CFC Tiles Research Articles

Surface compositional study of Be and T contaminated CFC tiles from JET

In fusion devices, in the region next to high temperature plasma, material could be eroded from plasma-facing materials in one location and is transported to other, sometimes remote, locations throughout the device. These plasma-created materials are so-called mixed materials. Previous studies on JET samples have shown that compositional changes in the depth profiles of the mixed-material surfaces can be correlated to the operational history of the machine. SEM and detailed EDX analysis of CFC Be and T contaminated samples coming from the JET divertor (shutdown 2007) have been carried out in the SCKCEN hot cell facilities. The main aim of these measurements is to have a clear image of the erosion/deposition pattern in those samples for each element. The processes for material migration and the morphology of the deposits are discussed. XPS/AES measurements will also be carried out to understand the possible compound processes, mainly focussed on Be2C formations.

Performance and statistical quality assessment of CFC tile bonding on the pre-series elements of the Wendelstein 7-X divertor

Extensive high heat flux (HHF) testing of pre-series IV targets was performed to establish the industrial process for the ongoing production of the actively water-cooled target elements which will be needed for the installation of the Wendelstein 7-X (W7-X) divertor. Finally, 890 components covered with about 18,000 CFC tiles will be installed. The examinations of the elements with 10 MW/m 2 cycling up to 10,000 pulses, 16 MW/m 2 cycling and screening tests up to 32 MW/m 2, confirm the robustness of the design and in particular of the applied CFC bonding technology. The results of the IR examination of the initial tests have been assessed statistically. The paper presents a detailed statistical analysis based on the Six-Sigma method of the surface temperature increase of the CFC tiles tested for 100 cycles at 10 MW/m 2. Assuming that the series elements will behave in a similar fashion to the pre-series elements this statistical assessment can also be performed for the series elements.

Heat load measurements on the JET first wall during disruptions

Disruptions in tokamaks lead to high heat loads onto the plasma facing components (PFC). Two processes, of particular concern for the first wall integrity, have been studied in dedicated experiments at JET: (1) During the thermal quench, it is measured (using fast IR thermography) that 5% of the plasma stored energy is deposited onto the outer and inner poloidal limiters. More surprisingly, during the current quench, about 10% of the magnetic energy is deposited onto the outer and inner poloidal limiters via plasma wall interaction. (2) Very localised heat loads due to runaway electrons, generated in disruptions triggered by massive injection of argon and neon, are measured onto the JET upper dump plate. The temperature increase measured on the CFC tiles scales with the square of the runaway current.

Raman study of CFC tiles extracted from the toroidal pump limiter of Tore Supra

The structure of six tiles extracted from the erosion and deposition zones (thin and thick deposition) of the Tore Supra toroidal pump limiter (TPL) have been analyzed in the framework of the DITS campaign using micro-Raman spectroscopy. This post-mortem analysis gives information on both carbon structure and D content. We have found that the carbon structure is most often similar to that of plasma-deposited hard amorphous carbon layers. The role of the surface temperature during the discharge in the D content is investigated: in all locations where the temperature does not reach more than 500°C the D content seems to be roughly uniform with D/D+C≈20%.

Heat transfer characteristics of the first wall with graphite sheet interlayer

Aiming at a design study of the first wall structure for next fusion devices and cost-down, the heat transfer characteristics of the first wall structure with CFC tiles bolted to a heat sink were investigated with a variety of graphite sheets and fixing-bolt torque conditions. The heat load tests of the mock-up were performed with the heat fluxes of 1 and 3 MW/m 2 on JAEA electron beam irradiation system. The experimental results showed that the first wall structure with 0.1-mm thickness × 3 PGSs (Pyrolytic Graphite Sheet; Panasonic Co., Ltd.) had the highest heat transfer performance. Furthermore, the first wall structure with the PGSs withstood the heat fluxes of 1 MW/m 2 × 100 s and 3 MW/m 2 × 20 s. The maximum surface temperatures of the CFC tile were 500 °C, 800 °C, respectively. These results indicated that the bolted tile structure with the graphite sheets can be utilized as the first wall at the long pulse operation with the heat flux of several MW/m 2.

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.

Pre-series and testing route for the serial fabrication of W7-X target elements

The fabrication of the actively cooled high-heat flux divertor of the WENDELSTEIN 7-X stellarator (W7-X) requires the delivery of 890 target elements, which are designed to withstand a stationary heat flux of 10 MW/m 2. The organization of the manufacturing and testing route for the serial fabrication is the result of the pre-series activities. Flat CFC Sepcarb ® NB31 tiles are bonded to CuCrZr copper alloy cooling structure in consecutive steps. A copper layer is active metal cast to CFC tiles, and then an OF-copper layer is added by hot isostatic pressing to produce bi-layer tiles. These tiles are bonded by electron beam welding onto the cooling structure, which was manufactured independently. The introduction of the bi-layer technology proved to be a significant improvement of the bond reliability under thermal cycling loading. This result is also the consequence of the improved bond inspections throughout the manufacturing route performed in the ARGUS pulsed thermography facility of PLANSEE. The repairing process by electron beam welding of the bonding was also qualified. The extended pre-series activities related to the qualification of fabrication processes with the relevant non-destructive examinations aim to minimize the risks for the serial manufacturing and to guarantee the steady-state operation of the W7-X divertor.

Development and qualification of a bulk tungsten divertor row for JET

A bulk tungsten divertor row has been developed in the frame of the ITER-like Wall project at JET. It consists of 96 tiles grouped in 48 modules around the torus. The outer strike point is located on those tiles for most of the ITER-relevant, high triangularity plasmas. High power loads (locally up to 10–20 MW/m 2) and erosion rates are expected, even a risk of melting, especially with the transients or ELM loads. These are demanding conditions for an inertially cooled design as prescribed. A lamella design has been selected for the tungsten, arranged to control the eddy and halo current flows. The lamellae must also withstand high temperature gradients (2200 to 220 °C over 40 mm height), without overheating the supporting carrier (600–700 °C maximum). As a consequence of the tungsten emissivity, the radiative cooling drops appreciably in comparison with the current CFC tiles, calling for interleaved plasma scenarios in terms of performance. The compromise between shadowing and power handling is discussed, as well as the consequences for operation. Prototypes have been exposed in TEXTOR and in an electron beam facility (JUDITH-2) to the nominal power density of 7 MW/m 2 for 10 s and, in addition, to higher loads leading to surface temperatures above 2000 °C.

Testing of tungsten coatings in JET for the ITER-like wall

A CFC tile 5 from the JET outer divertor, and CFC tiles from neutral beam shine-through and re-ionisation regions were coated with tungsten and exposed during the 2005-7 JET campaigns in preparation for the ITER-like wall project. Approximately 1.6 microns of coating were eroded from the tile 5 during high-delta discharges when the outer strike-point is on the tile. The coatings on the other tiles were unaffected by NB-heating and divertor discharges, however a tile mounted near the centre of the Inner Wall Guard Limiter lost all its coating from the surface within 10 mm of the tile leading edge; this probably occurred during the ramp-up phase of JET discharges.

Preliminary assessment of possible carbide formation on Be and T contaminated CFC tiles from JET

The fusion plasma, with a typical temperature of 10 keV, has to be brought into contact with a physical wall in order to remove the helium produced and drain the excess energy in the fusion plasma. The fusion plasma is far too hot to be brought into direct contact with a physical wall. It would degrade the wall and the debris from the wall would extinguish the plasma. Therefore, schemes are developed to cool down the plasma locally before it impacts on a physical surface. The resulting plasma wall-interaction in ITER is facing several challenges including surface erosion, material redeposition and tritium retention. In JET facility both beryllium and carbon are used together as plasma facing materials, and the resultant surfaces show considerable mixing between the two elements. In this work a first XPS characterisation has been performed on the surface of several PFC (plasma facing components) samples obtained from various JET tiles from the Mk II GB divertor in JET. The substrates of the tiles are CFC, and the deposited films have various D/C ratios depending on the position in the tokamak. Ion beam analysis shows that the D/C ratio is ∼1 in the shadowed areas and Be/C ∼1–2 in the plasma-exposed regions of the inner divertor. These values could give rise to the formation of beryllium carbides, and indications of a carbide component are found in the XPS spectra. The processes for material migration and the morphology of the deposits are discussed.

Industrial scale 10 μm W coating of CFC tiles for ITER-like Wall Project at JET

As a result of the R&D phase of the ITER-like Wall Project, combined magnetron sputtering and ion implantation (CMSII) technique was selected for 10 μm W coating of approx. 1000 CFC tiles for the new JET first wall. This technique involves simultaneous magnetron sputtering and high energy ion bombardment. A high voltage pulse discharge is superposed over the magnetron deposition and by this way positive ions are accelerated, bombarding initially the substrate and then the coating itself during its growth. Based on this method, industrial equipment with a deposition chamber of Φ800 mm × 750 mm and 24 magnetrons was designed, manufactured and commissioned. The coating productivity is about 1 m 2/week. Tungsten coatings with a thickness of up to 17 μm and multilayer structures Mo/W/Mo/W with a thickness of ∼25 μm were produced and successfully tested at 100 pulses of 16.5 MW/m 2 for 1.5 s.

Evaluation of the erosion on the CFC tiles of the ITER divertor by means of FE calculations

In order to evaluate by finite element (FE) calculations the possibility of erosion instability on the CFC tiles of the vertical target of the ITER divertor an original methodology has been developed using the CAST3M CEA finite element code. The used approach provides a series of steady state calculation steps, the mesh being updated at each step of the iteration in order to take into account the erosion between two steps. Both physical and chemical sputtering together with sublimation have been included in the code to describe the loss of material by the thermal and particle loads envisaged for ITER normal operation regime. This model has been validated by comparison with analytical or other code results and then applied to the ITER divertor vertical target (CFC monoblock geometry). The calculations have pointed out the possibility of some erosion instabilities when a monoblock has a neighbour one with reduced conductivity or with interface defects. However it was shown that when applying normal condition loads (10 MW/m 2) for 10,000 s and then off-normal loads (20 MW/m 2) for 10 s the total erosion was 1 mm. Based on these first results no modification of the monoblock acceptance criteria is necessary. This way of calculation can be easily transposed in other FE codes and extensively used for checking various ITER operative scenarios.

Detritiation of JET Tiles by Laser Cleaning

The retention of tritium (T) by carbon based deposits on tokamak surfaces is of increasing concern to the fusion community as the scale of tritium retention by this mechanism could be a limiting factor for the operation of fusion reactors, such as ITER. Hence there is a need to investigate ways of mitigating T retention and also for detritiating surfaces by either desorption of T or removal of tritiated deposits. The results of the removal of codeposits from CFC tiles by pulsed laser ablation are reported here. The results show that it is possible to completely remove a 300μm thick hydrogen isotope rich carbon film at a rate of 12x10-3m2/hr by this method and that with optimisation of the laser parameters there is scope to improve the treatment rates to provide a useful tool for managing T inventory in tokamaks.

Flexible small size radiofrequency plasma torch for Tokamak wall cleaning

A plasma torch aiming to assist the detritiation process of Tokamak tiles in gaps and zones with difficult access, like in the divertor regions, was designed and built up. The torch operation is based on the expansion of a radiofrequency discharge constrained to burn in a limited space. The plasma torch is small size, flexible and compatible with mounting on a robotic arm, and can be operated stable both in argon and nitrogen at low and atmospheric pressure. The capability of the source in cleaning surfaces was proved by experiments on material removal from graphite and CFC tiles.

Pre-brazed casting and hot radial pressing: A reliable process for the manufacturing of CFC and W monoblock mock-ups

ENEA is involved in the European International Thermonuclear Experimental Reactor (ITER) R&D activities and, in particular, for the manufacturing of high heat flux plasma-facing components (HHFC), such as the divertor targets, the baffles and the limiters. During last years, ENEA has manufactured actively cooled mock-ups by using different technologies, namely brazing, diffusion bonding and hot isostatic pressing (HIPping). A new manufacturing process has been set up and tested. It was successfully applied for the manufacturing of W armoured monoblock mock-ups. This technique is the HRP (hot radial pressing) based on performing a radial diffusion bonding between the cooling tube and the armour tile by pressurizing only internal tube and by keeping the joining zone in vacuum at the required bonding temperature. The heating is obtained by a standard air furnace. The HRP technique is now used for the manufacturing of CFC armoured monoblock components. For this purpose, some issues have to be faced, like the low CFC tensile strength, the pure copper interlayer between the heat sink and the armour necessary to mitigate the stress at the joint interface, and the low wettability of the pure copper on the CFC matrix. This paper reports the research path followed to manufacture a medium scale vertical target CFC and W armoured mock-up by HRP. A casting of a soft copper interlayer between the tube and the tile was obtained by a new technique: the pre-brazed casting (PBC, ENEA patent). Some preliminary mock-ups with three NB31 CFC tiles were successfully manufactured and tested to thermal fatigue using electron beam facilities. They all reached at least 1000 cycles at 20 MW/m 2 without suffering any damage. The manufactured medium scale vertical target mock-up is now under testing at the FE2000 (France) facility. These activities were performed in the frame of ITER-EFDA contracts.

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.

Results of the examinations of the W7-X pre-series target elements

The target elements of the WENDELSTEIN 7-X (W7-X) divertor are designed to sustain a stationary heat flux of 10 MW/m 2 and to remove a maximum power of 100 kW. CFC Sepcarb ® NB31 tiles are bonded to a water-cooled CuCrZr heat sink in two steps: active metal casting (AMC ®) of an AMC ®-copper interlayer to CFC tiles, electron beam welding (EBW) or hot isostatic pressing (HIP) of the AMC ®-NB31 tiles to CuCrZr. The fabrication of the whole amount of CFC NB31 has been completed. The key target of the pre-series phase is the qualification of this bond based on a series of examinations. The introduction of silicon during the AMC ® process significantly improved the strength of the joint between CFC and AMC ®-copper. The strength of the bond is preserved after either EBW or HIP processes. High heat flux testing carried out in the ion beam facility GLADIS exhibited a too high percentage of defective tiles. Pre-series activities have been extended to reduce the stress concentration at the interface between tiles and heat sink by optimizing the design.

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.

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.

Tungsten and beryllium armour development for the JET ITER-like wall project

For the ITER-like wall project at JET the present main chamber CFC tiles will be exchanged with Be tiles and in parallel a fully tungsten-clad divertor will be prepared. Therefore three R&D programmes were initiated: Be coatings on Inconel as well as Be erosion markers were developed for the first wall of the main chamber. High heat flux screening and cyclic loading tests carried out on the Be coatings on Inconel showed excellent performance, above the required power and energy density. For the divertor a conceptual design for a bulk W horizontal target plate was investigated, with the emphasis on minimizing electromagnetic forces. The design consisted of stacks of W lamellae of 6 mm width that were insulated in the toroidal direction. High heat flux tests of a test module were performed with an electron beam at an absorbed power density up to 9 MW m−2 for more than 150 pulses and finally with increasing power loads leading to surface temperatures in excess of 3000 °C. No macroscopic failure occurred during the test while SEM showed the development of micro-cracks on the loaded surface. For all other divertor parts R&D was performed to provide the technology to coat the 2-directional CFC material used at JET with thin tungsten coatings. The W-coated CFC tiles were subjected to heat loads with power densities ranging up to 23.5 MW m−2 and exposed to cyclic heat loading for 200 pulses at 10.5 MW m−2. All coatings developed cracks perpendicular to the CFC fibres due to the stronger contraction of the coating upon cool-down after the heat pulses.