ECRH for JET: A feasibility study

For JET to fulfil its mission in preparing ITER operation, the installation of an electron cyclotron resonance heating system on JET would be desirable. The study described in this paper has investigated the feasibility of installing such a system on JET. The principal goals of such a system are: current drive over a range of radii for NTM stabilization, sawtooth control and current profile tailoring and central electron heating to equilibrate electron and ion temperatures in high performance discharges. The study concluded that a 12 gyrotron, 10 MW, system at the ITER frequency (170 GHz) adapted for fields of 2.7–3.3 T would be appropriate for the operation planned in JET. An antenna allowing toroidal and poloidal steering over a wide range is being designed, using the ITER upper launcher steering mechanism. The use of ITER diamond windows and transmission line technology is suggested while power supply solutions partially reusing existing JET power supplies are proposed. Detailed planning shows that such a system can be operational in about 5 years from the time that the decision to proceed is taken. The cost and required manpower associated with implementing such a system on JET has also been estimated.     

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.     

Tiles chamfering and power handling of the MK II HD divertor

The JET high triangularity (δ, HD) divertor is an upgrade of the present JET divertor consisting of two modified toroidal segments which are: a new load bearing septum replacement plate (LB-SRP) tile located in the center of the divertor and a high field gap closure (HFGC) tile protecting inboard diagnostic cabling. The aim of the upgrade is to allow high power operation and a wider range of plasma triangularities at the divertor poloidal null. This paper describes the optimisation of the tile chamfering (including edge shadowing) and the power handling evaluation for a set of 12 planned plasma configurations given by the JET team and on two sets of mechanical tile tolerances issued by the JET drawing office. The PROTEUS code (magnetic equilibrium by finite element) is used to calculate the various field line angles, which are inputs for the chamfering angle calculation process. After calculating the chamfering angle values of each face, a checking exercise has been realised on the 3D CATIA models of the tiles by putting them at their extreme tolerance positions and validating if the shadowing is ensured for a angle calculated to take into account the worst possibilities. With the final chamfering angle value for each face, the power handling of the tiles has been estimated with finite element calculations. Power handling is given either by the critical time to reach 1800 °C at the tile surface for a total injected power of 40 MW or by the maximum total injected power allowable for a 10 s power pulse without exceeding 1800 °C. The estimated power handling gives promising results in regard to the JET EP project objectives.     

The JET technology program in support of ITER

This paper presents an overview of the current and planned technological activities at JET in support of ITER operation and safety. The scope is very broad and it ranges from analysis of components from the ITER-like Wall (ILW) to determine material erosion and deposition, dust generation and fuel retention to neutronics measurements and analyses. Preliminary results are given of the post-mortem analyses of samples exposed to JET plasmas during the first JET-ILW operation in 2011–2012, and retrieved during the following in-vessel intervention. JET is the only fusion machine capable of producing significant neutron yields, up to nearly 10¹⁹ n/s (14.1 MeV) in DT operations. Recently, the technological potential of a new DT campaign at JET in support of ITER has been explored and the outcome of this assessment is presented. The expected 14 MeV neutron yield, the use of tritium, the preparation and implementation of safety measures will provide a unique occasion to gain experience in several ITER relevant technological areas. A number of projects and experiments to be conducted in conjunction with the DT operation have been identified and they are described in this paper.     

A shared memory based interface of MARTe with EPICS for real-time applications

The Multithreaded Application Real-Time executor (MARTe) is a multi-platform C++ middleware designed for the implementation of real-time control systems. It currently supports the Linux, Linux + RTAI, VxWorks, Solaris and MS Windows platforms. In the fusion community MARTe is being used at JET, COMPASS, ISTTOK, FTU and RFX in fusion [1].The Experimental Physics and Industrial Control System (EPICS), a standard framework for the control systems in KSTAR and ITER, is a set of software tools and applications which provide a software infrastructure for use in building distributed control systems to operate devices.For a MARTe based application to cooperate with an EPICS based application, an interface layer between MARTe and EPICS is required. To solve this issue, a number of interfacing solutions have been proposed and some of them have been implemented. Nevertheless, a new approach is required to mitigate the functional limitations of existing solutions and to improve their performance for real-time applications.This paper describes the design and implementation of a shared memory based interface between MARTe and EPICS.     

Commissioning and operation of 130 kV/130 A switched-mode HV power supplies with the upgraded JET neutral beam injectors

The design features, on-site testing, commissioning and operation are described of two new 130 kV/130 A HV power supply units serving four upgraded 130 kV/60 A positive ion neutral injectors (PINIs) on JET. Both units were factory tested at full power and pulse length into dummy resistive load. Following on-site installation, the factory tests were repeated. The transition from dummy-load testing to PINI operation required full integration of the HVPS within the overall JET control system, and rigorous testing of the co-ordinated actions and protections of all PINI power supplies (filament and arc for plasma source and negative suppression grid). The implementation of these functions is described. Extensive use was made of parasitic integrated test pulses, where the other PINIs could be operated normally, with the HVPS energised under full remote control together with the corresponding PINI plasma sources, but with the HVPS connected to dummy load. The amount of NB operation time dedicated to commissioning was thereby minimised, yet gave a high degree of confidence of readiness for HV energisation of the PINI, and first beam operation followed less than 24 h from HV connection to the PINI. The routine operating experience and performance, including load protection characteristics of the new HVPS units are also described.     

New signal processing methods and information technologies for the real time control of JET reactor relevant plasmas

A general trend in the experimental programmes of present day Tokamaks, and of JET in particular, is the constant increase in the number of parameters to be controlled in real time, to satisfy the machine protection requirements on the one hand and to improve performance on the other. Since the amount of data collected is also increasing at least at a rate compatible with the Moore law, significant developments are required in the field of real time algorithms particularly for magnetic reconstructions, disruption prediction and image processing. A new real time equilibrium code called EQUINOX, using internal and external measurements of the magnetic fields, has been qualified on JET. It can provide reconstructed accurate equilibria about every 50 ms on a 2 GHz PC. An advanced disruption predictor, based on machine learning tools, has been deployed using inputs selected with a genetic algorithm. Its success rate remains of the order of 94% for up to 170 ms before the occurrence of the disruption. Nonextensive entropies, which are more sensitive to long range correlations, seem to be useful in detecting vibrations in the videos of JET cameras, both visible and infrared.     

Commissioning of the ITER-like ICRF antenna for JET

The new JET ion cyclotron resonance frequency (ICRF) ITER-like antenna (ILA), which was assembled during 2006, was commissioned on the JET RF testbed prior to installation on the JET torus. The 4 resonant double loops (RDL) of the ILA were tested at high power at 42 MHz up to 42 kV for 5 s in 10 min intervals. Low power matching studies using a saltwater load placed in front of the ILA have allowed testing and optimizing proposed matching algorithms on single RDLs, paired RDLs and finally on the full array. The upper limit of the frequency range of the ILA appears to be limited to 47–49 MHz due to the effect on the electrical lengths of the connection between the capacitors and the conjugate T point. Capacitor position scans have allowed obtaining the necessary data to confirm the RF model of the RDL which is necessary for the scattering matrix arc detection. The latter is deemed necessary in order to detect arcs at the low impedance conjugate T of the circuit. The antenna was installed onto JET during August 2007 and commissioning on plasma started May 2008. At present the commissioning of the ILA on JET is ongoing in a series of dedicated experimental campaigns.     

Development and application of high volume remote handling systems in support of JET and ITER

The Joint European Torus (JET) Remote Handling System has evolved from a small scale maintenance capability to one of high efficiency large volume installations. The Enhanced Performance 2 shutdown 2010–2011 for example, required the replacement of many thousands of components ranging from about 100 g to 130 kg in weight. The scale of this type of operation and the necessity to maximise operational availability intensified the demands for high productivity whilst maintaining the necessary high standards for precision, reliability, cleanliness, and operational security.This paper discusses the developments in design, control, maintenance, preparation and operation of the current state of the art remote handling facilities at JET. It explores how the experience of over 20,000 h of operations has developed the applied methodology and how this could be appropriate to ITER and other facilities requiring complex remote maintenance, where extensive, high productivity remote handling operations will be essential. It also discusses the advances that have been made in management and presentation of operational data within the command, control and human machine interfaces (HMI) systems, along with the supporting operational databases.     

Optimization of MGI in JET using the TOKES code and mitigation of RE damage for the first wall

Simulations of massive gas injection (MGI) into the JET plasma using noble gas (NG) have been performed using the TOKES code. The results of these simulations have been verified by comparison with JET experiment for disruption control. Further simulations performed showed that the amount of NG for triggering the thermal quench can be reduced 40 times comparing with the JET experiment. Such small amount of injected NG should generate runaway electrons (RE) with rather large probability. For mitigation of the wall damage with these RE, a special sacrificial diaphragm consisting of several W ‘nails’ with a characteristic thickness of 2 cm has been proposed. The diaphragm erosion with RE beam has been roughly estimated from above as less than 0.5 cm per disruption. Taking into account the plasma shielding may decrease erosion ∼10 times.     

Recent contribution of JET to the ITER physics

In recent years the JET scientific programme has focussed on addressing physics issues essential for the consolidation of design choices and the efficient exploitation of ITER in parallel to qualifying ITER operating scenarios and developing advanced control tools. This paper reports on recent achievements in the following areas: mitigation of edge localised modes (ELMs), effects of toroidal field (TF) ripple, advanced tokamak scenarios, material migration and fuel retention. Active methods have been developed to mitigate ELMs without adversely affecting confinement. A systematic characterisation of the edge plasma, pedestal energy and ELMs, and their impact on plasma-facing components as well as their compatibility with material limits has been performed. The unique JET capability of varying the TF ripple from its normal low value δB_T = 0.08% up to δB_T = 1% has been used to elucidate the role of TF ripple on confinement and ELMs. Increased TF ripple in ELMy H-mode plasmas is found to have a detrimental effect on plasma stored energy and density, especially at low collisionality. The development of ITER advanced tokamak scenarios has been pursued. In particular, β_N values above the ‘no-wall limit’ (β_N ～ 3.0) have been sustained for a resistive time. Gas balance studies combined with shot-resolved measurements from deposition monitors and divertor spectroscopy have confirmed the strong role of fuel co-deposition with carbon in the retention mechanism through long-range migration and also provided further evidence for the important role of ELMs in the material migration process within the JET inner divertor leg.     

Performance of upgraded JET neutral beam injectors

The JET neutral beam injection (NBI) system is undergoing an upgrade of both beam power and pulse duration, which will be completed in 2011. In order to obtain an early assessment of the performance of the upgraded injectors, two positive ion neutral injectors (PINIs) with modified ion source and accelerator configuration were installed on Octant 8 Neutral Injector Box and successfully commissioned in summer 2009. Both PINIs were routinely delivering ～2 MW of deuterium neutral beam power during the JET experimental campaign in autumn 2009. These early tests allowed us to predict with confidence that the JET NBI upgrade objective of injecting 34 MW of total deuterium neutral beam power into the JET plasma will be achieved.     

Fusion technology activities at JET: Latest results

The JET Task Force Fusion Technology (TF-FT) was launched in 2000 to use the unique capabilities, facilities and operating experience at JET to provide significant contributions to the research programme on both JET and ITER. This paper presents the most recent results obtained within the JET TF-FT programme.The Tritium (T) retention measurements have confirmed high surface but little bulk T concentrations on the MKII-SRP divertor tiles and T thermal desorption tests confirmed the necessity to reach at least 600 °C. From the 2007 shutdown the MKII-HD (more ITER like) divertor has revealed some slight changes in the nature of the erosion/deposition. In order to improve analysis, time resolution devices such as quartz micro-balances and rotating collectors have been located beneath the divertor for deposition and plasma physics correlations. Due to improvement of dedicated models and technologies, in situ laser techniques for detritiation and characterisation/removal have provided encouraging results on quantitative characteristics (composition, thickness, adherence, temperature) of deposited films on plasma facing components. A particular effort on temperature control of the new metallic ITER-like wall (ILW) that is presently being installed in JET has been pursued with active laser infrared thermography. JET TF-FT also contributes to the operator strategy to comply with the safety agency requirements for T management. Recent results on two major topics purification of tritiated water and development of the ³He method for the determination of the T concentration in waste drums are presented. Finally, this paper also presents some activities in preparation of the ILW for the pre-characterisation of marker tiles and the refurbishment of diagnostics for deposition characterisation.     

The impact of thermal fatigue and carbidization on the W coatings deposited on CFC tiles for the ITER-like Wall project at JET

Since August 2011 JET operates with the ITER-like wall comprising bulk Be tiles, bulk W tiles and W coated CFC tiles with a thickness of 10–15 μm and 20–25 μm. In order to evaluate behavior of the W coatings to a cyclic thermal loading relevant to JET operation, high heat flux (HHF) tests have been carried out up to 5100 pulses with an electron beam facility at peak temperatures of 1000 °C, 1250 °C and 1450 °C. The pulse duration was 24 s. Optical inspections of the W layer performed periodically by interrupting the test revealed small delaminations with the size of 50–500 μm. The dependence of the delamination percentage on the number of pulses can be seen as a degradation curve for each particular W coating. In this way the thermo-mechanical properties of the W coatings can be characterized quantitatively. Thermal fatigue and carbidization of the tungsten due to the diffusion of the carbon from the substrate have been recognized as mechanisms for degradation of the coatings. Tungsten carbides have been identified by using TEM (transmission electron microscopy) diffraction analysis on FIB (focused ion beam) prepared cross-section samples subjected to HHF tests. Nano-pores developed at the CFC–Mo and Mo–W interfaces during the tests might be also responsible for the degradation of the coating.     

A web application for poloidal field analysis on HL-2M

Recently, many web tools [1], [2], [3] in fusion society have been designed and demonstrated, which has been proved to be powerful and convenient to fusion researchers. Many physicists and engineers need a tool to compute the poloidal magnetic field for some purposes (for example, the calibration of magnetic probes for EFIT, the field null structure analysis for control, the design of some plasma diagnostic systems), so to develop a powerful and convenient web application for the calculation of magnetic field and magnetic flux produced by PF coils is very important. In this paper, a web application tool for poloidal field analysis on HL-2M with a totally original framework is presented. This web application is full of dynamic and interactive interface, and can run in any popular browser (IE, safari, firefox, opera), on any hardware (smart phone, PC, ipad, Mac) and operating system (ios, android, windows, linux, Mac OS). No any plugins is needed. The three layers (jQuery + PHP + Matlab) of this framework are introduced. The front top client layer is developed by jQuery code. The middle layer, which plays a role of a bridge to connect the server and client through socket communication, is developed by PHP code. The behind server layer is developed by Matlab, which compute the magnetic field or magnetic flux through a Special Function called Complete Elliptic Integral, and returns the results in the client favorite way, either by table or by JPG image. The field null structure and the vertical and radial field structure calculated by this tool are introduced with details. The idea to design a web tool with jQuery + PHP + Matlab framework may apply to other machines. The address for this application is http://dp.swip.ac.cn/hl2m.     

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.     

Application of HDF5 in long-pulse quasi-steady state data acquisition at high sampling rate

A new high sampling rate quasi-steady state data-acquisition system has been designed for the microwave reflectometry diagnostic of EAST experiments. In order to meet the requirements of long-pulse discharge and high sampling rate, it is designed based on PXI Express technology. A high-performance digitizer National Instruments PXIe-5122 with two synchronous analog input channels in which the maximum sampling rate is 100 MHz has been adopted. Two PXIe-5122 boards at 60 MSPS and one PXIe-6368 board at 2 MSPS are used in the system and the total throughput is about 500 MB/s. To guarantee the large amounts of data being saved continuously in the long-pulse discharge, an external hard-disk data stream enclosure NI HDD-8265 in which the capacity of sustained speed of reading and writing is 700 MB/s. And in RAID-5 mode its storage capacity is 80% of the total.The obtained raw data firstly stream continuously into NI HDD-8265 during the discharge. Then it will be transferred to the data server automatically and converted into HDF5 file format. HDF5 is an open source file format for data storage and management which has been widely used in various fields, and suitable for long term case. The details of the system are described in the paper.     

Feasibility study of an intense D–T fusion source: “The New Sorgentina”

The international community agrees on the importance to build a large facility devoted to test and validate materials to be used in harsh neutron environments. Such a facility, proposed by ENEA, reconsiders a previous study known as “Sorgentina” but takes into account new technological development so far attained. The “New Sorgentina” Fusion Source (NSFS) project is based upon an intense D–T 14 MeV neutron source achievable with T and D ion beams impinging on 2 m radius rotating targets. NSFS produces about 1 × 10¹³ n cm⁻² s⁻¹ over about 50 cm³. Larger volumes of lower neutron flux will be available (e.g. for TBM experiments) as well as collimated channels to study some features of the ITER neutron camera. The NSFS facility will overcome problems related to the ion source and accelerating system, by means of an upgraded version of the JET–PINI ion beams. NSFS has to be intended as an European facility that may be realized in a few years, once provided a preliminary technological program devoted to the operation of the ion source in continuous mode, target heat loading/removal, target and tritium handling, inventory as well as site licensing. In this contribution, the main characteristics of NSFS project will be presented.     

A distributed timing and synchronization system for EAST

The distributed timing and synchronization system (DTSS) plays an important role in Experimental Advanced Superconducting Tokamak (EAST), which is one of the national key fusion research facilities in China. This system synchronizes each subsystem of EAST by using reference clock and trigger. A prototype DTSS module has been developed based on PXI bus and RIO (reconfigurable I/O) devices. The DTSS can provide reference clock in frequency up to 80 MHz. The trigger can be pre-defined from 1 ms to 6872 s with 10 ns accuracy. In addition, this system can acquire, process signals, and send output or command to other systems. The DTSS has been successfully applied to 2010 fall EAST experiment, and the results confirmed its accuracy and reliability. After the analysis of system requirement, the architecture of the DTSS and the technical implementation based on PXI are presented in this paper.     

High-speed,multi-input,multi-output control using GPU processing in the HBT-EP tokamak

We report on the design of a new plasma control system for the HBT-EP tokamak that utilizes a graphical processing unit (GPU) to magnetically control the 3D perturbed equilibrium state [1] of the plasma. The control system achieves cycle times of 5 μs and I/O latencies below 10 μs for up to 96 inputs and 64 outputs. The number of state variables is in the same order. To handle the resulting computational complexity under the given time constraints, the control algorithms are designed for massively parallel processing. The necessary hardware resources are provided by an NVIDIA Tesla M2050 GPU, offering a total of 448 computing cores running at 1.3 GHz each. A new control architecture allows control input from magnetic diagnostics to be pushed directly into GPU memory by a D-TACQ ACQ196 digitizer, and control output to be pulled directly from GPU memory by two D-TACQ AO32 analog output modules. By using peer-to-peer PCI express connections, this technique completely eliminates the use of host RAM and central processing unit (CPU) from the control cycle, permitting single-digit microsecond latencies on a standard Linux host system without any real-time extensions.