitoring, alarm handling, data archiving, plant visualization (HMI) and remote experiment functions. CIS (Central Interlock System) and CSS (Central Safety System) also form the upper level of the hierarchy to supervising and integrating all plant system interlock and safety functions. Plant system I&C forms the lower level of the hierarchy, and provide dedicated plant data acquisition, plant status monitoring, plant control and plant protection functions to perform individual plant system operation under the supervision of central I&C systems.

Emphases in JT-60 experiments are placed on (1) lower-hybrid (LH) current drive characteristics with a multi-junction type launcher, and (2) the confinement study with combination of neutral beam injection LH current drive and pellet injection. The new multi-junction LH launcher provides a sharp N/sub /// spectrum with high directivity for N/sub ///=1-3.4. The current drive efficiency and the radial distribution of high energy electron production show clear correlation with injected N/sub ///: the current drive efficiency has the maximum at low N/sub ///( approximately 1.3) while flattening of plasma current is more effective in large N/sub ///. A broad radial distribution of high energy electron current and approximately 30% reduction in sawtooth inversion radius were obtained by high N/sub /// ( approximately 2.5) LH injection. To fully suppress the sawtooth activity, low N/sub /// ( approximately 1.3) injection was found to be more effective. Improved energy confinement has been obtained with hydrogen pellet injection. Energy confinement time was enhanced up to 40% relative to usual gas fuelled discharges. The discharge has a strongly peaked electron density profile with ne(0)/(ne) approximately 5 and ne(0) approximately 2.0*1020 m-3. The improved discharges are characterized by a strongly peaked pressure profile within the q=1 magnetic surface, and degrades when a large sawtooth recovers or the pressure gradient may reach a critical value. When large (3 mm, 4 mm) and fast (2.2 km/s) pellets were injected, 30% energy confinement improvement was obtained even during the NB heating of 14 MW. Further investigations of IDC characteristics have been made. The oxygen impurity lines from the main plasma and the main radiative loss drop first. Then the plasma stored energy starts to rise. The particle recycling is reduced around the main plasma, and is localized in the neighborhood of the X-point with a time lag of approximately 0.2 sec. Eventually the discharge shows a significant remote radiative cooling power at the divertor region.

Initial ohmic heating experiments in JT-60 were performed for a three month period of April-June 1985. A maximum plasma current of 1.6 MA was obtained for both divertor and limiter discharges. Low-q discharges of qeff=2.5 and high density discharges of 4.8*1019 m-3 in line-averaged density were obtained in the divertor configuration. In divertor discharges radiated loss from the main plasma can be kept at 20-30% of the ohmic input.

This paper describes the central control system of the JT-60 tokamak. This control system not only has to organize plasma operation, but also has to utilize devices under isolated operation. Therefore the control system has been designed in a special way.

Essential divertor functions — density and energy control — have been investigated in the JT-60 divertor with metal walls. Neutral pressure in the divertor chamber increases in proportion to n2e and high recycling state for particle exhaust is realized. At ne ≤ 6 × 1019m−3, wall pumping by the divertor plates is dominant in particle exhaust. However, at ne ≥ 6 × 1019m−3, particle exhaust by active divertor pumping systems becomes essential. Because of the effectiveness of the divertor, radiation loss in the main plasma is reduced to 5–10% of the absorbed power and impurity concentrations are significantly suppressed at very low level (Zeff = 1.5-2.0). In the experiments with graphite walls, short periods of H-mode phases were found in the outside X-point divertor discharges although improvement in energy confinement is limited to within 10%. In limiter discharges, combination of high vessel temperature (210–300 °C) and low current limiter discharges (1.5 MA) without gas puffing is effective as a wall conditioning in obtaining reproducible high density discharges. The maximum parameters, Ip = 3.2 MA, ne = 1.2 × 1020m−3, qeff = 2.2 at Bt = 4.8T, a Murakami parameter of 7.5, and a stored energy of 3.1 MJ have been attained. The maximum density is limited by the occurence of the MARFE. During high power neutral beam heating, enhanced carbon influx was observed possibly due to chemical sputtering. The wall pumping by the inner wall is still effective in decreasing the electron density from the high density region of ne ~ 1 × 1020m−3. An empirical energy confinement scaling of JT-60 has been drawn in terms of an offset linear function as τE = 0.19 I1.9PPabs + 0.062 a1.8p

From June to October 1987, JT-60 achieved fusion product (ne(0). tau E*.Ti(0)) of 6*1019 m-3.keV.s with hydrogen plasma at plasma current of 2.8 to 3.1 MA with neutral beam power of approximately 20 MW. The central electron density of 1.3*1020 m-3 was obtained at plasma current of 3 MA with 13 approximately 20 MW neutral beam power and the confinement time reached 0.14-0.18 s. An offset linear scaling law like the Shimomura-Odajima scaling on confinement time will be able to reproduce experimental data better than that of the Goldston type scaling. With low beam energy injection approximately 40 keV, confinement degradation was found. Many short periods (0.05 approximately 0.1 s) of H-mode phase were found in outside X-point divertor discharges with NB or NB+RF(LH or IC) heating power above 16 MW. However, improvement in energy confinement time was limited to 10 %. The ballooning/interchange stability analyses were also made for the outside X-point divertor equilibrium in connection with H0-phase capability. Heating powers of 9.5 MW and 1.9 MW were obtained by LHRF, ICRF injection, respectively. In combined LHRF and NB heating, the incremental energy confinement time of 0.064 s was obtained, which is the same level of that of NB heating only. In combined NB and on-axis ICRF heating of low ne discharge, an incremental energy confinement time of 0.21 s was obtained, which is three times as long as those of NB or ICRF heating only. High energy beam ions were accelerated by ICRF in the central region of the plasma.

In the last year, ohmic heating experiments were performed in JT-60 for about three months; May–June 1985 and March 1986. The major experimental effort has been concentrated on the divertor configuration in anticipation of H-mode operation of JT-60 in high power supplementary heating experiments starting in August 1986. Stable discharges with a divertor have been obtained at Ip = 1.85 MA and ne = 5.7 × 1019m−3. Radiated power from the main plasma in divertor discharges is reduced to one third of that in material limiter discharges. The electron density has a broad radial profile and the effective safety reaches qeff = 2.5. The neutral gas pressure in the divertor chamber becomes significantly high and the control of particle recycling has been attempted.

Real-time plasma control is planned to be executed in JT-60 experiments. The object of which is to obtain the Q=l plasma, protecting the JT-60 machine from the electromagnetic force, heat load, etc. In this report we present the design of this control system.

The control system of JT-60 consists of several computer-CAMAC combinations, timing systems and hard-wired interlock systems. The functions of the control system are roughly represented by discharge control including real-time feedback control, plant operation support and safety interlock.

Burning plasma simulation experiments were performed for bum control study on ELMy H-/L-mode plasmas and reversed shear (RS) plasmas with an internal transport barrier in JT-60U. In a burning plasma simulation scheme, two neutral beam (NB) groups were used: one that simulates alpha-particle heating and another that simulates external heating. For the alpha-particle heating simulation, the heating power proportional to the deuterium-deuterium (D-D) neutron yield rate was injected. The behavior of the part of the NB heating simulating alpha-particle heating was varied by increasing the proportional gain relating the applied power to the measured neutron yield rate in both ELMy H-mode and RS plasmas, while the part of the NB power in the role of external heating was held constant i.e., no-bum-control case. Above a certain value of the proportional gain, a runaway effect was triggered where excursive increases in the neutron yield rate and stored energy were observed. With burn control, where the stored energy was controlled at a constant value by a feedback control system using the external heating, the runaway was not triggered, and the neutron yield rate was kept at a constant value in the L-mode plasmas. Zero-dimensional calculation indicated that the runaway triggered by increasing the proportional gain well simulates the runaway triggered by improved confinement. The limitations came from differences between deuterium-tritium and D-D plasmas, such as the dominant reaction for the neutron yield and the temperature dependence of the fusion reaction rate, which were discussed together with improvement on the burning plasma simulation scheme.

CODAC orchestrates the activity of 80–120 Plant Systems in normal ITER operation using a set of dedicated networks. Interlock systems protect ITER from potentially damaging operating off-normal conditions, using a segregated network. Safety systems protect the personnel and the environment and the nuclear safety systems will be subject to licensing. The principal challenges to be met in the design and implementation of CODAC include: complexity, reliability, transparent access respecting security, a high experiment data rate and data volume since ITER is an experimental reactor, scientific exploitation from multiple participant team experiment sites and the long 35-year-period for construction and operation. Two non-technical challenges drive the CODAC design. Firstly all technical plant systems follow an “in-kind” procurement model, as opposed to a direct commercial contractual relationship with suppliers. Secondly, the period of procurement will, itself, be long compared with the evolution of technologies in the fields of control and data acquisition.This paper addresses the various phases from definition of procurement agreements through to on-site commissioning of the project.