Fusion technology activities at JET in support of the ITER program

Among the technological activities performed at JET in support of the scientific objectives of both JET and ITER, a significant effort is devoted to the investigation of the erosion, transport and deposition of wall materials, and of their fuel retention properties. With the analysis of wall tiles retrieved in the 2010 shutdown, the full characterization of the previous JET carbon wall is obtained. In order to confirm the expectations on properties of the new ITER-Like Wall (ILW) installed in 2011, a large number of marker tiles and profiled tiles have been prepared and installed both in the main wall and in the divertor. These will be retrieved from the vessel during a short shutdown at the end of 2012 and analyzed. The major changes introduced by the new ILW materials in JET required also a new nuclear characterization of the machine. Neutronics measurements have been performed to obtain the neutron/γ-ray field changes inside and outside the JET machine. The experimental data are also used to validate neutronics codes used in ITER design. A new calibration of neutron detectors, scheduled in the 2012 shutdown and adopting the same procedure as in ITER, has been prepared based on extensive neutronics calculations.     

Operational impact on the JET ITER-like wall in-vessel components

The JET ITER-like Wall (ILW) provides the same plasma facing component configuration as ITER during its active phase: beryllium in the main chamber and tungsten in the divertor. Moving from a carbon-based wall to an all metals wall requires some operational adjustment. The reduction in radiation at the plasma edge and in the divertor can lead to high power loads on the plasma facing components both in steady state and in transients and requires the development of radiative scenarios and the use of massive gas injection to mitigate disruptions. These tools are even more important now because an all metal wall is much less forgiving to thermal overloading the carbon based wall used to be. Here the impact of the first 11 months of operation on the ILW plasma facing components is discussed.     

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.     

First experimental results with the Current Limit Avoidance System at the JET tokamak

The Current Limit Avoidance System (CLA) has been recently deployed at the JET tokamak to avoid current saturations in the poloidal field (PF) coils when the eXtreme Shape Controller is used to control the plasma shape. In order to cope with the current saturation limits, the CLA exploits the redundancy of the PF coils system to automatically obtain almost the same plasma shape using a different combination of currents in the PF coils. In the presence of disturbances it tries to avoid the current saturations by relaxing the constraints on the plasma shape control. The CLA system has been successfully implemented on the JET tokamak and fully commissioned in 2011. This paper presents the first experimental results achieved in 2011–2012 during the restart and the ITER-like wall campaigns at JET.     

The installation, testing and performance on the jet coils of the enhanced radial field amplifier (ERFA)

The ERFA is a major part of the upgrade to the plasma vertical stabilisation system for JET. As well as improvements to the plasma controller, there was a requirement for a new power supply with increased voltage and current capability over its predecessor the Fast Radial Field Amplifier (FRFA). The ERFA had to be factory tested, installed at JET, power and signal connections made, all signals tested and then installed adjacent to FRFA. The connection to the JET coils had to be achieved in a planned 7-week pause in operation dedicated to this installation activity and perform to its full capability from the JET restart. The ERFA project achieved all of its aims and, following a minor upgrade, met or exceeded the performance specifications. This paper covers the site installation, signal testing, and power tests on dummy load leading to the final acceptance tests on the JET coils.     

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.     

Shape Control with the eXtreme Shape Controller During Plasma Current Ramp-Up and Ramp-Down at the JET Tokamak

The eXtreme Shape Controller (XSC) has been originally designed to control the plasma shape at JET during the flat-top phase, when the plasma current has a constant value. During the JET 2012 experimental campaigns, the XSC has been used to improve the shape control during the transient phases of plasma current ramp-up and ramp-down. In order to avoid the saturation of the actuators with these transient phases, a current limit avoidance system has been designed and implemented. This paper presents the experimental results achieved at JET during the 2012 campaigns using the XSC.     

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.     

First plasma operation of the enhanced JET vertical stabilisation system

A project dedicated to the enhancement of the JET vertical stabilization system was launched in 2006, including an upgrade of the Power Supply of the Radial Field Amplifier, of hardware and software of the vertical stabilization control system. The main aim was to double the JET capability in stabilising high current plasmas when subject to perturbations, in particular large Edge Localised Modes. We present here the results of first plasma operation with the new Enhanced Radial Field Amplifier and its data acquisition and control system, focussing on the benefits of an approach based on phased commissioning, modelling and offline algorithm validation.     

Electromagnetic analysis of breakdown conditions in JET

This paper presents the breakdown studies carried out in the framework of JET Enhancement Projects for Plasma Control Upgrade (PCU) and Enhanced Radial Field Amplifier (ERFA), to obtain plasma formation with different sets of coil turns in the radial field circuit. The electromagnetic conditions to reach the plasma breakdown in the JET machine are strongly dependent on the properties of JET iron core and the effects of the eddy currents driven by the transient electric field on the present passive structures. The study has been carried out by using a linearized dynamic model of JET provided by 2D axisymmetric finite element code CREATE-L [R. Albanese, et al., Nucl. Fusion 38 (1998) 723–738]. The dynamic simulations have been compared with the experimental data. A new fast visible camera has been installed and has been used for the first time at JET for studies of plasma breakdown. The images show, coherently with the model, that the avalanche evolves dynamically towards a region where the stray field is perpendicular to the first wall.     

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.     

Developments in remote metrology at JET

The need to maximise the operational availability of fusion devices has driven the enhancements in accuracy, flexibility and speed associated with the inspection techniques used at JET. To this end, the remote installation of the ITER-Like Wall (ILW) tiles, conduits and embedded diagnostics has necessitated the adoption of technologies from other industries for their use in conjunction with the JET Remote Handling (RH) system. The novel adaptation of targetless stereophotogrammetry, targeted single-camera photogrammetry and gap measurement techniques for remote applications has prompted a range of challenges and lessons learnt both from the design process and operational experience.Interfacing Commercial Off-The-Shelf (COTS) components with the existing RH equipment has highlighted several issues of relevance to the developing ITER RH system. This paper reports results from the stereophotogrammetry and the single-camera photogrammetry surveys, allowing analysis of the effectiveness of the RH system as a platform for in-vessel measurement. This includes scrutiny of the accuracy achieved with each technique as well as the impact on the in-vessel Configuration Management Model (CMM). The paper concludes with a summary of key recommendations for the ITER RH system based on the experience of remote metrology at JET.     

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.     

Geometry and expected performance of the solid tungsten outer divertor row in JET

At JET new plasma-facing components for the main chamber wall and the divertor are being designed and built to mimic the expected ITER plasma wall conditions in the deuterium-tritium operation phase. The main wall elements at JET will be made of beryllium and the divertor plasma-facing surface will be made of tungsten. Most of the divertor tiles will consist of tungsten-coated Carbon Fibre Composite (CFC) material. However one toroidal row in the outer divertor will be made of solid, inertially cooled tungsten. The geometry of these solid tungsten divertor components is optimized within the boundary conditions of the interfaces and the constraints given by the electrodynamical forces. Shadowing calculations as well as rough field line penetration analysis is used to define the geometry of the tungsten lamella stacks. These calculations are based on a set of magnetic equilibria reflecting the operation domain of current JET plasma scenarios. All edges in poloidal and toroidal direction are shadowed to exclude near perpendicular field line impact. In addition, the geometry of the divertor structure is being optimized so that the fraction of the plasma wetted surface is maximised. On the basis of the optimized divertor geometry, performance calculations are done with the help of ANSYS to assess the maximum power exhaust possible with this inertially cooled divertor row.     

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.     

Remote handling installation of diagnostics in the JET Tokamak

The requirement for an upgrade of the diagnostics for the JET ITER Like Wall (ILW) while maintaining personnel exposure to contamination as low as reasonably practicable or ALARP, has necessitated the development of a bespoke set of diagnostic components. These components, by virtue of their design and location, require a versatile yet comprehensive suite of remote handling tools to undertake their in-vessel installation. The installation of the various diagnostic components is covered in multiple tasks. Each task requires careful assessment and design of tools that can successfully interface with the components and comply with the handling and installation requirements. With remote maintenance a requirement, the looms/conduits were designed to be modular with connections which are electrically connected when the module is fitted or conversely disconnected when removed. The shape of each complex and often bulky component is verified during the design phase, to ensure that it can be delivered and installed to its specified location in the torus. This is done by matching the kinematic capabilities of the remote handling system and the path of the component through the torus by using a state of the art virtual reality system.     

ITER vertical stabilization system

As part of the ITER design review, a reassessment of the specifications underlying the design of the vertical stabilization system (VS) was performed.Recent results from experiments, aimed at the evaluation of the feasibility of the ITER reference scenarios, have raised several concerns regarding mostly the ramp-up and ramp-down phases of the pulse. The main issue is the value of the internal inductance li which may reach values outside the range 0.7–1, considered as reference for the ITER control system design. Similar concerns apply to the low current L-mode plasmas, needed to the exploitation of the machine towards the development of the 15 MA pulse. The performance of the reference vertical stabilization system, under the revised conditions may be marginal, in particular if the effect of plasma generated noise on the velocity measurement is considered.A reliable and robust VS is mandatory to guarantee the operation of ITER at the reference elongation and plasma current values. To avoid de-scoping of the machine mission, several solutions have been proposed to improve the VS performances, ranging from an upgrade of the maximum voltage available to the present external coils system, to the introduction of in-vessel passive and/or active conductors.The paper presents an overview of the modelling and experimental effort aimed at the assessment of the baseline ITER VS and analyses the proposed solutions to improve the system performance.     

High level integration of remote handling control systems at JET

To reduce the timescale of the JET Enhanced Performance 2 (EP2) shutdown, two multi-jointed Booms instead of one will be used for maintenance and upgrades inside the JET vessel. To fully utilize this new configuration, the control systems of the Booms have been modified at a high level to allow quick and safe interactions between them. This paper will discuss how the control systems of the Booms have been integrated to exploit the increased mechanical functionality of the Octant 1 Boom, and will demonstrate how this has improved safety, utility and efficiency for the remote handling operators during the EP2 shutdown. Other operational streamlining functions will be mentioned, as well as a look to the future of Remote Handling at JET.     

Improvement of divertor triple probe system and its measurements under full graphite wall on EAST

Full graphite wall of experimental advanced superconducting tokamak (EAST) has been developed in the spring of 2008. A new divertor triple probe diagnostics system (DTPDs) is built for EAST during this upgrade. The tip shape and connected structure of the probe are optimized for variational magnetic field directions and DTPDs maintenance. The experiment has been carried out with a full graphite wall for EAST, and near double-null diverted plasma is achieved successfully. The evolutions of electron temperature, density, particle flux and power densities along the divertor targets have been obtained with DTPDs.     

Design,manufacture and initial operation of the beryllium components of the JET ITER-like wall

The aim of the JET ITER-like wall project was to provide JET with the plasma facing material combination now selected for the DT phase of ITER (bulk beryllium main chamber limiters and a full tungsten divertor) and, in conjunction with the upgraded neutral beam heating system, to achieve ITER relevant conditions.The design of the bulk Be plasma facing components had to be compatible with increased heating power and pulse length, as well as to reuse the existing tile supports originally designed to cope with disruption loads from carbon based tiles and be installed by remote handling. Risk reduction measures (prototypes, jigs, etc.) were implemented to maximize efficiency during the shutdown. However, a large number of clashes with existing components not fully captured by the configuration model occurred.Restarting the plasma on the ITER-like Wall proved much easier than for the carbon wall and no deconditioning by disruptions was observed. Disruptions have been more threatening than expected due to the reduced radiative losses compared to carbon, leaving most of the plasma magnetic energy to be conducted to the wall and requiring routine disruption mitigation. The main chamber power handling has achieved and possibly exceeded the design targets.