Carbon Fiber Composite Tiles Research Articles

Acceptance criteria for the ITER divertor vertical target

In the frame of the toroidal pump limiter fabrication for Tore Supra, CEA developed a large experience of infrared test for acceptance of high heat flux components armoured with carbon fibre composite flat tiles. The test is based on a thermal transient induced by an alternative hot/cold water flow in the heat sink structure. The tile surface temperature transients are compared with those of a reference element, the maximum difference for each tile leading to a value called Δ T ref_max. This method is proposed for the commissioning of plasma facing components for the ITER divertor vertical target. This paper describes the determination of the best acceptance criteria for the “monoblock” geometry of the carbon part. First, it has been shown that the location and the extension of the defects could reliably be determined by monitoring both top and lateral surfaces during the test. Second, it was possible to fix an acceptance method based on Δ T ref_max. Samples with calibrated defects are now under fabrication to validate the results.

Electrostatic dust detection on remote surfaces

The inventory of dust in next-step magnetic fusion devices will be regulated for safety reasons, however diagnostics to measure in-vessel dust are still in their infancy. Advances in dust particle detection on remote surfaces are reported. Grids of interlocking circuit traces with spacing in the range 125–25 μm were biased to 30 V. Impinging dust creates a short circuit and the resulting current pulse was recorded. The detector response was measured with particles scraped from a carbon fiber composite tile and sorted by size category. The finest 25 μm grid showed a sensitivity more than an order of magnitude higher than the 125 μm grid. The response to the finest particle categories (5–30 μm) was two orders of magnitude higher than the largest (125–250 μm) category. Longer duration current pulses were observed from the coarser particles. The results indicate a detection threshold for fine particles below 1 μg/cm 2.

Assessment of the heating technique as a possible ex situ detritiation method for carbon wall materials from fusion machines

Detritiation by heat treatment of graphite or carbon fibre composite (CFC) tiles retrieved from the Tokamak Fusion Test Reactor (TFTR) was assessed using the standard and well proved technique of full combustion combined with liquid scintillation analysis. Complete carbon tiles were heated in a oven to 500 °C in air for one hour at Princeton Plasma Physics Laboratory (PPPL) and sent to Tritium Laboratory Karlsruhe (TLK) for surface and depth profile analysis. Coring/full combustion technique performed on the treated and untreated tiles, both graphite and CFC, has shown that only the tritium held close to the surface (few tens μm) is efficiently released by baking in air while the bulk tritium is almost unaffected. Therefore, baking the tile under air, even at 500 °C, does not detritiate the bulk.

Non-destructive tritium measurements of Mk IIA divertor tile by BIXS

β-Ray-induced X-ray spectrometry was applied to determine the amounts and depth profiles of tritium captured by graphite tiles used in the PTE1 campaign and by carbon fiber composite tiles in the DTE1 campaign of JET. Although the X-ray spectra observed for both carbon tiles consisted of characteristic and bremsstrahlung X-rays, the intensities and shapes of the X-ray spectra were quite different between these tiles. The analyses of the X-ray spectra showed that the amount of tritium retained in the latter tile was on average 15 500 times greater than that in the former tile, and the tritium depth profile extended to 60 μm for the latter tile compared to 5 μm for the former tile. In addition, characteristic X-rays corresponding to Cl, Cr, Fe, Ni and Mo appeared in the latter tile, where Ni showed the highest concentration, indicating that considerable erosion and deposition of vacuum vessel materials took place in the DTE1 campaign.

Tritium depth profiles in 2D and 4D CFC tiles from JET and TFTR

Carbon fibre composite (CFC) is currently the candidate material for the vertical target tiles in the International Thermonuclear Experimental Reactor (ITER) divertor because of its superior thermomechanical properties. However, its affinity for hydrogen isotopes and their co-deposition with eroded carbon may severely limit ITER plasma operations. Recently tritium depth profiles in divertor tiles retrieved from the Joint European Torus have been obtained by the coring/full combustion technique. The results revealed that a surprisingly large fraction (up to 61%) of the retained tritium had diffused deep into the bulk of the tile, most probably between the woven sheets of the CFC. Additionally, the coring/full combustion technique has shown that only the surface tritium (few ten μm) is efficiently released by air baking while the bulk tritium is almost not affected. Baking the tile under air even at 500 °C does not detritiate the bulk.

Thermal Response of Tritiated Codeposits from JET and TFTR to Transient Heat Pulses

High heat flux interactions with plasma facing components have been studied at microscopic scales. The beam from a continuous wave neodymium laser was scanned at high speed over the surface of graphite and carbon fiber composite tiles that had been retrieved from TFTR and JET after DT plasma operations. The tiles have a surface layer of amorphous hydrogenated carbon that was codeposited during plasma operations and laser scanning has released more than 80% of the codeposited tritium. The temperature rise of the codeposit was much higher than that of the manufactured material and showed an extended time history. The peak temperature varied dramatically (e.g. 1436°C compared to >2300°C) indicating strong variations in the thermal conductivity to the substrate. A digital microscope imaged the codeposit before, during and after the interaction with the laser and revealed 100-micron scale hot spots during the interaction. The heat flux produced a pattern of beads on a mixed Be/C deposit. Heat pulse durations of order 100 ms resulted in brittle destruction and material loss from the surface, whilst a duration of ≈ 10 ms showed minimal changes to the codeposit. These results show that reliable predictions for the response of deposition areas to off-normal events such as ELMs and disruptions in next step devices need to be based on experiments with tokamak generated codeposits.

Material properties and consequences on the quality of tore supra plasma facing components

In the frame of the CIEL project, which is an upgrade of the in-vessel system of the Tore Supra tokamak, a toroidal pump limiter is under fabrication. The elementary components are made of N11 carbon fibre composite (CFC), CuCrZr alloy, OFHC copper (for compliant layer), low cobalt stainless steel and Ni (as an interlayer between CuCrZr and stainless steel). The properties of the materials were carefully analysed at the beginning of the fabrication as they can have consequences on the final quality of the actively cooled high heat flux components. However: (1) during the manufacture, an abnormal phase was detected in the stainless steel with a risk of cracks at installation welding, (2) the properties of CuCrZr material were rather scattered, at the reception, among the different batches, (3) a low percentage (3–6%) of CFC tiles were faulty suggesting the CFC physical properties scattered significantly to fall out of the margins.

Development and Operational Experiences of the JT-60U Tokamak and Power Supplies

The design of the JT-60U tokamak, the configuration of the coil power supplies, and the operational experiences gained to date are reviewed. JT-60U is a large tokamak upgraded from the original JT-60 in order to obtain high plasma current, large plasma volume, and highly elongated divertor configurations. All components inside the toroidal magnetic field coils, such as vacuum vessel, poloidal magnetic field coils, divertor, etc., were modified. Various technologies and ideas were introduced to develop these components; for example, a multi-arc double skin wall structure for the vacuum vessel and a functional poloidal magnetic field coil system with taps for obtaining various plasma configurations. Furthermore, boron-carbide coated carbon fiber composite (CFC) tiles were used as divertor tiles to reduce erosion of carbon-base tiles. Later, a semiclosed divertor with pumps, for which cryo-panels originally used for NBI units were converted, was installed in the replacement of the open divertor. These development and operational results provide data for future tokamaks. Major failures experienced in the long operational period of JT-60U, such as water leakage from the toroidal magnetic field coil, fracture of carbon tiles, and breakdown of a filter capacitor, are described. As a maintenance issue for tokamaks using deuterium fueling gas, a method for reducing radiation exposure of in-vessel workers is described.

The effect of oxygen on the release of tritium during baking of TFTR D–T tiles

A series of tests involving 10 h baking under the current ITER design conditions (240 °C with 933 Pa O 2) was performed using a cube of a carbon fiber composite tile that had been used in Tokamak Fusion Test Reactor (TFTR) during its deuterium–tritium burning operation. The removal rate of the codeposits was about 3 μm/h near the surface and 0.9 μm/h in the deeper region. Total amount of tritium released from the cube during 10 h baking was 202 MBq, while remaining tritium in the cube after baking was 403 MBq. Thus 10 h baking at 240 °C with 933 Pa O 2 removed 1/3 of tritium from the cube. After 10 h baking, the tritium concentration on the cube surface also dropped by about 1/3. In addition, some tritium was released from another cube of the tile during baking at 240 °C in pure Ar, and a rapid increase of tritium release was observed when the purging gas was shifted from pure Ar to Ar–1%O 2. When a whole TFTR tile was baked in air at 350 °C for 1 h and then at 500 °C for 1 h, the ratios of tritium released were 53 and 47%, respectively. Oxygen reacted with carbon to produce carbon monoxide during baking in air.

High heat load test of CFC divertor target plate with screw tube for JT-60 superconducting modification

A flat carbon fiber composite (CFC) tile mock-up with screw tubes, which have helical fins like a nut, was fabricated aiming at further improvement of the heat removal performance of the cost-effectively manufactured divert or target for JT-60SC (modified JT-60 as a superconducting coil tokamak). The heat removal performance of the mock-up was successfully demonstrated on the JAERI Electron Beam Irradiation Stand. The estimated heat transfer coefficient of the screw tube at the non-boiling region was roughly three times higher than that of the smooth tube. This corresponds to 1.5 times that of the swirl tube. A heat cycle test of 10 MW/m 2 showed that the mock-up with the screw tubes could withstand for 1400-cycles. These results indicate that the divertor target plate with the flat CFC tile and the screw tube can be a promising candidate for the JT-60SC divertor target.

The effect of UV light irradiation on the removal of tritium from the codeposited carbon–tritium layer in fusion reactors

A series of experiments have been performed at the Karlsruhe Tritium Laboratory to examine the effects of UV light on tritium removal from the codeposited layers on carbon fiber composite tiles removed from the JET fusion device. Codeposited layers were exposed to UV light from a mercury lamp (254 nm) for various temperatures using air and argon atmospheres. Samples were also heated externally with no UV light to compare tritium release with and without UV exposure

Experience feedback from high heat flux component manufacturing for Tore Supra

Tore Supra is involved in flat tile carbon armoured plasma facing components (PFCs) since 1985. In 1997, a third generation of components, based on the original concept developed with Plansee Company, called active metal casting (AMC®), has been launched. Since 1998, 660 elementary components for the toroidal pump limiter (TPL) are in production. The route of the manufacture is rather complex and many controls were requested all along the fabrication to insure a high reliability of the elements. One of the main controls is the final infrared (IR) test allowing to determine the quality of the bonding between the carbon fibre composite (CFC) tiles and the heat sink made of copper–chromium–zirconium alloy (CuCrZr). Although results for the first batch of elements were as expected (less than 5% rejected at the final test), unexpected defects appeared with the followings batches. Investigations on the fabrication processes underlined the importance of having a better heat treatment of the pieces in copper alloy (CuCrZr), however this was not sufficient to completely explain the observed defects.

Concepts and prototype elements of the low- Z wall protection for the stellarator W7-X

Stationary operation, envisaged at WENDELSTEIN 7-X, demands for an active cooled wall protection system. To minimise the radiation losses of the plasma during long pulse operation the inner surface of the vacuum vessel has to be covered completely with materials having low atomic numbers. Unlike from the target plates (receiving a high heat flux of up to 10 MW/m 2) that consist of Carbon Fibre Composite (CFC) brazed on a cooling structure, a more simple solution for the Plasma Facing Components (PFC) on the basis of CFC tiles clamped on a cooling structure or larger stainless steel (SS) -panels covered with B 4C will be developed for the less loaded baffle areas of the divertor and the 120 m 2 wall protection. The choice of the components to be installed in the various areas for wall protection will be determined mainly by the conditions of the physics of the plasma, additional loads by plasma heating (mainly neutral beam injection, NBI and electron cyclotron resonance heating, ECRH), demands by the diagnostics and the geometrical restrictions at different locations in the vacuum vessel.

Development of actively cooled components for the Tore Supra toroidal pump limiter

Actively cooled plasma facing components armoured with carbon fibre composite flat tiles have been developed for Tore Supra Tokamak since 1994. After fabrication of prototypes successively tested in a high heat flux test bed the special bond between the carbon tiles and the hardened copper of the heat sink was particularly studied. Since 1997, 660 elements are in fabrication. The infrared acceptance test underlines the importance of the heat treatment during the manufacture and revealed some unexplained cracked tiles.

Development and material testing of OF-Cu /DS-Cu/OF-Cu triplex tube (dispersion strengthened copper clad with oxygen free-copper) and trial fabrication of a vertical target mock-up for ITER divertor

For the divertor target of the International Thermonuclear Experimental Reactor (ITER), an OF-Cu/DS-Cu/OF-Cu triplex-structured cooling tube has been newly fabricated through powder metallurgy and drawing. The triplex structure comprised an aluminium oxide (0.5 mass%) - dispersion strengthened copper core (DS-Cu) clad with oxygen free copper (OF-Cu), a compliant layer for joining to the carbon fiber composite (CFC) tiles, and with an inner skin which tightly grasps a twisted INCONEL tape to assist heat transfer. Physical and mechanical properties of the DS-Cu core after heat treatment at 850°C for 600 s were investigated. Also, CFC brazability, fabricability and feasibility of the triplex tube for cooling channels for the divertor target were studied: A large scale vertical target mock-up of a 1500 mm length, 35 mm width and 3000 mm radius curved front face, has been fabricated with nearly 50 pieces of “saddle”-shaped one-dimensional (1D)-CFC tiles were brazed on to 1500 mm long triplex tubes set in grooves of OF-Cu heat sink blocks joined to a stainless-steel back plate. The mock-up was tested under 20 MW/m 2 for 15 s for 1000 cycle thermal loadings, which simulated transient heat loadings of a vertical target of an ITER divertor.

Evidence of damage in carbon fibre composite tiles joined to a metallic heat sink under high heat flux fatigue

The two years experience with Active Metal Casting (AMC) flat bonds shows that this technology is suitable for the heat fluxes expected in Tore Supra (10 MW/m 2). Tests were pursued up to 3330 cycles, with elements still functional. At higher heat fluxes, fatigue damage is observed, but the bond resists remarkably well with no tile detachment. Examination of such deliberately damaged bonds showed distributed cracking, proving the absence of any weak link. The limitations to those higher heat fluxes are more related to the design and the base materials than to the bond itself.

Development of a compact W-shaped pumped divertor in JT-60U

In JT-60U, the modification to a W-shaped pumped divertor will be completed in May 1997, aiming to realize sufficient reduction in heat flux to the targets and good H-mode confinement simultaneously. W-shaped geometry is optimized not only for forming radiative divertor plasmas and reducing the back flow of neutral particles but also for allowing various experimental configurations. Toroidally and poloidally segmented divertor plates, dome and baffles are arranged in a W-shaped poloidal configuration. The pumping speed can be changed during a shot by variable shutter valves in the three pumping ports under the outer baffle. The net throughput is enough for particle control in the steady radiative operations with high power NBI heating. Carbon fiber composite (CFC) tiles are used for the divertor targets and the divertor throat where large heat flux is expected. Gaps between two adjacent segments are carefully sealed to suppress the leak of neutral gas from the exhaust duct below the divertor and baffles. The strength of the whole structure is confirmed by an electromagnetic force analysis and structural analysis carried out for disruptions of 3 MA discharges with a halo current.

Investigation of plasma facing components in JT-60U operation

The mechanical fracture of three carbon fiber composite (CFC) first wall tiles was observed. This damage was probably caused by the electromagnetic force due to halo current during disruption. The required current to break the CFC tile is estimated to be 25 kA. The broken tile was rotated poloidally around the plasma with a speed of about 10 m/s during the following discharge. A possible driving force of this rotation might be the electromagnetic force due to the scrape-off layer (SOL) current. The required current to rotate the piece of the broken tile is 1 kA. These results indicate that electromagnetic interaction between SOL plasma and the plasma facing component is important in the research on the plasma wall interactions in fusion devices.

Brazing of the Tore Supra actively-cooled Phase III limiter

The head of the water-cooled Tore Supra Phase III Limiter is a bank of 14 round OFHC copper tubes, curved to fit the plasma radius, onto which several hundred pyrolytic graphite (PG) tiles and a lesser number of carbon fiber composite tiles are brazed. The small allowable tolerances for fitting the tiles to the tubes and mating of compound curvatures made the brazing and fabrication extremely challenging. The paper describes the fabrication process with emphasis on the procedure for brazing. In the fixturing for vacuum furnace brazing, the tiles were each independently clamped to the tube with an elaborate set of “window frame” clamps. Braze quality was evaluated with transient heating tests. Some rebrazing was necessary.

Thermo-mechanical tests of a CFC divertor mock-up

Thermo-mechanical tests have been performed on a divertor mock-up consisting of a metallic tube armoured with five carbon fibre composite tiles. The tube is inserted inside the tiles and brazed with TiCuSil braze (monoblock concept). The tube material is TZM, a molybdenum alloy, and the armour material is SEP CARB N112, a high conductivity carbon-carbon composite. Using special surface preparation consisting of laser drilling, small (∼- 500 μm) holes in the composite have been made to increase the surface wetted by the braze and the resistance. The mock-up has been tested at the JAERI 400 kW electron beam test facility JEBIS. The aim of the test was to assess the performance of the mock-up in screening and thermal fatigue tests with particular attention to the behaviour of the armour to heat sink joint.