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.     

The WEST project: Testing ITER divertor high heat flux component technology in a steady state tokamak environment

The WEST project recently launched at Cadarache consists in transforming Tore Supra in an X-point divertor configuration while extending its long pulse capability, in order to test the ITER divertor technology. The implementation of a full tungsten actively cooled divertor with plasma facing unit representative of ITER divertor targets will allow addressing risks both in terms of industrial-scale manufacturing and operation of such components. Relevant plasma scenarios are foreseen for extensive testing under high heat load in the 10–20 MW/m² range and ITER-like fluences (1000 s pulses). Plasma facing unit monitoring and development of protection strategies will be key elements of the WEST program.WEST is scheduled to enter into operation in 2016, and will provide a key facility to prepare and be prepared for ITER.     

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 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.     

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.     

RF sources for ITER Ion Cyclotron H&CD system

The Ion Cyclotron Heating and Current Drive (IC H&CD) system for ITER will provide 20 MW to the plasma. The associated Radio Frequency (RF) source system has to be compliant with all operation modes foreseen in that frame. Their specifications are fully described in this paper and constraints on IC RF source components are detailed, in particular concerning the final stage tube of the amplifier. Results of tests performed under a collaborative work at the National Institute for Fusion Science (NIFS) facility are presented. Consequences on the procurement process by ITER India (II) are deduced.     

5 MW CW supply system for the ITER gyrotrons Test Facility

ECH (Electron Cyclotron Heating) for ITER will deliver into the plasma 20 MW of RF power. The procurement of the RF sources will be shared equally between the three following partners: Europe, Japan and Russia. Moreover, Europe decided to develop a RF source capable of 2 MW CW of RF power, based on the design of a coaxial gyrotron with a depressed collector. In order to be able to develop and test these RF sources, a Test Facility (TF) has been built at the CRPP premises in Lausanne (CH).The present paper will first remind the main operation conditions considered to test safely a gyrotron. The power supplies parameters allowing to fulfill these conditions will be reviewed. The core of the paper content will describe the newly installed Main High Voltage Power Supply (MHVPS), to be connected to the gyrotron cathode and capable of ?60 kV/80 A-CW. The principle, the characteristics, the on-site test results will be described at the light of the requirements imposed by the gyrotron testing. Particular aspects of the installation and commissioning on-site will be highlighted in comparison with the ITER environment. The synchronized operation of the MHVPS and the BPS (Body Power Supply) on dummy load, piloted through the TF remote control, will be presented and commented.Since the TF supply structure has been built integrating the particular conditions and requirements expected for ITER, a conclusion will summarize the performances obtained at the light of these criteria.     

The negative ion source test facility ELISE

The ITER neutral beam system is using inductively coupled radio frequency (RF) ion sources, that have demonstrated the required ITER parameters on (small) sources with extraction areas up to 200 cm². As a next step towards the full size ITER source IPP is presently constructing the test facility ELISE (“Extraction from a Large Ion Source Experiment”) operating with a “half-size” source which has approximately the width but only half the height of the ITER source. The modular driver concept is expected to allow a further extrapolation to the full size in one direction to be made. The main aim of this experiment is to demonstrate the production of a large uniform negative ion beam with ITER relevant parameters in stable conditions up to one hour.Plasma operation of the source is foreseen to be performed continuously for 1 h; extraction and acceleration of negative ions up to 60 kV is only possible in pulsed mode (10 s every 180 s) due to limitations of the existing IPP HV system. The design of the source and extraction system implements a high experimental flexibility and a good diagnostic access while still staying as close as possible to the ITER design. The main differences are the source operating in air and the use of a large gate valve between the source and the target chamber.ELISE is expected to start operation at the end of 2011 and is an important step for the development of the ITER NBI system; the experience gained early will support the design as well as the commissioning and operating phases of the PRIMA NBI test facilities and the ITER neutral beam system.     

The JET ITER-like wall experiment: First results and lessons for ITER

The JET programme is strongly focused on preparations for ITER construction and exploitation. To this end, a major programme of machine enhancements has recently been completed, including a new ITER-like wall, in which the plasma-facing armour in the main vacuum chamber is beryllium while that in the divertor is tungsten—the same combination of plasma-facing materials foreseen for ITER. The goal of the initial experimental campaigns is to fully characterise operation with the new wall, concentrating in particular on plasma-material interactions, and to make direct comparisons of plasma performance with the previous, carbon wall. This is being done in a progressive manner, with the input power and plasma performance being increased in combination with the commissioning of a comprehensive new real-time protection system. Progress achieved during the first set of experimental campaigns with the new wall, which took place from September 2011 to July 2012, is reported.     

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.     

Design and testing of ITER ECH & CD transmission line components

Initial testing on the Japan Atomic Energy Agency Gyrotron Test Stand of ITER-relevant TL components, has shown reasonable efficiencies, but identified that trapped modes between closely located miter bends, as well as mode conversion at miter bends can lead to excessive heating of the connecting waveguides. General Atomics has designed, built, and will test components to address this issue as well as ITER relevant components that have not been tested at the levels of 1 MW, 170 GHz, for extended pulse lengths. Some of the components that will be tested are ultra low loss miter bends, dc breaks, polarizers, power monitors, bellows, waveguide switches, waveguide cooling clamps, etc. Details of the components and test results will be presented.     

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.     

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.     

Gyrotron and power supply development for upgrading the electron cyclotron heating system on DIII-D

An upgrade of the electron cyclotron heating system on DIII-D to almost 15 MW is being planned which will expand it from a system with six 1 MW 110 GHz gyrotrons to one with ten gyrotrons. A depressed collector 1.2 MW 110 GHz gyrotron is being commissioned as the seventh gyrotron. A new 117.5 GHz 1.5 MW depressed collector gyrotron has been designed, and the first article will be the eighth gyrotron. Two more are planned, increasing the system to ten total gyrotrons, and the existing 1 MW gyrotrons will subsequently be replaced with 1.5 MW gyrotrons.Communications and Power Industries completed the design of the 117.5 GHz gyrotron, and are now fabricating the first article. The design was optimized for a nominal 1.5 MW at a beam voltage of 105 kV, collector potential depression of 30 kV, and beam current of 50 A, but can achieve 1.8 MW at 60 A. The design of the collector permits modulation above 100 Hz by either the body or the cathode power supply, or both, while modulation below 100 Hz must use only the cathode power supply.General Atomics is developing solid-state power supplies for this upgrade: a solid-state modulator for the cathode power supply and a linear high voltage amplifier for the body power supply. The solid-state modulator has series-connected insulated-gate bipolar transistors that are switched at a fixed frequency by a pulse-width modulation regulator to control the output voltage. The design of the linear high voltage amplifier has series-connected transistors to control the output voltage, which was successfully demonstrated in a proof-of-principle test at 2 kV. The designs of complete power supplies are progressing.The design features of the 117.5 GHz 1.5 MW gyrotron and the solid-state cathode and body power supplies will be described and the current status and plans are presented.     

FAST: A European ITER satellite experiment in the view of DEMO

Fusion Advanced Studies Torus (FAST) aims to contribute to the exploitation of ITER and to explore innovative DEMO technology. FAST has been designed to study, in an integrated scenario: (a) relevant plasma-wall interaction problems, with a large power load (P/R ～ 22 MW/m; P/R² ～ 12 MW/m²) and with a full metallic wall; (b) to tackle operational problems in regimes with relevant fusion parameters; (c) to investigate the non-linear dynamics of fast particles (alpha like) in burning plasmas. FAST will operate on a wide parameters range, namely in high performance H-mode (BT ～ 8.5 T; IP ～ 8 MA) as well as in advanced Tokamak operation up to full non-inductive current scenario (IP ～ 2 MA). The main heating is based on 30 MW ICRH, but the ports have been designed to allocate up to 20 MW of 1 MeV NNBI. Helium gas at 30 K is used for cooling of the full machine, a preliminary analysis shows the possibility of realizing FAST with a complete superconductor set of coils. An innovative active system is under development to reduce and to control the magnetic ripple. Tungsten (W) or liquid lithium (L–Li) has been chosen for the divertor material plates and the code EDGE2D has been used to optimize the divertor geometry.     

Risk Mitigation for ITER by a Prolonged and Joint International Operation of JET

Prolonged operation of the Joint European Torus (JET) in a set-up involving all ITER partners will be beneficial for ITER. Experiments at JET with its ITER-like wall and using a D–T plasma mixture will help to mitigate risks in the ITER research plan. Training of the ITER operators, technicians and engineers at JET will safe valuable time when ITER comes into operation. Moreover, the way in which the future ITER experiments will be organized can already be experienced at JET, by imposing a similar organisational structure. This paper will present arguments in favour of an extension of JET and additionally briefly discuss a number of enhancements that will make experiments on JET even more relevant for ITER.     

Infrared thermography inspection of the ITER vertical target qualification prototypes manufactured by European industry using SATIR

High heat flux loaded components which will be installed in the ITER Divertor require a heat flux removal capability in the range 5–10 MW/m² at steady-state and up to 20 MW/m² in transients. Within the ITER plasma facing components procurement context, each party should demonstrate its technical capability to carry out the manufacturing with the required quality. This is achieved through the successful manufacturing and testing of medium-size qualification prototypes. Each Qualification Prototype consists of three high heat flux units mounted onto an actively cooled supporting structure. Currently, the SATIR method has been identified by the ITER Organization as the basic test to decide upon the final acceptance of the ITER Divertor components. SATIR testing was performed on each CFC part of European HHF units prior to the insertion of the twisted tape and prior to assembling the units onto the steel support structure. The paper deals with SATIR results of all qualification prototypes manufactured by European industry.     

Engineering evolution of the FAST machine

FAST (Fusion Advanced Studies Torus) is a proposal for a Satellite Facility which can contribute the rapid exploitation of ITER and prepare ITER and DEMO regimes of operation, as well as exploit innovative plasma facing component systems for DEMO. FAST is a compact (Ro = 1.82 m, a = 0.64 m, triangularity δ = 0.4) and cost effective machine able to investigate, with integration capability, non linear dynamics effects of alpha particle behaviour in burning plasmas. FAST operates in high performance H-mode (B_T up to 8.5 T; I_P up to 8 MA), as well as in advanced tokamak mode (I_P = 3 MA), and in full non inductive current mode (I_P = 2 MA). Helium gas at 30 K is used for cooling the resistive copper magnets. This allows for a pulse duration up to 170 s at 3 MA/3.5 T. The vacuum vessel (VV), segmented into 20-degree modules, is capable to accommodate a 40 MW RF power system. The machine has been designed to house a 10 MW Negative Neutral Beam Injection (NNBI) system. Tungsten (W) or liquid lithium (L-Li) have been chosen as the divertor plate materials, and argon or neon as the impurities to be injected for mitigating the thermal loads.     

The 70 kA pyrobreaker for ITER magnet back-up protection

The paper presents the results of development and testing of an explosively actuated circuit-breaker (the so-called pyrobreaker) designed and manufactured at the Efremov Institute [1]. In accordance with the ITER specifications this switch will be used for continuous operation with DC currents up to 70 kA and shall be capable, on command, to transfer this current to a resistive load under a voltage up to 10 kV in less than 1 ms.A number of current commutation tests have been carried out on several prototypes [2]. The last experimental campaign has demonstrated reliable operation of the pyrobreaker with 20% safety margin for the interrupted current and 100% margin for the recovery voltage relative to the ITER requirements.Besides, peak current withstand tests have been performed with pulse currents up to 420 kA generated by the unipolar current generator available at the Efremov Institute.