P. Chappuis

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-Destructive Testing of Divertor Components

This task within the EU RD (2) blind non-destructive round robin test of the prototype; (3) HHF test in FE200 electron beam (EB) facility; (4) post-fatigue blind non-destructive round robin test; (5) destructive examination. The general final conclusion was that the NDT techniques can reliably detect and locate defects having dimensions well below those, which could impair the thermal fatigue lifetime.

Manufacturing and testing of a prototypical divertor vertical target for ITER

Abstract After an extensive R&D activity, a medium-scale divertor vertical target prototype has been manufactured by the EU Home Team. This component contains all the main features of the corresponding ITER divertor design and consists of two units with one cooling channel each, assembled together and having an overall length and width of about 600 and 50 mm, respectively. The upper part of the prototype has a tungsten macro-brush armour, whereas the lower part is covered by CFC monoblocks. A number of joining techniques were required to manufacture this component as well as an appreciable effort in the development of suitable non-destructive testing methods. The component was high heat flux tested in FE200 electron beam facility at Le Creusot, France. It endured 100 cycles at 5 MW/m 2 , 1000 cycles at 10 MW/m 2 and more then 1000 cycles at 15–20 MW/m 2 . The final critical heat flux test reached a value in excess of 30 MW/m 2 .

Design, fabrication and testing of an improved high heat flux element, experience feedback on steady state plasma facing components in Tore Supra

Abstract Actively cooled plasma facing components (PFC) have been developed and used in Tore Supra since 1985. One of the main technological problem is due to the expansion mismatch between graphite armour and metallic heat sink material. A first technology used graphite tiles with or without a reinforcement and a compliant layer, brazed with titanium copper–silver (TiCuAg) alloy. The next technology used carbon fiber material (CFC) tiles with a 2 mm pure copper compliant layer, since the good mechanical strength of the CFC allowed the reinforcement layer to be suppressed. No destructive inspection during the manufacturing procedure was found to be essential to insure a good reliability of the elements. A recent technology was developed for the new actively cooled toroidal pump limiter of Tore Supra (designed to remove up to 15 MW during 1000 s with incident heat flux up to 10 MW m −2 ). This new technology uses an active metal casting (AMC) of copper onto CFC tiles whose surface is prepared by a laser treatment. Copper compliant layer is then electron-beam welded to the heat sink. This technology is silver free and could be ITER relevant. The first bonding is X-ray controlled and the second one ultrasonic (US) tested. Finally all the elements are controlled by an infrared imaging technique during a hot water test.

TORE SUPRA experience of copper chromium zirconium electron beam welding

The specification of TORE SUPRA to perform quasi steady state plasma operation has induced several R&D studies on the actively cooled structures of the plasma facing components and the associated assembling processes of the materials. Various industrial copper alloys have been characterized and tested to select the most optimized grade for EB welding. Welding samples of representative geometry were also analysed.

In-service experience feedback of the Tore Supra actively cooled inner first wall

Abstract Over 12 000 plasma shots (some of them with up to 8 MW of additional power and some as long as 60 s) have been achieved in Tore Supra, with a significant number of them limited by the inner first wall. This actively water-cooled wall is covered with brazed graphite tiles. High power, high energy experiments have shown that reliability of the graphite tile-heat sink joint and accurate alignment of the wall are needed. This paper summarizes the experience gained with this component, and the developments in progress to improve the performance of such an inner first wall.

Non destructive testing of actively cooled plasma facing components by means of thermal transient excitation and infrared imaging

SATIR is a new test-bed installed at Tore Supra to perform non destructive examination of actively cooled plasma facing components. Hot and cold water flow successively in the cooling tube of the component and the surface temperature is recorded with an infrared camera. Defects are detected by a slower temperature response above unbrazed areas. The connection between temperature differences and defect sizes is the main difficulty. It is established by tests of standard defects and thermal transient calculations of defective geometries. SATIR has been in use for two years and has proved to be very valuable to test industrial components as well as prototypes.

TORE SUPRA first-wall description and behaviour during the initial plasma experiments

Abstract TORE SUPRA is a large-size, high-field superconducting tokamak ( B φ = 4.5 T , R = 2.37 m , a = 0.8 m ). It has been completed in the beginning of 1988 and is starting its experimental operation. The plasma-wall interaction programme will be mainly devoted to obtaining very long pulses in the next future, i.e. the control of impurity content, power and particule exhaust. The design of the first-wall items takes this into account. The inner wall is made of graphite tiles brazed on stainless steel tubes, the outer wall of stainless steel panels. Moreover, three graphite limiters and six ergodic divertor modules have already been installed. Each of these elements is water-cooled ( T water = 150–230 °C, p = 3.6 Mpa ). Vessel pumping is achieved by means of three turbomolecular pumps providing 9000 1/s H2 pumping speed. A 250 °C baking capability of the vessel is also available. First operations were mainly devoted to conditioning and carbonizing the whole plasma chamber, with RG discharges, allowing to work in homogeneous conditions. The first tokamak discharges are studied for a better knowledge of plasma control and stability in this configuration. Then, power deposition will be particularly investigated so as to prepare for the next steps: additional power and pump limiter operation.

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

Abstract 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.

Steady-state heat and particle removal with the actively cooled Phase III outboard pump limiter in Tore Supra

Abstract Tore Supras Phase III outboard pump limiter (OPL) is a modular actively-cooled mid-plane limiter, designed for heat and particle removal during long pulse operation. During its initial operation in 1993, the OPL successfully removed about 1 MW of power during ohmicly heated shots of up to 10 s duration and reached (steady state) thermal equilibrium. The particle pumping of the Phase III OPL was found to be about 50% greater than the Phase II OPL which had a radial distance between the last closed flux surface and the entrace of the pumping throat of 3.5 cm compared with only 2.5 cm for the Phase III OPL. This paper gives examples of power distribution over the limiter from IR measurements of surface temperature and from extensively calorimetry (34 thermocouples and 10 flow meters) and compares the distributions with values predicted by a 3D model (HF3D) with a detailed magnetic configuration (e.g., includes field ripple).