Search

ITER will operate with a tungsten divertor, a material featuring surface morphology changes when exposed to helium plasmas, in particular the formation of the so called tungsten fuzz under specific conditions. Investigating interactions between tungsten plasma facing components and helium plasmas in a tokamak environment is therefore a key point to consolidate predictions for the ITER divertor performance and lifetime. To this end, a dedicated helium campaign was performed in the full tungsten WEST tokamak, cumulating ∼2000 s of repetitive L mode discharges. It is shown that conditions for tungsten fuzz formation, as derived from linear devices experiments (incident helium energy E inc > 20 eV, helium fluence >1024 He/m2, surface temperature T surf > 700 °C), were met in the outer strike point (OSP) area of the inertially cooled tungsten divertor elements of WEST. Preliminary inspection of the components after the campaign did not show visible signs of surface modification, but points to significant net erosion in the OSP area. An extensive post mortem analysis is now ongoing to confirm these first indications. These results underline that the complex balance between erosion/redeposition (in particular linked to impurities) and tungsten fuzz formation needs to be taken into account in tokamak conditions.

As part of ITPA-Integrated Operational Scenario activities, this contribution reviews recent experimental characterizations of radio-frequency (RF)-induced scrape-off layer (SOL) modifications of various tokamaks worldwide and of the LArge Plasma Device at UCLA. The phenomenology, as observed using a large variety of measurement techniques, is consistent with expectations from RF sheath rectification. Emphasis is then put on the complex three-dimensional (3D) spatial patterns of the RF–SOL interaction, in relation to the magnetic topology and the spatial distribution of RF currents over the metallic structures surrounding the RF wave launchers. Dependence on the local plasma parameters in the antenna vicinity is also briefly addressed. The final part discusses implications for future devices.

A universal energy partition mechanism between ions and electrons has been confirmed to exist in the scrape-off layer of both the WEST tokamak and the W7-X stellarator. A peaked plasma density structure induced by an infinite magnetic connection length structure is observed to destroy this dependence and enhance the local ion and electron temperature ratio in the stellarator. A theoretical analysis reveals that the ratio of ion and electron parallel heat conduction is predominant in determining this universality, while electrons would further be cooled in the density-peaked region.

This paper summarizes the emissivity measurements performed on the plasma-facing units (PFU) of the WEST lower divertor during the first phase of WEST running with a mix of actively cooled ITER-like PFUs made of bulk tungsten (W) and inertially cooled PFUs made of graphite with a coating of tungsten. In situ assessments of the emissivity and laboratory measurements after removing the W-coated graphite and ITER-grade PFUs from the WEST device are shown. The measurements exhibit a complex pattern with strong emissivity variation as a function of space and time mainly explained with the variation of magnetic equilibrium (strike point location) as well as the plasma performances during the experimental campaigns. The exposed ITER-grade PFU exhibits sharp spatial variation of the emissivity from 0.05 to 0.85 at a monoblock scale (12 mm) at the transition of the erosion (strike point location) and deposition (next to the strike point location) areas on the high-field side. On the low-field side, the emissivity varies from 0.12 at the strike point location, to 0.2 a few cm away in the low-field-side direction. This emissivity range after exposure is much higher than the emissivity variation of unexposed PFU with emissivity from 0.09 to 0.15. In situ observation performed on the W-coated graphite PFU shows a rapid evolution, typically a few pulses, of the emissivity in the inner and outer strike point location. The whole spatial distribution is discussed as well as its variation due to the plasma operation from the start-up of WEST to the removal of the W-coated graphite components.

International audience; The paper discusses the characterization of the three high power steady-state and load-resilient ICRF launchers of WEST before their installation in the tokamak. These launchers have been characterized and validated in low-power experiments (milliwatt range) as well as in experiments at the nominal RF voltages and currents in the TITAN vacuum chamber ($\sim$30 kV and 915 A peak). The successful commissioning of two of the launchers during the WEST C3 campaign at $\sim$1 MW power level is illustrated. Manual and real-time controlled impedance-matching of the launchers are discussed, as well as the validation of their load-resilience. Furthermore, several redundant and complementary protection systems have been validated and are reviewed in the paper.

The new WEST configuration of Tore Supra facility leads to control challenges and the need to handle events of a modern diverted and metallic machine (vertical stabilization, impurity control, …). To address them, a new Plasma Control System (PCS) has been built based on the ASDEX upgrade (AUG) real-time framework called DCS (Discharge Control System). This contribution summarizes the work done during the 5 years of the project from the definition of the PCS concepts to its operation during the WEST campaigns. The integration into the Tore Supra control infrastructure is detailed as well as the different real-time control processes. The efficiency and the versatility of the PCS are illustrated by several examples of plasma operation.

Using a recently installed impurity powder dropper (IPD), boron powder (μm) was injected into lower single null (LSN) L-mode discharges in WEST. IPDs possibly enable real-time wall conditioning of the plasma-facing components and may help to facilitate H-mode access in the full-tungsten environment of WEST. The discharges in this experiment featured I p = 0.5 MA, B T = 3.7 T, q 95 = 4.3, t pulse = 12–30 s, n e,0 ∼ 4 × 1019 m−2, and P LHCD ∼ 4.5 MW. Estimates of the deuterium and impurity particle fluxes, derived from a combination of visible spectroscopy measurements and their corresponding S/XB coefficients, showed decreases of ∼50% in O+, N+, and C+ populations during powder injection and a moderate reduction of these low-Z impurities (∼50%) and W (∼10%) in the discharges that followed powder injection. Along with the improved wall conditions, WEST discharges with B powder injection observed improved confinement, as the stored energy W MHD, neutron rate, and electron temperature T e increased significantly (10%–25% for W MHD and 60%–200% for the neutron rate) at constant input power. These increases in confinement scale up with the powder drop rate and are likely due to the suppression of ion temperature gradient (ITG) turbulence from changes in Z eff and/or modifications to the electron density profile.

Sheared flows are known to reduce turbulent transport by decreasing the correlation length and/or intensity of turbulent structures. The transport barrier that takes place at the edge during improved regimes such as H mode, corresponds to the establishment of a large shear of the radial electric field. In this context, the radial shape of the radial electric field or more exactly of the perpendicular E × B velocity appears as a key element in accessing improved confinement regimes. In this paper, we present the radial profile of the perpendicular velocity measured using Doppler back-scattering system at the edge of the plasma, dominated by the E × B velocity, during the first campaigns of the WEST tokamak. It is found that the radial velocity profile is clearly more sheared in lower single null configuration (with the B × ∇B magnetic drift pointing toward the active X-point) than in upper single null configuration for ohmic and low current plasmas (B = 3.7 T and q 95 = 4.7), consistently with the expectation comparing respectively ‘favourable’ versus ‘unfavourable’ configuration. Interestingly, this tendency is sensitive to the plasma current and to the amount of additional heating power leading to plasma conditions in which the E × B velocity exhibits a deeper well in USN configuration. For example, while the velocity profile exhibits a clear and deep well just inside the separatrix concomitant with the formation of a density pedestal during L–H transitions observed in LSN configuration, deeper E r wells are observed in USN configuration during similar transitions with less pronounced density pedestal.

International audience; Infra-red (IR) thermography is a widely used tool in fusion devices to monitor and to protect the plasma-facing component (PFC) from excessive heat loads. However, with the use of all-metal walls in fusion devices, deriving surface temperature from IR measurements has become more challenging. In this paper, an overview of infra-red measurements in the metallic tokamaks WEST and ASDEX Upgrade (AUG) is reported and the techniques carried out in the modeling and experimental fields to deal with this radiative and fully reflective environment are presented. Experimental characterizations of metallic samples in laboratory and experiments in WEST and AUG reveal that the behavior of both the emission and the reflectance can vary significantly with surface roughness, machining process and as the plasma operation progress. In parallel, the development of a synthetic IR diagnostic has allowed for a better interpretation of the IR images by assessing the reflection patterns and their origin. This has also proven that small-scale change in the emission pattern of beveled PFC can be confused with abnormal thermal events. Numerical solutions to evaluate the contribution of the reflections associated with a variable emissivity in a fully reflective and radiative environment are finally presented.

International audience; WEST is a full W tokamak with an extensive set of diagnostics for heat load measurements especially in the lower divertor. It is composed by infrared (IR) thermography, thermal measurement with thermocouples (TC) and fiber Bragg grating (FBG) embedded few mm below the surface and flush mounted Langmuir probes (LP). A large database including different magnetic equilibrium and input power is investigated to compare the heat load pattern (location, amplitude of the peak and heat flux decay length) on the inner and outer strike point regions as function of the continuous progress achieved in WEST : from the first ohmic diverted plasma (obtained during the second experimental campaign C2 in 2018) up to the high power (up to 8 MW total injected) and high energy (up to 90 MJ injected energy in lower single null configuration) steady state experiments performed in the last experimental campaign (C4 in 2019). Concerning the peak location, a good agreement (

International audience; A two-equation model to self-consistently determine cross-field fluxes in the edge and scrape-off layer region of diverted plasma is used to complete 2D mean-field edge transport description of plasma wall interaction. Inspired by the Reynolds Average Navier-Stokes simulations for neutral fluids, this model is based on the local evolution of the turbulent kinetic energy κ and its dissipation rate ε. These two equations are algebraically derived for RANS modeling and are very slightly modified and adapted to describe self-consistent plasma turbulent transport. The general features of the model are discussed and bridged to the well-known predatorprey and quasilinear models commonly used to investigate plasma transport. Specific closures are proposed based on the interchange turbulence. Results of the 1D model are confronted to experimental evidence by analyzing the computed SOL width and comparing the results to the existing scaling law for L-mode plasmas. Introducing a dependence on the shear of large scale flows, typically the zonal flows, 1D simulations can exhibit an H-mode like transition when increasing the input power, generating an increased stored energy thanks to an interface barrier located at the separatrix. Further 2D plasma-wall interaction simulations for WEST are analyzed that show a good match with the experimental profiles, as well as a ballooned transport driving turbulent transport in the divertor SOL and nearly no transport in the private flux region. The SOL width of WEST is also recovered. These results show the remarkable capability of the κ-ε model to capture key aspects of the physics of turbulent transport throughout the plasma knowing that a unique scalar free parameter is available to tune cross field transport in the whole 2D cross section of the plasma.

International audience; The WEST experiment is currently operating with tungsten plasma-facing components and testing ITER-like divertor monoblocks. In order to support WEST experiments interpretation, numerical analyses were carried out. Starting from WEST experimental data, realistic background plasma conditions were reproduced through SolEdge-EIRENE and used as input for ERO2.0 simulations to investigate tungsten migration. Tungsten contamination due to the different plasma-facing components was modelled under different plasma conditions, highlighting a non-negligible contribution of tungsten coming from the tokamak main chamber. Tungsten penetration factor was computed and used as an indication for tungsten screening by the background plasma at the different tokamak plasma-facing components. Simulations showed the main chamber components to be very weakly screened. Light impurities charge was showed to influence not only tungsten sputtering, but also its probability to enter the confined plasma. Simulations results indicated that even when the tungsten source is not heavily influenced by self-sputtering, contamination of the confined plasma can be strongly impacted by it in low density background plasma conditions. Finally, a one-to-one comparison between tungsten visible spectroscopy at the lower divertor from experimental data and from synthetic diagnostics was performed, showing that it is possible to reproduce a realistic lower divertor signal following experimental evidence on light impurities asymmetry between the targets.

In the full tungsten environment of WEST, during its first phase of operation, around 25% of the pulses exhibited a rapid central electron temperature collapse. In its first phase, WEST plasmas were mostly heated by lower hybrid current drive (LHCD) and ion cyclotron resonance heating (ICRH). In this publication, the collapsing pulses are analysed to understand the key actuators at play. Experimentally, an initial slow reduction of central electron temperature due to a density increase is observed, while the central tungsten profile is flat and constant in time. Then, radiative collapse occurs: the core tungsten profile peaks rapidly, at the same time, the central hard x-ray channel measurement decreases indicating a change in core LHCD absorption. Integrated modelling is used to explore the causality chain. To capture the collapse speed, both, tungsten core peaking and reduction of central LHCD absorption are required. When central LHCD power absorption is reduced, core electron and ion temperature profiles flatten which reduces the tungsten neoclassical thermal screening and leads to the observed core tungsten accumulation.

International audience; In 2018, Ion Cyclotron Resonant Frequency (ICRF) waves were for the first time applied to the WEST plasmascenarios. In ICRF-only plasmas at medium density, or on top of a low level of Lower Hybrid (LH) power, the coupledICRF power increases the plasma energy content. However in discharges with large LH power at high core density, nearlyall the applied ICRF power gets radiated, mainly in the plasma bulk. Both the energy content and conducted power decrease.Two peculiarities of WEST, that may combine, are presently invoked to explain this phenomenology:1) Fast ion ripplelosses, evidenced between TF coils on the baffle, likely degrade the ICRF heating efficiency. 2) RF-enhanced W sources,evidenced on several Plasma-Facing Components (PFCs), likely over-contaminate the high-power plasmas.

This study compares experimental observations on the two challenges of Ion Cyclotron Resonance Heating (ICRH), namely the coupling of waves to the plasma and the enhanced impurity sputtering in EAST and WEST medium size tokamaks.In WEST, Lower Hybrid (LH) power helps improving ICRH coupling. In both machines experiments reveal that fueling from the midplane not only helps to couple waves from nearby antennas like in other devices, but also has an impact on the scrape-off layer (SOL) density in regions that are not magnetically connected to the valves. Localized midplane nozzle valves allow similar (in WEST) or better (in EAST) coupling compared to poloidally distributed valves. Core density control requirements for long-pulse operation, in particular in L-mode regime, however limit the amount of gas that can be injected.During ICRH impurities can contaminate the plasma up to a level detrimental for the operation, e.g. up to 100% of ICRH power can be radiated on WEST in certain high power conditions. In WEST, tungsten (W) production measured by visible spectroscopy increases on all the observed objects during ICRH compared to a reference phase without ICRH. On some components (antenna side limiters, baffle, divertor) the rise is larger than with a similar LH power. The relative contribution of each object and physical process (RF-sheaths, fast ion ripple losses) to core contamination yet remains poorly known. Comparing antenna limiters with W-coating vs low-Z materials would help quantifying the role of these components. In EAST, the core W content, measured by EUV spectroscopy in presence of divertor sources only, is correlated with the total injected power, either from ICRH or LH. Since 2018 the LH guard limiter tiles were W-coated. Their contribution to the core W content appears more important than divertor sources when two-strap ICRH antennas magnetically connected to W components at the midplane is powered, compromising high performance operations.

A series of L-mode discharges in upper single null configuration with more than 30 s duration were performed in the WEST tokamak with the strike line on actively cooled tungsten-coated divertor components. This series of discharges accumulated 20 min of plasma and a strike point fluence of $\approx 2\times {10}^{25}\,{{\rm{m}}}^{-2}$ in two days of operation in attached conditions. The discharges showed good density control, no tungsten accumulation and overall very benign thermal behaviour of the upper divertor. A subset of the discharges were seeded with nitrogen either in the divertor or midplane region to study plasma–wall interaction and associated ammonia formation in this weakly pumped scenario with all-tungsten plasma-facing components. While ammonia signals remained too weak for useful analysis, a strong dependence of nitrogen plasma penetration efficiency and residence time on the injection location was found. This hints towards a currently underestimated plasma-near nitrogen reservoir in these tokamak discharges in divertor configuration.

The work describes a novel approach to the design of a fast, multichannel measurement system for plasma diagnostics [A. Wojenski et al., Fusion Eng. Des. 123, 727 (2016)]. Its main scope is to provide measurements of soft X-ray (SXR) emission during plasma phenomena at the W-Environment in Steady-state Tokamak (WEST), especially for monitoring and tracing tungsten impurities. This paper describes the vertical Gas Electron Multiplier (GEM) camera installed at the WEST [M. Chernyshova et al., J. Instrument. 10, P10022 (2015)]. The designed GEM detector readout board has more than 100 channels, resulting in high-performance requirements for the data acquisition and processing system. The novel system construction approach is that the unit works on the raw signals providing a high quality of the data, especially in the scope of pileup effect analysis. In the case of doubtful results, the source data can be easily reviewed offline. The data selection and transmission are done in Field-Programmable Gate Arrays (FPGAs) on the custom boards with the custom Peripheral Component Interconnect (PCI)-Express Gen2 switch that allows us to register signals from multiple FPGAs and then process the data by complex algorithms [G. Kasprowicz et al., J. Fusion Energy 38, 480 (2019)]. The firmware is replaceable and different working modes can be applied (some under verification): global trigger mode, high-speed data serialization, and extended signal registration. Low level optimized central processing unit software for data readout was also designed [P. Linczuk et al., J. Instrum. 14, C05001 (2019)]. The installation of the system is described due to complex system components’ distribution. The first results of the successful acquisition of the plasma at the WEST are discussed. The corresponding SXR energy and topology spectra were computed. Those are the first technical measurements of the system to ensure verification of data quality.

Before the installation of the first ICRF launcher in WEST, dedicated RF, cold and hot leak tests have been performed in the TITAN test bed [ 1 , 2 ] (Bernard et al., 2011a,b). All the water circuits have been controlled in order to avoid any in-vessel leak. The impedance matching has been checked at low level (mW range) and the high RF voltage standoff has been validated at the nominal value of 27 kV peak. During the WEST C3a campaign, the commissioning of the first antenna has been successfully performed. About 0.6 MW of power has been coupled to the plasma, and the launcher’s load-resilience has been proved.

In this work we present an analysis of fast-camera imaging of Alcator C-Mod's divertor region and of the plasma cross section in the W7-X stellarator. In C-Mod, the dynamics of intermittent edge plasma filaments are imaged at 397,100 frames per second in an oblique view. An automatic pipeline was developed to reconstruct the filament motion in the R, Z plane of the divertor using EFIT field line projections. Next, we added our fast-framing camera to W7-X's diagnostic suite for the 2017 run campaign, the first with a divertor. It has a tangential view of the machine cross section, including part of a divertor module. The view is coupled to the camera via a coherent fiber bundle. The spatial resolution is 2–4 cm and the typical plasma brightness limits the fastest usable frame rate to 30k frames/s. We observe filaments in the plasma boundary that rotate around the magnetic axis during radiative collapse; filaments were observed in steady state in the limiter configuration of W7-X by Kocsis et al. [1]. We also observe for the first time a 1–2 kHz quasi-coherent fluctuation in divertor target emissions in the W7-X “standard” magnetic configuration. Finally, a rough analysis of edge gas fueling efficiency using fast camera data and S/XB coefficients finds that fueling is less efficient at densities higher than 4 × 1019 m−3. Keywords: C-Mod, W7-X, Filaments, Blobs, Turbulence, Fluctuation

International audience; Threshold ionization mass spectrometry (TIMS) is one of two methods envisioned in ITER to quantify the helium (He) fusion product in the exhaust pumping lines during plasma discharges. We present the first demonstration of another potential application of TIMS in a tokamak environment, namely, the analysis of deuterium (D) and He outgassing following a plasma discharge i.e. during the post-discharge. This method has been tested with sub-second temporal resolution in WEST during its first He plasma discharges in the so-called He changeover experimental campaign. The calibration method of TIMS using a D plasma discharge is presented while the uncertainties related to TIMS during rapid pressure variations, i.e. upon plasma breakdown and plasma termination, are discussed. The first results obtained with TIMS during consecutive D and He plasma discharges in the full tungsten (W) tokamak WEST are reported. It is found that the time evolutions for He and D outgassing in the post-discharge are markedly different. On one hand, He outgassing is instantaneous and decays within 60 s until the He signal gets below detection level. On the other hand, D outgassing can reach a maximum up to several tens of seconds after the termination of the plasma and this outgassing can last for about 10 min. These striking differences should be related to different retention and outgassing from WEST plasma facing components, presently constituted of actively-cooled ITER-like W units and inertially cooled W-coated graphite. Potential mechanisms at the origin of the different outgassing behavior for D and He in W plasma facing components are discussed in light of a systematic analysis of the He and D gas balance and a macroscopic rate equation modeling of the D outgassing from the divertor strike points.

International audience; The contamination of core plasma by high-Z impurities, especially tungsten (W), is the main reason of very high level of radiated power in WEST experiments. Intrinsic light impurities, mainly oxygen and carbon, play a dominant role in the sputtering of W on plasma facing components. In this contribution, we present a detailed analysis of WEST experiments supported by numerical modeling performed with the transport code SOLEDGE-EIRENE providing a clear picture of light impurities transport and poloidal distribution. Moreover, making use of SOLEDGE-ERO2.0 simulations, possible strategies to reduce core contamination due to W penetration are presented. .

International audience; Nitrogen (N2) will be used in ITER to enhance the radiative fraction to ~90%, thereby cooling the edge plasma and preventing damage to the plasma-facing components. However, the reactivity of N2 with hydrogen isotopes can lead to the formation of tritiated ammonia (NT3). This should be considered in terms of the in-vessel tritium inventory, the regeneration of the cryo pumps, and the processes in the ITER de-tritiation plant. In the 'W' Environment in Steady-state Tokamak (WEST), a series of long L-mode discharges (~50 s), with a constant N2 seeding from the outer strike point region has been performed on the upper actively cooled divertor. In the absence of active pumping, the N2 balance shows steady-state retention during plasma discharge, and is partially (~35%) released in between discharges. Although a significant amount of N2( 18.65 Pa m3) has been injected, the wall still exhibited N2 pumping capabilities. Under these conditions, as long as this N2 reservoir is not saturated, there is not enough N available for the detectable threshold of ND3 formation to be reached. In these WEST experiments, no ammonia is detected during the pulse or after the pulse in the outgassing phase. These results are consistent with and complementary to the N2 seeded experiments performed in the Joint European Torus (JET) with its ITER-like wall and in the Axially Symmetric Divertor Experiment (ASDEX) upgrade.

The evaluation of the impact of plasma-facing components damage on subsequent plasma operation is an important issue for ITER. During the first phase of the operation of WEST, a few ITER-like divertor plasma-facing units (PFUs) have been installed on the lower divertor. One PFU was pre-damaged under electron beam gun thermal loading, before its installation in WEST, and the subsequent evolution of the damage was studied after the WEST plasma exposure. This paper presents the procedure followed to get the pre-damaged PFU. It consists of the characterization of the response of tungsten samples representative of WEST PFU under high heat flux (HHF) loading, the selection of damage (namely, small cracks, crack network, crack network, and W melt droplets). Finally, according to the WEST plasma loading conditions, the blocks with damage within the PFU and the position of the pre-damaged PFU on the WEST lower divertor are attributed. The first results obtained after an initial plasma exposure in WEST lead to the assessment, as expected with regard to the heat loading conditions, that no major surface aspect modification was found. This result emphasized the possibility of implementing as pre-damaged small droplets of melted tungsten in a high heat-loaded zone for a future WEST experimental campaign.

International audience; Three identical new WEST Ion Cyclotron Resonance Heating (ICRH) antennas have been designed, assembled then commissioned on plasma from 2013 to 2019. The WEST ICRH system is both load-resilient and compatible with long-pulse operations. The three antennas have been successfully operated together on plasma in 2019 and 2020. The load resilience capability has been demonstrated and the antenna feedback controls for phase and matching have been developed. The breakdown detection systems have been validated and successfully protected the antennas. The use of ICRH in combination with Lower Hybrid has triggered the first high confinement mode transitions identified on WEST.

Neutron production rates in fusion devices are determined not only by the kinetic profiles but also the fast ion slowing-down distributions. In this work, we investigate the effect of magnetic configuration on neutron production rates in future deuterium plasmas in the Wendelstein 7-X (W7-X) stellarator. The neutral beam injection, beam and triton slowing-down distributions, and the fusion reactivity are simulated with the ASCOT suite of codes. The results indicate that the magnetic configuration has only a small effect on the production of 2.45 MeV neutrons from thermonuclear and beam-target fusion. The 14.1 MeV neutron production rates were found to be between $1.49 \times 10^{12}$ $\mathrm{s}^{-1}$ and $1.67 \times 10^{12}$ $\mathrm{s}^{-1}$, which is estimated to be sufficient for a time-resolved detection using a scintillating fiber detector, although only in high-performance discharges., 17 pages, 10 figures. Version submitted to Plasma Physics and Controlled Fusion

The Collective Thomson Scattering (CTS) diagnostic measures the scattering spectrum of incident radiation off collective fluctuations in plasmas. In Wendelstein 7-X (W7-X) the diagnostic uses a 140 GHz heating gyrotron as a source of the probing radiation. At this frequency, the CTS spectra are heavily affected by the electron cyclotron emission, and the microwave beam propagation is restricted at typical W7-X plasma parameters. The diagnostic was successfully commissioned in the last experimental campaign and demonstrated ion temperature measurements. However, the signal-to-noise ratio was too low for measuring other quantities such as the fast-ion velocity distribution function or the fuel ion ratio. Currently, the W7-X CTS diagnostic is undergoing an upgrade to a frequency of 175 GHz. This will increase the sensitivity of the diagnostic, since the noise due to electron cyclotron emission will be reduced, and it will relax the constraints on microwave beam propagation in W7-X. Here we present the salient features of the upgraded CTS system and discuss its prospects for both thermal-ion and fast-ion measurements.

Within the framework of R&D activities on the Wendelstein 7-X (W7-X) stellarator machine, the assessment of heat loads onto the plasma facing components (PFCs) is an important aspect. So far, W7-X was operated in short pulses without water cooling of the PFCs. Presently, the device is being prepared for future operation phases with water cooling. The target plates, which receive the highest heat loads, are monitored by a thermography system. The rest of the first wall (heat shield) is not designed to receive convective particle loads, and it is in part poorly monitored by the infrared cameras. Recent studies have shown that for many plasma configurations there are locations on the heat shield which do receive significant convective heat flux. This is in particular true for the heat shields adjacent to the target plates (called baffles) and for configurations with finite plasma pressure, where the magnetic configuration is modified by plasma currents. On the other side, the heat flux limit to the baffle tiles had to be reduced from 0.50 to 0.25 MW/m2. In this work an evaluation of the heat flux to targets and baffles in plasma configurations with finite plasma pressure is presented. To this end, Field Line Diffusion (FLD) calculations have been performed to obtain the heat load pattern distributions for the considered magnetic configurations. The results have been assessed statistically to achieve a measure of certainty in the prediction of an overload in a certain location. The possibility of overloads onto the baffles due to plasma radiation has also been investigated. The results of the entire analysis show that local temperature monitoring by thermocouples in a rather limited number of locations will be sufficient to avoid heat flux overloads of the baffles and heat shields in all magnetic configurations considered.