A core fluctuation diagnostic based on the phase-contrast imaging (PCI) technique has been designed for the JT-60SA tokamak, with the assistance of a synthetic diagnostic coupled to a gyrokinetic code. Using a tangentially viewing geometry, this system would be able to resolve small-scale microturbulence as well as macroscopic fluctuations, with good spatial and temporal resolution, throughout the plasma cross-section and in all plasma regimes. The spatial resolution will be optimal (<5% of the minor radius) in the pedestal region and near the magnetic axis. The accessible wave-number range will cover the main ion-scale instabilities predicted to be at play, and optionally also the electron-scale ones. The new superconducting tokamak JT-60SA, which is due to begin operating in 2021, will be the largest tokamak ever built and the most significant intermediate step in magnetic-confinement fusion before the inception of ITER operations. Turbulence is the primary cause of anomalous transport, one of the primary limiting factors in controlled nuclear fusion. Understanding and possibly controlling turbulence thus remain paramount to the fusion quest, and JT-60SA offers the possibility of performing such studies for the first time in a true reactor-relevant environment, providing also a unique opportunity for code validation. This proposal is supported by a direct modeling effort, employing the gyrokinetic code gene and a PCI synthetic diagnostic. Linear and nonlinear flux-tube simulations have been performed for a typical high-performance scenario, to identify the spectral and spatial areas of interest and illustrate the potential for fruitful theory-experiment comparison. While only electrostatic simulations could be carried to full completion with the computing resources deployed for this study, the work has highlighted the importance of including electromagnetic effects for proper comparisons.

The Tokamak à Configuration Variable (TCV) tokamak is in the midst of an upgrade to further its capability to investigate conventional and alternative divertor configurations. To that end, modular and removable gas baffles have been installed to decrease the coupling between the divertor and the plasma core. The baffles primarily seek to suppress the transit of recycling neutrals to closed flux surfaces. A first experimental campaign with the gas baffles has shown that the baffled divertor remains compatible with a wide range of configurations including snowflake and super- X divertors. Plasma density ramp experiments reveal an increase of the neutral pressure in the divertor by up to a factor ×5 compared to the unbaffled divertor and thereby qualitatively confirm simulations with the SOLPS-ITER code that were used to guide the baffle design. Together with a range of new and upgraded divertor diagnostics, the baffled TCV divertor is now used to validate divertor models for ITER and next step devices with particular emphasis on geometric variations.

In the frame of the TCV Tokamak upgrade, two 84/126 GHz/2 s dual frequency gyrotrons designed by SPC and KIT and manufactured by THALES will be added to the existing EC-System. The first unit has been delivered to EPFL-SPC and tested. In the commissioning configuration, a matching optics unit (MOU) is connected to the gyrotron window. The RF is then coupled to the HE11 mode of a 63.5mm corrugated waveguide and dissipated in a load procured by CNR after 4m of waveguide and 2 miter bends. Owing to the flexible triode gun design giving the possibility to adjust the pitch angle parameter, the specifications were met at both frequencies. At 84 GHz (TE17,5 mode), a power of 0.930 MW was measured in the calorimeter, with a pulse duration of 1.1 s. At the high frequency (126 GHz, TE26,7 mode), a power of 1.04 MW was reached for a pulse length of 1.2 s. Accounting for the load reflection and the ohmic losses in the various subcomponents of the transmission line and the tube, it is estimated that the output power at the gyrotron window is in excess of 1 MW at both frequencies, with an electronic efficiency of 32% and 34% at 84 GHz and 126 GHz respectively. The gyrotron behavior is remarkably robust and reproducible, and the pulse length is limited by external systems that will be improved shortly.

The operational range and the reactor relevance of the TCV experiments are being enhanced by two sets of major upgrades. The first includes the installation of neutral beam injection (NBI) and new electron cyclotron (EC) auxiliary heating sources, to reach ITER relevant beta values and vary the electron to ion temperature ratio. A 15?30?keV, 1 MW tangential NBI system has been operational on TCV since 2015. A second beam of 1 MW, 50?60?keV ion energy, also aligned tangentially but opposite to the first beam, is foreseen to approach beta limits, vary the applied torque through zero and probe suprathermal ion physics. For the EC power, two 0.75 MW gyrotrons at the second harmonic have been installed. The next step will add two 1 MW dual frequency gyrotrons, one of which is currently being commissioned. These heating upgrades will increase the total available power for high-density plasmas from 1.25 MW to 5.0 MW. The rest of the upgrade consists of installing an in-vessel structure to form a divertor chamber of increased closure, to reach higher neutral divertor densities and impurity compression and thereby extend TCV divertor regimes toward more reactor relevant conditions for conventional and advanced divertor configurations. Graphite gas baffles will be installed inside the TCV vessel to delineate divertor and main chamber regions. The first set of baffles features 32 tiles on the high and 64 tiles on the low-field side, with geometry guided by simulations performed using the SOLPS-ITER code. The baffles are expected to be effective for a wide range of divertor configurations, including snowflake and super-X divertors, yet maintain plasma close to the inner wall for improved passive stabilization. The baffle dimensions may be varied in the future to modify the divertor closure. Control of the plasma, neutral and impurity densities will be achieved by a combination of toroidally distributed gas injection valves and impurity seeding, and a possible addition of cryo-condensation pumps. Significant diagnostic developments will be undertaken, to better characterize the divertor plasma, measure power and particle deposition at the strike points, and, specifically, improve our physics understanding of the detachment process.

The first unit of the dual-frequency gyrotron, 84-126GHz/1MW/2s, for the upgrade of the TCV ECH system has been delivered and is presently being commissioned. During a first phase, long-pulse operation (T RF >0.5s) has been achieved and powers in excess of 0.93MW/1.1s and 1MW/1.2s have been measured in the evacuated RF-load at the two frequencies, 84GHz (TE 17,5 mode) and 126GHz (TE 26,7 mode), respectively. Considering the different rf losses in the experimental setup, the power level generated in the gyrotron cavity is in excess of 1.1MW and 1.2MW, with a corresponding electronic efficiency of 35% and 36%. These values are in excellent agreement with the design parameters and would likely lead to a gyrotron total efficiency higher than 50% in case of implementation of a depressed collector. The gyrotron behavior is remarkably reliable and robust with the pulse length extension to 2s presently only limited by external auxiliary systems.

The EC-system of the TCV tokamak is progressively being upgraded with the addition of two MW-class dual-frequency gyrotrons (84 and 126 GHz/2s/1MW). In order to connect the two gyrotrons to the existing low field side and top launchers, new waveguide routing from gyrotron hall to TCV tokamak was designed and dedicated Matching Optics Units (MOU) have been developed. The internal optics of the system have been determined aiming at optimal coupling to the HE11 waveguides. The laws of quasi-optics were used to find quadratic surfaces to shape an incoming Gaussian beam representative of the gyrotron output into a beam matching the proper field distribution at the waveguide entrance and with HE11 content compatible with the system requirements. A solution with one flat (movable) mirror and two shaping mirrors was found and characterized with the physical optics code GRASP. The resulting field distribution is then truncated and projected onto the HE11 component to evaluate the design solution (coupled power at the waveguide entrance >98.4% and HE11 content >96.8% for both frequencies). The model and the results of this analysis will be presented and compared to a model based on the Rayleigh–Sommerfeld scalar diffraction integral. GRASP was also used to evaluate preliminary misalignment effects in terms of coupled power to the waveguide and the MOU design moved to the manufacturing phase. In parallel to the gyrotron integration and to extend the level of flexibility of the TCV EC-system, a modular closed divertor chamber is developed, requiring the X3 top-launcher to be redesigned. Preliminary antenna conceptual design studies including new curvature to cope with the requirements of modularity and flexibility will be presented.

As part of an ongoing divertor upgrade of the TCV tokamak it is planned to add gas baffles to form a divertor chamber of variable closure. The baffles promise to increase the compression of neutral particles in the divertor and, thereby, extend the research on the TCV divertor towards more reactor relevant, highly dissipative divertor regimes. It is foreseen to construct the baffles entirely of polycrystalline graphite that was used for the existing TCV protection files. The thermal considerations of the baffle design are based on the heat loads expected during normal operation, where even an extremely large increase in the power carrying plasma channel towards the baffle over the entire 2 seconds duration of a TCV discharge gives no cause for concern. An electromagnetic analysis considers halo currents flowing through the baffles, which can occur during disruptions, as a worst-case scenario. It is found that a halo current of 250 kA results in an average vertical force in the baffles of up to 950 kN/m(3). The fixture of the baffle files to the vacuum vessel is designed for a maximum tensile stress of 31 MPa and maximum compressive stress of 60 MPa that remains a factor of two below their respective material limits. The obtained results of the thermal, electromagnetic and structural analysis thus validate the proposed baffle design.

Innovative 3D high-frequency magnetic sensors have been designed and manufactured in-house for installation on the Tokamak a Configuration Variable (TCV), and are currently routinely operational. These sensors combine the Low Temperature Co-fired Ceramic (LTCC) and the thick-film technologies, and are in various aspects similar to the majority of the inductive magnetic sensors currently being procured for ITER (290 out of 505 are LTCC-1D). The TCV LTCC-3D magnetic sensors provide measurements in the frequency range up to 1MHz of the perturbations to the toroidal (quasi-parallel: delta B-TOR(similar to)delta B-PAR), vertical (quasi-poloidal: delta B-V(ER)similar to delta B-PO(L)), and radial (delta B-RAD) magnetic field components, the latter being generally different from the component normal to the Last Closed Flux-Surface (delta B-NOR). The LTCC-3D delta B-RAD measurements improve significantly on the corresponding data with the saddle loops, which are mounted onto the wall and have a bandwidth of (similar to)3 kHz (due to the wall penetration time). The LTCC-3D delta B-TOR measurements (not previously available in TCV) provide evidence that certain MHD modes have a finite delta B-P(AR) at the LCFS, as recently calculated for pressure-driven instabilities. The LTCC-3D delta B-PO(L) measurements allow to cross-check the data obtained with the Mirnov coils, and led to the identification of large EM noise pick-up for the Mirnov DAQ. The LTCC-3D data for delta B-POL agree with those obtained with the Mirnov sensors in the frequency range where the respective data acquisition overlap, routinely up to 125kHz, and up to 250kHz in some discharges, when the EM noise pick-up on the Mirnov DAQ is removed. Finally, we look at what lessons can be learnt from our work for the forthcoming procurement, installation and operation of the LTCC-1D sensors in ITER.

A variable configuration V-band heterodyne Doppler back-scattering diagnostic has been recently made operational in the tokamak à configuration variable. This article describes the hardware setup options, flexible quasi-optical launcher antenna, data-analysis techniques, and first data. The diagnostic uses a fast arbitrary waveform generator as the main oscillator and commercial vector network analyzer extension modules as the main mm-wave hardware. It allows sweepable single or multi-frequency operation. A flexible quasi-optical launcher antenna allows 3D poloidal (10°-58°) and toroidal (-180° to 180°) steering of the beam with 0.2° accuracy. A pair of fast HE11 miter-bend polarizers allow flexible coupling to either O or X mode and programmable polarization changes during the shot. These have been used to measure the magnetic-field pitch angle in the edge of the plasma by monitoring the backscattered signal power. Ray-tracing simulations reveal an available k⊥ range between 3 and 16 cm-1 with a resolution of 2-4 cm-1. Perpendicular rotation velocity estimates compare well against ExB plasma poloidal rotation estimates from charge exchange recombination spectroscopy.

The recently completed MST2 upgrade to the Thomson scattering (TS) system on TCV (Tokamak à Configuration Variable) at the Swiss Plasma Center aims to provide an enhanced spatial and spectral resolution while maintaining the high level of diagnostic flexibility for the study of TCV plasmas. The MST2 (Medium Sized Tokamak) is a work program within the Eurofusion ITER physics department, aimed at exploiting Europe's medium sized tokamak programs for a better understanding of ITER physics. This upgrade to the TCV Thomson scattering system involved the installation of 40 new compact 5-channel spectrometers and modifications to the diagnostics fiber optic design. The complete redesign of the fiber optic backplane incorporates fewer larger diameter fibers, allowing for a higher resolution in both the core and edge of TCV plasmas along the laser line, with a slight decrease in the signal to noise ratio of Thomson measurements. The 40 new spectrometers added to the system are designed to cover the full range of temperatures expected in TCV, able to measure electron temperatures (Te) with high precision between (6 eV and 20 keV) . The design of these compact spectrometers stems originally from the design utilized in the MAST (Mega Amp Spherical Tokamak) TS system located in Oxfordshire, United Kingdom. This design was implemented on TCV with an overall layout of optical fibers and spectrometers to achieve an overall increase in the spatial resolution, specifically a resolution of approximately 1% of the minor radius within the plasma pedestal region. These spectrometers also enhance the diagnostic spectral resolution, especially within the plasma edge, due to the low Te measurement capabilities. These additional spectrometers allow for a much greater diagnostic flexibility, allowing for quality full Thomson profiles in 75% of TCV plasma configurations.

The Swiss Plasma Center (SPC) is planning a divertor upgrade for the TCV tokamak. The upgrade aims at extending the research of conventional and alternative divertor configurations to operational scenarios and divertor regimes of greater relevance for a fusion reactor. The main elements of the upgrade are the installation of an in-vessel structure to form a divertor chamber of variable closure and enhanced diagnostic capabilities, an increase of the pumping capability of the divertor chamber and the addition of new divertor poloidal field coils. The project follows a staged approach and is carried out in parallel with an upgrade of the TCV heating system. First calculations using the EMC3-Eirene code indicate that realistic baffles together with the planned heating upgrade will allow for a significantly higher compression of neutral particles in the divertor, which is a prerequisite to test the power dissipation potential of various divertor configurations.

The importance of localized injection of Electron Cyclotron waves to control Magneto-HydroDynamic instability is well assessed in tokamak physics and the set of four Electron Cyclotron (EC) Upper Launchers (UL) in ITER is mainly designed for this purpose. Each of the 4 ULs uses quasi-optical mirrors (shaping and planes, fixed and steerable) to redirect and focus 8 beams (in two rows, with power close to 1 MW per beam coming from the EC transmission lines) in the plasma region where the instability appears. Small beam dimensions and maximum beam superposition guarantee the necessary localization of the driven current. To achieve the goal of MHD stabilization with minimum EC power to preserve the energy confinement in the outer half of the plasma cross section, optimization of the quasi-optical design is required and a guideline of a strategy is presented. As a result of this process and following the guidelines indicated, modifications of the design (new mirrors positions, rotation axes and/or focal properties) will be proposed for the next step of an iterative process, including the mandatory compatibility check with the mechanical constraints.

The measurement of the plasma current in ITER plays an outstanding role as it is part of the machine protection and is a safety-relevant measurement: it will be used in relation with regulatory limits to show that the operation remains within the safe envelope defined in the ITER license. The Continuous External Rogowski (CER) is an inductive sensor designed for current measurements and located in the casing of 3 Toroidal Field Coils (TFCs). After the completion of the design of the CER, 4 prototypes of the sensor were manufactured and R&D activities were performed under a Grant with the European Domestic Agency (F4E-GRT-012). The work was carried out between 2010 and 2011 by the ITERMAG consortium comprising 3 laboratories: CRPP (Switzerland) as leader, CEA (France) and RFX (Italy). The R&D campaign on CER prototypes consisted in the measurement of about 100 parameters to characterize the CER in terms of electrical, thermal, mechanical and also of vacuum compatibility. From these results, electromagnetic modeling of the CER response was performed. It is demonstrated that the CER fulfills ITER requirements. However, the vacuum compatibility of the prototype has to be improved and solutions to cope with this issue are proposed.

The high-frequency (HF) magnetic sensors for ITER are currently based on a conventional, Mirnov-type pickup coil, with an effective area in the range 0.03 < (NA)(EFF) (m(2)) <0.1; the sensor is required to provide measurements of magnetic instabilities with magnitude around vertical bar delta B/B-theta vertical bar similar to 10(-4) in the 10-kHz to 2-MHz frequency range. The physical, mechanical, and electrical properties of one representative ITER HF pickup coil design have been analyzed with particular attention to the manufacturing and assembly process for the winding pack, as its integrity was found to be of concern when performing a coupled electromagnetic, structural, and thermal analysis of the sensor. Three different options for the guiding grooves in that design have been tested, using copper and tungsten for the winding pack, but none of them has been convincing enough due to the likelihood of breakages of the thin grooving and of the tungsten wire itself Hence, alternative designs still based on a conventional Mirnov-type pickup coil have been explored, and a nonconventional Mirnov-type pickup coil was produced using direct laser cutting of a Type 316 stainless steel hollow tube, avoiding the difficulties encountered during the winding operations for conventional Mirnov-type sensors. This process of manufacturing appears to be acceptable for HF magnetic sensors of Mirnov-type design in ITER, and it is recommended for future prototyping studies, as the effective area of our first prototype, (NA)(EFF) similar to 0.01 m(2), was well below the ITER requirement. The electrical characteristics and the frequency response of all these prototypes were evaluated up to 8 MHz, with the results in good agreement with model calculations. The conventional Mirnov-type prototypes behave as expected in terms of their main electrical properties and should satisfy the present measurement performance requirements. Finally, a direct measurement of the effective area of these sensors has shown that the geometrical value is a sufficiently correct estimate of its actual value at low frequencies (<10 kHz) when the winding pack closely follows the nominal shape of the coil itself

This paper is an overview of work carried out on the design of the ITER high-frequency magnetic diagnostic coil (HF sensor). In the first part, the ITER requirements for the HF sensor are presented. In the second part, the ITER reference design of the HF sensor has been assessed and showed some potential weaknesses, which led us to the conclusion that alternative designs could usefully be examined. Several options have been explored, and are presented in the third part: (a) direct laser cutting a metallic tube, (b) stacking of plane windings manufactured from a tungsten plate by electrical discharge machining, (c) coil using the conventional spring manufacture. In the fourth part, sensors using the low temperature co-fired ceramic technology (LTCC) are presented: (d) monolithic 1D magnetic flux sensors based on LTCC technology, and (e) monolithic 3D magnetic flux sensors based on the same LTCC technology. The solution which showed the best results is the monolithic 3D magnetic flux sensor based on LTCC. (C) 2011 Elsevier B.V. All rights reserved.

The ITER high-frequency (HF) magnetic sensor is currently intended to be a conventional, Mirnov-type, pickup coil, designed to provide measurements of magnetic instabilities with magnitude as low as vertical bar delta B vertical bar similar to 10(-4) G at the position of the sensors and up to frequencies of at least 300 kHz. Previous prototyping of this sensor has indicated that a number of problems exist with this conventional design that are essentially related to the winding process and the differential thermal expansion between the metallic wire and the ceramic spacers. Hence, a nonconventional HF magnetic sensor has been designed and prototyped in-house in different variants using low-temperature co-fired ceramic (LTCC) technology, which involves a series of stacked ceramic substrates with a circuit board printed on them with a metallic ink (silver in our case). A method has then been developed to characterize the electrical properties of these sensors from the direct-current range up to frequencies in excess of 10 MHz. This method has been successfully bench-marked against the measurements for the built sensors and allows the electrical properties of LTCC prototypes to be predicted with confidence and without the need of actually building them, which therefore significantly simplifies future research and development (R&D) activities. When appropriate design choices are made, LTCC sensors are found to meet in full the volume occupation constraints and the requirements for the sensor's electrical properties that are set out for the ITER HF magnetic diagnostic system. This nonconventional technology is therefore recommended for further R&D and prototyping work, particularly for a three-dimensional sensor, and possibly using materials more suitable for use in the ITER environment, such as palladium and platinum inks, which could remove the perceived risk of transmutation under the heavy neutron flux that we may have with the Au (to Hg, then to Pb) or the Ag (to Cd) metallic inks currently used in LTCC devices.

The ITER high-frequency (HF) magnetic diagnostic system has to provide essential measurements of MHD instabilities with |δBMEAS/BPOL| ∼ 10−4 (∼1 G) for frequencies up to 2 MHz to resolve toroidal mode numbers (n) in the range |n| = 10 to |n| = 50. A review of the measurement requirements for HF MHD instabilities in ITER was initiated during the TW4 work-program and led to significant interest for physics and real-time control issues in measuring modes with |δBMEAS| as low as ∼10−3 G at the position of the sensors, with |n| ≤ 30 and poloidal mode numbers |m| ∼ 2|n| up to |n| ∼ 15, for a frequency range extending up to ∼500 kHz. We have examined the ability of the current ITER design for the individual sensors and the diagnostic system as a whole to meet these needs, and have explored what adjustments to the design (of the individual sensors and/or of the system as a whole) or to the requirements would be needed to meet them when considering different hypothesis for the financial costs and risk management over the ITER life-time. First, we find that the proposed diagnostic layout, with 168 sensors in total, does not meet the more stringent measurement requirements and risk management criteria: these can only be met by a revision of the design requiring 350–500 sensors, depending on different costing and risk management options. Second, we find that the current design for the ITER HF Mirnov-type pick-up coil could be usefully revised.

This paper presents the multiple set of requirements for the ITER magnetic diagnostic systems and the current status of the various R&D activities performed by the EU partners.

The measurement performance of the baseline system design for the ITER high-frequency magnetic diagnostic has been analyzed using an algorithm based on the sparse representation of signals. This algorithm, derived from the SparSpec code [S. Bourguignon et al., Astron. Astrophys., 462, 379 (2007)] has previously been extensively bench marked on real and simulated JET data. To optimize the system design of the ITER high-frequency magnetic diagnostic, we attempt to reduce false detection of the modes and to minimize the sensitivity of the measurement with respect to noise in the data, loss of faulty sensors, and the displacement of the sensors. Using this approach, the original layout design for the ITER high-frequency magnetic diagnostic system, which uses 168 sensors, is found to be inadequate to meet the ITER measurement requirements.Based on this analysis, and taking into account the guidelines for the risk mitigation strategies that are given in the ITER management plan, various attempts at optimization of this diagnostic system have been performed. A revised proposal for its implementation has been developed, which now meets the ITER requirements for measurement performance and risk management. For toroidal mode number detection, this implementation includes two arrays of 50 to 55 sensors and two arrays of 25 to 35 unevenly spaced sensors each on the low-field side and two arrays of 25 to 35 unevenly spaced sensors each on the high-field side. For poloidal mode number detection, we propose six arrays of 25 to 40 sensors each located in nonequidistant machine sectors, not covering the divertor region and, possibly, poloidal angles in the range 75 < vertical bar theta vertical bar(deg) < 105, as this region is the most sensitive to the details of the magnetic equilibrium. In this paper we present the general summary results of this work, for which more details and an overview of our test calculations are reported in the companion paper.

This paper reports the mechanical and electrical tests performed for the prototyping of the ITER high-frequency magnetic sensor and the analysis of the measurement performance of this diagnostic. The current design for the sensor is not suitable for manufacturing for ITER due to the high likelihood of breakages of the un-guided tungsten wire during the winding. A number of alternative designs and manufacturing processes have been investigated, with the Low Temperature Co-fired Ceramic technology giving the best results. The measurement performance of the baseline system design for the high-frequency magnetic diagnostic cannot meet the intended ITER requirements due to its intrinsic spatial periodicities.