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Since the 2018 IAEA FEC Conference, FTU operations have been devoted to several experiments covering a large range of topics, from the investigation of the behaviour of a liquid tin limiter to the runaway electrons mitigation and control and to the stabilization of tearing modes by electron cyclotron heating and by pellet injection. Other experiments have involved the spectroscopy of heavy metal ions, the electron density peaking in helium doped plasmas, the electron cyclotron assisted start-up and the electron temperature measurements in high temperature plasmas. The effectiveness of the laser induced breakdown spectroscopy system has been demonstrated and the new capabilities of the runaway electron imaging spectrometry system for in-flight runaways studies have been explored. Finally, a high resolution saddle coil array for MHD analysis and UV and SXR diamond detectors have been successfully tested on different plasma scenarios.

This article investigates the triggering of ELMs on JET by injection of frozen pellets of isotopes of Hydrogen. A method is established to determine the probability that a specific pellet triggers a particular ELM. This method allows clear distinction between pellet-ELM pairs that are very likely to represent triggering events and pairs that are very unlikely to represent such an event. Based on this, the pellet parameters that are most likely to affect the ability of pellets to trigger ELMs have been investigated. It has been found that the injection location is very important, with injection from the vertical high field side showing a much higher triggering efficiency than low field side (LFS) injection. The dependence on parameters such as pellet speed and size and the time since the last ELM is also seen to be much stronger for LFS injection. Finally, the paper illustrates how improvements to the pellet injection system by streamlining the pellet flight lines and slightly increasing the pellet size has resulted in a significantly improved ability to deliver pellets to the plasma and trigger ELMs.

Plasma Event TRiggering for Alarms (PETRA) is a flexible new system at JET for detecting plasma events to trigger exception handling responses. It provides a platform for all existing and new techniques which involve processing JET real-time data to identify when to change the trajectory of a plasma shot, and reduces the technical cost and risk in demonstrating new techniques. At the same time as providing a protection function to the machine, it still supports additions and modifications to the event detector library during the course of an experimental campaign. It has successfully demonstrated novel techniques, and has computational capacity to continue supporting research into further event detection triggers.

International audience; The paper presents an analysis of disruptions occurring during JET-ILW plasma operations covering the period from the start of ILW (ITER-like wall) operation up to completion of JET operation in 2016. The total number of disruptions was 1951 including 466 with deliberately induced disruptions. The average rate of unintended disruptions was 16.1 %, which is significantly above the ITER target at 15 MA. The pre-disruptive plasma parameters are: plasma current Ip = (0.82-3.38) MA, toroidal field BT = (0.98-3.4) T, safety factor q95 = (1.52-9.05), plasma internal inductance li = (0.58-1.86), Greenwald density limit fraction FGWL = (0.04-1.61), with 720 X-point plasma pulses from a subset of 1420 unintended disruption shots. Massive gas injection (MGI) has been routinely used in protection mode both to terminate pulses when the plasma is at risk of disruption and to mitigate against disruption effects. The MGI was mainly triggered by the n = 1 locked mode (LM) amplitude exceeding a threshold or by the disruption itself, namely, either dIp/dt (specifically, a fast drop in Ip) or the toroidal loop voltage exceeding threshold values. For mitigation purposes, only the LM was used as a physics precursor and threshold on the LM signal was used to trigger the MGI prior to disruption. Long lasting LM (? 100 ms) do exist prior to disruption in 75% of cases. However, 10% of non-disruptive pulses have a LM which eventually vanished without disruption. The plasma current quench (CQ) may result in 3D configurations, termed as asymmetrical disruptions, which are accompanied by sideways forces. Unmitigated vertical displacement events (VDEs) generally have significant plasma current toroidal asymmetries. Unmitigated non-VDE disruptions also have large plasma current asymmetries presumably because there is no plasma vertical position control during the CQ and so they too are subject to large vertical displacements. MGI is a reliable tool to mitigate 3D effects and correspondingly sideways forces during the CQ. The vessel structure loads depend on the force impulse and force time behaviour, including their rotation. The toroidal rotation of 3D configuration may cause resonance with the natural frequencies of the vessel components in large tokamaks such as ITER. The JET-ILW amplitude-frequency interdependence of toroidal rotation of 3D configurations is presented.

In a tokamak device, magnetic diagnostics play a key role in: (i) the understanding of plasma physics, (ii) control and (iii) safe operation (Ariola and Pironti [ 1 ]). JET plasmas have magnetic fluctuations from a large variety of MHD instabilities (Wesson, 2000) and is equipped with fast magnetic Mirnov coils (Giovannozzi et al., 2014; Nave et al., 2004; Heeter et al., 2000) for spectral and mode number analysis. By the end of 2016 (end of experimental campaign C36B), JET had lost almost all pick-up coils used for MHD analysis (fast Mirnov coils). In order to restore the JET MHD modes analysis capability for the coming DT campaign, 25 + 2 faulty in-vessel coils were refurbished. A new coil design was implemented to try to diminish the failure rate, mitigating the possible cause that led to those failures. New sensors will use GLIDCOP® (Boyer et al., 2015) wire and will be remotely-handleable (RH) compatible.

In vessel Mirnov coils are an essential diagnostic in present day tokamaks. Their use in ITER and future Fusion reactors presents some disadvantages linked to the high radiation environment. Furthermore large Electro Magnetic (EM) forces can be experienced by the coil, due to the pulsed operation of the tokamak device (Van Nieuwenhove and Vermeeren, 2003, Vayakis et al., 2011 [ 1 , 2 ]), and disruptions (Gerasimov et al., 2015 [ 3 ]). Since the operation with the ITER-like wall, JET has experienced severe faults in the high-bandwidth Ti wire coils. During 2016-17 new coils have been designed and installed. These can be replaced using remote handling, and they use Cu alloy wire. The presented work includes the failure analysis and modelling, motivating the design differences between old and new coils. The latter will provide valuable information on the long term effects of EM loads during disruptions, as well as chemical degradation processes that will be encountered for ITER High Frequency (HF) coils, which are characterized by the same materials.

This paper describes the activities of installation and commissioning of the magnetic diagnostic enhancement for JET. The new system includes four sub-sets of probes located in the vessel in two opposite octants, for a total of 48 magnetic pick-up coils. The main goal of this enhancement is to improve the accuracy of the reconstruction of the plasma equilibrium and the performance of the real time feedback control of the plasma shape. After a brief description of the system, the paper focuses on the installation activities, accomplished during the 2007 shut-down using the JET remote handling system. The paper then concentrates on the analysis and interpretation of the data collected during the functional commissioning of the new system, carried out during the restart phase of the machine and the first part of the 2008 experimental campaigns, which proved the accuracy of the new signals and the achievement of the project goals.

Preliminary TECXY code simulation results of scrape-off layer conditions in DEMO1 reactor design are presented. The values for snowflake (SF) and single null divertor configurations are discussed. The SF divertor configuration used here, although not fully optimised yet, already promises better performance than standard single null divertor as far as heat load mitigation due to interactions with neutrals. (© 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

This paper deals with the electromagnetic aspects related to a series of experiments aimed at the identification and control of vertical instabilities in the JET Tokamak. The first part of the work is devoted to model the transient magnetic fields produced by the poloidal field circuits and the plasma edge localized modes (ELMs). The second part describes the approach used to design an ELM independent plasma speed estimator. The design of the estimator had to be developed with the constraint of using the present controller parameters, so as to minimize the impact on JET operation.

A new system of magnetic probes was installed during the 2005 shutdown and was commissioned during the 2005/2006 restart phase of JET. The system is made of traditional pick-up coils and Hall sensors, all located outside the vessel. Direct field measurements from Hall sensors are used to correct the drift of the integrators of the pick-up coils signals. This feature will be crucial for future ITER-like devices, where long-lasting flat top phases are expected, in a high neutron yield and a high temperature environment. After a general overview of the system, the paper describes the major manufacturing and installation issues, including the construction of the supports and probes as well as the acceptance tests before and after installation. The functional commissioning of the system includes the integration of the new signals in the JET Data Acquisition System and the analysis of several discharges with and without plasma. The critical aspects of the assessment of the reliability of the signals are shown and commented on.

The aim of this work is to provide reliable linearized models for the design and assessment of a new magnetic control system for MAST (Mega Ampère Spherical Tokamak) using rtEFIT, which can easily be exported to MAST Upgrade. Linearized models for magnetic control have been obtained using the 2D axisymmetric finite element code CREATE L. MAST linearized models include equivalent 2D axisymmetric schematization of poloidal field (PF) coils, vacuum vessel, and other conducting structures. A plasmaless and a double null configuration have been chosen as benchmark cases for the comparison with experimental data and EFIT reconstructions. Good agreement has been found with the EFIT flux map and the experimental signals coming from magnetic probes with only few mismatches probably due to broken sensors. A suite of procedures (equipped with a user friendly interface to be run even remotely) to provide linearized models for magnetic control is now available on the MAST linux machines. A new current driven model has been used to obtain a state space model having the PF coil currents as inputs. Dynamic simulations of experimental data have been carried out using linearized models, including modelling of the effects of the passive structures, showing a fair agreement. The modelling activity has been useful also to reproduce accurately the interaction between plasma current and radial position control loops. © 2013 Euratom-ENEA Association sulla Fusione.

The JET enhancement project Plasma Control Upgrade (PCU) aimed at increasing the capabilities of the plasma vertical stabilization (VS) system. One of the activities of this project was devoted to the development of simple but sufficiently accurate models of the VS system so as to address the main design choices, use the simulation tools as reliable test-beds, and provide an adequate support to the engineering design and commissioning of the new Enhanced Radial Field Amplifier (ERFA). This paper illustrates some of the main achievements of the modelling activity, which gave rise to a closed loop model of the VS system, including plasma, PF coils and passive structures. In particular the paper deals with the selection of the set of turns to be used in the control coils and with the estimation of the eddy current effects on the VS system. The latter analysis addressed an upgrade of the converter units of ERFA, successfully implemented during its commissioning on plasma in August 2009.

Fusion Advanced Studies Torus (FAST) conceptual study has been proposed as possible European ITER Satellite. This facility is aimed at exploring and preparing ITER operation scenarios as well as helping DEMO design and R&D. This paper presents the plasma position and shape control studies, focusing on the machine flexibility in terms of both operational space and physics that can be investigated. The optimization of a copper shell position inside the vacuum vessel provides a slowing down of the growth rate of the vertical instability around 13s−1. A sensitivity analysis of each poloidal field (PF) circuit and their effects on plasma shape is carried out. A SVD based PID control scheme has been used for controlling the plasma current, shape and position.

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.

The Fusion Advanced Studies Torus (FAST) conceptual study has been proposed as possible European ITER Satellite. This facility is aimed at exploring and preparing ITER operation scenarios as well as helping DEMO design and R&D. This paper presents the plasma position and shape control studies, focusing on the machine flexibility in terms of both operational space and physics that can be investigated. A model based sensitivity analysis is carried out, showing the dependence of the poloidal field (PF) system on the main parameters of the plasma. The analysis illustrates the capabilities in terms of plasma shape configurations, as for instance the possibility to achieve plasma with a ratio R/a ∼ 4, and the different influence of the vertical field circuits on the outer plasma shape, the up-down asymmetry of the machine, and the most demanding phases of a pulse, including for instance the strike points sweeping during the steady state of the reference scenario. These obtained results are applicable for the design of the divertor zone geometry and of the power supply amplifiers.

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. (C) 2011 EURATOM. Published by Elsevier B.V. All rights reserved.

In the framework of the ITER project, an investigation has been conducted on an alternative vessel and blanket design, aimed at reducing cost and production risk. The modifications proposed have a strong impact on plasma control since they affect the main conducting structures surrounding the plasma column, providing passive stabilization but at the same time shielding the field generated by the active coils to control the plasma motion and shape. An extensive analysis was performed to assess the plasma vertical controllability and the modified requirements to the in-vessel vertical stability coils system as well as to the external Poloidal Field coils system. A similar analysis was aimed at assessing the performance of the shape control system in presence of the modified structures. The effect on plasma breakdown was also evaluated in terms of maximum initial loop voltage, quality of magnetic null and the flux loss related to the breakdown delay that was quantified under the same hypothesis employed by ITER for the baseline design. Furthermore, the modified design presents issues for the magnetic diagnostic system, related to the shielding of the probes by the eddy currents, which were analysed with a 3D model. The results of the analyses performed have some general interest in particular regarding the influence on plasma stability of 3D structures with close proximity to the plasma. The present paper aims at giving an overview of the analyses that have been carried out and a summary of the results in terms of impact of the modified design on plasma control and scenario, and in general an evaluation of the role of passive structure in plasma vertical stability and shape control.

An investigation of pellet pacing and triggering of Edge Localized Modes (ELMs) was carried out in the frame of ELM mitigation studies aimed at reducing their damaging effects on the plasma-facing components (PFCs). The divertor power load footprint of triggered ELMs was compared with gas puffing controlled ELMs. Small pellets, corresponding to a few per cent of the target plasma particle inventory, were used to minimize the fueling effect and the total particle throughput. There is no evidence that pellets can reduce the divertor power load with respect to gas fueling when operating at the same ELM frequency. The line average density and the energy confinement time remained constant when the gas was progressively substituted by pellets. The launch from the Vertical High Field Side (VHFS) confirmed to be more efficient in ELM triggering than from the Low Field Side (LFS) while the power load footprint remained the same both in time evolution and in spatial distribution when changing the injection geometry. © 2015 Elsevier B.V.

The intrinsic limited toroidal electric field (0.3 V m-1) in devices with superconducting poloidal coils (ITER, JT-60SA) requires additional heating, like electron cyclotron (EC) waves, to initiate plasma and to sustain it during the burn-through phase. The FTU tokamak has contributed to studying the perspective of EC assisted plasma breakdown. Afterward, a new experimental and modeling activity addressing the study of assisted plasma start-up in a configuration close to the ITER one (magnetic field, oblique injection, and polarization) has been performed and is presented here. These experiments have been supported by a 0D code, BKD0, developed to model the plasma start-up and linked to a beam tracing code computing, in a consistent way, EC absorption. The FTU results demonstrate the role of polarization conversion at the inner wall reflection. Dedicated experiments also showed the capability of EC power to sustain plasma start-up in the presence of strong error field (12 mT), with a null outside the vacuum vessel. The BKD0 code, applied to FTU data, has been used to determine the operational window of sustained breakdown as a function of toroidal electric field and neutral pressure. Experimental results in agreement with the BKD0 simulations support the use of the code to predict start-up in future tokamaks, like ITER and JT60SA. © 2015 EURATOM.

Neon injection in FTU can cause a spontaneous increase of the line-average density by a factor 2. The recent experiments were devoted to characterize the plasma response to the neon injection at different densities and plasma currents. A qualitative estimate from UV spectroscopy measurements indicates that the density behaviour cannot be attributed simply to the stripped electrons from the puffed impurity, but a modification of particle transport should be invoked in order to explain the spontaneous rise and the higher peaking. JETTO transport and GENE gyrokinetic codes analyses, as well as a calculation of the electron diffusion coefficients D and pinch velocity U, contribute to feature the peaking effect. © 2015 EURATOM.

The Fusion Advanced Studies Torus (FAST) conceptual study has been proposed [A. Pizzuto on behalf of the Italian Association, The Fusion Advanced Studies Torus (FAST): a proposal for an ITER Satellite facility in support of the development of fusion energy, in: Proceedings of 22nd IAEA Fusion Energy Conference, Geneva, Switzerland, October 13-18, 2008; Nucl. Fusion, submitted for publication] as possible European ITER Satellite facility with the aim of preparing ITER operation scenarios and helping DEMO design and R&D. Insights into ITER regimes of operation in deuterium plasmas can be obtained from investigations of non linear dynamics that are relevant for the understanding of alpha particle behaviours in burning plasmas by using fast ions accelerated by heating and current drive systems. FAST equilibrium configurations have been designed in order to reproduce those of ITER with scaled plasma current, but still suitable to fulfil plasma conditions for studying burning plasma physics issues in an integrated framework. In this paper we report the plasma scenarios that can be studied on FAST, with emphasis on the aspect of its flexibility in terms of both performance and physics that can be investigated. All plasma equilibria satisfy the following constraints: (a) minimum distance of 3 energy e-folding length (assumed to be 1 cm on the equatorial plane) between plasma and first wall to avoid interaction between plasma and main chamber; (b) maximum current density in the poloidal field coils, transiently, up to around 30 MA/m2. The discharge duration is always limited by the heating of the toroidal field coils that are inertially cooled by helium gas at 30 K. The location of the poloidal field coils has been optimized in order to: minimize the magnetic energy; produce enough magnetic flux (up to 35 Wb stored) for the formation and sustainment of each scenario; produce a good field null at the plasma break-down (BP/BT < 2 × 10-4 at low field, i.e. BT = 4 T and ET = 2 V/m for at least 40 ms). Plasma position and shape control studies will also be presented. The optimization of the passive shell position slows the vertical stability growth time down to 100 ms

A new ex-vessel magnetic diagnostics was installed in JET during the 2005 shutdown with the objectives of (i) providing experimental data for better modeling of the iron in the axisymmetric codes for plasma equilibrium reconstruction and (ii) testing the reliability of direct field measurements from Hall probes. The latter is of great interest for future ITER-like devices, where long lasting flat top phases are expected, in a high neutron yield and a high temperature environment. The experimental data achieved during the restart phase were analyzed with the aid of simulation codes. Each individual signal was assessed for inclusion in the general JET diagnostics system. The system is ready for use in the 2006 experimental campaigns.

The assessment of JET magnetics has been carried out using a statistical technique on a large set of sensor data. The results highlight that improved accuracy in the magnetic field reconstruction can be obtained by making an efficient use of the information brought in by the available sensors, in conjunction with the new magnetics that will be installed during the 2004 shutdown (JET Enhancement on Magnetic Diagnostics). Moreover, the better estimation of the magnetic configuration provides an improved capability to locate the plasma boundary and determine its shape.

Experimental fusion devices requires flexible control systems with a modern architecture, which allows the controller to be distributed and modular. The aforementioned requirements are all fulfilled by MARTe, a multi-platform framework for the development of low-latency hard-real-time control system already used with success in many European machine, it was decided to adopt it as the basis of the new FTU Plasma Position Current Density Control (PPCDC) system and the other coupled realtime systems. The main rationale to revamp the FTU control system was to use new technologies and to easily test different control solutions. MARTe has been proved effective from both the points of view, being platform independent, and having a modular architecture which completely separate the control algorithms from the rest of the infrastructure. We report on the new controller deployed at FTU. In particular, after a brief introduction on the machine, we illustrate the structure of the feedback system, together with a detailed analysis and appropriate experimental examples, of the various GAMs (modules) which make up the controller. ﾩ 2014 EURATOM-ENEA Association.

Montreux, 17-21 June 2002, ECA Vol. 26B, P-1.031 (2002)

The behaviour of the SOL in FAST for the new quasi snowflake configuration of the divertor is compared with that of the conventional single null case for three main scenarios: reference, advanced and extreme H-mode. The flexible, quick and versatile 2D code TECXY is used. The main physics processes occurring at the edge are carefully taken into account but the neutral dynamics is simplified. Even though the status close to detachment conditions cannot be detailed, new phenomena generated by the configuration are clearly highlighted. The heat load is strongly mitigated to a level that is much easier to tackle with the present technology. Mitigation is always stronger than expected from the magnetic topology only, especially in the high density regimes. This is attributed to the much longer time that the SOL particles spend in the quite cold region of the X point where mostly the connection length of the magnetic lines is strongly increased and hence to the much more intense interaction of the plasma particle with the background neutrals. © 2013 Elsevier B.V.

The capability of predicting the density limit of a magnetically confined burning plasma is of crucial importance to establish the ultimate performance of a fusion power plant. The Greenwald density limit, commonly used as an empirical scaling law, predicts that the maximum achievable central line-averaged density is given by the relation (n) over bar (G) = k (J) over bar, where (J) over bar is the average plasma current density and k is the plasma elongation. However, several experiments have pointed out that such a limit can be overcome in the presence of peaked density profiles. This paper proposes a new empirical scaling law for a limiter tokamak operating in the low-energy confinement mode (L-mode) concerning the case of peaked density profiles associated with the presence of multifaceted asymmetric radiation from the edges. This result is based on dedicated experiments performed on the Frascati Tokamak Upgrade (FTU) under extremely clean machine conditions (Z(eff) = 1.0-1.5), in which the high-density domain is explored in a wide range of values of plasma current (I-p = 500-900 kA) and toroidal magnetic field (B-T = 4-8 T). It is found that the maximum achievable central line-averaged density essentially depends on the toroidal magnetic field only and does not depend on the average plasma current density: the behaviour is explained in terms of density profile peaking in the high-density domain. As a confirmation that the limit is an edge limit, it is also shown that a Greenwald-like scaling (i.e. depending on the current density) actually holds for the edge line-averaged density (at r/a similar or equal to 4/5).

The Electron Cyclotron (EC) fast launcher for real time control experiments has been installed on FTU and characterized to be fully integrated in a real-time MHD control module under development. The launcher scheme is based on a two module sys- tem, symmetric with respect to the equatorial plane of FTU, with a front steering concept and the launched beams are real-time controllable both in poloidal and toroidal directions. Specific design parameters, defined by the FTU MHD dynamics (typically island size and q profile changes), are the beam dimensions with zooming capabilities, the steering range and mirror speed with the most demanding requirement on poloidal speed of the Steering Mirror (SM) = 1 in 10 ms. A set of tests has been done to verify the system perfor- mance. High power tests of the launcher have been done on a 500 kA, 0.5 10 20 m -3 and 5.3 T plasma with 2 300 kW of EC power delivered to the plasma. The steering mechanism was tested under the automatic control system and showed a dynamic response in line with the requirements. Results of these tests will be presented in the paper.

One of the main problems in tokamak fusion devices concerns the capability to operate at a high plasma density, which is observed to be limited by the appearance of catastrophic events causing loss of plasma confinement. The commonly used empirical scaling law for the density limit is the Greenwald limit, predicting that the maximum achievable line-averaged density along a central chord depends only on the average plasma current density. However, the Greenwald density limit has been exceeded in tokamak experiments in the case of peaked density profiles, indicating that the edge density is the real parameter responsible for the density limit. Recently, it has been shown on the Frascati Tokamak Upgrade (FTU) that the Greenwald density limit is exceeded in gas-fuelled discharges with a high value of the edge safety factor. In order to understand this behaviour, dedicated density limit experiments were performed on FTU, in which the high density domain was explored in a wide range of values of plasma current (Ip = 500-900 kA) and toroidal magnetic field (B T = 4-8 T). These experiments confirm the edge nature of the density limit, as a Greenwald-like scaling holds for the maximum achievable line-averaged density along a peripheral chord passing at r/a ≃ 4/5. On the other hand, the maximum achievable line-averaged density along a central chord does not depend on the average plasma current density and essentially depends on the toroidal magnetic field only. This behaviour is explained in terms of density profile peaking in the high density domain, with a peaking factor at the disruption depending on the edge safety factor. The possibility that the MARFE (multifaced asymmetric radiation from the edge) phenomenon is the cause of the peaking has been considered, with the MARFE believed to form a channel for the penetration of the neutral particles into deeper layers of the plasma. Finally, the magnetohydrodynamic (MHD) analysis has shown that also the central line-averaged density at the onset of the MHD activity depends only on the toroidal magnetic field. © 2013 IAEA, Vienna.

The impact of the 3D effects on the dynamic response of the plasma in the JET tokamak is assessed on the time scale of interest for plasma magnetic control and stabiliza tion. These effects are due to the limbs of the magnetic circuit and to the eddy currents in the 3D metallic structures. The plasma behaviour is taken into account by a specific matrix of modifi ed inductances interacting with the remaining part of the syst em via a virtual coupling surface.

The Fusion Advanced Studies Torus (FAST) conceptual study has been proposed [A. Pizzuto on behalf of the Italian Association, The Fusion Advanced Studies Torus (FAST): a proposal for an ITER Satellite facility in support of the development of fusion energy, in: Proceedings of 22nd IAEA Fusion Energy Conference, Geneva, Switzerland, October 13–18, 2008; Nucl. Fusion, submitted for publication] as possible European ITER Satellite facility with the aim of preparing ITER operation scenarios and helping DEMO design and R&D. Insights into ITER regimes of operation in deuterium plasmas can be obtained from investigations of non linear dynamics that are relevant for the understanding of alpha particle behaviours in burning plasmas by using fast ions accelerated by heating and current drive systems. FAST equilibrium configurations have been designed in order to reproduce those of ITER with scaled plasma current, but still suitable to fulfil plasma conditions for studying burning plasma physics issues in an integrated framework. In this paper we report the plasma scenarios that can be studied on FAST, with emphasis on the aspect of its flexibility in terms of both performance and physics that can be investigated. Plasma position and shape control studies will also be presented. The optimization of the passive shell position slows the vertical stability growth time down to 100 ms.

Since the 2012 IAEA-FEC Conference, FTU operations have been largely devoted to runaway electrons generation and control, to the exploitation of the 140GHz electron cyclotron (EC) system and to liquid metal limiter elements. Experiments on runaway electrons have shown that the measured threshold electric field for their generation is larger than predicted by collisional theory and can be justified considering synchrotron radiation losses. A new runaway electrons control algorithm was developed and tested in presence of a runaway current plateau, allowing to minimize the interactions with plasma facing components and safely shut down the discharges. The experimental sessions with 140GHz EC system have been mainly devoted to experiments on real-time control of magnetohydrodynamic (MHD) instabilities using the new EC launcher with fast steering capability. Experiments with central EC injection have shown the onset of 3/2 and 2/1 tearing modes, while EC assisted breakdown experiments have been focused on ITER start-up issues, exploring the polarization conversion at reflection from inner wall and the capability to assure plasma start-up even in presence of a large stray magnetic field. A new actively cooled lithium limiter has been installed and tested. The lithium limiter was inserted in the scrape-off layer, without any damage to the limiter surface. First elongated FTU plasmas with EC additional heating were obtained with the new cooled limiter. Density peaking and controlled MHD activity driven by neon injection were investigated at different plasma parameters. A full real-time algorithm for disruption prediction, based on MHD activity signals from Mirnov coils, was developed exploiting a large database of disruptions. Reciprocating Langmuir probes were used to measure the heat flux e-folding length in the scrape-off layer, with the plasma kept to lay on thea internal limiter to resemble the ITER start-up phase. New diagnostics were successfully installed and tested, as a diamond probe to detect Cherenkov radiation produced by fast electrons and a gamma camera for runaway electrons studies. Laser induced breakdown spectroscopy measurements were performed under vacuum and with toroidal magnetic field, so demonstrating their capability to provide useful information on the surface elemental composition and fuel retention in present and future tokamaks, such as ITER. © 2015 EURATOM.

The magnetic vacuum topology reconstruction using magnetic measurements is essential in controlling and understanding plasmas produced in magnetic confinement fusion devices. In a wide range of cases, the instruments used to approach the problem have been designed for a specific machine and to solve a specific plasma model. Recently, a new approach has been used for developing new magnetic software called FELIX. The adopted solution in the design allows the use of the software not only at JET but also at other machines. In order to reduce the analysis and debugging time the software has been designed with modularity and platform independence in mind. This results in a large portability and in particular it allows using the same code both offline and in real-time. One of the main aspects of the tool is its capability to solve different plasma models of current distribution. Thanks to this feature, in order to improve the plasma magnetic reconstruction in real-time, a set of different models has been run using FELIX. FELIX is presently running at JET in different real-time analysis and control systems that need vacuum magnetic topology.

This paper deals with the problem of the equilibrium reconstruction in a tokamak. We firstly describe two approaches for the shape reconstruction: the former based on a nonlinear least square minimization with Finite Element Method modelling taking into account the Grad Shafranov equilibrium equation; the latter based on linear multifilament models accounting for the presence of iron, eddy currents in the passive structures, and plasma. A possible unified treatment for shape and current profile reconstruction using both internal and external measurements is then discussed: the nonlinear least square minimization procedure used for shape identification is modified to account for internal measurements and profile shape fitting. A linearized approach is finally analysed making use of Singular Value Decomposition theory. Some results on the JET tokamak are shown and discussed.

The aim of this paper is to present some new tools used to review the capability of the ITER Poloidal Field (PF) system in controlling the broad range of plasma configurations presently forecasted during ITER operation. The attention is focused on the axi-symmetric aspects of plasma magnetic configuration control since they pose the greatest challenges in terms of control power and they have the largest impact on machine capital cost. Some preliminary results obtained during ongoing activities in collaboration between ENEA/CREATE and EFDA are presented. The paper is divided in two main parts devoted, respectively, to the presentation of a procedure for the PF current optimisation during the scenario, and of a software environment for the study of the PF system capabilities using the plasma linearized response. The proposed PF current optimisation procedure is then used to assess Scenario 2 design, also taking into account the presence of axisymmetric eddy currents and possible variations of poloidal beta and internal inductance. The numerical linear model based tool derived from the JET oriented eXtreme Shape Controller (XSC) tools is finally used to obtain results on the strike point sweeping in ITER. © 2007 Elsevier B.V. All rights reserved.