H. Bolt

Dust characterization and analysis in Tore-Supra

Dust produced by the last 1000 shots in Tore-Supra has been characterized. The measured mean equivalent diameter is 2.7 μm, comparable to other tokamaks. The dust surface density varies from 2000 mg/m2 on the bottom of the vessel to 15 mg/mm2 on the remaining surfaces. The carbon dust sampling in Tore-Supra revealed some nano-structures which have been identified as fullerenes.

Non-metallic materials for plasma-facing first wall components: Status and further development

During the last few years, systematic investigations of the response of the plasma-facing material to the plasma-wall interaction have been carried out. The goal of this work is to find materials which, on the one hand, withstand the plasma attacks, and, on the other hand, do not contaminate the plasma. The available results allow first systematic conclusions on the status of materials development and on the way that has still to be covered to reach acceptable solutions for this problem area. In this paper, the available test facilities are presented, together with the target values on which the material tests are aiming. The first systematic test results are discussed for non-metallic materials and the conclusions that can be drawn on the status of the already reached position. This is done on the basis of the materials work performed at the Institute for Reactor Materials in cooperation with the Institute for Plasma Physics of the Nuclear Research Center, Julich.

Simulation of tokamak runaway-electron events

High energy runaway-electron events, which can occur in tokamaks when the plasma hits the first wall, are a critical issue for the materials selection of future devices. Runaway-electron events are simulated with an electron linear accelerator to better understand the observed runaway-electron damage to tokamak first wall materials and to consider the runaway-electron issue in further materials development and selection. The electron linear accelerator produces beam energies of 20–30 MeV at an integrated power input of up to 1.3 kW. Graphite, SiC + 2% AlN, stainless steel, molybdenum and tungsten have been tested as bulk materials. To test the reliability of actively cooled systems under runaway-electron impact, layer systems of graphite fixed to metal substrates have been tested. The irradiation resulted in damage to the metal compounds but left graphite and SiC + 2% AlN without damage. Metal substrates of graphite-metal systems for actively cooled structures suffer severe damage unless thick graphite shielding is provided.

DISRUPTION SIMULATION EXPERIMENTS IN ELECTRON AND LASER BEAM FACILITIES

To design next step thermonuclear fusion devices a profound data base is needed for materials which are considered as prime candidates for the plasma facing side of the torus. Here especially reliable data on the performance of these materials under short heat pulses with an energy deposition of several MJ/m2 are required. Up to now apart from theoretical data only very limited exerimental results are available. Both laser and electron beam facilities have recently been used to perform disruption simulation tests on graphites, pyrographite, carbon carbon composites, and boron doped carbon materials; the deposited energies were 6 and 9 MJ/m2, resp.. Emphasis was laid on the quantification of the erosion depth during a single disruption event. To get additional information on the damaging mechanism in dependence on the specific material parameters a careful characterization of the damaged surfaces was performed using different analytical methods.

Energy deposition during disruption simulation experiments in a plasma accelerator

Plasma accelerators are often proposed as test beds for disruption simulation experiments on plasma facing materials for fusion applications. The incident energy fluxes and the discharge duration are of similar order to those expected during disruptions in ITER. In this work the VIKA facility of the D.V. Efremov Institute was used for calorimetric measurements and for material tests. The calorimetric measurements showed that below the threshold for the ablation of material only a small fraction of the total power of the plasma is deposited in the target. Carbon materials which were exposed to energy depositions above the ablation threshold exhibited a nearly proportional increase of the erosion with increasing total power. From these experiments it is concluded that the shielding effect of the material surface from the incident power is mainly due to the formation of a dense cloud of recycled gas in front of the target and to a smaller extent due to the ablation of target material.

Behaviour of boron-carbide materials in TEXTOR and under electron beam irradiation

Abstract Boron-carbon materials have been exposed to high heat fluxes in the TEXTOR tokamak and in the JUDITH electron beam facility. The materials were a B 4 C Coat Mix bulk material, a conversion-type B 4 C coating, and a plasma-sprayed B 4 C coating on graphite EK98. The boron-carbide Coat Mix limiter showed, due to the low thermal conductivity of 11 W/m K, surface melting already at a heat flux of about 1.5 kW/cm 2 after 0.4 s. Both in the tokamak experiments and the electron beam tests the surface temperature stayed close to the melting temperature under continued heating. The conversion coated limiter and specimens had a very irregular coating thickness and surface structure. At heat fluxes of 2.6 kW/cm 2 melting of the coating was reached after 0.5 s. The thermal behaviour and evaporation of impurities into the plasma were comparable to the Coat Mix limiter but shifted to higher heat fluxes. Both limiters fractured under the applied heat loads. The plasma sprayed coating showed early melting of the coating and agglomeration of the molten carbide to droplets on the graphite.

OPERATION EXPERIENCES WITH JT-60U PLASMA FACING COMPONENTS AND EVALUATION TESTS OF B4C-OVERLAID CFC/GRAPHITES

Erosion of carbon fiber composite divertor tiles of JT-60U has been reduced significantly by the precise alignment and insitu taper-shaping of tile edges. The divertor tiles are coated with redeposited carbon films. None of graphite first wall tiles has been broken. Evaluation tests of B4C-converted and -coated CFC/graphite have been performed from viewpoints of high heat fluxdurability, thermal shock, deuterium retention, and erosion yields. JT-60U in-pile test has also been carried out. The results exhibit satisfactory performance for the divertor plate and first wall of JT-60U.

Behaviour and structural changes of redeposited material due to plasma-wall interaction

During plasma-wall interaction in tokamaks erosion and redeposition processes substantially influence plasma parameters as well as the properties of the first wall. During disruptions or run-away-electron accidents hot spots will form on limiters or other protective installations of the first wall. In the case of metallic components beside erosion and evaporation melting processes are becoming essential: liquid metal droplets with diameters in the range of some 10μm up to some 100 μm are deposited onto the vacuum vessel and the limiters. Another important redeposition process is the formation of thin films by atomic condensation during normal operation. Influenced by high heat fluxes during successive shots these films are forced to undergo structural changes which can result in the formation of metallic agglomerates with mean diameters in the range of some microns.

Evaluation of cooling concepts and specimen geometries for high heat flux tests on neutron irradiated divertor elements

Abstract To assess the lifetime and the long term heat removal capabilities of plasma facing components in future thermonuclear fusion reactors, neutron irradiation and subsequent high heat flux (HHF) tests will be of great importance. The effect of neutron damage will be simulated in material test reactors. To investigate the behaviour of components subjected to heat loads during normal and off-normal conditions, a 60 kW electron beam teststand (Julich Divertor Test Facility in Hot Cells (JUDITH)) has been installed in a hot cell which can be operated by remote handling techniques. In this facility inertially cooled test coupons can be handled as well as small actively cooled mock-ups. The irradiation of complete divertor modules however requires greater efforts. To investigate specific problems such as neutron induced changes in the performance of the braze interface, small test coupons without active cooling capabilities may be suitable. ITER relevant temperature gradients and resulting stress fields can be established during transient heating by electron beam methods. The only reasonable alternatives to this procedure are actively cooled test specimens or small divertor modules. A special clamping mechanism for test samples with an integrated coolant channel has been developed and tested. This experimental procedure is an attractive and cost effective alternative comparing large scale tests on complete divertor modules.

Quantitative determination of plasma induced material damage in plasma facing metallic alloys — electron beam simulation experiments

Abstract Electron beam simulation experiments allow a systematic quantitative determination of thermal shock loading of plasma facing components during plasma wall interaction. Cronifer 1810 (1.4311) as a typical alloy (Fe-base) used for the first wall of todays tokamaks is tested in single shot electron beam experiments with respect to its thermal shock behaviour. Automatic quantitative image analysis is applied for the determination of the amounts of removed material, melted material (size of melt crater), and crack length in the samples. The results are presented primarily in three-dimensional diagrams to allow an empirical quantification of the relation: energy deposition time ( = pulse length)-deposited power density ( = beam power density)-resp. type of material damage.