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A novel application of strip and pixel silicon radiation detectors for study and characterization of run-away electron events in tokamaks is presented. Main goal was to monitor runaway electrons both directly and indirectly. The strip detector was placed inside the tokamak vacuum chamber in order to monitor the run-away electrons directly. Whereas the pixel detector was placed outside the tokamak chamber behind a pin hole for monitoring the run-away electrons indirectly via radiation produce by interaction of the electrons with the plasma facing material. Results obtained using the silicon detectors are compared with already existing diagnostic methods consisting of scintillation devices detecting X-rays and photo-neutrons, providing the same results in the observable comparisons. Tests with the pixel detector proved that the pinhole camera is able to extract spatial information of interaction point (a place where the runaway electrons hit on the facing material) and the strip detectors indicate presence of additional signal from throughout the discharge. The performed experiments are innovative, illustrating possible development of new and easy to use diagnostic method.

The COMPASS-U tokamak is designed at IPP CAS Prague. It will be a compact device that should achieve high toroidal magnetic field and high wall temperatures. In order to observe fast phenomena, photodiode arrays combined with beryllium filters and a pinhole are planned to be used as Soft X-Ray (SXR) detectors. Operational temperatures of photodiodes are typically below 100 °C and planned first wall temperatures as high as 500 °C for the later phases of operation. Therefore, an actively cooled support structure housing the detector has to be designed. Moreover, in order to obtain good inversion results the layout of tomographic diagnostics systems have to be optimised to provide the best possible coverage of the plasma. The proposed SXR system consists of two sets of detectors with different filter thickness enabling the use of the two filter method for electron temperature estimation. Both sets will share the same line of sight layout for better comparability. Optimisation of this layout is adapted for limited options of detector placement and extensive testing of tomography capabilities is performed. Results of these tests are shown for one of considered layouts.

Within the EUROfusion MST1 work package, a series of experiments has been conducted on AUG and TCV devices to disentangle the role of plasma fueling and plasma shape for the onset of small ELM regimes. On both devices, small ELM regimes with high confinement are achieved if and only if two conditions are fulfilled at the same time. Firstly, the plasma density at the separatrix must be large enough (), leading to a pressure profile flattening at the separatrix, which stabilizes type-I ELMs. Secondly, the magnetic configuration has to be close to a double null (DN), leading to a reduction of the magnetic shear in the extreme vicinity of the separatrix. As a consequence, its stabilizing effect on ballooning modes is weakened. Peer Reviewed Article escrit per 365 autors/autores: Labit, B.; Eich, T.; Harrer, G. F.; Wolfrum, E.; Bernert, M.; Dunne, M. 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This paper presents two scenarios used for generation of a runaway electron (RE) beam in the COMPASS tokamak with a focus on the decay phase and control of the beam. The first scenario consists of massive gas injection of argon into the current ramp-up phase, leading to a disruption accompanied by runaway plateau generation. In the second scenario, injection of a smaller amount of gas is used in order to isolate the RE beam from high-temperature plasma. The performances of current control and radial and vertical position feedback control in the second scenario were experimentally studied and analysed. The role of RE energy in the radial position stability of the RE beam seems to be crucial. A comparison of the decay phase of the RE beam in various amounts of Ar or Ne was studied using absolute extreme ultraviolet (AXUV) tomography and hard x-ray (HXR) intensity measurement. Argon clearly leads to higher HXR fluxes for the same current decay rate than neon, while radiated power based on AXUV measurements is larger for Ne in the same set of discharges.

The role of the COMPASS tokamak in research of generation, confinement and losses of runaway electron (RE) population is presented. Recently, two major groups of experiments aimed at improved understanding and control of the REs have been pursued. First, the effects of the massive gas injection (~1021 Ar/Ne particles) and impurity seeding (~1018 particles) were studied systematically. The observed phenomena include generation of the post-disruption RE beam and current conversion from plasma to RE. Zero loop voltage control was implemented in order to study the decay in simplified conditions. A distinctive drop of background plasma temperature and electron density was observed following an additional deuterium injection into the RE beam. With the loop voltage control the parametric dependence of the current decay rate dI/dt can be studied systematically and possibly extrapolated to larger facilities. Second, recent results of experiments focused on the role of the magnetic field in physics of RE were analysed. In this contribution, special attention is given to the observed effects of the resonant magnetic perturbation on the RE population. The benefits of the RE experiments on COMPASS was reinforced by diagnostic enhancements (fast cameras, Cherenkov detector, vertical ECE etc) and modelling efforts (in particular, coupling of the METIS and LUKE codes).

The purpose of the ITER Radial Neutron Camera (RNC) is the measurement of the plasma neutron emissivity profile [neutrons∙s−1 m-3] for burn control purposes. The present RNC design consists of 22 collimated detector systems providing a set of line-integrated neutron measurements [neutrons∙s−1 m-2] with full coverage of the plasma poloidal cross-section; the neutron emissivity can be recovered from the line-integrated measurements by means of dedicated reconstruction techniques (1D spatial inversion, 2D tomography). The present paper focuses on the evaluation of the improvement in the RNC 2D reconstruction of the neutron emissivity obtained by using the total neutron yield value provided by an independent diagnostic as additional constraint in the tomography procedure. The analysis was performed using a tomography code based on the Minimum Fisher Regularization (MFR). A clear improvement of the neutron emissivity reconstruction has been observed when the total neutron yield constraint is considered. The improvement is seen in: extension of the spatial region in which the accuracy of the reconstruction is better than 10 %; better reconstruction of peaked emissivity profiles; and robustness against measurements noise and line of sight data loss.

The Joint European Torus (JET) presently operates with the ITER-like wall (ILW) including tungsten divertor. Analyses of the soft X-ray radiation (SXR) data present an important tool for studies of transport of tungsten into JET plasmas. SXR spatial distribution can be measured by horizontal (H) and two vertical (V,T) pinhole cameras. Tomographic reconstruction of SXR emissivity from their data is challenging due to different spectral sensitivities of the cameras which can cause significant inconsistencies due to spectral properties of tungsten radiation. In this contribution an advanced correction based on evaluation of the sensitivity differences for individual lines of view of the SXR detectors is presented. The adjustments take into account atomic data of tungsten as well as the SXR intensity perturbation due to tungsten influx. First results are presented and discussed.

The significant role of magnetic perturbations in mitigation and losses of runaway electrons (REs) was documented in dedicated experimental studies of RE at the COMPASS tokamak. REs in COMPASS are produced both in low density quiescent discharges and in disruptions triggered by massive gas injection (MGI). The role of the RE seed produced in the beginning of the discharge on the subsequent RE population proved significant. Modulation of the RE losses by MHD instabilities was observed at several characteristic frequencies, as well as by magnetic field oscillations related to power supplies. Magnetic islands seem to suppress the losses as the HXR signal is low and coherent with the island rotation frequency. Moreover, periods of increased losses of REs observed in the current quench (CQ) and early RE beam plateau phase of the MGI disruptions seem to be linked to the bursts of magnetic perturbation, and to the observation of filaments in the fast visible camera images.

Machine learning tools have been used since a long time ago to study disruptions and to predict their occurrence. On the other hand, the challenges posed by the quality and quantities of the data available remain substantial. In this paper, methods to optimize the training data set and the potential of kernels-based advanced machine learning tools are explored and assessed. Various alternatives, ranging from appropriate selection of the weights to the inclusion of artificial points, are investigated to improve the quality of the training data set. Support vector machines (SVM), relevance vector machines (RVMs), and one-class SVM are compared. The relative performances of the different approaches are initially assessed using synthetic data. Then they are applied to a relatively large database of JET disruptions. It is shown that in terms of final results, the optimization of the training databases proved to be very productive. Further, the RVM algorithm performs well when it is trained on a small set of discharges compared to the traditional methods.

International audience; The tokamak WEST (Tungsten Environment in Steady-State Tokamak) will start operating by the end of 2016 as a test bed for the ITER divertor components in long pulse operation. In this context, radiative cooling of heavy impurities like tungsten (W) in the Soft X-ray (SXR) range [0.1 keV; 20 keV] is a critical issue for the plasma core performances. Thus reliable tools are required to monitor the local impurity density and avoid W accumulation. The WEST SXR diagnostic will be equipped with two new GEM (Gas Electron Multiplier) based poloidal cameras allowing to perform 2D tomographic reconstructions in tunable energy bands. In this paper tomographic capabilities of the Minimum Fisher Information (MFI) algorithm developed for Tore Supra and upgraded for WEST are investigated, in particular through a set of emissivity phantoms and the standard WEST scenario including reconstruction errors, influence of noise as well as computational time.

International audience; Tomography inversion has been used routinely for processing outputs of plasma radiation diagnostics. Various tomographic algorithms have been developed, with those based on Tikhonov regularization being among the fastest while still providing reliable results. This paper presents a further speed optimization of the minimum Fisher Tikhonov regularization algorithm based on reducing iteration cycles used during the calculation. Ten to twentyfold speed gain is achieved compared to the original implementation. Robustness of the new method is demonstrated using both artificially generated data sets and real data from the soft X-ray diagnostics at the COMPASS tokamak. The advantage gained by the optimization is investigated in particular with respect to the possibility of real-time control of the plasma position; the option of impurity control is also discussed.

The COMPASS tokamak, a divertor device with clear H-mode and ITER-relevant geometry, has been re-installed in IPP Prague after its transport from CCFE in UK. The first plasma was achieved in December 2008. Many new diagnostic tools with both high temporal and spatial resolutions have been developed to address the scientific programme of COMPASS focused on H-mode physics and pedestal investigations. In the paper, an overview of existing and in near future planned diagnostics (magnetic, spectroscopy, microwave, probe and beam/particle) on COMPASS is given including their basic technical specifications and achievements.

In ITER, alpha particle loss measurements remain difficult and further progress is needed. A number of standard measurement methods have been proposed but techniques capable of operating in the fusion reactor conditions need development. The charged particle activation method is a novel concept demonstrated only recently on JET and its feasibility for performing measurements in ITER is being investigated. This design study presents the first detailed activation calculations in order to derive the diagnostic response and expected performance. The calculations were carried out using the activation inventory code FISPACT-2007 and the EAF-2007 nuclear data libraries. The results indicate that ITER Baseline requirements may be fulfilled and in situ measurements may be performed in an equatorial port plug. The main constraints and parameters are described. Finally, research needs for further diagnostic developments are outlined.

The paper concerns measurements of runaway electrons (REs) which are generated during discharges in tokamaks. The control of REs is an important task in experimental studies within the ITER-physics program. The NCBJ team proposed to study REs by means of Cherenkov-type detectors several years ago. The Cherenkov radiation, induced by REs in appropriate radiators, makes it possible to identify fast electron beams and to determine their spatial- and temporal-characteristics. The results of recent experimental studies of REs, performed in two tokamaks - COMPASS in Prague and FTU in Frascati, are summarized and discussed in this paper. Examples of the electron-induced signals, as recorded at different experimental conditions and scenarios, are presented. Measurements performed with a three-channel Cherenkov-probe in COMPASS showed that the first fast electron peaks can be observed already during the current ramp-up phase. A strong dependence of RE-signals on the radial position of the Cherenkov probe was observed. The most distinct electron peaks were recorded during the plasma disruption. The Cherenkov signals confirmed the appearance of post-disruptive RE beams in circular-plasma discharges with massive Ar-puffing. During experiments at FTU a clear correlation between the Cherenkov detector signals and the rotation of magnetic islands was identified. © Published under licence by IOP Publishing Ltd.

In metallic tokamaks, the interplay between particle transport and MagnetoHydroDynamic (MHD) activity might lead to impurities accumulation and finally to disruption. Studying such phenomena is thus essential if stationary discharges are to be achieved. Measuring the soft X-ray (SXR) radiation ([0.1 keV; 20 keV]) of magnetic fusion plasmas is a standard way of accessing valuable information on particle transport and MHD. Generally, like at Tore Supra (TS), the analysis is performed with a 2D tomographic system composed of several cameras equipped with silicon barrier diodes (SBD). On WEST the installation of an upper divertor masks many of the actual TS vertical diodes so that no proper tomography is possible. This paper presents the design of a new SXR diagnostic for the WEST project developed in collaboration with IPPLM (Poland) and the Warsaw University of Technology, based on a triple gas electron multiplier (GEM) detector. Preliminary simulations performed to size and position the detector and its electronics inside the vertical thimble are also presented, in particular estimation of magnetic field and temperature variation affecting GEM spatial resolution and signal quality. As a conclusion, perspectives about tomographic capabilities of the new system for studying impurity transport are given.

Ion cyclotron resonance frequency (ICRF) heating has been an essential component in the development of high power H-mode scenarios in JET-ILW. The steps that were taken for the successful use of ICRF heating in terms of enhancing the power capabilities and optimizing the heating performance in view of core impurity mitigation in these experiments will be reviewed.

Since the ITER-like wall in JET (JET-ILW) came into operation, dedicated ITER-like plasma current (I p) ramp-up (RU) and ramp-down (RD) experiments have been performed and matched to similar discharges with the carbon wall (JET-C). The experiments show that access to H-mode early in the I p RU phase and maintaining H-mode in the I p RD as long as possible are instrumental to achieve low internal plasma inductance (l i) and to minimize flux consumption. In JET-ILW, at a given current rise rate similar variations in l i (0.7–0.9) are obtained as in JET-C. In most discharges no strong W accumulation is observed. However, in some low density cases during the early phase of the I p strong core radiation due to W influx led to hollow electron temperature (T e) profiles. In JET-ILW Z eff is significantly lower than in JET-C. W significantly disturbs the discharge evolution when the W concentration approaches 10−4; this threshold is confirmed by predictive transport modelling using the CRONOS code. I p RD experiments in JET-ILW confirm the result of JET-C that sustained H-mode and elongation reduction are both instrumental in controlling l i.

For ITER-relevant runaway electron studies, such as suppression, mitigation, termination and/or control of runaway beam, obtaining the runaway electrons after the disruption is important. In this paper we report on the first achieved discharges with post-disruptive runaway electron beam, entitled "runaway plateau", in the COMPASS tokamak. The runaway plateau is produced by massive gas injection of argon. Almost all of the disruptions with runaway electron plateaus occurred during the plasma current ramp-up phase. Comparison between the Ar injection discharges with and without plateau has been done for various parameters. Parametrisation of the discharges shows that COMPASS disruptions fulfill the range of parameters important for the runaway plateau occurrence. These parameters include electron density, electric field, disruption speed, effective safety factor, maximum current quench electric field. In addition to these typical parameters, the plasma current value just before the massive gas injection surprisingly proved to be important., Comment: Submitted for the special issue on "Energetic Electrons in Space and Laboratory Plasmas" in Journal of Plasma Physics

Two-dimensional (2D) spatial profile and the temporal evolution of the 14 and 2.5 MeV neutron emissivities from D–D and D–T fusion reactions were studied using the measurements of the upgraded neutron profile monitor during the last trace tritium experiments in JET. The JET neutron profile monitor provides unique capability for 2.5 and 14 MeV neutrons line-integrated measurements simultaneously. A systematic comparison of D–D and D–T neutron emissivity was performed. The tritium concentration or fuel ratio (nT/nD) was analysed for a set of 34 ELMy-H mode discharges with tritium puff. Tritium concentration is deduced with a method based on the ratio of D–T 14 MeV and D–D 2.5 MeV neutron emissivities in order to exploit the maximum information available from neutron data. With the help of a tomography algorithm recently developed at JET, 2D spatial profiles of the tritium concentration in the plasma were obtained. These profiles can be used to perform transport studies. Tritium core confinement is clearly seen to increase with plasma density for the set of discharges studied. Differences in the shape of these profiles are also found between low and high density plasmas. Shortly after tritium puffing, 2D spatial profiles of the tritium concentration exhibit typical hollow profiles and in some cases transient poloidal asymmetric features have been observed in 2D images.

Extensive analysis of disruptions in JET has helped advance the understanding of trends of disruption-generated runaway electrons. Tomographic reconstruction of the soft x-ray emission has made possible a detailed observation of the magnetic flux geometry evolution during disruptions. With the aid of soft and hard x-ray diagnostics runaway electrons have been detected at the very beginning of disruptions. A study of runaway electron parameters has shown that an approximate upper bound for the conversion efficiency of pre-disruptive plasma currents into runaways is about 60% over a wide range of plasma currents in JET. Runaway generation has been simulated with a test particle model in order to verify the results of experimental data analysis and to obtain the background for extrapolation of the existing results onto larger devices such as ITER. It was found that close agreement between the modelling results and experimental data could be achieved if in the calculations the post-disruption plasma electron temperature was assumed equal to 10 eV and if the plasma column geometry evolution is taken into account in calculations. The experimental trends and numerical simulations show that runaway electrons are a critical issue for ITER and, therefore, the development of mitigation methods, which suppress runaway generation, is an essential task.

At JET, the NE213 liquid scintillator is being validated as a diagnostic tool for spectral measurements of neutrons emitted from the plasma. Neutron spectra have to be unfolded from the measured pulse-height spectra, which is an ill-conditioned problem. Therefore, use of two independent unfolding methods allows for less ambiguity on the interpretation of the data. In parallel to the routine algorithm MAXED based on the Maximum Entropy method, the Minimum Fisher Regularisation (MFR) method has been introduced at JET. The MFR method, known from two-dimensional tomography applications, has proved to provide a new transparent tool to validate the JET neutron spectra measured with the NE213 liquid scintillators. In this article, the MFR method applicable to spectra unfolding is briefly explained. After a mention of MFR tests on phantom spectra experimental neutron spectra are presented that were obtained by applying MFR to NE213 data in selected JET experiments. The results tend to confirm MAXED observations.

JET (Joint European Torus) is the largest tokamak in the world and the only fusion facility able to operate with Tritium, the fusion fuel, and Beryllium, the ITER first wall material. JET also features the most complete remote handling equipment for invessel maintenance. As a multinational research center, JET provides logistic experience in preparing for operation of the global facility, tokamak ITER. Experiments on JET are focused on ITER-relevant studies, in particular on detailing the operational scenarios (EL My H-modes and advanced regimes), on enhancing the heating systems, on developing diagnostics for burning plasmas etc. Pioneering real-time control techniques have been implemented that maximize performance and minimize internal disturbances of JET plasmas. In helium plasmas, ion cyclotron heating (ICRH) created fast α-particles, mimicking their populations in future burning plasmas. The recent successful Trace Tritium campaign provided important new data on fuel transport. Current enhancements on JET include a new ITER-like ELM-resilient high power ICRH antenna (7 MW) and over twenty new diagnostics that will further extend the JET scientific capabilities and push the facility even closer to the ITER parameters. A special mention is given to the involvement of the fusion experts from Association EURATOM-IPP.CR, who have been actively participating in the collective use of JET facility for more than three years.

A rapid algorithm for tomographic inversion is presented that allows post-discharge reconstruction of the soft x-ray (SXR) emissivity evolution in the Tokamak a configuration variable (TCV). Simultaneously, it determines the centre of gravity of the emissivity that may serve as a reliable measurement of the plasma position, that is independent of magnetic measurements. Tests of the algorithm were performed using shaped phantom datasets and all past SXR measurements were processed to obtain plasma position statistics. Multiple linear regression is applied in matching the results with the magnetic axis data. Systematic functionalities including magnetic field dependence and drift of the magnetic axis are observed. The joint resolution limit of magnetic and SXR position measurements at TCV is derived from the residual discrepancies and equals 2.2 and 3.1 mm in the radial and vertical directions, respectively.

Recently, an experimental campaign has been launched on TCV with the aim of exploring and extending the limits of the operating space. The vertical position control system has been optimized, with the help of extensive model calculations, in order to allow operation at the lowest possible stability margin. In addition, the growth rate of the axisymmetric instability has been minimized by choosing optimum values for the plasma triangularity and squareness and by operating close to the current limit imposed by the n = 1 external kink mode. These measures have allowed us to reach record values of elongation, k = 2.8, and normalized current, I-N = 3.6, in a tokamak with standard aspect ratio, R/a = 3.5.

Sawtooth inversion radii and profile peaking factors of a large variety of ohmic and ECH heated L mode plasmas, including, elongations up to 2.6 and triangularities between -0.5 and 0.75, have been investigated in the TCV tokamak. In ohmic plasmas, normalized inversion radii and electron temperature profile peaking factors (corrected for sawtoothing effects) depend solely on the parameter (j)/q(0)j(0), irrespective of plasma shape. With ECH this parameter remains the main scaling parameter. Density profiles are well described as functions of poloidal flux, in agreement with turbulent equipartition theories. Parameter conversions are also provided that allow the observed scalings to be expressed using the conventional scaling variables q(95), delta(95) and kappa(95).

Direct measurements of fast electrons, which are produced in high-temperature plasma and escape from tokamak-type facilities, are of particular interest for ITER and future fusion devices, where intense runaway electrons (RE) can significantly damage the first wall components. Therefore, the runaway control and mitigation based on credible measuring methods should be developed already in present devices. A team from the National Centre for Nuclear Research (NCBJ), Poland, developed special probes equipped with Cherenkov-type detectors for measurements of the fast electrons within edge plasmas of tokamaks. Studies of the fast runaway electrons were extensively carried out at the COMPASS tokamak at the Institute of Plasma Physics (IPP) in Prague during experimental campaigns in 2014–2016. In order to investigate an electron-beam energy distribution a three-channel probe equipped with the Cherenkov-type detectors sensitive to electrons of different energies has been constructed. The measurements performed by means of these detectors showed that the first fast electron peak appears usually in the current ramp-up phase, even before the hard X-rays (HXR) pulse. Some electron signals can also be observed during subsequent HXR emissions. However, the most distinct electron peaks in all energy channels appear mainly during the plasma disruption. A correlation of Cherenkov signals with the MHD activity was also studied.

The behaviour of tungsten in the core of hybrid scenario plasmas in JET with the ITER-like wall is analysed and modelled with a combination of neoclassical and gyrokinetic codes. In these discharges, good confinement conditions can be maintained only for the first 2?3?s of the high power phase. Later W accumulation is regularly observed, often accompanied by the onset of magneto-hydrodynamical activity, in particular neoclassical tearing modes (NTMs), both of which have detrimental effects on the global energy confinement. The dynamics of the accumulation process is examined, taking into consideration the concurrent evolution of the background plasma profiles, and the possible onset of NTMs. Two time slices of a representative discharge, before and during the accumulation process, are analysed with two independent methods, in order to reconstruct the W density distribution over the poloidal cross-section. The same time slices are modelled, computing both neoclassical and turbulent transport components and consistently including the impact of centrifugal effects, which can be significant in these plasmas, and strongly enhance W neoclassical transport. The modelling closely reproduces the observations and identifies inward neoclassical convection due to the density peaking of the bulk plasma in the central region as the main cause of the accumulation. The change in W neoclassical convection is directly produced by the transient behaviour of the main plasma density profile, which is hollow in the central region in the initial part of the high power phase of the discharge, but which develops a significant density peaking very close to the magnetic axis in the later phase. The analysis of a large set of discharges provides clear indications that this effect is generic in this scenario. The unfavourable impact of the onset of NTMs on the W behaviour, observed in several discharges, is suggested to be a consequence of a detrimental combination of the effects of neoclassical transport and of the appearance of an island.

During the initial operation of the JET ITER-like wall, particular attention was given to the characterization of the Ion Cyclotron Resonance Frequency (ICRF) heating in this new metallic environment. In this contribution we compare L-modes plasmas heated by ICRF or by Neutral Beam Injection (NBI). ICRF heating as expected led to a much higher centrally peaked power deposition on the electrons and due to the central fast ion population to stronger sawtooth activity. Surprisingly, although a higher bulk radiation was observed during the ICRF phase, the thermal plasma energy was found similar for both cases, showing that a higher radiation inside the separatrix was not incompatible with an efficient central heating scheme. The higher radiation was attributed to the presence Tungsten (W). Tomographic inversion of SXR emissions allowed a precise observation of the sawtooth effect on the radiation pattern. W concentration profiles deconvolved from SXR emission showed the flattening of the profiles due to sawtooth for both heating and the peaking of the profiles in the NBI case only hinting for extra transport effect in the ICRF case.

In 2011/12, JET started operation with its new ITER-Like Wall (ILW) made of a tungsten (W) divertor and a beryllium (Be) main chamber wall. The impact of the new wall material on the JET Ion Cyclotron Resonance Frequency (ICRF) operation was assessed and also the properties of JET plasmas heated with ICRF were studied. No substantial change of the antenna coupling resistance was observed with the ILW as compared with the carbon wall. Heat-fluxes on the protecting limiters close the antennas quantified using Infra-Red (IR) thermography (maximum 4.5 MW/m2 in current drive phasing) are within the wall power load handling capabilities. A simple RF sheath rectification model using the antenna near-fields calculated with the TOPICA code can well reproduce the heat-flux pattern around the antennas. ICRF heating results in larger tungsten and nickel (Ni) contents in the plasma and in a larger core radiation when compared to Neutral Beam Injection (NBI) heating. Some experimental facts indicate that main-chamber W components could be an important impurity source: the divertor W influx deduced from spectroscopy is comparable when using RF or NBI at same power and comparable divertor conditions; the W content is also increased in ICRF-heated limiter plasmas; and Be evaporation in the main chamber results in a strong and long lasting reduction of the impurity level. The ICRF specific high-Z impurity content decreased when operating at higher plasma density and when increasing the hydrogen concentration from 5% to 20%. Despite the higher plasma bulk radiation, ICRF exhibited overall good plasma heating efficiency; The ICRF power can be deposited at plasma centre and the radiation is mainly from the outer part of the plasma. Application of ICRF heating in H-mode plasmas started, and the beneficial effect of ICRF central electron heating to prevent W accumulation in the plasma core could be observed.

Équipe 107 : Physique des plasmas chauds; International audience; In 2011/12, JET started operation with its new ITER-Like Wall (ILW) made of a tungsten (W) divertor and a beryllium (Be) main chamber wall. The impact of the new wall materials on the JET Ion Cyclotron Resonance Frequency (ICRF) operation is assessed and some important properties of JET plasmas heated with ICRF are highlighted. A similar to 20% reduction of the antenna coupling resistance is observed with the ILW as compared with the JET carbon (JET-C) wall. Heat-fluxes on the protecting limiters close the antennas, quantified using Infra-Red thermography (maximum 4.5MW/m(2) in current drive phasing), are within the wall power load handling capabilities. A simple RF sheath rectification model using the antenna near-fields calculated with the TOPICA code can reproduce the heat-flux pattern around the antennas. ICRF heating results in larger tungsten and nickel (Ni) contents in the plasma and in a larger core radiation when compared to Neutral Beam Injection (NBI) heating. The location of the tungsten ICRF specific source could not be identified but some experimental observations indicate that main-chamber W components could be an important impurity source: for example, the divertor W influx deduced from spectroscopy is comparable when using RF or NBI at same power and comparable divertor conditions, and Be evaporation in the main chamber results in a strong reduction of the impurity level. In L-mode plasmas, the ICRF specific high-Z impurity content decreased when operating at higher plasma density and when increasing the hydrogen concentration from 5% to 15%. Despite the higher plasma bulk radiation, ICRF exhibited overall good plasma heating performance; the power is typically deposited at the plasma centre while the radiation is mainly from the outer part of the plasma bulk. Application of ICRF heating in H-mode plasmas has started, and the beneficial effect of ICRF central electron heating to prevent W accumulation in the plasma core has been observed.

The operating space of the TCV tokamak has been extended to higher elongation, kappa = 2.8, and higher normalized current, I-N = I-p/aB = 3.6MA/mT, than were previously attainable. This has been achieved by optimizing the parameters of the vertical position control system and by optimizing the plasma shape. Experimental current and beta limits were found to be consistent with ideal MHD stability calculations, using measured plasma shapes and profiles. The sawtooth period was measured as a function of elongation and sawtooth oscillations are found to disappear, at high elongation, when the internal inductance drops below l(i) similar to 0.69. MHD events at high elongation have been analysed, using magnetic fluctuation measurements. The electron energy confinement time at high elongation, 2.2 < kappa < 2.7, shows a strong dependence on density and power flux through the plasma surface, but no statistically significant dependence on K. Finally, we have applied off-axis, second harmonic, X-mode electron cyclotron resonance heating (ECRH) in order to flatten the current profile and improve vertical stability. This gave access to previously inaccessible domains in kappa-IN space.

Diagnostics of fusion products, in particular of energetic charged particles, is a challenging task that requires further development. In this respect, activation probes were tested in JET, TEXTOR and ASDEX-U tokamaks as a novel and robust diagnostics, compatible with harsh conditions in future fusion reactors. The activation probe consists of a set of pure, well-characterized materials in which measureable amounts of radioisotopes are produced from nuclear reactions due to energetic particles. The production of each radioisotope can be modelled by an energy dependent yield curve derived from the specific nuclear reaction cross section data. In this contribution, the JET activation probe data will be reviewed with focus on possible spectral reconstruction of the incident fusion particles. A method will be outlined to reliably unfold neutron and proton energies, based on Tikhonov regularisation. First results of the neutron and proton energy spectra unfolded from the activation data will be presented. Although the analyses of the proton data are more complex due to several geometrical factors, the initial results demonstrate the potential of this robust diagnostic method, provided that the ill-conditioned task of unfolding is carefully implemented.

This contribution reviews an important example of current developments in diagnostic systems and data analysis tools aimed at improved understanding and control of transport processes in magnetically confined high temperature plasmas. The choice of tungsten for the plasma facing components of ITER and probably also DEMO means that impurity control in fusion plasmas is now a crucial challenge. Soft X-ray (SXR) diagnostic systems serve as a key sensor for experimental studies of plasma impurity transport with a clear prospective of its control via actuators based mainly on plasma heating systems. The SXR diagnostic systems typically feature high temporal resolution but limited spatial resolution due to access restrictions. In order to reconstruct the spatial distribution of the SXR radiation from line integrated measurements, appropriate tomographic methods have been developed and validated, while novel numerical methods relevant for real-time control have been proposed. Furthermore, in order to identify the main contributors to the SXR plasma radiation, at least partial control over the spectral sensitivity range of the detectors would be beneficial, which motivates for developments of novel SXR diagnostic methods. Last, but not least, semiconductor photosensitive elements cannot survive in harsh conditions of future fusion reactors due to radiation damage, which calls for development of radiation hard SXR detectors. Present research in this field is exemplified on recent results from tokamaks COMPASS, TORE SUPRA and the Joint European Torus JET. Further planning is outlined.

MeV particle loss measurements from fusion plasmas, in particular alpha particles, remain difficult in large fusion devices and further R&D is needed for ITER. This paper describes the first attempt to measure 3 MeV escaping fusion protons emitted from TEXTOR tokamak plasmas using activation technique. This technique was successfully demonstrated, initially, in 2006 on the JET tokamak. An ion camera equipped with a collimator and several types of activation detectors was installed inside the TEXTOR vacuum vessel to perform these measurements. After irradiation, the detectors were analyzed using ultra low level gamma-ray spectrometry at the HADES underground laboratory. 3 MeV escaping fusion protons were detected in larger number −∼6 times more - compared to earlier measurements using this technique on JET. Another major progress was the reduction of the cooling time by a factor of 50, which made possible to detect radionuclides with half-life of less than 90 min.

This paper is focused on the soft x-ray (SXR) tomography system setup at Tore Supra (DTOMOX) and the recent developments made to automatically get precise information about plasma features from inverted data. The first part describes the main aspects of the tomographic inversion optimization process. Several observations are made using this new tool and a set of shape factors is defined to help characterizing the emissivity field in a real-time perspective. The second part presents a detailed off-line analysis comparing the positions of the magnetic axis obtained from a magnetic equilibrium solver, and the maximum of the reconstructed emissivity field for ohmic and heated pulses. A systematic discrepancy of about 5 cm is found in both cases and it is shown that this discrepancy increases during sawtooth crashes. Finally, evidence of radially localized tungsten accumulation with an in-out asymmetry during a lower hybrid current drive pulse is provided to illustrate the DTOMOX capabilities for a precise observation of local phenomena.

Tomography is a technique which is widely applied to fusion plasmas as it can provide improved understanding of plasma emissivity distributions. It is challenging because of the sparse nature of data available from the measured plasma projections. An optimised version of robust and fast tomographic algorithm based on the Tikhonov regularisation constrained to Minimum Fisher Information is presented in this contribution. A new regularisation matrix enforcing preferential emissivity smoothness along magnetic flux surfaces is introduced. The paper also details application of advanced numerical methods which lead to a substantial decrease in computation time. Subsequent implementation of fast presolvers of the inverse problem further contributes to the algorithm's efficiency and also an improved stability of the tomography reconstruction. Finally, reliability and performance of the tomography algorithm is exemplified by the reconstruction of soft X-ray data evolution following tungsten ablation into a JET plasma. The resulting speed of reconstruction is compared to other referenced tomographic algorithms.

The contribution focuses on plasma tomography via the minimum Fisher regularisation (MFR) algorithm applied on data from the recently commissioned tomographic diagnostics on the COMPASS tokamak. The MFR expertise is based on previous applications at Joint European Torus (JET), as exemplified in a new case study of the plasma position analyses based on JET soft x-ray (SXR) tomographic reconstruction. Subsequent application of the MFR algorithm on COMPASS data from cameras with absolute extreme ultraviolet (AXUV) photodiodes disclosed a peaked radiating region near the limiter. Moreover, its time evolution indicates transient plasma edge cooling following a radial plasma shift. In the SXR data, MFR demonstrated that a high resolution plasma positioning independent of the magnetic diagnostics would be possible provided that a proper calibration of the cameras on an x-ray source is undertaken.

Notwithstanding the advances of the past decades, significant developments are still needed to satisfactorily diagnose "burning plasmas." D-T plasmas indeed require a series of additional measurements for the optimization and control of the configuration: the 14 MeV neutrons, the isotopic composition of the main plasma, the helium ash, and the redistribution and losses of the alpha particles. Moreover a burning plasma environment is in general much more hostile for diagnostics than purely deuterium plasmas. Therefore, in addition to the development and refinement of new measuring techniques, technological advances are also indispensable for the proper characterization of the next generation of devices. On JET an integrated program of diagnostic developments, for JET future and in preparation for ITER, has been pursued and many new results are now available. In the field of neutron detection, the neutron spectra are now routinely measured in the energy range of 1-18 MeV by a time of flight spectrometer and they have allowed studying the effects of rf heating on the fast ions. A new analysis method for the interpretation of the neutron cameras measurements has been refined and applied to the data of the last trace tritium campaign (TTE). With regard to technological upgrades, chemical vapor deposition diamond detectors have been qualified both as neutron counters and as neutron spectrometers, with a potential energy resolution of about one percent. The in situ calibration of the neutron diagnostics, in preparation for the operation with the ITER-like wall, is also promoting important technological developments. With regard to the fast particles, for the first time the temperature of the fast particle tails has been obtained with a new high purity Germanium detector measuring the gamma emission spectrum from the plasma. The effects of toroidal Alfven eigenmodes modes and various MHD instabilities on the confinement of the fast particles have been determined with a combination of gamma ray cameras, neutral particle analyzers, scintillator probe, and Faraday cups. From a more technological perspective, various neutron filters have been tested to allow measurement of the gamma ray emission also at high level of neutron yield. [doi:10.1063/1.3502038]

In ITER, α particle loss measurements will be required in order to understand the alpha particle physics. Techniques capable of operating in a fusion reactor environment need further development. Recent experimental studies on JET demonstrated the potential of nuclear activation to measure the flux of escaping MeV ions. New results from MeV ion induced activation of metallic, ceramic, and crystal samples placed near the plasma edge are reported. Activation products were measured as function of orientation with respect to the magnetic field as well as function of the distance to the plasma. Sample activity was measured using ultralow-level gamma-ray spectrometry. Distribution of 14.68 MeV fusion proton induced activation products is strongly anisotropic in agreement with simulations and falls off sharply with increasing distance to the plasma. Prospects for using the technique in ITER are discussed.

The integrated spectroscopic system for visible plasma radiation, soft X-ray, and bolometric measurements has been designed for the COMPASS tokamak. This diagnostic allows tomographic reconstruction at a few microseconds time base and features an improved spatial resolution about 1 cm in the pedestal region, where the highest gradient of plasma pressure drives turbulent structures and their propagation towards the wall. In combination with other developed diagnostics on COMPASS, it represents an ideal tool for a measurement and characterization of these radiating plasma structures, which are responsible for an extremely high particle/energy transport across magnetic field. The design of the new integrated multi-channel spectroscopic diagnostics is reviewed and corresponding technical challenges are described. Additionally, the methods, which will be used for data processing, are briefly mentioned and target physics is discussed.

JET's recent experimental programme proved that 'burning plasma diagnostics', i.e. neutron, alpha particle, He ash, and fuel mixture measurements, can provide very useful information about crucial physical aspects of great reactor relevance. First of all, several of these diagnostics can improve the diagnostic capability of the ion fluid significantly. During TTE spatially resolved neutron measurements at JET were essential in obtaining the isotopic composition and the transport of the hydrogen isotopes, allowing a direct comparison between the measured transport coefficients and the neoclassical theory. The neutron emission profiles can also give crucial indications for assessing the merits of various heating schemes and their current drive capability. Neutron spectroscopy in its turn provides a clear and direct measurement of the temperature and the velocity distribution of the fuel ions. For example, the dependence of the toroidal velocity from the ion cyclotron radiofrequency heating phasing was clearly seen during TTE. The requirements of accurate neutron measurements are also promoting considerable research in detector technology, in particular in the fields of compact spectrometers and solid state detectors. 'Burning plasma' diagnostics can also strongly contribute to the physics of energetic particles and their interaction with the main plasma. gamma-Ray spectroscopy is now an established method to determine the spatial localization, to visualize the trajectories of the alpha particles and the fast deuterons and to obtain estimates of their slowing down. A completely new method to detect the energetic particles, exploiting the line intensity ratio of extreme ultraviolet radiation emitted by suitable extrinsic impurities, is also being pursued. This technique allows investigating the energy range below 600 keV, extremely interesting for the study of wave-particle interactions.

The TCV tokamak (R = 0.88 in, a < 0.25 in, BT < 1.54 T) is equipped with six 0.5 MW gyrotron sources operating at 82.7 GHz for second harmonic X mode ECH. By distributing the ECCD current sources over the discharge cross section, fully driven stationary plasmas with I-p = 210 kA, n(e0) = 2 x 10(19) m(-3), T-e0 approximate to 4 keV, were obtained for the full discharge duration of 2 s. Highly peaked electron temperature profiles with T-e0 up to 12 keV were obtained in central counter-current drive scenarios with off-axis ECH. Absorption measurements using a 118 GHz gyrotron have demonstrated the importance of suprathermal electrons for third harmonic absorption. A coupled heat-particle transport phenomenon known as 'density pumpout', which leads to the expulsion of particles from the plasma core, has been linked to the presence of m = I modes, suggesting that it is due to the existence of locally trapped particles associated with the loss of axisymmetry. Highly elongated discharges have been developed with ohmic heating (kappa < 2.8) and off-axis ECH. The latter exhibit considerably improved vertical stability due to current profile broadening. A 'gateway' for ELMy H modes has been discovered, which allows stationary ohmic ELMy H mode operation over a wide range of elongation, triangularity and density. Divertor detachment experiments suggest the existence of recombination pathways other than three body or radiative processes.

Fully noninductive, steady-state electron cyclotron current drive (ECCD) has been demonstrated for the first time in experiments carried out in the tokamak a configuration variable (TCV) [O. Sauter , Phys. Rev. Lett. 84, 3322 (2000)]. By appropriately distributing six 0.45 MW ECCD sources over the discharge cross section, fully noninductive, stable, and stationary plasmas with I-p up to 210 kA were obtained for the full discharge duration of 1.9 s, corresponding to more than 900 energy confinement times and more than 10 current redistribution times at an average current drive efficiency eta (20CD)=0.01[10(20) A W-1 m(-2)]. These experiments have also demonstrated for the first time the steady recharging of the ohmic transformer using ECCD only. The effect of localized off-axis electron cyclotron heating (ECH) and EC current drive (ECCD) (co- and counter-) is investigated showing that locally driven currents amounting to only 1% of I-p significantly alter sawtooth periods and crash amplitudes. An improved quasi-stationary core confinement regime, with little or no sawtooth activity, has been obtained by a combination of off-axis ECH and on-axis CNTR-ECCD. (C) 2001 American Institute of Physics.

In ITER, magnetic fusion will explore the burning plasma regime. Because such burning plasma is sustained by its own fusion reactions, alpha particles need to be confined (Hazeltine 2010 Fusion Eng. Des. 7–9 85). New experiments using d(3He,p)α and d(d,p)t fusion reaction products were performed in JET. Fusion product loss was measured from MHD-quiescent plasmas with a charged particle activation probe installed at a position opposite to the magnetic field ion gradient drift (see figure 1)—1.77 m above mid-plane—in the ceiling of JET tokamak. This new kind of escaping ion detector (Bonheure et al 2008 Fusion Sci. Technol. 53 806) provides for absolutely calibrated measurements. Both the mechanism and the magnitude of the loss are dealt with by this research. Careful analysis shows measured loss is in quantitative agreement with predictions from the classical orbit loss model. However, the comparison with simulated loss radial profile, although improved compared with previous studies in TFTR, Princeton, US (Zweben et al 2000 Nucl. Fusion 40 91), is not fully satisfactory and potential explanations for this discrepancy are examined.

A new method combining a free-form solution for the neutron emissivity and the ratio method (Bonheure et al 2006 Nucl. Fusion 46 725-40) is applied to the investigation of tritium particle transport in JET plasmas. The 2D neutron emissivity is calculated using the minimum Fisher regularization method (MFR) (Anton et al 1996 Plasma Phys. Control. Fusion 38 1849, Mlynar et al 2003 Plasma Phys. Control. Fusion 45 169). This method is being developed and studied alongside other methods at JET. The 2D neutron emissivity was significantly improved compared with the first MFR results by constraining the emissivity along the magnetic flux surfaces. 1D profiles suitable for transport analysis are then obtained by subsequent poloidal integration. In methods on which previous JET publications are based (Stork et al 2005 Nucl. Fusion 45 S181, JET Team (prepared by Zastrow) 1999 Nucl. Fusion 39 1891, Zastrow et al 2004 Plasma Phys. Control. Fusion 46 B255, Adams et al 1993 Nucl. Instrum. Methods A 329 277, Jarvis et al 1997 Fusion Eng. Des. 34-35 59, Jarvis et al 1994 Plasma Phys. Control. Fusion 36 219), the 14.07 MeV D-T neutron line integrals measurements were simulated and the transport coefficients varied until good fits were obtained. In this novel approach, direct knowledge of tritium concentration or the fuel ratio n(T)/n(D) is obtained using all available neutron profile information, e. g both 2.45 MeV D-D neutron profiles and 14.07MeV D-T neutron profiles (Bonheure et al 2006 Nucl. Fusion 46 725-40). Tritium particle transport coefficients are then determined using a linear regression from the dynamic response of the tritium concentration n(T)/n(D) profile. The temporal and spatial evolution of tritium particle concentration was studied for a set of JET discharges with tritium gas puffs from the JET trace tritium experiments. Local tritium transport coefficients were derived from the particle flux equation Gamma = -D del n(T) + Vn(T), where D is the particle diffusivity and V the convection velocity. Tritium transport in the plasma core approaches the neoclassical level in some cases. For the set of plasmas analysed and within the uncertainty, tritium particle confinement follows the ITER IPB98(y, 2) particle scaling confinement.

Ion cyclotron resonance frequency (ICRF) heating has been an essential component in the development of high power H-mode scenarios in the Jet European Torus ITER-like wall (JET-ILW). The ICRF performance was improved by enhancing the antenna-plasma coupling with dedicated main chamber gas injection, including the preliminary minimization of RF-induced plasma-wall interactions, while the RF heating scenarios where optimized for core impurity screening in terms of the ion cyclotron resonance position and the minority hydrogen concentration. The impact of ICRF heating on core impurity content in a variety of 2.5 MA JET-ILW H-mode plasmas will be presented, and the steps that were taken for optimizing ICRF heating in these experiments will be reviewed.

The Tokamak Configuration Variable (TCV) tokamak (R = 0.88 m, a < 0.25 m, B < 1.54 T) programme is based on flexible plasma shaping and heating for studies of confinement, transport, control and power exhaust. Recent advances in fully sustained off-axis electron cyclotron current drive (ECCD) scenarios have allowed the creation of plasmas with high bootstrap fraction, steady-state reversed central shear and an electron internal transport barrier. High elongation plasmas, kappa = 2.5, are produced at low normalized current using far off-axis electron cyclotron heating and ECCD to broaden the current profile. Third harmonic heating is used to heat the plasma centre where the second harmonic is in cut-off. Both second and third harmonic heating are used to heat H-mode plasmas, at the edge and centre, respectively. The ELM frequency is decreased by the additional power. In separate experiments, the ELM frequency can be affected by locking to an external perturbation current in the internal coils of TCV. Spatially resolved current profiles are measured at the inner and outer divertor targets by Langmuir probe arrays during ELMs. The strong, reasonably balanced currents are thought to be thermoelectric in origin.