The goal of the European plasma Volumetric Neutron Source (VNS), which is a 14 MeV n-source, is to test and validate technological solutions of breeding blankets in an environment representative of a future fusion power plant such as DEMO. One of the assessed magnetic configurations is a tokamak device with a radius of ti 2.5 m that produces a D-T fusion power of ti 30 MW. The architecture of VNS integrates a single null divertor located at the bottom of the vacuum vessel. The ITER-like divertor is designed to be actively water-cooled for stationary operation and is made up with 36 cassette modules. Each module has a cassette body with a set of plasma facing components (PFCs) with tungsten as plasma facing material. The PFC arrangement is a dome positioned in the central part with reflector plates on both sides located between inboard and outboard target striking surfaces. The selected PFC technological solutions are: the plasma facing units (PFUs) of the targets are armored with tungsten monoblocs bonded onto a cooling pipe made of CuCrZr with an inserted twisted tape, the PFUs of the dome are made up with tungsten flat tiles bonded onto a CuCrZr hypervapotron cooling structure. This paper presents the status of the development of the divertor conceptual design. Preliminary analyses confirmed that the design is compatible with plasma scenarios foreseen for VNS operation.

The qualification of in-vessel components for a fusion power plant requires a test environment with a high flux of 14 MeV neutrons over a sufficiently large surface and volume. Performance testing and qualification of the complex design and technologies of fusion nuclear components is needed, in particular that of the tritium breeding blanket (BB). Testing in relevant conditions over a relevant time will also allow gaining the necessary confidence regarding the build-up and control of tritium inventories inside the BB, which will be an important radioactive source. An option of such a volumetric neutron source (VNS) is a beam-driven tokamak. A feasibility study of the main machine components and associated plant systems is described in this article. The machine has a major radius of 2.53 m, a single-null divertor configuration, and four tangential 120 keV beamlines that generate a fusion power of approximately 30 MW and provide current drive for a steady-state plasma scenario. The plasma is small with a minor radius of a = 0.55 m to maximize the neutron wall load, up to 0.5 MW/m(2), similar to what is targeted in ITER. Approximately 25 m(2) are available for blanket testing including 4 port plugs, which offer flexibility regarding the test module operating conditions and the implementation of instrumentation. Given the small plasma, much of the tokamak's volume is made up by the neutron shielding structures that are similarly sized as in ITER. To reduce the construction risk, ITER-like concepts were adopted for many components. In some cases, however, lessons learned from ITER led to the development of customized or innovative concepts. Due to the modest fusion power the plasma will burn <1 kg of tritium per year, which can be provided from external sources.

In the context of EUROfusion activities for the development of the DEMO reactor project, the divertor design is a major challenge. It must sustain very high heat, ion particle and neutron fluxes allowing, at the same time, the shielding of the vacuum vessel and the vacuum pumping for reducing the plasma pollution. The conceptual divertor design is based on the use of EUROFER97 for the divertor cassette body, while tungsten monoblocks bonded to CuCrZr pipes are used for plasma-facing targets. EUROFER97 was selected considering its reduced long-term activation and superior creep and swelling resistance under neutron irradiation. However, depending on the operating temperature under neutron irradiation, a pronounced shift of the Ductile to Brittle Transition Temperature (DBTT) is expected. At the same time, for the plasma-facing targets, the coolant temperature has to be identified such to allow sufficient heat removal capacity at the strike point. This study explores alternative cooling conditions for the divertor system that are able to ensure the fulfillment of functional and system requirements and to allow for divertor cassette body re-use during plant lifetime. The main aim is to identify the best water cooling thermal-hydraulic conditions avoiding material embrittlement (for EUROFER 97) and softening/hardening (for copper alloy pipes). At the same time, the goal is to reduce the inventories (enthalpy of the cooling circuit) and the radwaste at the end of divertor lifetime.

After the successful short pulse operation phase 1 (OP1) of the stellarator Wendelstein 7-X with maximal plasma energies of 200 MJ, the upcoming long pulse OP2 aims at stepwise higher energies up to 18 GJ. A series of the stainless steel wall protection panels is positioned behind the divertor pumping gaps for average stationary heat loads of 100 kW/m². These panels were produced by electron beam welding of stainless steel parts to build the housing of the cooling channels. This technology was more demanding than foreseen due the complicated 3D weld seams required by the shaping of the panels. An innovative technology which is more adapted to the panel geometry is additive manufacturing. It intends to bring significant advantages by printing the housing as a single piece without welds.The paper introduces the thermal-hydraulic and structural analyses of an additive manufactured panel mock-up performed with ANSYS V2022R2. The purpose of the calculation is to compare the thermal, hydraulic and structural performance of the improved design based on additive manufacturing and the present one. The welding process dictated the design of the cooling channel that induced local recirculation areas, additional local pressure drops, non well cooled areas for weld seams integration. In addition the weld strength assessment required special codes and methods, which cost more time and efforts. Without these constraints, additive manufacturing offers the opportunity of more freedom for the design of the cooling channel to reach better heat transfer performance of the panel.

A challenging aspect in view of the realization of a future magnetic confinement fusion reactor is the design and manufacture of highly loaded divertor target plasma-facing components (PFCs) which have to sustain intense particle, heat and neutron fluxes. In this context, tungsten-copper (W-Cu) composites are currently being investigated as potentially advanced heat sink materials for PFCs. The development and manufacture of W-Cu composite pipes, which are bonded to tungsten monoblocks, poses new challenges in terms of manufacturing and accuracy.A new approach is proposed to produce the W-Cu composite pipe. The proposed alternative is to directly infiltrate tungsten fibers with galvanic copper instead of by copper infiltration in an furnace. The paper presents the status of the development to better assess the impact of various parameters affecting the galvanic process. Different braid thicknesses (3 and 6 layers) and their orientation (70° and 80°) have been tested. The effects of the electrolytic bath circulation, the variation in the braid material (porosities), and the electrolytic exchange by using pulsating direct current on the diffusion in the braid have been investigated. The influence of the passivation mode has also been carefully analyzed because in the case of a strong passivation copper only grows too quickly outwards from the core and not onto the tungsten braid.

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.

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.

Electron cyclotron resonance heating (ECRH) is a powerful and flexible plasma heating technique that serves as the main heater at Wendelstein 7-X (W7-X) and will be used at ITER for start-up, heating, current drive and mitigation of plasma instabilities. In the case of poor or degraded microwave absorption, which is expected in the O2-mode heating scenario, a significant part of the beam directly hits the wall, leading to local overheating and potential damage. The ECRH shine-through power is mostly reflected onto the targets; only a small fraction is really absorbed through ohmic losses (typically 3% for graphite at 140 GHz). The ohmic losses do not only depend on the material properties and the frequency, but also on the polarization of the wave and the angle of incidence. This paper presents a thermographic analysis of ECRH experiments at W7-X, including heat load and temperature simulations of the first wall that include ECRH shine through. Two O-mode ECRH experiments with both a high temperature rise of the first wall and different angles of beam incidence on the wall's surface are depicted. One experiment has 775 kW of power modulation (5 Hz) with mixed polarization (45% O-mode, 55% X-mode) and an EC beam angle almost normal to the first wall. The second has 550 kW of steady EC power with O-mode polarization, a shallow beam angle and increased power absorption by the material. It is shown that infrared thermography is a useful tool for measuring shine-through power and protecting wall components.

The study focuses on the CFD analysis of a possible prototype of water-cooled divertor element for the experimental stellarator-type fusion machine Wendelstein W7-X. The coupled solid-fluid analysis was performed on the selected type of divertor target element. The singlefluid modelling approach was used to explore the effect of the cooling water on the local wall temperature distribution in the cooling channel. Comparison of the wall temperature profile with the local coolant saturation temperature at critical location is presented to detect local superheating regions that may lead to the evolution of boiling flow. The effect of two different heat flux profiles, material of the tiles (tungsten or tungsten heavy alloy) and the effect of the twisted tape was investigated. Maximum temperatures of target materials were compared. The results also included analysis of cooling channel parameters (e.g. superheating, water temperature distribution, pressure drop) and temperature distribution in the cooling channel wall. It is interesting to note that the maximum velocity region is not necessarily located just below the heated wall, but rather near the bottom of the pipe in the colder flow region. The single-fluid results show the necessity to perform realistic simulations in the two-phase flow regime.

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.

After the successful short pulse operation phase 1 (OP1) of the stellarator Wendelstein 7-X with maximal plasma energies of 200 MJ, the upcoming long pulse OP2 aims at stepwise higher energies up to 18 GJ. With the knowledge of partly unexpected experimental heat load distributions on plasma facing components (PFC), their allowable loads were re-evaluated which consequently lead to some adaptation work. The divertor target modules TM5 h and TM6 h were loaded more than expected during OP1. This triggered a detailed transient and stationary analysis which revealed a high sensitivity of the thermal response to the heating power distribution. At the baffles, cracks were detected just before assembly for OP1 at the braze joints between CuCrZr heat sinks which carry the graphite tiles, and the stainless steel (SS) cooling pipes. Due to large temperature differences between parts of the pipes and support structure the braze becomes heavily stressed during the high thermal loads in OP2. Multiple optimization analyses led to the decision to release thermomechanical stresses by loosening the connections between the heat sinks and steel structures of the modules. New stainless steel wall protection panels to be positioned behind the divertor pumping gaps for stationary OP2-loads of 100 kW/m(2) are being manufactured. Calculations were performed to optimize the intricate cooling channels and to lead as well as confirm the mechanical design for all working cases, the CFD calculation, static strength, ratcheting performance and fatigue assessment are performed in sequence. The findings of these analyses lead also to adaptation of the plasma operating instructions for OP2.

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.

Experiments were performed during the first divertor operational phase (OP1.2a) of the Wendelstein 7-X stellarator to verify predictions of potential overload conditions corresponding to certain high-power long-pulse OP2 scenarios. A potential solution to this overload is the installation of new divertor components called scraper elements, which are designed to intercept heat flux that would otherwise be incident on low-rated divertor edges. Heat flux measurements were obtained in a series of magnetic configurations designed to mimic the magnetic topology evolution caused by net toroidal current and beta, which is not directly accessible in OP1.2a. The experimental flux patterns are qualitatively reproduced in position and magnitude for by field line diffusion simulations using ad hoc cross-field diffusivities near the value used to design the scraper element. however, some important differences are observed, including a shift towards the pumping gap and low-rated components. Potential sources of discrepancy such as toroidal current evolution and error fields are discussed. A shift in the experimental heat flux pattern due to increasing toroidal current is observed in a 12s discharge.

Simulations of heat fluxes to the plasma facing components in the Wendelstein 7-X stellarator will be tested in its next operational phase. The simulations consist of core transport calculations that determine the evolution of the kinetic profiles and the toroidal current, which modifies the fluxes to the divertor, as the magnetic geometry changes. An additional divertor component, the scraper element, was designed to protect the edges of the primary divertor throughout this evolution during certain high-power long-pulse operational scenarios. The effect of unknown parameters of the heat flux calculations, namely, the cross-field thermal diffusivity and the magnetic field structure, is explored. The predicted scaling of the heat flux widths and magnitudes is presented, along with a new method of calculating the 3-D magnetic field structure required to perform the flux calculations.

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.