Divertor Target Elements Research Articles

Migration and balance of 13C in the stellarator Wendelstein 7-X after 13CH[formula omitted] local injection into the island divertor

A 13C tracer experiment was carried out to investigate impurity transport and migration in the stellarator Wendelstein 7-X (W7-X). About 4.0 × 1022 molecules of 13C marked methane were injected through the divertor gas injection system during plasma operation. After the experiment, the divertor target elements were extracted to be analyzed by post-mortem methods. In particular, the results of the Nuclear Reaction Analysis are presented, and the 13C balance and predominating contributors to migration are discussed.A two-dimensional 13C deposition pattern on 12C graphite modules was obtained with over 6000 single measurements. The post-mortem analysis results extrapolations show a high local deposition of around 25% of the 13C within a 5 cm radius around the injection holes embedded in a graphite target module of one of the ten divertor units. Additionally, a striation is detected in the toroidal direction along the outer strike line and the net carbon deposition zone along the intersecting magnetic island containing about 23% of the 13C within the same divertor unit where injections were made. The residual 52% of 13C are globally migrated across the machine. In total, ∼ 90% of the injected 13C could be extrapolated by post-mortem analysis. Thus, the 13C balance is closed. The distribution indicates a non-negligible amount of global transport and the strong influence of the magnetic topology on migration. In contrast to the strike line pathway, the transport along the magnetic island only takes place counterclockwise in toroidal direction. These results are used to provide benchmark data for ERO2.0 impurity tracing simulations and to be a reference for impurity migration processes in a stellarator.

Distributions of deposits and hydrogen on the upper and lower TDUs3 target elements of Wendelstein 7-X

Distributions of deposits and hydrogen (H) on the graphite divertor target elements TM4h4 and TM3v5 in the test divertor units 3 (TDUs3) of Wendelstein 7-X (W7-X) are studied. The TM4h4 and TM3v5 are located at the magnetically symmetric positions in the upper and lower divertor. The microstructure of the deposition layer is characterized by a transmission electron microscope (TEM) combined with a focused ion beam (FIB). Metallic deposits such as iron (Fe), molybdenum (Mo), chromium (Cr) are detected in the deposition layer by energy-dispersive x-ray spectroscopy (EDS). The depth-resolved distribution patterns of boron (B) and metallic deposits on upper and lower horizontal (h) divertor target elements TDUs3-TM4h4 as well as upper and lower vertical (v) divertor target elements TDUs3-TM3v5 are clarified by glow discharge optical emission spectrometry (GDOES). Results for both TDUs3-TM4h4 and TDUs3-TM3v5 show that the B deposition regions exhibit higher H retention due to the co-deposition with deposits. On the other hand, up-down asymmetries in B deposition caused by particle drift exist on both TDUs3-TM4h4 and TDUs3-TM3v5. The B deposition amount on upper TDUs3-TM4h4 is 40% smaller than that on lower TDUs3-TM4h4. While for the vertical target elements, the B deposition amount on upper TDUs3-TM3v5 is 35% larger than that on lower TDUs3-TM3v5. Meanwhile, a shift of around 3 cm in B deposition peaks is observed on upper and lower TDUs3-TM4h4 and TDUs3-TM3v5. Results of numerical simulation of carbon deposition/erosion profiles on the target elements using ERO2.0 code and power flux measured by infrared cameras are shown and compared with the above mentioned B profiles.

Investigation of boron distribution and material migration on the W7-X divertor by picosecond LIBS

One set of horizontal target elements of the Test Divertor Units (TDU), retrieved from the Wendelstein 7-X (W7-X) vessel after the end of second divertor Operation Phase (OP1.2B) in Hydrogen (H), were investigated by picosecond Laser-Induced Breakdown Spectroscopy (ps-LIBS). The Boron (B) distribution, H pattern and the material erosion/deposition pattern on these target elements were analyzed with high depth resolution and mapped in the poloidal direction of W7-X. From the spectroscopic analysis, B, H, Carbon (C) and Molybdenum (Mo) were clearly identified. A non-uniformly distributed B pattern on these divertor target elements was determined by the combination of B layer deposition during the three boronizations and W7-X plasma operation with multiple erosion and deposition steps of B. Like the TDU, the analyzed target elements are made of fine grain graphite, but have two marker layers which allow us to determine the material migration via the ps-LIBS technique. Two net erosion zones including one main erosion zone with a peak erosion depth of 6.5 μm and one weak erosion with a peak erosion of 1.3 μm were determined. Between two net erosion zones, a net deposition zone with width of 135 mm and a thickness up to 3.5 μm at the peak deposition location was determined by the ps-LIBS technique. The B distributions are correlated with the erosion/deposition pattern and the operational time in standard magnetic configuration of W7-X in the phases after the boronizations. The thickness of the containing B layer on these target elements also correlates with the erosion/deposition depth, in which the thickness of the containing B layer varies spatially in poloidal direction between 0.1 μm and 6 μm. Complementary, Focused Ion Beam combined with Scanning Electron Microscopy (FIB-SEM) was employed also to verify and investigate the deposition layer thicknesses at typical net erosion and net deposition zones as well as to identify the three boronizations in depth.

Thermal-hydraulics modeling for prototype testing of the W7-X high heat flux scraper element

Long-pulse operation of the Wendelstein 7-X (W7-X) stellarator experiment is scheduled to begin in 2020. During this operational phase the vessel will be equipped with water-cooled plasma facing components to allow for longer pulse durations. Certain simulated plasma scenarios have been shown to produce heat fluxes that could surpass the technological limits of the divertor target elements during steady-state operation. To reduce the heat load on the target elements, the addition of a “scraper element” (SE) is under investigation. The SE is composed of 24 water-cooled carbon fiber reinforced carbon composite monoblock units. Multiple full-scale prototypes have been tested in the GLADIS high heat flux test facility. Previous computational studies revealed discrepancies between the simulations and experimental measurements. In this work, single-phase thermal-hydraulics modeling was performed in ANSYS CFX to identify potential causes for such discrepancies. Possible explanations investigated were the effects of a non-uniform thermal contact resistance and a potential misalignment of the monoblock fibers. While the difference between the experimental and computational results was not resolved by a non-uniform thermal contact resistance, the computational results provided insight into the potential performance of a W7-X monoblock unit. Circumferential temperature distributions highlighted the expected boiling regions of such a unit. Furthermore, simulations revealed that modest angles of fiber misalignment in the monoblocks result in asymmetries at the unit edges and provide temperature differences similar to the experimental results.

Summary of the production of the divertor target elements of Wendelstein 7-X

The realization of the 19.6m2 highly heat loaded part of the divertor of Wendelstein 7-X (W7-X) requires the installation of 890 target elements. A target element is made of a CuCrZr copper alloy heat sink armored with carbon fibre reinforced carbon CFC NB31 tiles. The industrial production of the target elements by the Austrian company PLANSEE SE needed 5 years. The successful delivery of a total of 973 target elements (armored with ∼16,200 tiles) was based on an efficient quality assurance throughout the production by the manufacturer: 44 different quality examinations for 82 manufacturing steps. It was followed by an intensive quality assessment by IPP: visual inspections, dynamic pressure tests, He-leak testing in vacuum oven, high heat flux testing. The quality of the delivered elements, and in particular the reliability of the bonding between CFC tiles and heat sink, was confirmed by high heat flux testing on the basis of a statistical approach. The essential recovery of CuCrZr properties after ageing processes was confirmed by the measurement of hardness and electrical conductivity. 104 target elements (10.7%) were accepted after repair.

Local copper coating of the connectors of the divertor target elements of Wendelstein 7-X

Water connectors of the divertor target elements of Wendelstein 7-X (W7-X) are made of CuCrZr with a Ni (270) transition to stainless steel 316L. A small He leak rate was always detected in the welding area between 316L and Ni of some target elements. To seal these connectors, the coating onto round tubes with very high purity oxygen-free Cu CW009A was developed. This technology was validated with tests reproducing expected operation conditions in W7-X. Metallography examinations indicated the good adhesion without pores or cracks on connectors. The main characteristics of the coating are: a ∼1μm thick Ni interlayer, a 0.3mm thick and 20mm long Cu coating. The previous un-tight target elements passed the He leak testing in vacuum oven after coating.

Multiphysics Analysis of the Wendelstein 7-X Actively Cooled Scraper Element

The Wendelstein 7-X stellarator experiment is scheduled to start operation in mid-2015, and to move to steady-state operation in 2019. During this steady-state operation, certain plasma scenarios have been shown to produce heat fluxes that exceed the technological limits on the edges of the divertor target elements. The addition of a so-called scraper element (SE) in the ten divertor locations is being investigated in order to reduce the heat load on these divertor target edges. The ANSYS commercial multiphysics package is used to model the performance of the SE under predicted operational conditions. Computational fluid dynamics (CFD) modeling is performed to analyze the hydraulic and thermal characteristics of the water-cooled SE under thermal loading using the ANSYS CFX software. This multiphysics modeling is performed for the entire SE to ensure that the total pressure drop in the cooling water circuits, the increase in water temperature, and the peak temperature in the CFC all satisfy the design requirements. Because the contour of the SE surface must be machined to a sub-millimeter precision, it is important to determine the amount of thermal expansion experienced by the entire SE. The thermal-hydraulic results are imported into ANSYS Mechanical to perform the thermal-structural analysis. The thermal deformation of the SE is examined to confirm that the component’s position will remain within its operational limits.

Experience gained with the 3D machining of the W7-X HHF divertor target elements

The high heat flux (HHF) divertor of W7-X consists of 100 target modules assembled from 890 actively water-cooled target elements protected with CFC tiles. The divertor surface will be built up of individually 3D machined target elements with 89 individual element types. To date 300 of the 890 target elements have been 3D machined with a very good accuracy. To achieve this successful result, a prototyping phase has been conducted to qualify the manufacturing route and to define the acceptance criteria with measures taken to minimize the risk of unacceptable damage during the manufacturing. After the 3D-machining, during the incoming inspection, copper infiltration from the interface between the CFC tiles and the CuCrZr heat sink to the plasma facing surface was detected in a small number of elements.

Status of High Heat Flux Components at W7-X

The actively water-cooled in-vessel components (IVCs) of the stellarator Wendelstein 7-X consist of the divertor, the first wall protection components, the port liners, each designed for different loading conditions, and the associated pipework, the control coils, the cryo-pump system, the Glow discharge electrodes, and a set of diagnostics. The divertor, designed for high heat fluxes (HHFs), is a set of 10 target and baffle units arranged along the plasma surface. The design and production of these HHF components is a challenging task. The divertor target elements, which are based on flat carbon-carbon fiber composite tiles bonded via active metal casting onto CuCrZr cooling structures required intensive development and testing to reach a reliable performance; removing, under stationary conditions, 10 MW/m2. Industrially manufactured high quality target elements have been delivered and assessed, and the process of incorporating them into assembly units, so-called modules, has begun. The time scale for the completion of the HHF divertor has been held for the last four years and the final delivery of the HHF divertor is still planned in 2017. In parallel to the realization of the divertor, most of the remaining IVCs have been defined, developed, designed, and fabricated and the installation of many of these components has begun. Some of these components can also be expected, for a short period of time, to receive high heat loads approaching those of the divertor. These components will be described, in detail, from conception to realization.

Summary of Research and Development Activities for the Production of the Divertor Target Elements of Wendelstein 7-X

The realization of the high heat flux divertor of the stellarator Wendelstein 7-X requires the production of 890 target elements. Since the divertor has to operate for 30 min with an input continuous wave (cw) plasma power of 10 MW, the target elements need to be actively water cooled. A target element is made of a CuCrZr copper alloy heat sink armored with carbon fiber reinforced carbon (CFC) NB31 tiles. It is designed to remove a stationary heat flux of 10 MW/m <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> on its main area and 5 MW/m <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> at the end adjacent to the pumping gap. Research and development activities proved to be essential for the preparation and the support of the on-going serial fabrication. The most critical issue is the stable production of the bonding between the CFC tiles and the heat sink as the divertor surface consists of ~ 16 000 tiles. The fast unexpected failure of one tile may reduce the whole divertor performance and, consequently, restrict the stationary machine operation. Two main research and development phases allow the design improvements of the target elements compatible with industrial and economic constrains of the series production as well as the development of quality controls such as a visual inspection system able to easily detect and fast record cracks at the interface between CFC tiles and heat sink.

Pre-qualification of brazed plasma facing components of divertor target elements for ITER like tokamak application

Qualification of tungsten (W) and graphite (C) based brazed plasma facing components (PFCs) is an important R&D area in fusion research. Pre-qualification tests for brazed joints between W–CuCrZr and C–CuCrZr using NDT (IR thermography and ultrasonic test) and thermal fatigue test are attempted. Mockups having good quality brazed joints of W and C based PFCs were identified using NDT. Subsequently, thermal fatigue test was performed on the identified mockups. All brazed tiles of W based PFC mockups could withstand thermal fatigue test, however, few tiles of C based PFC mockup were found detached. Thermal analyses of mockups are performed using finite element analysis (ANSYS) software to simulate the thermal hydraulic condition with 10 MW/m 2 uniform heat flux. Details about experimental and computational work are presented here.

Fabrication and characterization of tungsten and graphite based PFC for divertor target elements of ITER like tokamak application

The development of the fabrication technology of macro-brush configuration of tungsten (W) and carbon (graphite and CFC) plasma facing components (PFCs) for ITER like tokamak application is presented. The fabrication of qualified joint of PFC is a requirement for fusion tokamak. Vacuum brazing method has been employed for joining of W/CuCrZr and C/CuCrZr. Oxygen free high conductivity (OFHC) copper casting on W tiles was performed followed by machining, polishing and ultrasonic cleaning of the samples prior to vacuum brazing. The W/CuCrZr and graphite/CuCrZr based test mockups were vacuum brazed using silver free alloys. The mechanical shear and tensile strengths were evaluated for the W/CuCrZr and graphite/CuCrZr brazed joint samples. The micro-structural examination of the joints showed smooth interface. The details of fabrication and characterization procedure for macro-brush tungsten and carbon based PFC test mockups are presented.

Upgraded acceptance criteria from transient thermography control for the W7-X divertor target elements

The commissioning of plasma-facing component fields needs advanced non-destructive methods to detect in a reliable way the defects, which can impair the component performances and/or integrity during operation. Within this framework, CEA developed a dedicated non-destructive examination method based on active infrared thermography (SATIR facility) to inspect the bonding between armour material and metallic heat sink. Used with successful in the commissioning of the toroidal pump limiter of Tore Supra, this technique was applied in the frame of the pre-series activities of the Wendelstein 7-X high heat flux divertor elements to assess the bonding quality of the delivered components. This paper presents the methodology adopted to define an acceptance criterion based on SATIR test bed possibly applied for a serial inspection of the Wendelstein 7-X elements. Using the well-tried acceptance test based on the DTref_max parameter, the new method includes advanced data post-processing techniques from thermo-signal SATIR and a data merging method to help the decision-making and to optimise the reliability of the binary response expected for a final decision in terms of acceptance test.

High heat flux tests of the WENDELSTEIN 7-X pre-series target elements: Experimental evaluation of the thermo-mechanical behaviour

The high heat flux (HHF) testing of WENDELSTEIN 7-X pre-series target elements is an indispensable step in the qualification of the manufacturing process. A set of 20 full scale pre-series elements was manufactured by PLANSEE SE to validate the materials and manufacturing technologies prior to the start of the series production. The HHF tests were performed in the ion beam test facility GLADIS. All actively water-cooled elements were tested for about 100 cycles at 10 MW/m 2 (10–15 s pulse duration). Several elements were loaded with even higher cycle numbers (up to 1000) and heat loads up to 24 MW/m 2. Hot spots were, observed at the edges of several tiles during the HHF tests indicating local bonding problems of the CFC. The thermo-mechanical behaviour under HHF loading has been evaluated and compared to the FEM predictions. The measured temperatures and strains confirm the chosen FEM approach. This allows a component optimisation to achieve a successful series production of the W7-X divertor target elements.

Power load limits of the WENDELSTEIN 7-X target elements—comparison of experimental results and design values for power loads up to the critical heat flux

The power load limits of the WENDELSTEIN7-X divertor target elements were experimentally evaluated with heat loads considerably exceeding the expected operating conditions. The water-cooled elements are designed for steady-state heat flux of 10 MW m−2 and to remove a power load up to 100 kW. The elements must allow a limited operation time at 12 MW m−2 steady-state and should not fail for short pulses of up to 15 MW m−2 for cooling conditions in the subcooled nucleate boiling regime. In the framework of the qualification phase, pre-series target elements were loaded up to 24 MW m−2 without loss of CFC tiles. A critical heat flux at the target of 31 MW m−2 was achieved. The paper discusses the results of the tests performed at the high heat flux test facility GLADIS. The experimental results compared to transient nonlinear fine element method (FEM) calculations confirm a high thermal safety margin of the target design sufficient for plasma operation in W7-X.