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The stellarator Wendelstein 7-X is being prepared for long pulse operation. This includes diagnostics for investigation of plasma wall interaction processes. A versatile optical observation system has been developed for local characterization of the divertor plasma and the divertor target surface. The optical systems consist of two endoscopes each with perpendicular fields of view and the opportunity of tomographic reconstruction. Mirror based optics has been chosen in order to assure good optical properties independent of the wavelength. A narrow field of view allows for high spatial resolution while rotation of the first mirror covers the full poloidal divertor sections. An integrated shutter mechanism and a vacuum window far back minimize coating of optical components. For assessment of change of light transmission, a relative calibration function is implemented. The output light is split into wavelength ranges. Both, cameras equipped with narrow band filters as well as spectrometers are connected. The first endoscope was mounted at W7-X after successfully passing mechanical, optical and functional tests.

During its second operational campaign (OP 1.2a) W7-X operated with an edge island divertor. Much higher densities (1020m−3) than in to the previous, limiter campaign were achieved. In order to asses this improved performance and the nature of transport with regard to the expected higher densities and other parameters such as the electron and ion temperature, the electric field and plasma flow, measurements at the plasma edge were conducted with reciprocating probe, mounted on a Manipulator. The probe contains a set of Langmuir pins, Mach probes, compensation coils and a set of 3D pick up coils for the measurement of the local magnetic field and fluctuations, and an ion sensitive probe. It has been observed, in addition to differences due to the magnetic configurations, that the magnetic topology is sensitive to the plasma conditions, especially to changes due to a evolving toroidal plasma current. Measurements from divertor cameras showed that an increasing toroidal current in the standard configuration broadened the heat flux and shifted the strike line considerably. These down-stream observation can be complemented with the measurements at the mid-plane with the manipulator mounted probes. During the experiments the edge island remnant was identified with the measured edge plasma profiles and compared with the predictions from the field line tracing code. Those electron densities and temperatures have been used as an input for understanding edge transport with the EMC3-EIRENE modeling. The results from EMC3-EIRENE also help improve the caculated density profiles, since the modelling provides a profile of the effective charge Zeff. Keywords: W7-X, Langmuir probes, Magnetic islands, Plasma edge transport

The investigation of plasma wall interactions in terms of erosion processes, fuel retention and new plasma facing components exhibits a scientific gap as discharges in tokamaks and stellarators last only seconds up to a minute. Currently, linear plasma machines are bridging this gap, while Wendelstein 7-X (W7-X) will provide steady-state discharges of up to 30 min in its second operation phase (OP2), in which a divertor manipulator is envisaged to expose samples to reactor-relevant plasma fluxes and fluences. With a design following the limiter lock system at TEXTOR and the DIM-II at ASDEX Upgrade (AUG), the divertor manipulator at W7-X aims to study plasma surface interactions ex-situ by daily/weekly probe exchange and in-situ by embedded diagnostics and an observation system.The design, layout and capabilities of the embedded and observing diagnostics determine the accessible parameters for the in-situ investigation of the exposed probe materials, ranging from plasma edge measurements in the manipulator shaft to intersecting the magnetic island structure on the manipulator head. A versatile laser and observation system adds active surface analysis techniques. To prevent local overheating and excess thermal stress, essential diagnostics for temperature control are identified and combined with first considerations of the manipulator head shaping, based on the field line tracing tool of the W7-X web-service. Keywords: Fusion, Wendelstein 7-X, Plasma diagnostic, Divertor manipulator, Plasma surface interaction

The optimized superconducting stellarator device Wendelstein 7-X (with major radius , minor radius , and plasma volume) restarted operation after the assembly of a graphite heat shield and 10 inertially cooled island divertor modules. This paper reports on the results from the first high-performance plasma operation. Glow discharge conditioning and ECRH conditioning discharges in helium turned out to be important for density and edge radiation control. Plasma densities of with central electron temperatures were routinely achieved with hydrogen gas fueling, frequently terminated by a radiative collapse. In a first stage, plasma densities up to were reached with hydrogen pellet injection and helium gas fueling. Here, the ions are indirectly heated, and at a central density of a temperature of with was transiently accomplished, which corresponds to with a peak diamagnetic energy of and volume-averaged normalized plasma pressure . The routine access to high plasma densities was opened with boronization of the first wall. After boronization, the oxygen impurity content was reduced by a factor of 10, the carbon impurity content by a factor of 5. The reduced (edge) plasma radiation level gives routinely access to higher densities without radiation collapse, e.g. well above line integrated density and central temperatures at moderate ECRH power. Both X2 and O2 mode ECRH schemes were successfully applied. Core turbulence was measured with a phase contrast imaging diagnostic and suppression of turbulence during pellet injection was observed.

The Multi-Purpose-Manipulator (MPM) situated at the outboard mid-plane is a versatile carrier system for diagnostic probes at Wendelstein 7-X (W7-X). It is able to probe into the last scrape-off (SOL) layer close to the last closed flux surface. The manipulator is also has the capability to puff gas for well-localized fueling/impurity seeding into the SOL. In the previous operational campaign of OP 1.2a and 1.2b both hydrogen and methane were injected from the manipulator mounted probes. The effect of hydrogen puffing on the main plasma was observed to be much stronger than that of the methane puffing. This difference is related to the difference in ionization lengths between the two gas species. To support this assumption a puffing of a similar amount of methane and hydrogen was modelled with ERO.

In this study, the ablation mass features related to reflectivity of bulk Molybdenum (Mo) were investigated by a Ti: Sa 6 fs laser pulse at central wavelength 790 nm. The ablated mass removal was determined using Confocal Microscopy (CM) technique. The surface reflectivity was calibrated and measured by a Lambda 950 spectrophotometer as well as a CCD camera during laser ablation. The ablation mass loss per pulse increase with the increasing of laser shots, meanwhile the surface reflectivity decrease. The multi-pulses (100 shots) ablation threshold of Mo was determined to be 0.15 J/cm2. The incubation coefficient was estimated as 0.835. The reflectivity change of the Mo target surface following multi-pulses laser ablation were studied as a function of laser ablation shots at various laser fluences from 1.07 J/cm2 to 36.23 J/cm2. The results of measured reflectivity indicate that surface reflectivity of Mo target has a significant decline in the first 3-laser pulses at the various fluences. These results are important for developing a quantitative analysis model for laser induced ablation and laser induced breakdown spectroscopy for the first wall diagnosis of EAST tokamak.

In the stellarator Wendelstein 7-X with its twisted 3D magnetic field geometry, studies of material migration with respect to first wall components become very important in view of the envisioned long-pulse operation. A variety of erosion/deposition samples were installed on the plasma-facing components exposed at three different nominal heat load levels between 0.1 and 10 MW/m2. After the first successful operation phase in divertor configuration, all the probes at higher and lower load levels were removed, whereas at the intermediate load levels, 352 out of 30 000 screws have been exchanged at selected locations along the toroidal and poloidal directions. The exchanged probes have been analyzed by various measurement techniques. At the higher load levels where the probes were installed within the divertor, heavy erosion has been observed presumably at the strike line positions. Both, erosion and deposition phenomena have been found on the screw heads. The optical reflection measurement profile of the whole plasma vessel show the deposition patterns at similar locations in all the five modules. At the low load level, the Si-wafer probes are under investigation.

International audience; The Wendelstein 7-X (W7-X) optimized stellarator fusion experiment,which went into operation in 2015, has been operating since 2017 with anun-cooled modular graphite divertor. This allowed first divertor physicsstudies to be performed at pulse energies up to 80 MJ, as opposed to 4MJ in the first operation phase, where five inboard limiters wereinstalled instead of a divertor. This, and a number of other upgrades tothe device capabilities, allowed extension into regimes of higher plasmadensity, heating power, and performance overall, e.g. setting a newstellarator world record triple product. The paper focuses on the firstphysics studies of how the island divertor works. The plasma heat loadsarrive to a very high degree on the divertor plates, with only minorheat loads seen on other components, in particular baffle structuresbuilt in to aid neutral compression. The strike line shapes andlocations change significantly from one magnetic configuration toanother, in very much the same way that codes had predicted they would.Strike-line widths are as large as 10 cm, and the wetted areas alsolarge, up to about 1.5 m2, which bodes well for futureoperation phases. Peak local heat loads onto the divertor were ingeneral benign and project below the 10 MW m?2 limit of thefuture water-cooled divertor when operated with 10 MW of heating power,with the exception of low-density attached operation in the high-iotaconfiguration. The most notable result was the complete (in all 10divertor units) heat-flux detachment obtained at high-density operationin hydrogen.

For the study of electrostatic and magnetic properties of filaments characterizing the edge region of the stellarator experiment W7-X, a specifically designed insertable probe head was constructed within the framework of EUROfusion WP.S1 work package in collaboration between Consorzio RFX, IPP Greifswald and Forschungszentrum Mich. The probe head, named High Resolution Probe (HRP), was conceived to be installed on the mid-plane multi-purpose fast reciprocating manipulator on W7-X. Electromagnetic filamentary turbulent structures are found to characterize the edge region of different magnetic configurations including Reversed Field Pinch, stellarator and tokamak, where strong currents are associated also to ELM filamentary structures. The study of those phenomena in W7-X stellarator is of particular interest as the electromagnetic features of filaments are expected to become more relevant with the increase of the local plasma beta. In particular the aim is to provide information about the presence and the features of parallel current density associated to filamentary turbulent or ELM-like structures. Furthermore, the possibility to measure the time evolution of the flow radial profiles using the Mach probe array was considered as a further interesting part of the study, given the strong interplay expected between the turbulent fluctuations and the average flows. Further important information provided is the radial propagation of turbulent flux. The contribution will present the design development, the R&D studies and the applied solutions for the sensors embedded in the probe head. In particular, the presence of 140 GHz ECRH plasma environment represents one of the main challenges for reliable magnetic fluctuation measurements. First measurements were performed during the W7-X experimental campaign OP1.2b, where electromagnetic features of turbulence were measured with the probe, at different magnetic configurations.

Wendelstein 7-X (W7‐X) is the world’s largest superconducting nuclear fusion experiment of the optimized stellarator type. In the first Operation Phase (OP1.1) helium and hydrogen plasmas were studied in limiter configuration. The heating energy was limited to 4 MJ and the main purpose of that campaign was the integral commissioning of the machine and diagnostics, which was achieved very successfully. Already from the beginning a comprehensive set of diagnostics was available to study the plasma. On the path towards high-power, high-performance plasmas, W7‐X will be stepwise upgraded from an inertially cooled (OP1.2, limited to 80 MJ) to an actively cooled island divertor (OP2, 10 MW steady-state plasma operation). The machine is prepared for OP1.2 with 10 inertially cooled divertor units, and the experimental campaign has started recently. The paper describes a subset of diagnostics which will be available for OP1.2 to study the plasma edge, divertor and scrape-off layer physics including those already available for OP1.1, plus modifications, upgrades and new systems. The focus of this summary will be on technical and engineering aspects, like feasibility and assembly but also on reliability, thermal loads and shielding against magnetic fields.

Probe manipulators are a versatile addition to typical plasma edge diagnostics. Equipped with material samples they allow for detailed investigation of plasma–wall interaction processes, such as material erosion, deposition or impurity transport pathways. When combined with electrical probes, a study of scrape-off layer and plasma edge density, temperature and flow profiles as well as magnetic topologies is possible. A mid-plane manipulator is already in operation on Wendelstein 7-X. A system in the divertor region is currently under development. In the present paper we discuss the critical issue of heat and power loads, power redistribution and experimental access to the complex magnetic topology of Wendelstein 7-X. All the aforementioned aspects are of relevance for the design and operation of a probe manipulator in a device like Wendelstein 7-X. A focus is put on the topological region that is accessible for the different coil current configurations at Wendelstein 7-X and the power load on the manipulator with respect to the resulting different magnetic configurations. Qualitative analysis of power loads on plasma-facing components is performed using a numerical tracer particle diffusion tool provided via the Wendelstein 7-X Webservices.

In the first experimental campaign of Wendelstein 7-X (W7-X), a combined probe head mounted on the multi-purpose manipulator has been used to measure the scrape-off layer (SOL) turbulence characteristics. The preliminary experimental results are summarized to illustrate the SOL turbulence properties in the limiter configuration on W7-X. In a standard limiter configuration, significant electrostatic fluctuations can be found in the near SOL, and the dominant frequency of fluctuation power is below 100 kHz. The auto-correlation spectrum power law decay factor is α ≈ −1 below 40 kHz and α ≈ –2 between 50 and 200 kHz. A broadband spectrum appears between 240 and 380 kHz with a low spectral power density, but a high cross-correlation coefficient. The statistical characteristics of turbulence are calculated by the two-point cross-correlation technique. A clear poloidal dispersion relation is found in the spectrum S(kθ, f), propagating along the ion diamagnetic drift direction with a group velocity (below 100 kHz) about 0.56 km/s in the near SOL in the laboratory frame. The poloidal correlation length is around 5–10 mm in SOL. The turbulence phase velocity is about 0.5–1 km/s when close to the last closed flux surface, which is comparable with the poloidal E × B drift speed.

A new ultraviolet-visible-near infrared (UV-VIS-NIR) overview spectroscopy has been developed for routinely monitoring divertor recycling, intrinsic and extrinsic impurity fluxes on Wendelstein 7-X (W7-X). The overview spectroscopy is part of a large integrated system which allows several diagnostic systems to share the same field of view through an endoscope. During the first divertor operation phase on W7-X, a temporary view port was setup for the overview spectroscopy in order to investigate the divertor plasmas and provide reference data for the commissioning of the whole integrated endoscope system. The overview spectroscopy uses a 5-channel spectrometer to cover the wavelength range 300-1100 nm. The reciprocal linear dispersion of the spectrometer are in the range of 0.04-0.19 nm/pixel. The spatial coverage of the system is 10 cm in the vertical direction and 20 cm in the toroidal direction. The first full spectral survey identified H, He, C, O, Fe, Ne, N and Ar lines. The temporal evolutions of hydrogen and impurity radiation with maximum 10 ms resolution were obtained. Spectra fall in the range of filter transmission curves have been investigated for the filter camera systems which will be commissioned along with the endoscope in the next operation phase.

Within the framework of fusion technology new materials for the first wall are investigated. These materials must withstand reactor relevant high neutron fluxes which change structure and properties of the material and radio-activate it. The impact of neutron damage on plasma-material interaction with respect to thermal and particle loads is investigated in customized linear plasma generators in a hot cell. In order to avoid emission of possibly radioactive contamination and to get stable conditioning a remote handling system including lock system for the samples was developed for the Julich plasma generator (JULE-PSI). All mountings and operations like moving and fixing the sample are designed such that maintenance and revisions are minimized at a high level of operational safety and availability. For instance, actuators are arranged outside the vacuum as much as possible. Furthermore in case of problems the radioactive sample can be removed from the vessel in a simple and reliable way. Concepts and detailed solutions are presented.

A broadband multilayer dielectric mirror coating of TiO2/SiO2 was manufactured on stainless steel substrate of grade 1.4429 with a mean specular reflectivity approaching 98% over the wavelength range of 450–670 nm. The coating was tested under thermal conditions as expected for in-port mirrors in ITER, exceeding 1000 h at Tmirror = 260 °C with repeated thermal cycles in vacuum. They were also exposed to water vapour atmosphere at 120 °C. Reflectivity measurements were conducted at elevated temperature and room temperature in between tests. No significant change in specular reflectivity after excursions to 350 °C was found and reflectivity was measured to be stable at elevated temperature. Local defects developed during thermal testing in the form of blisters and flakes with up to 200 μm diameter. The blisters grew and flaked off over several days at elevated temperature. Exposure to water vapour and thermal gradients of +100 K h−1 and −200 K h−1 did not accelerate subsequent generation or growth of defects. Coatings with different interlayer were exposed to elevated temperature to study the impact on adhesion and defect generation.

The First Mirror (M1), as part of the ITER core CXRS diagnostics, is responsible for acquisition and transportation of the optical signal from the plasma to the corresponding spectrometer. The M1 is the most vulnerable component of this system working in severe conditions caused by its location in the direct view of the plasma. Based on the numerical analysis an optimized arrangement of baffles was found for the optical channel. It is shown that such measures like the deeper M1 positioning in the shielding, duct configuration and baffles implementation made it possible to reduce heat loads on the M1 by a factor of about 10 2 –10 3 .

To prolong a lifetime of the ITER first diagnostic mirrors some protective shutters can be engaged. A concept of an elastic shutter that operates frictionless in vacuum has been studied at the Forschungszentrum Julich, Germany. Under actuation two shutter arms (∼2 m long) bend laterally between two pairs of limiting bumpers thus shielding the optical aperture or opening it for measurements. To increase the shutter efficiency the transition time between its open and closed states can be minimized. This demands a fast shutter that operates in fractions of a second and exhibit essentially dynamic behavior, like impacts with the bumpers that cause the shutter arms’ bouncing and oscillations. The paper presents numerical studies of the shutter dynamic behavior using the explicit and implicit 3D FE transient structural modeling. Simple 1D analytical model was developed to predict the shutter impact kinetic energy that mostly determines its further dynamic response. The structure sensitivity to different parameters was studied and ways for its optimization were laid down. A parametric shutter mockup with easily changeable mechanical characteristics was manufactured. A test program aimed for further shutter optimization, basing on the analysis performed and engaging powerful capabilities of the parametric shutter mockup is discussed in the paper.

Optical and laser-based diagnostics in ITER will use mirrors to transmit plasma radiation and laser light to the corresponding detectors and cameras. Mirrors will be sputtered by the fast plasma particles and contaminated by impurities leading to the degradation of the reflectivity and hampering the performance of corresponding diagnostics. Dedicated measures were proposed to minimize the impurity deposition comprising the use of shutters and fins inside diagnostic ducts to trap impurities on their way toward the mirror located in the end of these ducts. Modeling results predict at least 7-fold suppression of the deposition for the duct having four fins located at the distance of a half of a diameter from each other. The Mirror Station (MS) was designed to validate modeling predictions and to study the suppression of deposition inside of diagnostic ducts. The MS contained cylindrical and cone-shaped tubes of different lengths with smooth and shaped geometry of ducts. The MS was exposed in the midplane port of TEXTOR for about 3960 s of plasma operation. After exposure, no drastic suppression of deposition was observed in the cylindrical ducts with fins. In the conical tubes no deposition was detected outlining the advantages of a cone form.

Disruptions in tokamaks seem to be unavoidable. Consequences of disruptions are (i) high heat loads on plasma-facing components, (ii) large forces on the vacuum vessel, and (iii) the generation of runaway electron beams. In ITER, the thermal energy of the plasma needs to be evenly distributed on the first wall in order to prevent melting, forces from vertical displacement events have to be minimized, and the generation of runaway electrons suppressed. Massive gas injection using fast valves is a concept for disruption mitigation which is presently being explored in many tokamaks. Fast disruption mitigation valves based on an electromagnetic eddy current drive have been developed in Julich since the 1990s and models of various sizes have been built and are in operation in the TEXTOR, MAST, and JET tokamaks. A disruption mitigation valve for ITER is of necessity larger with an estimated injected gas volume of ∼20 kPa m3 [7] for runaway electron suppression and all materials used have to be resistant to much higher levels of neutron and gamma radiation than in existing tokamaks. During the last 5 years, the concept for an ITER prototype disruption mitigation valve has been developed up to the stage that a fully functional valve could be built and tested. Special emphasis was given to the development and functional testing of some critical items: (i) the injection chamber seal, (ii) the piston seal, (iii) the eddy current drive, and (iv) a braking mechanism to avoid too fast closure of the valve, which could damage the injection chamber seal. The concept of the valve will be introduced and detailed solutions and testing results for the critical items presented.

The ITER core charge exchange recombination spectroscopy diagnostics (cCXRS) occupies the vacuum vessel upper port #3 and includes, in its generic version, the following in-vessel components: an optical mirror system, a shutter, the diagnostic first wall and the neutron shielding block. The most vulnerable diagnostic mirror is obviously the first one (M1) directly observing the plasma. The M1 reference option is made of a single crystalline molybdenum (ScMo). The paper indicates major aspects influencing the first mirror design and identifies the most reasonable and reliable concept for cCXRS M1 at present. The applicability of the option presented is determined by many reasons, and especially, by the ITER generic upper port plug and its customization flexibility. The largest dimension of the mirror polished face is ∼300 mm. Such large ScMo workpieces are currently not available on the market. The mirror should be designed as an assembly of several ScMo pieces joined together. The M1 design is supported by multifield thermal, electromagnetic and structural analyses. The performed study confirms the feasibility of the proposed solutions. At the same time, the paper indicates numerous technological issues of the M1 unit to be solved in future.

An ICRH antenna system is under construction to be ready for use in the operational phase 1.2 of W7-X. A two strap antenna, with a surface adapted to the 3D shape of the Last Closed Flux Surface (LCFS) of the standard magnetic configuration (m/n = 5/5), will be installed in the equatorial plane on the low field side of W7-X. The antenna system is optimised for plasma heating and wall conditioning in presence of the magnetic field in the frequency range 25–38 MHz. Each strap is short-circuited on one end, is pre-matched using a tunable capacitor on the other end with connection to the RF power source about halfway in the poloidal direction. To allow optimal coupling to different magnetic configurations in W7-X the antenna can be moved radially over a distance of 350 mm. The cooling circuits in the antenna head are designed to sustain cw W7-X plasma operation in a retracted position and 2 MW ICRH pulses with maximum 10 s duration every 300 s close to the LCFS.

Wendelstein 7-X (W7-X), started operation in December 2015 with a limiter configuration. In conjunction with the multi-purpose manipulator, a carrier for fast reciprocating probe systems, the combined probe has been installed. This combined probe is able to measure the local electron temperatures and densities, magnetic field, the electric field and the plasma flow. These parameters are very useful in ascertaining the edge plasma perfomance. In addition, the field line tracing feature of the W7-X webservices was used to calculate the connection length along the path of the probe, for each configuration.

In the Wendelstein 7-X stellarator with its twisted magnetic geometry the investigation of plasma wall interaction processes in 3D plasma configurations is an important research subject. For the upcoming operation phase i.e. OP1.2, three different types of material probes have been installed within the plasma vessel for the erosion/deposition investigations in selected areas with largely different expected heat load levels, namely, ≤10 MW m−2 at the test divertor units (TDU), ≤500 kW m−2 at the baffles, heat shields and toroidal closures and ≤100 kW m−2 at the stainless steel wall panels. These include 18 exchangeable target elements at TDU, about 30 000 screw heads at graphite tiles and 44 wafer probes on wall panels, coated with marker layers. The layer thicknesses, surface morphologies and the impurity contents were pre-characterized by different techniques and subjected to various qualification tests. The positions of these probes were fixed based on the strike line locations on the divertor predicted by field line diffusion and EMC3/EIRENE modeling calculations for the OP1.2 plasma configurations and availability of locations on panels in direct view of the plasma. After the first half of the operation phase i.e. OP1.2a the probes will be removed to determine the erosion/deposition pattern by post-mortem analysis and replaced by a new set for the second half of the operation phase, OP1.2b.

The Multi-Purpose Manipulator (MPM) for Wendelstein 7-X, at the Max-Planck-Institute for Plasma Physics in Greifswald (MPIPP), is used to transport electrical probes and targets to the edge of the inner vessel. It is a lock system, which is attached to the outer port in the equatorial plane of the cryostat vessel. At the parking position, the tip of the probe coincides with the inner vessel wall; a fully controlled movement into the edge plasma for all magnetic field configurations is feasible. The whole functionality of the MPM system is maintained by a distributed control system (DCS) based on Siemens PCS7, which is the recommended standard for machine and diagnostic control at W7-X. Beside standard scenarios like target exchange and generation of ultra-high vacuum conditions a high variety of parameter selections for stroke depth, velocity or acceleration is selectable in the sequence control system within safety limits. A sophisticated error handling facilitates a reliable remotely controlled operation without manual access over longer periods.

A group of edge diagnostics and modelling has been developed for investigation of synergy between 3D edge physics and plasma-wall interactions on Wendelstein 7-X (W7-X). Two endoscopes have been designed for visible and ultraviolet spectroscopy and tomography of the plasma edge, along with infrared thermography of the divertor tiles. 2D profiles of impurities (e.g. helium, carbon) will be measured by two endoscopes viewing the island divertor region in the plasma edge with a spatial resolution of

The investigation of edge plasmas at the stellarator W7-X requires a flexible tool for integration of a variety of different diagnostics as e.g. electrical probes, probing magnetic coils, material collection, or material exposition probes, and gas injection. A multi-purpose manipulator (MPM) system has been developed and attached to the W7-X vessel before the operational phase 1.1. The system was designed as user facility for many diagnostics, which can be mounted on a unique interface without breaking the W7-X vacuum. The manipulator system, located in the equatorial plane, transports the inserted diagnostic probe to the edge of the inner vacuum vessel. From there, the probe can be moved over a maximum distance of 350 mm to different positions inside the plasma with a maximum acceleration and deceleration of 30 m/s2. Acceleration, speed, and stroke depth are individually adjustable and programmable by a PLC system within predetermined limits. The MPM system can be equipped with multifunctional probes and is prepared for cooling/heating of the probe head, gas injection, and a flexible setting of the electrical diagnostic. In the paper, the MPM components, their functionalities, and the interface to the probe head are described. The scope of physical parameters is presented for the development of diagnostic probes applicable in the MPM user facility. This also includes the operational parameters for the movements and the electrical environment for measurements. The operation of the MPM during the OP1.1 is demonstrated on the application of a multipurpose probe head, including electric, magnetic, and thermal probes.

The development of components for fusion devices includes numerous simulations to validate their design compatibility with specified loads according to applicable codes and standards. The strategy of multifield finite element analyses is determined by the iterative nature of the design process. The level of structural complexity dictates the analytical approaches and modeling techniques. Even though nowadays powerful software exploiting the performance of modern computers allows structures to be simulated in a very detailed way, direct meshing of CAD models with all their complexity is often not beneficial for nonlinear and multifield analyses. Reasonable modeling simplifications, submodeling, and integration with analytical modeling and testing can optimize the analysis process. This paper gives examples of simplified modeling for engineering purposes with a clear understanding of the analysis goal and the role of simulations in the design process, such as for the W7-X stellarator busbar system. Engineering methods to perform multifield analyses including electromagnetic transients for ITER diagnostic port plug components are also presented. An engineering material model is discussed for calculating the superconducting coil steel structures operating in the inelastic range.

Fast shutters can play important role for the ITER diagnostics. They protect diagnostic mirrors, especially the first ones closest to the plasma, between measurements, during dwell time and baking. In a nominal mode, as it is assumed, for example, for the ITER upper port plug #3 diagnostics, its shutter stays open ∼1 s and closed ∼21 s. The principal idea of the shutter concept is its ability to make fast transitions between the open and closed positions within fractions of second. A pneumatic actuator produces a pressure force to open or close the shutter. Due to the fast transient nature of the shutter operation, complicated by the parts’ impact interaction, the FE codes using the explicit time integration scheme have an advantage over the implicit ones. The shutter operation is modeled using the explicit solver. Since the shutter dynamic behavior strongly depends on the actuator flood time, different time profiles of the actuator pressure rise (drop) has been verified, including a fast transient with duration of ∼0.1 s. The results of the implicit and explicit solvers are compared. Their pros and cons are pointed out. The system damping, estimations of energy loss and ways to specify damping in FE models are discussed in the paper.

For the ITER core charge exchange recombination spectroscopy (CXRS) upper port plug #3 (UPP) diagnostics, a fast shutter is necessary for the first optical mirrors. The shutter concept for the reference CXRS layout has been well developed and studied focusing on the structure dynamic behavior. This paper presents two shutter concepts for the new baseline layout. Concept-1 is similar to the reference concept, including the helium actuator attached to the diagnostic shielding module (DSM) in the primary vacuum. Concept-2 considers the actuator outside the primary vacuum. It is attached to the back side of the UPP and connected to the shutting mechanism with a push–pull rod. Assembly and disassembly, actuator location, and actuation type are key issues for concept assessment. For a feasibility study, two different leak cases for the shutter concepts have also been analyzed. Concept-1 requires a specialized actuator design compatible with the in-vessel conditions. The actuator size is limited by the lack of available space in the DSM. This option needs complex procedures to access the actuator for maintenance. Concept-2 allows the use of standard pneumatic/hydraulic actuators and other industrial components. Space allocation is needed inside the UPP to accommodate the push–pull rod system. Concept-2 allows independent maintenance procedures for the actuator(s) and the shutting mechanism. It was generally concluded that both options are suitable for the new baseline layout, and are to be studied in detail.

The core charge exchange recombination spectroscopy (cCXRS) upper port plug diagnostic system serves to control and optimize plasma performance. The system contains a labyrinth of optical mirrors delivering the visible light from the plasma to the end spectrometers. Harsh ITER near plasma environment, resulting in high thermal fluxes and severe electromagnetic (EM) loads, dictates a necessity of a robust design for the plug components. The structural response of the plug components to combined loading coming from different sources has to be calculated and analyzed. The global ANSYS EM model featuring the reference cCXRS layout has been upgraded to be in line with the recent ITER generic upper port plug. Impact of different parameters on the plug EM loading was studied. The dynamic structural behavior of the plug (new baseline layout) under the EM forces has been analyzed. The plug lateral deflection is about 9 mm and does not exceed the ITER limit of 12 mm. The cooling and supporting system of the cCXRS first mirrors, located near the plasma, have to cope with significant EM loads and high heat fluxes. The engineering ‘express’ approach, that was recently effectively used to calculate the eddy currents and EM forces for the cCXRS fast shutter, has been further validated for the mirrors. The paper outlines approaches to a multifield analysis of the first mirror.

Fusion alpha particle diagnostics for ITER remain a challenging task. Standard escaping alpha particle detectors in present tokamaks are not applicable to ITER and techniques suitable for fusion reactor conditions need further research and development [1] , [2] . The activation technique is widely used for the characterization of high fluence rates inside neutron reactors. Tokamak applications of the neutron activation technique are already well developed [3] whereas measuring escaping ions using this technique is a novel fusion plasma diagnostic development. Despite low alpha particle fluence levels in present tokamaks, promising results using activation technique combined with ultra-low level gamma-ray spectrometry [4] were achieved before in JET [5] , [6] . In this research work, we use new advanced detector materials. The material properties beneficial for alpha induced activation are (i) moderate neutron cross-sections (ii) ultra-high purity which reduces neutron-induced background activation and (iii) isotopic tailoring which increases the activation yield of the measured activation product. Two samples were obtained from GERDA [7] , an experiment aimed at measuring the neutrinoless double beta decay in 76Ge. These samples, made of highly pure (9 N) germanium highly enriched to 87% in isotope Ge-76, were irradiated in real D–D fusion plasma conditions inside the TEXTOR tokamak. Comparison of the calculated and the experimentally measured activity shows good agreement. Compared to previously investigated high temperature ceramic material [8] , this candidate detector offers better prospects for signal to background S/B ratio, energy resolution and particle selectivity due to a unique alpha particle signature. Applicability to ITER is discussed. Finally, research needs for further development of this diagnostic technique are outlined.

Wendelstein 7-X, a superconducting optimized stellarator built in Greifswald/Germany, started its first plasmas with the last closed flux surface (LCFS) defined by 5 uncooled graphite limiters in December 2015. At the end of the 10 weeks long experimental campaign (OP1.1) more than 20 independent diagnostic systems were in operation, allowing detailed studies of many interesting plasma phenomena. For example, fast neutral gas manometers supported by video cameras (including one fast-frame camera with frame rates of tens of kHz) as well as visible cameras with different interference filters, with field of views covering all ten half-modules of the stellarator, discovered a MARFE-like radiation zone on the inboard side of machine module 4. This structure is presumably triggered by an inadvertent plasma-wall interaction in module 4 resulting in a high impurity influx that terminates some discharges by radiation cooling. The main plasma parameters achieved in OP1.1 exceeded predicted values in discharges of a length reaching 6 s. Although OP1.1 is characterized by short pulses, many of the diagnostics are already designed for quasi-steady state operation of 30 min discharges heated at 10 MW of ECRH. An overview of diagnostic performance for OP1.1 is given, including some highlights from the physics campaigns.

The stellarator fusion experimental device Wendelstein 7-X (W7-X) is presently under assembly at the Greifswald branch of the Max-Planck-Institut fur Plasmaphysik, Germany. The superconducting magnet system consists of 50 nonplanar coils, 20 planar coils, a superconducting bus system, and 14 current leads. It is organized in seven electrical circuits with ten coils, the bus system for the interconnection, and two current leads each. The magnet system is cooled by supercritical helium. It is enclosed in a cryostat with an outer diameter of 16 m formed by the plasma vessel and the outer vessel. There are five different types of nonplanar coils having dimensions of 3.5 × 2.5 × 1.5 meters in maximum and a weight of about 5.5 tons. The two different types of the planar coils are nearly circular coils with a diameter of up to 4.5 m and a weight of about 3 tons. The superconducting bus bar system connects the coils to each other and provides the connection to the current leads inside the cryostat. All types of coils and the bus system use the W7-X superconductor, a forced flow cable-in-conduit superconductor with 243 copper stabilized NbTi strands with an outer aluminum-alloy jacket. The current leads provide the transfer of the electrical current from the room temperature bus bar system outside the cryostat to the superconducting parts inside the cryostat. Their special feature is the upside-down orientation with the cold end at the top.

The upper port of the ITER vacuum vessel contains the core charge exchange spectroscopy (cCXRS) port plug diagnostic system. Electromagnetic (EM) loads caused by plasma transients are critical for the port plug. An ANSYS 3D EM model of the ITER magnet system, addressing main port plug features (conceptual design 2009), was developed by Forschungszentrum Julich (FZJ). EM forces on main plug components have been calculated and main integral mechanical moments are represented. Two external contracts on independent global EM models were launched and are being successfully benchmarked against the FZJ model. In support of the cCXRS design its components are structurally analyzed under EM loading. Four different approaches to calculate EM loads for detailed structural models have been used and are described. One of them, called an “express” analysis, allows a prompt choice between different design solutions. The “air” elements, which should fill gaps between assembled conducting parts in the electromagnetic FE models, are neglected in this approach. This may save up to 90% of time for building and calculating complex EM models. Applicability and use of this method for small critical plug components like a fast shutter are demonstrated in this paper.

The high heat flux ion beam test facility MARION (Material Research ION Beam test Stand) has been used for ion source conditioning and material testing within the IEK-4 for several years. Mainly small samples have been tested in the past by perpendicular irradiation. In these experiments only 25% of the available power density of MARION (>80 MW/m 2 ) was used so far. The activation of these resources for material testing was realized in the first test phase of the new tungsten divertor element for the JET tokamak. The enhancement of MARION by a new protection scraper is shown. During the last experiments the need of a more flexible handling of the sample itself and of the facility became clear. Based on those experiences an upgrade concept was developed to prepare MARION for future demands allowing the handling of much bigger prototypes. Additionally main emphasis is given to the access to improved (surface) diagnostics for all orientations of the target.

The first mirror (M1) as an important part of the CXRS diagnostic system is responsible for acquisition and transportation of the optical signal of required quality from the ITER plasma to detectors. At the same time, the M1 is the most vulnerable component of this system working in drastic and severe conditions caused by its location in the direct view of the plasma. To minimize impurities deposition and guarantee the required lifetime for M1 its temperature shall be maintained within the operational range of 350 ± 50 °C. The computation of M1 of the reference CXRS design demonstrates that, firstly, helium is the best candidate among the analyzed gases (He, N2, Ar) to be used as a coolant to maintain the M1 temperature in the required operational range of 350 ± 50 °C giving the highest heat transfer coefficient and the lowest hydraulic losses at moderate flow rates (vgas = 40–60 m/s) and, secondly, optimization of the cooling channels configuration together with an increase in the gas velocity and/or with heat transfer intensifiers makes it possible to provide the required M1 temperature conditions.

New optical configurations for the ITER core CXRS system offer the possibility of longer ducts between the first mirror and the plasma. This has led to a renewed optimization of the availability. using a simple model of the degradation of the first mirror that starts with the conditions of (a) the required measurement performance and (b) the geometry of the port plug. It is found that for a fully passive system the design should strive for the longest duct length possible. Given known data, this will result in a diagnostic lifetime still substantially shorter than ITER lifetime. When an option of cleaning the first mirror is introduced (assuming this is a feasible option) the optimum is less straightforward, because the lifetime of the second mirror then also becomes important. The optimum then depends on the ratio between the cleaning interval and the ITER lifetime. Options are presented for various sets of assumptions. Finally practical limitations of supporting subsystems (cleaning system, shutter, calibration system) may influence the final design. Examples of such limitations with their impact are presented. (C) 2011 Published by Elsevier B.V.

In the frame of the ITER-like Wall (ILW) for the JET tokamak, a divertor row made of bulk tungsten material has been developed for the position where the outer strike point is located in most of the foreseen plasma configurations. In the absence of active cooling, this represents a formidable challenge when one considers the temperature reached by tungsten (TW,surf > 2000 °C) and the vertical gradient ∂T/∂z = 5 × 104 K/m. As the development is drawing to an end and most components are in production, actual 1:1 prototypes are exposed to an ion beam with a power density around 7 MW/m2 on the plasma-facing surface. Advantage is taken of the flexibility of the Marion facility to bombard the tungsten stack under shallow angles of incidence (∼6°) with a powerful beam of ions and neutrals (>70 MW/m2 on axis). The shallow angles are important, with respect to the toroidal wetted surface, for properly simulating the expected performance under actual tokamak conditions. The M arion tests have been used to validate for a few typical cases the thermal calculations that were steadily developed along with the tungsten tile and, at the same time, to gather information on the actual temperatures of individual components. The latter is an important factor to a finer estimation of the power handling capabilities.

Transient electromagnetic (EM) analysis is presented focusing on main components of the ITER core charge exchange recombination spectroscopy (cCXRS) port plug developed by Forschungszentrum Julich (FZJ), ITER-NL and UKAEA(CCFE) in 2009. The cCXRS primary function is to transfer the light in the visible part of spectra emitted by interaction of the plasma ions with a diagnostic neutral beam. The TYPHOON software package has been used for the EM analysis. The code is dedicated for simulation of transient electromagnetic processes using a shell approach in the integral–differential formulation to represent conducting structures with a set of multi-connected shells arbitrary located in a space. The advantage of the shell approach is a higher flexibility in modelling detailed structures as compared with widely used 3D models. On the other hand, the shell approach requires ultimate care in modelling relatively thick structures. These issues are discussed in the paper. Two vertical displacement events (VDE) which seem to result in the largest EM loads on the main cCXRS components have been agreed with FZJ and simulated. Transient electromagnetic processes caused by different sources have been considered separately, and then superimposed to obtain the total solution. Three types of transient processes for each type of VDE have been analyzed: (1) due to variations of a toroidal plasma current, shape and position and due to variations of poloidal field coils (PFC) and central solenoid (CS) currents, (2) due to variations of the Halo current and (3) due to variations of a toroidal magnetic flux of plasma. The analysis covers two options for electrical contact between the main shell (MS) of the port plug and the blanket shield module (BSM). The results are supposed to be used for benchmarking with independent 3D EM models developed for the upper port plug.

Controlling the heat and particle fluxes in the plasma edge and on the plasma facing components is important for the safe and effective operation of every magnetically confined fusion device. This was attempted on Wendelstein 7-X in the first operational campaign, with the modification of the magnetic configuration by use of the trim coils and tuning the field coil currents, commonly named iota scan. Ideally, the heat loads on the five limiters are equal. However, they differ between each limiter and are non-uniform, due to the (relatively small) error fields caused by the misalignment of components. It is therefore necessary to study the influence of the configuration changes on the transport of heat and particles in the plasma edge caused by the application of error fields and the change of the magnetic configuration. In this paper the up-stream measurements conducted with the combined probe are compared to the downstream measurements with the DIAS infrared camera on the limiter.

Diagnostic plug for the ITER core charge exchange recombination spectroscopy (core CXRS) is located in the upper port 3. It transfers the light emitted by interaction of plasma ions with the diagnostic neutral beam (DNB). Conceptual design study of the core CXRS port plug has indicated several challenging technical problems: (1) likely too short lifetime of the first mirror, (2) quite contradictory requirements to the first mirror holder, (3) harsh environmental conditions for the "shutter", that is a movable element protecting the first mirror, (4) a task to combine a sufficient structural integrity and nuclear shielding capability of the plug with a wide enough optical path, (5) excessive electromagnetic loads caused by the halo current and applied at the plug as a whole. This paper describes possible design solutions for the listed technical problems.

The simulation of the thermal response of the JET bulk tungsten divertor modules to real operational conditions demonstrates that the analyzed design provides more favourable conditions than before (≤600 °C) for the temperature-vulnerable materials of the stack attachment (esp. Inconel spring discs), thereby increasing the attachment lifetime. The computation shows that the JET divertor unit can withstand up to five consecutive cycles, 10 s long, of an evenly distributed heat loading (averaged heat flux of ∼6.5 MW/m2) with half-an-hour intervals before the springs reach their uppermost temperature limit. However, if we take into account: (i) the real heat load distribution over the upper lamellae surface (shadowing effect), (ii) an initial temperature of the structure of 200C, it may be that the temperature approaches 600C after a single full-energy pulse.

Forschungszentrum Julich has taken over the design, manufacturing and assembly of the superconducting busbar system for the stellarator Wendelstein 7-X. This includes the busbars itself, the support structure consisting of supports and clamps, and the joints for electrical and hydraulic connection of the busbars and coil terminals. Apart from providing the required electrical connection scheme, the busbar system has to be designed for relevant electrical and mechanical loads. Numerous interfaces and geometric boundary conditions define the confined space to accommodate the busbars and their support elements. This article describes how the individual challenges to engineering have been met in the course of the project. This includes design concepts and the method for iterative design of supports with respect to the individual load distribution caused by the supports itself.

Within the ITER-like wall project at Joint European Torus (JET), the original plasma facing tiles in the divertor region made of carbon fibre composite (CFC) will be replaced by tungsten coated CFC and bulk tungsten tiles. The main constraint of the bulk tungsten concept is to accommodate the power and energy handling requirements, the electromagnetic (EM) forces and the mechanical requirements of the existing remote handling system. Through a number of intermediate design options the “lamella” option has been developed, consisting of plasma facing tiles, an inconel wedge holding the tiles and an inconel interface plate attaching the wedge to the JET CFC base plate. In order to minimize eddy currents the wedge must be equipped with slits and the W-lamellae are isolated from each other. Defined electrical contacts from the lamellae via the wedge to the base plate define the path of the halo currents. The pairs of lamellae are isolated from each other by insulating spacers and tie rods keep the stack of tungsten lamellae and ceramic-coated spacers together. Eight tungsten lamella stacks are attached to the wedge via eight “rails.” This paper reports on calculation of the electromagnetic (EM) loads in the block components and of the stress–strain state of the block, subjected to the worst combination of EM loads, by 3D finite element (FE) electromagnetic and stress analyses. As a result of these studies the level of initial pre-tension of the joint elements has been checked against possible detachment. The bending of the load bearing components has proved to be moderately low compared with the structural material allowable limits.

The ITER-like Wall project (ILW) for JET aims at providing the plasma chamber of the tokamak with an environment of mixed materials which will be relevant for the actual first wall construction on ITER. Tungsten plays a key role in the divertor cladding. For the central tile, also called LB-SRP for “load-bearing septum replacement plate”, bulk tungsten is envisaged in order to cope with the high heat loads expected (up to 10 MW/m 2 for 10 s). The outer strike-point in the divertor will be positioned on this tile for the most relevant configurations. Forschungszentrum Julich (FZJ) has developed a conceptual design based on an assembly of tungsten blades or lamellae. An appropriate interface with the base carrier of JET, on which modules of two tiles are positioned and fixed by remote handling procedures, is a substantial part of the integral design. Important issues are the electromagnetic forces and expected temperature distributions. Material choices combine tungsten, TZM™, Inconel ® and ceramic parts. The completed design has been finalised in a proposal to the ILW project, with utmost ITER-relevance.

Wendelstein 7-X (W7-X) is a large optimized stellarator (B=2.5T, V=30m3) aiming at demonstrating the reactor relevance of the optimized stellarators. In 2015 W7-X will begin its first operation phase (OP1.1) with five inertially cooled inboard limiters made of graphite. Assuming the heat loads can be spread out evenly between the limiters, 1 second discharges at 2 MW of ECRH heating power could be run in OP1.1. The expected plasma parameters will be sufficient to demonstrate the readiness of the installed diagnostics and even to run a first physics program. The diagnostics available for this first operation phase, including some special limiter diagnostics, and their capabilities are being presented. A shorter version of this contribution is due to be published in PoS at: 1st EPS conference on Plasma Diagnostics

The CXRS (Charge-eXchange Recombination Spectroscopy) diagnostic for the core plasma of ITER will be designed to provide observation of the dedicated diagnostic beam (DNB) over a wide radial range, roughly from a normalised radius r/a = 0.7 to close to the plasma axis. The collected light will be transported through the Upper Port Plug #3 (UPP3) to a bundle of fibres and ultimately to a set of remote spectrometers. The design is particularly challenging in view of the ITER environment of particle, heat and neutron fluxes, temperature cycles, electromagnetic loads, vibrations, expected material degradation and fatigue, constraints against tritium penetration, integration in the plug and limited opportunities for maintenance. Moreover, a high performance (étendue × transmission, dynamic range) is expected for the port plug system since the beam attenuation is large and the background light omnipresent, especially in terms of bremsstrahlung, line radiation and reflections. The present contribution will give an overview of the current status and activities which deal with the core CXRS system, summarising the investigations which have taken place before entering the actual development and design phase.