Tungsten Tiles Research Articles

Current state-of-the-art manufacturing technology for He-cooled divertor finger

A divertor concept for DEMO has been investigated at Karlsruhe Institute of Technology (KIT) which has to withstand a heat flux of 10MW/m2. The design utilizes small finger module composed of a small tungsten tile brazed on a thimble made from tungsten alloy. The divertor finger is cooled by helium jet impingement at 10MPa and 600°C. The subject of this paper is technological studies on machining and braze joining the divertor components. Goal of this task, which is considered an important R&D issue, is to find out appropriate manufacturing methods to ensure high functionality and high reliability of the divertor as well as to meet the economic aspect. One of the major requirements for manufacturing is micro-crack-free surface of tungsten parts, since crack propagations in tungsten were observed in the previous high-heat-flux tests at Efremov. Different manufacturing methods and the corresponding results are discussed in the following report.

Evaluation of the scavenging effect by low temperature laboratory plasmas driven with radiofrequency

Proposed materials for ITER are beryllium for first wall and tungsten tiles (plus carbon fibre composites) for the divertor. The use of carbon materials at the divertor plates of ITER can be problematic due to the formation of tritium-containing co-deposits coming from the products of chemical erosion and physical sputtering. Scavenging effect in the hidden regions of the divertor has been proposed as a possible technique to eliminate the formation of amorphous carbon films. Nitrogen gas has been proposed as a good scavenger. Although a complete suppression of the formation of amorphous carbon films by nitrogen was achieved in laboratory experiments (in a low temperature plasma), only a moderate reduction (<30–40%) was seen at JET in type-III ELM discharges (Tabares et al 2005 J. Nucl. Mater. 337–339 867–71). This justifies further analyses for an understanding of physicochemical processes for the deposition/inhibition of carbon films. In this work, experiments in RF plasma discharges at low pressures of CH4/N2 admixtures were carried out. The formation of amorphous carbon films was investigated under different physical conditions, with high and low energetic ions on powered and floating electrodes, respectively. Optical emission spectroscopy of ion and neutral species was employed as the main diagnostic.

Analysis, verification, and benchmarking of the transient thermal hydraulic ITERTHA code for the design of ITER divertor

An analytical study for the International Thermonuclear Experimental Reactor Thermal Hydraulic Analysis code (ITERTHA) is carried out for a copper divertor with a 5 mm tungsten tile. The influence of the incident heat flux, swirl-tape insertion in cooling channels as well as the coolant flow velocity on the divertor thermal response is analyzed and discussed. The ITERTHA code results are verified by the commercial finite element code, COSMOS. The heat transfer coefficients at the nodes located on the cooling channel-wall are determined outside COSMOS code by the same methodology used in ITERTHA. A good agreement is achieved under different incident heat fluxes. The ITERTHA code is also benchmarked against the thermal-hydraulic calculation of the outer divertor of the Fusion Ignition Research Experiment, FIRE for an incident heat flux of 20 MW/m 2 and coolant flow velocity of 10 m/s in a cooling channel of 8 mm diameter with swirl-tape inserts of 2 ratio and 1.5 mm thickness. The results show excellent agreement for both steady and transient states and prove the successful implementation of both the hydraulic and heated diameters of the swirl-tape channels in the used heat transfer correlations.

Hydrogen interaction with Al2O3-coated tungsten under plasma irradiation

The effect of amorphous films of Al2O3 on deuterium trapping in polycrystalline tungsten under plasma irradiation (150 eV/D) and high fluence (up to 4 × 1024 D/m2) was studied by the methods of nuclear reaction product spectroscopy and thermal desorption spectroscopy. It was determined that the hydrogen permeation blocking properties of aluminium oxide lead to a great increase in deuterium trapped in tungsten. With the complete erosion of the film, this effect continues but with a lower amplitude. Extrapolation of this property to beryllium oxide implies that the presence of beryllium oxide on the surface of the tungsten tiles of the ITER leads to a considerable increase in the tritium inventories accumulated in them.

Modeling tungsten and carbon sputtering by carbon at elevated temperatures

The planned use of carbon strike plates adjacent to tungsten tiles in the ITER divertor will lead to W erosion by sputtered C ions. The W sputtering yield by C decreases with increasing surface concentration of the implanted C. However with increasing temperature, the C self-sputtering yield increases, reducing the rate of C surface coverage. This study attempts to model the temperature dependent C self-sputtering process and its impact on W erosion by using modified surface binding energies (SBEs) in TRIDYN. The SBEs are benchmarked against C self-sputtering yields measured in this study and W irradiation by C is simulated and compared to previous experimental results. Good agreement between simulation and experiments is seen at 670 K, while small deviations are observed at 770 K due to surface roughening. At 870 K, significant deviation is observed which can be attributed to enhanced surface roughening as well as diffusion effects.

Localized tungsten deposition in divertor region in JT-60U

Tungsten local transport in the diverter region emitted from the outer divertor was studied in JT-60U. Tungsten-coated CFC tiles were installed in the outer divertor position, on which outer strike points did not normally sit. A neutron activation method to measure tungsten surface density was used for the first time. On the outer wing tile, W existed only in the top surface, while on the inner divertor tile W was embedded in the thick carbon deposition layer. Poloidal distribution in the same toroidal section as the tungsten tiles showed dense W deposition near the inner strike points, the dome top and outer wing of the dome. Carbon ions by puffing 13CH4 at the same toroidal position of the W-tiles were also deposited near the inner strike points. Toroidal distribution on the outer wing showed significant localization of W deposition near the W tile array. In addition, toroidal asymmetry of W deposition on the outer wing probably due to the plasma parallel flow was observed.

Analysis and design of the beryllium tiles for the JET ITER-like wall project

Work is in progress to completely replace, in 2008/9, the existing JET CFC tiles with a configuration of plasma facing materials consistent with the ITER design. The ITER-like wall (ILW) will be created with a combination of beryllium (Be), tungsten (W), W-coated CFC and Be-coated inconel tiles, with the material depending on the local anticipated heat flux and geometry. Over 4000 tiles will be replaced and the ILW will accommodate additional heating up to at least 50 MW for 10 s. One of the objectives is to maintain or improve the existing CFC tile power handling performance which has been achieved in most cases by hiding bolt holes, optimising tile size and profile and introducing castellations on plasma facing surfaces. This paper describes the generic problems associated with the Be tiles (power handling capacity and disruption induced eddy currents) and illustrates the solution selected for the inner wall guard limiter (IWGL) where the present CFC tiles will be replaced with Be.

3D finite element electromagnetic and stress analyses of the JET LB-SRP divertor element (tungsten lamella design)

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.

R&D on full tungsten divertor and beryllium wall for JET ITER-like wall project

The ITER reference materials have been tested separately in tokamaks, plasma simulators, ion beams and high heat flux test beds. In order to perform a fully integrated material test JET has launched the ITER-like Wall Project with the aim of installing a full metal wall during the next major shutdown. As a result of R&D projects in 2005–2006, bulk tungsten tiles are foreseen at the outer horizontal target and tungsten coating at the other divertor tiles. In some regions of the main chamber, beryllium coated Inconel tiles and bulk beryllium tiles are utilised which include marker tiles as erosion diagnostics. This paper gives an overview of the R&D carried out in the frame of the ITER-like Wall Project on the development of an inertially cooled bulk tungsten tile design and the characterization of tungsten and beryllium coating technologies.

Fabrication of a He-cooled divertor module for DEMO reactor

One current design of the divertor for a fusion reactor like DEMO uses He-cooled thimble-like tungsten, which is covered by sacrificial tungsten tile. Each thimble has to be connected with a supporting unit made from ferritic steel. This paper describes the development of joining techniques between tungsten thimbles and steel supporting units. Paper also provides an evaluation of simple geometries up to more complex conical interlocks filled with cast copper. Four candidates tungsten alloys (WL10, W-single crystal, W–Cu composite and chemical vapour deposited (CVD) tungsten) were experimentally checked by ‘non-isothermal’ heating to characterize the thermal gradient in the range 600 °C (for joint) and more than 1000 °C (for thimble top) using a special testing procedure. Basing on the test results, several mock-ups were manufactured for future high heat flux testing in a helium loop.

Development of a helium-cooled divertor: Material choice and technological studies

Within the framework of the EU power plant conceptual study (PPCS), a He-cooled divertor concept has been investigated at Forschungszentrum Karlsruhe in cooperation with the Efremov Institute. The design goal is to remove a high heat load of at least 10 MW/m 2. The design is based on a modular construction of cooling finger unit that helps reduce thermal stresses. The divertor finger unit, which is cooled by high pressure helium, consists of a tungsten tile as thermal shield and sacrificial layer, and a thimble made of tungsten alloy. The success of this design depends strongly on the effectiveness of the cooling technology and on the availability of appropriate structural materials such as tungsten alloy and oxide-dispersion-strengthened (ODS) steel as well as the related fabrication and joining technology. Results of this investigation are discussed in the paper.

Examination of C/C flat tile mock-ups with hypervapotron cooling after high heat flux testing

Two C/C flat tile mock-ups with a hypervapotron cooling concept, have been successfully tested beyond ITER specification (3000 cycles at 15 MW/m 2, 300 cycles at 20 MW/m 2 and 800–1000 cycles at 25 MW/m 2) in two electron beam testing facilities [F. Escourbiac, et al., Experimental simulation of cascade failure effect on tungsten and CFC flat tile armoured HHF components, Fusion Eng. Des., submitted for publication; F. Escourbiac, et al., A mature industrial solution for ITER divertor plasma facing components: hypervapotron cooling concept adapted to Tore Supra flat tile technology, Fusion Eng. Des. 75–79 (2005) 387–390]. Both mock-ups provide a SNECMA SEPCARB ® NS31 armour, which has been joined onto the CuCrZr heat sink by active metal casting (AMC) and electron beam welding (EBW). No tile detachment or sudden loss of single tiles has been observed; a cascade-like failure of flat tile armours was impossible to generate. At the maximum cyclic heat flux load of 25 MW/m 2 all tested tiles performed well except one, which revealed already a clear indication in the thermographic examination at the end of the manufacture. Visual examination and analysis of metallographic cuts of the remaining tiles demonstrated that the interface has not been altered. In addition, the shear strength of the C/C to copper joints measured after the high heat flux (HHF) test has been found to be still above the interlamellar shear strength of the used C/C material. The high resistance of the interface is explained by a modification of the C/C to copper joint interface due to silicon originating from the used C/C material.

Overview of the ITER-like wall project**EURATOM/UKAEA Fusion Association, Abingdon, OX14 3DB, UK

Work is in progress to completely replace, in 2008/9, the existing JET CFC tiles with a configuration of plasma facing materials consistent with the ITER design. The ITER-like wall (ILW) will be created with a combination of beryllium (Be), tungsten (W), W-coated CFC and Be-coated inconel tiles, with the material depending on the local anticipated heat flux and geometry. It is part of an integrated package of JET enhancements whose aim is to develop an understanding of the ITER materials issues and develop the techniques required to operate with inductive and advanced scenarios as close as possible to ITER parameters. Over 4000 tiles will be replaced and the ILW will accommodate additional heating up to at least 50 MW for 10 s. This paper describes the scientific background to the project, the technical objectives, the material configuration selected, the R&D behind the practical realization of the objectives and the generic problems associated with the Be tiles (power handling capacity and disruption induced eddy currents). One of the objectives is to maintain or improve the existing CFC tile power handling performance which has been achieved in most cases by hiding bolt holes, optimizing tile size and profile and introducing castellations on plasma facing surfaces.

Development and testing of a bulk tungsten tile for the JET divertor

The ITER-like wall project has been launched to design, manufacture and test all the necessary components in view of their installation in a dedicated shutdown 2008–09. One of the R&D activities in the frame of the project is the development of a concept of an inertially cooled bulk W component for the JET divertor. Two options were studied as the concepts, (i) W/CFC brazing design which consists of W blocks on a carbon fibre composite (CFC) tile; (ii) W lamellae design, which is a mechanical assembly of several stacks of W blades. Both concepts were realized as test tiles and tested under high heat flux loading. The lamellae test tile showed a high performance fulfilling the requirements. Finally, the lamellae concept was chosen to be the primary solution and comprehensive numerical analyses were performed for the further optimization of the concept.

Material Integration Design for the Divertor Target Plate of EAST Device

In the next generation of fusion device in China, e.g., the Experimental Advanced Superconducting Tokamak (EAST), the divertor target will be exposed to high heat loads up to 5 MW/m2 for about 1000 s. An actively water-cooled target plate element with flat tungsten tile armored on CuCrZr heat sink was designed for EAST. A two-dimensional finite element method (FEM) code was used to analyze its thermal and mechanical properties under high heat flux of 10 MW/m2 for the selection of an appropriate cross section. To meet the integrated requirements of temperature and stress in the target element, twisted tapes have to be inserted into the cooling channels to strengthen the heat transfer efficiency, and a tungsten armor thickness of 4 mm and a distance of 2 mm from the interface to the vertex of the cooling channel were ultimately selected. The thermal and mechanical properties of two kinds of tungsten armor (sintered and plasma sprayed) were also analyzed and discussed in the FEM calculations. The designed structure can be used under the 5 MW/m2 heat load expected for normal operation of EAST device, but it would suffer from cracks/failure danger under higher heat load, up to 10 MW/m2.

Development of the plasma facing components for the dome-liner component of the ITER divertor

On the basis of the design and the specification of the dome-liner elaborated by EFDA, a manufacturing route based on high temperature brazing has been developed and proved by means of the fabrication and testing of several samples and mock-ups. The dome is protected with tungsten armour tiles joined onto heat sinks obtained from a bimetallic plate made of precipitation hardened copper–chromium–zirconium alloy and stainless steel realized by explosion bonding. The brazed joint between the tungsten tiles and the heat sink has been qualified by means of thermal fatigue tests on small-scale mock-ups in reactor relevant conditions. The properties of the explosion bonding joint between the front copper alloy plate to the rear steel backing has been assessed by means of an extensive metallurgical and mechanical test program according to the specification provided by EFDA. The dimensional stability during the fabrication route has been investigated by means of the realization of a relevant curved component that has been dimensionally tested after the completion of each step of the manufacturing route. The results of the experimental activity are presented and discussed in this paper.

Stress Analysis and Lifetime Evaluation of ITER Divertor High Heat Flux Components of the Hypervapotron Type

The main objectives of the work are the stress analysis and lifetime evaluation of the hypervapotron heat sink under loads corresponding to ITER conditions. Various design options of hypervapotron were considered. Developed 3D finite element models of the hypervapotron structure include a heat sink, tungsten tiles, a copper interlayer and stainless steel leg which connects the hypervapotron to the support structure. Coupled transient thermal and elastic-plastic analyses were done for all analyzed options of the hypervapotron. Lifetime analysis was performed from the low cycle fatigue and fatigue crack propagation points of view.

Stress Analysis and Lifetime Evaluation of ITER Divertor High Heat Flux Components of the Hypervapotron Type

The main objectives of the work are the stress analysis and lifetime evaluation of the hypervapotron heat sink under loads corresponding to ITER conditions. Various design options of hypervapotron were considered. Developed 3D finite element models of the hypervapotron structure include a heat sink, tungsten tiles, a copper interlayer and stainless steel leg which connects the hypervapotron to the support structure. Coupled transient thermal and elastic-plastic analyses were done for all analyzed options of the hypervapotron. Lifetime analysis was performed from the low cycle fatigue and fatigue crack propagation points of view.

Plasma operation with tungsten tiles at the central column of ASDEX Upgrade

Two rows of the graphite tiles (≈1.2 m 2) at the lower end of the central column of ASDEX Upgrade were coated with tungsten. Since no high heat fluxes are present in this region, the tiles were coated by a physical vapour deposition technique to a thickness of only 500 nm. The experimental campaign was started without wall conditioning by coating of the vacuum vessel. This allowed to measure the W-erosion and the W-concentration with an almost pure tungsten surface exposed to the plasma. The influx as well as the concentration were monitored by spectroscopic methods. The penetration probability for tungsten from the central column is deduced from direct laser ablation of the coating. The measurements were complimented by deposition probe measurements in the midplane and in the divertor. In all important discharge scenarios, which were performed already during the phase without wall conditioning, no concentrations above the detection limit of about 5×10 −6 were found. This result is supported by the deposition probe measurements, which showed no or very small amounts of W deposited in single discharges. The concentrations are at least a factor of 10 below the acceptable concentration in ASDEX Upgrade. Consequently, no influence on the overall plasma behaviour was found and W seems also to be suited for use on larger surfaces in ASDEX Upgrade. Extrapolation to ITER conditions yields concentrations, which will not prohibit successful operation.

Tungsten filament mock-ups for gas box liner

Carbon impurities create carbon films saturated by tritium and deuterium onto the gas box liner. The reference gas box liner formed by chevron-shaped tungsten tiles is capable of depositing and/or recombining only 5% of the carbon impurities. The innovative solution presented here increases within the range 10 2 to 10 3 the surface interacting with the pumped gas by using an array of filaments, either disordered (felt) or ordered (spiral). This paper describes the experiments carried out to prepare felt/spiral samples and indicates possible ways of fabricating a representative gas box liner (GBL) mock-up.