[en] The remote application of photogrammetry developed at JET proved that there has been no unacceptable change in shape or position of the divertor structure, therefore confirming its integrity. The flexibility of this technique was demonstrated when it was used for an unplanned, high accuracy, remotely targeted survey of the damaged KG6 outer waveguide. The ability to carry out an engineering analysis of large structures, using data captured remotely, is essential for the construction, operation and maintenance of fusion machines. This ability, using commercially available survey equipment, has been successfully demonstrated at JET. (authors)

[en] The superconducting magnet system of the WENDELSTEIN 7-X (W7-X) experiment consists of 50 non-planar and 20 planar coils which are connected by 121 bus bars in series of seven groups of ten coils each. The connection of the bus bars will be provided by 184 joints each with a maximum allowable resistance of 5 nΩ. To allow for possibly replacements and repair after installation all joints have to be designed as demountable joints. The principle of such a demountable joint was tested by two joints in SULTAN Test facility CRPP, Switzerland in 2002. Both joints showed resistances less than 1 nΩ at 20 kA in a magnetic field of 2 T. In the meantime efforts to improve the design of the casing and the design of the clamping parts have been carried out. Tools for the installation at W7-X and mounting procedures were developed using the test joints to ensure the best possible reproducibility. These improvements led to the requirement to repeat the measurement of joint resistance. This joint test was carried out in the test facility of SINTEZ of the Efremov Institute in St. Petersburg, Russia. This paper describes the design of the joint and the test facility and focus on the comparison of test results at SULTAN and SINTEZ. (author)

[en] The Poloidal Field (PF) coils of ITER (International Thermonuclear Experimental Reactor) will supply the necessary magnetic field to initiate, shape, control and shutdown burning plasmas. The PF coils use NbTi cable-in-conduit superconductors, which operate at maximum currents of the order of 45-60 kA and experience large variations of current and magnetic fields. In order to test full-scale NbTi superconductors at operational conditions similar to ITER, the European Team has been asked to design and manufacture a PF conductor insert (PFCI). The cable has been provided by the Russian Federation. The Insert will be tested in 2004 in the Central Solenoid Model Coil (CSMC) facility at JAERI Naka, Japan

[en] During the Remote Tile Exchange shutdown at JET, the purpose built Tile Carrier Transfer Facility (TCTF) has been successfully used for the remote removal and storage of activated, tritiated and beryllium contaminated torus components. The short boom, end effector and tine arrangement was also used during the installation of the new Gas Box Divertor. Tritium levels required the use of techniques and practices which were successful in confining contamination and allowed the declassification of work areas. A holding area and posting facilities enabled ancillary equipment / tool logistics to be managed efficiently. This article presents and describes all the equipment used and reports the operational experience. (authors)

[en] In the scope of the ITER EDA a Toroidal Field Model Coil (TFMC) has been manufactured accompanied by a thorough Quality Assurance (QA) test program. This large superconducting coil has been conceptually designed by the ITER European Home Team (EUHT) and manufactured by European industry. The coil is being completed and will be tested at the Forschungszentrum Karlsruhe in spring 2001. The race track shaped winding is made of a cable-in-conduit conductor in a circular 316LN stainless steel jacket. From this conductor five double pancake (DP) modules were fabricated. Results of conductor and DP manufacture were already presented at previous conferences and are therefore only summarized here. The paper concentrates on the subsequent manufacturing steps, namely the stacking of the DP modules, the insulation and impregnation of the winding pack, the outer joint manufacture by electron beam welding, the assembly of the winding pack with the stainless steel case, the mounting of the helium pipes, the sensors and the busbars. To assemble the coil into the TOSKA facility and to fit it to the EU-LCT coil a heavy Inter-Coil Structure (ICS) has been built, in which the TFMC will rest on four wedges

[en] The WENDELSTEIN 7-X (W7-X) stellarator, presently under construction in Greifswald, Germany, will be a 'fully-optimized' stellarator device with the aim to demonstrate the reactor potential of the HELIAS line at steady operation close to fusion relevant parameters. The in-vessel components of W7-X are designed for steady state operation with high heat flux divertor target plates designed to withstand 10 MW/m² power loading. The plasma vessel and the ports are further protected by a series of water cooled graphite tiles clamped on CuCrZr cooling structure and of stainless steel panels. Behind the divertor components are the cryo vacuum pumps and the sweep/control coils. For the first operation phase in 2014, an inertial cooled divertor of the same geometry as the high heat flux divertor will be installed as well as all in-vessel components except cryo vacuum pumps. Mostly, the components will be operated without water-cooling since during this phase the power will be restricted to 8 MW for 10s and 1 MW for 50s. This paper describes the selected technical solutions and the present status of the various in-vessel components of W7- X with a focus on the high heat flux divertor. (author)

[en] A short sample of the NbTi cable-in-conduit conductor (CICC) manufactured for the ITER PF insert coil has been tested in the SULTAN facility at CRPP. The short sample consists of two paired conductor sections, identical except for the sub-cable and outer wraps, which have been removed from one of the sections before jacketing. The test program for conductor and joint includes DC performance, cyclic load and AC loss, with a large number of voltage taps and Hall sensors for current distribution. At high operating current, the DC behavior is well below expectations, with temperature margin lower than specified in the ITER design criteria. The conductor without wraps has higher tolerance to current unbalance. The joint resistance is by far higher than targeted. (authors)

[en] A new divertor (MKII-HP) has been designed to be implemented in JET as part of a possible enhancement programme of the JET facility (JET EP). The aim is to handle up to 40 MW of injected power for 10 s with plasma triangularities up to 0.5 while keeping enough flexibility for other scenarios. The divertor is shaped to optimise the wetting fraction without exposing sharp edges or metallic parts and the general design allows for high halo currents

[en] The Poloidal Field Conductor Insert (PFCI) of the International Thermonuclear Experimental Reactor (ITER) has been designed in Europe and is being manufactured at Tesla Engineering, UK, in the frame of a Task Agreement with the ITER International Team. Completion of the PFCI is expected at the beginning of 2005. Then, the coil shall be shipped to JAERI Naka, Japan, and inserted into the bore of the ITER Central Solenoid Model Coil, where it should be tested in 2005 to 2006. The PFCI consists of a NbTi dual-channel conductor, almost identical to the ITER PF1 and PF6 design, about 45 m long, with a 50 mm thick square stainless steel jacket, wound in a single-layer solenoid. It should carry up to 50 kA in a field of about 6 T, and it will be cooled by supercritical He at around 4.5 K and 0.6 MPa. An intermediate joint, representative of the ITER PF joints and located at relatively high field, will be an important new item in the test configuration with respect to the previous ITER Insert Coils. The PFCI will be fully instrumented with inductive and resistive heaters, as well as with voltage taps, Hall probes, pick-up coils, temperature sensors, pressure taps, strain and displacement sensors. The test program shall be aimed at DC and pulsed performance assessment of conductor and intermediate joint, AC loss measurement, stability and quench propagation, thermalhydraulic characterization. Here we give an overview of the preparatory work towards the test, including a review of the coil manufacturing and of the available instrumentation, a discussion of the most likely test program items, and a presentation of the supporting modeling and characterization work performed so far. (authors)

[en] The in-vessel components of the WENDELSTEIN 7-X stellarator consist of the divertor components and the wall protection with its internal cooling supply. The main components of the open divertor are the vertical and horizontal target plates which form the pumping gap, the cryo-vacuum pumps and the control coils. The divertor volume is closed by graphite shielded baffle modules and with divertor closures. All these components are designed to be actively water-cooled. For the first commissioning phase planned in 2014, an inertial-cooled test divertor will be installed instead of the actively water-cooled high heat flux divertor. The wall protection consists of graphite-protected heat shields in the higher loaded areas and stainless steel panels in the lower loaded regions. The wall protection cooling circuits are connected through 80 supply-ports via so-called 'plug-ins'. It is envisaged to protect the diagnostic ports by panel-type port-liners. Special graphite-shielded port liners are used on the diagnostic injector and the neutral beam injector ports. The in-vessel components are mainly manufactured and tested at the Max-Planck-Institute fuer Plasmaphysik in its Garching workshop. Panels, high heat flux target elements and control coils are delivered by industrial partners. Manufacturing of the KiP ('Komponenten im Plasmagefaess') is in plan. Delivery of the components will be in time.