Ten cryo-vacuum pumps (CVPs) were installed in the subdivertor region of each island divertor in the stellarator Wendelstein 7-X (W7-X) and operated for the first time during the recently completed plasma campaign OP2.1. A pumping speed of 70 +/- 1 m(3)/s was measured during dedicated tests with known hydrogen gas injection. Based on a conductance model, the estimated pumping speed ranges from 86-93 m(3)/s for different sticking coefficients between 0.6 and 0.8. After completion of the initial tests the CVPs were operated successfully throughout the campaign, with regeneration performed once a week. Neutral gas pressures in the subdivertor in the range of 10(-4) mbar are well within the molecular flow regime and limit the particle exhaust capabilities of the CVPs. Simulations of the neutral gas pressure in the three-dimensional complex geometry of the subdivertor were performed using the DIVGAS code based on the direct simulation Monte Carlo method and a model implemented in the steady-state thermal package in ANSYS, which are in agreement with the measured values during plasma operation.

The stellarator Wendelstein 7-X (W7-X) will start operation with an actively water cooled divertor made of target elements armored with CFC (Carbon Fiber Reinforced Carbon) NB31 tiles in 2022. The next step (> 2030) is the installation of a metallic water cooled divertor. Research activities have been launched supported by EUROfusion to develop the next generation of metallic target element. The purpose of the conceptual design is to prepare the first prototyping phase. Stationary loading and water cooling conditions are: 10 MW/m2, Tin=30 degrees C, Pstatic=1 MPa, Vaxial=9m/s. Similar to the current divertor, the heat sink is made of CuCrZr . Two kinds of armor materials are considered: pure tungsten and W3.5Ni1.5Cu heavy alloy. One of the main constraints is to keep a weight similar to the CFC design to limit the divertor module weight for assembly. The first analyzed model is a straightforward adaptation of the simplest geometry of a CFC element. Thermal calculations show that the maximal temperatures remain within acceptable limits and 3.5 mm armor layer should be considered as the upper thickness limit. By reducing the distance between coolant and loaded surface the need for swirl flow should be assessed in more details. In addition to modelling, first trials for the industrial manufacturing of the CuCrZr heat sink, armor, bonding procedure have been launched.

Wendelstein 7-X, the world’s largest superconducting stellarator in Greifswald (Germany), started plasma experiments with a water-cooled plasma-facing wall in 2022, allowing for long pulse operation. In parallel, a project was launched in 2021 to develop a W based divertor, replacing the current CFC divertor, to demonstrate plasma performance of a stellarator with a reactor relevant plasma facing materials with low tritium retention. The project consists of two tasks: Based on experience from the previous experimental campaigns and improved physics modelling, the geometry of the plasma-facing surface of the divertor and baffles is optimized to prevent overloads and to improve exhaust. In parallel, the manufacturing technology for a W based target module is qualified.This paper gives a status update of project. It focusses on the conceptual design of a W based target module, the manufacturing technology and its qualification, which is conducted in the framework of the EUROfusion funded WPDIV program. A flat tile design in which a target module is made of a single target element is pursued. The technology must allow for moderate curvatures of the plasma-facing surface to follow the magnetic field lines. The target element is designed for steady state heat loads of 10 MW/m2 (as for the CFC divertor). Target modules of a similar size and weight as for the CFC divertor are assumed (approx. < 0.25 m2 and < 60 kg) using the existing water cooling infrastructure providing 5 l/s and roughly maximum 15 bar pressure drop per module.The main technology under qualification is based on a CuCrZr heat sink made either by additive manufacturing using laser powder bed fusion (LPBF) or by uniaxial diffusion welding of pre-machined forged CuCrZr plates. After heat treatment, the plasma-facing side of the heat sink is covered by W or if feasible by the more ductile WNiFe, preferably by coating or alternatively by hot isostatic pressing W based tiles with a soft OFE-Cu interlayer. Last step is a final machining of the plasma-exposed surface and the interfaces to the water supply lines and supports to correct manufacturing deformations.

For the upcoming long pulse operation phase 2 "OP2" of Wendelstein 7-X (W7-X), new water cooled non planar stainless steel panels have been manufactured to protect the wall of the plasma vessel behind the divertor pumping gap. Such a panel is made of a machined ground plate with channels which are covered by likewise machined sheets. The latter are electron-beam welded to the ground plates. There are 60 panels of 7 different types which are designed to remove a stationary heat flux of up to 100 kW/m(2). The specified water cooling conditions are: 2.5 MPa inlet pressure, 30 degrees C inlet temperature, and 0.27 l/s flow rate. A panel prototype has been manufactured to validate the design and manufacturing process. In order to verify the thermo-hydraulic calculations performed with ANSYS (R) CFX, a thermal loading test on the prototype was carried out in the SIR HEX test facility at KIT. The loaded surface of the prototype (about 0.11 m(2)) was black-finished, and thermocouples were installed at different positions. The prototype was placed in the vacuum tank and heated by an array of six infrared lamps (400 V, max. 16 kW per lamp). The deposited load of up to about 100 kW/m(2) was measured by calorimetry. The paper describes the loading experiment of the panel, and its results. A good agreement was found between calculations and test results, and thus the thermo-hydraulic panel design was well validated.

We present recent highlights from the most recent operation phases of Wendelstein 7-X, the most advanced stellarator in the world. Stable detachment with good particle exhaust, low impurity content, and energy confinement times exceeding 100 ms, have been maintained for tens of seconds. Pellet fueling allows for plasma phases with reduced ion-temperature-gradient turbulence, and during such phases, the overall confinement is so good (energy confinement times often exceeding 200 ms) that the attained density and temperature profiles would not have been possible in less optimized devices, since they would have had neoclassical transport losses exceeding the heating applied in W7-X. This provides proof that the reduction of neoclassical transport through magnetic field optimization is successful. W7-X plasmas generally show good impurity screening and high plasma purity, but there is evidence of longer impurity confinement times during turbulence-suppressed phases.

The installation of an actively water cooled divertor in the stellarator Wendelstein 7-X (W-7X) is mandatory to achieve stationary power and particle exhaust for pulse lengths up to 30 min. The highly loaded divertor area is made of 100 target modules distributed in ten divertor units. The target modules have mechanical support frames with attachment systems to the plasma vessel, and manifolds to distribute water equally between the target elements. A target element is designed to remove a stationary heat flux up to 10 MW/m2 and is made of a CuCrZr copper alloy heat sink armored with CFC NB31 tiles. The manufacturing process, assembly and quality assessment of the last 70 target modules has been successfully completed in the Integrated Technical Centre of IPPGarching. Some parts such as the target elements and manifolds were delivered by industry. The quality was assessed as follows: visual inspections, measurement of the 3-D CFC surface, dynamic pressure tests, He leak testing under pressure at different temperature (20 ?C, 160 ?C) in vacuum oven, high heat flux testing. The production of the water cooled divertor is now completed, and the mounting operation of the target modules in the plasma vessel of W7-X has started.

The net erosion and deposition pattern of carbon from the Test Divertor Unit (TDU) of the stellarator W7-X was determined. Special target elements with marker layers consisting of about 300 nm molybdenum and 5–10 μm carbon on top were used during the operational phase OP 1.2a. The thicknesses of the marker layers were determined by elastic backscattering spectrometry (EBS) using 2.5 MeV protons before and after plasma exposure and laser-induced breakdown spectroscopy on selected target elements after exposure. Scanning electron microscopy was used for investigating the surface morphology before and after exposure. Massive erosion of up to 20 μm carbon was observed at the strike line, in total 48 ± 14 g carbon were eroded from the 10 TDUs. The erosion was laterally non-uniform on the micro-scale. Strongly eroded surfaces were considerably smoother as compared to the original material. Only very little deposition of carbon is observed on the TDU: this means that the TDU is a large net erosion source.

The highly loaded surface of the actively water-cooled divertor of Wendelstein 7-X (W7-X) is made of 100 individual target modules. In each target module, a set of target elements is water-cooled in parallel and fed by manifolds. A target element is made of a CuCr1Zr copper alloy heat sink, armored with CFC NB31 tiles. He leak testing under pressure in a vacuum oven was performed for each individual target element and module. One series target element of type 5S did not pass this test after high heat flux testing. The leakage was the result of the combination of a porosity concentration due to the overlap of CFC-CuCr1Zr welds in the centre part along the element and the reduced CuCr1Zr thickness at this location due to machined cut-out in the cooling channel. The repair process, which was successfully applied to all 307 target elements of type 5S, was sealing the slit between two CFC tiles by electron beam welding. Two target modules of type 7H did not pass the integral leak test. Leaks were localized by sniffing the weld seam between the target element connectors and manifold pipes. The one leak of the first module was successfully repaired by hand-welding. The positions of the two leaks in the other module did not allow such a repair process. Drilling apertures through the neighbouring manifolds allowed a direct access to the leaking seams from inside. The openings allowed a stable and precise installation of an inside orbital welding anode. The repaired target module successfully passed the He leak test in oven after HHF testing.

The cryo-vacuum pump (CVP) system, consisting of 10 units distributed symmetrically inside the Wendelstein 7-X plasma vessel, will be installed together with the 10 units of the actively cooled high heat flux divertor. One pump each is located below the corresponding divertor, and positioned as close as possible to the flux line strike points in order to allow efficient control of plasma density, and for screening impurities. Each CVP is divided into two parts, interconnected by a transfer line, to ensure access for divertor diagnostic integration. The CVP panels are operated with supercritical helium at 3.3-3.8 K to pump discharge gases such as H-2 and D-2. They are protected against thermal radiation by black-oxide finished stainless steel chevrons cooled by liquid nitrogen at 77 K. In front of this LN2-shield is a water-baffle which protects the CVP against plasma and ECRH stray radiation. It consists of copper chevrons with zero overlap, mounted on a water-cooled steel pipe. These chevrons are coated with an Al2O3-TiO2 layer. In addition, an uncooled copper shield covers the gap between water-baffle and LN2 chevrons in order to prevent stray radiation to take this path to the He-cooled panel. The cryogenic fluids are supplied via a dedicated port plug-in which is thermally insulated by a LN2-cooled cryo-shield and superinsulation. All 10 CVPs are already manufactured and successfully leak tested under hot and cold conditions in the workshops of IPP Garching. This paper presents the design and the manufacturing technology of the CVPs and the adjacent periphery. Results of the quality assessment such as integral He leak testing at 160 degrees C and cryogenic temperatures (77 K) are also discussed.

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 quasi-symmetric fivefold modular Wendelstein 7-X (W7-X) stellarator consists of three groups of coil systems, i.e. superconducting magnet, trim coil and control coil systems. The control coil system contains ten identical 3D shaped control coils (CC) situated behind the baffle plates of corresponding divertor units, and is designated to rectify the error field and to sweep hot spots on the divertor target plates. The CC is wound from copper conductor with a square cross section of 16 mm x 16 mm and a water cooling hollow of empty set 8 mm. The control coil system was installed in W7-X in 2015, and the integral commissioning has been done in parallel with the completion of W7-X. During the operation phase (OP 1.2a) with limited plasma heating energy, a leakage in one of the CC cooling water plug-in was found and dictates a detailed transient thermal analysis of CC to determine the allowable operation time without cooling water flow. The paper presents the transient thermal analysis and is followed by a detailed finite element mechanical analysis with the consideration of temperature gradient loads, dead weight and electromagnetic forces. Moreover, the transient thermal and mechanical performance of actively cooled CC to be intensively operated during steady state operation phase (OP 2) are also analyzed and evaluated with the same FE model.

Heat load calculations have indicated the possible overloading of the ends of the water-cooled divertor facing the pumping gap beyond their technological limit. The intention of the scraper is the interception of some of the plasma fluxes both upstream and downstream before they reach the divertor surface. The scraper is divided into six modules of four plasma facing components (PFCs); each module has four PFCs hydraulically connected in series by two water boxes (inlet and outlet). A full-scale prototype of one module has been manufactured. Development activities have been carried out to connect the water boxes to the cooling pipes of the PFCs by tungsten inert gas internal orbital welding. This prototype was successfully tested in the GLADIS facility with 17 MW/m(2) for 500 cycles. The results of these activities have confirmed the possible technological basis for a fabrication of the water-cooled scraper.

Ten identical cryopumps (CVPs) are to be installed in corresponding divertor volumes of Wendelstein 7-X (W7-X) stellarator before commencing the steady-state phase of operation (OP 2). Each CVP is typically made of two units connected by a transfer line and is fed through a dedicated vacuum vessel feed through. The units consist of water baffle, liquid nitrogen (LN2) baffle, helium panel, and LN2-cooled housing. All components are expected to be well cooled with the available cooling capacity during long-pulse plasma operation in order to maintain the helium panel at about 4 K and hence ensure the desired absorption rate. The LN2-cooled housing has to minimize both the effect of electron-cyclotron resonance heating (ECRH) and the thermal radiation from backside of in-vessel components on the 80-K CVP shield elements. Moreover, the ECRH is unevenly distributed in W7-X, which requires analyses of several cases and sophisticated cooling scheme as described. This paper presents the thermal behavior of CVP components and is followed by the discussion of several important issues for assessment. In addition, eddy current and electromagnetic (EM) forces on CVP copper components are analyzed for the events of fast discharge (FD) of main superconducting coils, plasma current decay, alternating current in control coils, and FD of trim coils. Moreover, sharing of eddy currents between plasma vessel shell and attached CVP is estimated. Structural analysis taking temperature gradients and EM forces into account indicates that the mechanical performance is acceptable.

The cryo-vacuum pump (CVP) system controls the plasma density by condensing undesirable gases together with a set of turbo molecular pumps. One CVP will be installed under each of the ten units of the actively cooled divertor in Wendelstein7-X for the long-pulse operation up to 30-min duration. Each CVP is operated with supercritical Helium (ScHe) at 4 K and liquid nitrogen (LN2) at 80 K fed by a plug-in, which is installed inside a W7-X port of the plasma vessel (PV). The plug-in made of stainless steel provides for the vacuum boundary between the plasma chamber and the torus hall atmosphere. Four transfer lines (12 x 1 mm) are positioned inside the plug-in: two for the inlet/outlet of ScHe and two for the inlet/outlet of LN2, respectively. The design has to guarantee the feeding at the specified temperature of ScHe and LN2 while minimizing thermal losses and thermal interactions between pipes. Inside the plug-in the vacuum level is 10(-3) Pa at room temperature (RT) and 10(-5) Pa during operation. The pipes of the ScHe are shielded with a multilayer superinsulation. The cryogenic feed lines are protected with a cryo shield against thermal loads in the port as well as in the PV. The thermal analysis confirmed the efficiency of the thermal shielding design, which keeps temperatures of ScHe and LN2 pipes within operation limits. During baking, the PV is heated to 150 degrees C while the cryostat remains close to RT. The port and transfer lines are equipped with bellows to accommodate this situation. But the plug-in is rigid and the resulting bend induces mechanical loads on the flanges between port and plug-in. The mechanical analysis confirmed that the selected bolts and pretension conditions guarantee the tightness of the flange during baking.

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.

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/m(2). 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.

After completing the main construction phase of Wendelstein 7-X (W7-X) and successfully commissioning the device, first plasma operation started at the end of 2015. Integral commissioning of plasma start-up and operation using electron cyclotron resonance heating (ECRH) and an extensive set of plasma diagnostics have been completed, allowing initial physics studies during the first operational campaign. Both in helium and hydrogen, plasma breakdown was easily achieved. Gaining experience with plasma vessel conditioning, discharge lengths could be extended gradually. Eventually, discharges lasted up to 6 s, reaching an injected energy of 4 MJ, which is twice the limit originally agreed for the limiter configuration employed during the first operational campaign. At power levels of 4 MW central electron densities reached 3 x 10(19) m(-3), central electron temperatures reached values of 7 keV and ion temperatures reached just above 2 keV. Important physics studies during this first operational phase include a first assessment of power balance and energy confinement, ECRH power deposition experiments, 2nd harmonic O-mode ECRH using multi-pass absorption, and current drive experiments using electron cyclotron current drive. As in many plasma discharges the electron temperature exceeds the ion temperature significantly, these plasmas are governed by core electron root confinement showing a strong positive electric field in the plasma centre.

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.