[en] After definition of the constraints for superconductors to be used in Tokamak magnets the present state of art in conductor technology is compared with different conductor designs with these constraints. Analyzed are especially AC-loss performance and mechanical strength. Based on the results a conductor design is discussed. A first step for the industrial development of such a conductor is in progress with the fabrication of a conductor which will be used in a small torus assembly for technology tests

[en] The megawatt neutral injector being designed at Culham Laboratory will require pumps capable of handling pulsed hydrogen gas loads of more than 20 litre Torr sec^-1. Pressures of <10^-4 Torr need to be maintained in the presence of this gas load implying installed pumping speeds > 2 x 10⁵ litres sec^-1. Some fundamental concepts pertaining to cryopumps operating at relatively high pressures and some of the design parameters which need to be taken into consideration, are examined

[en] The conceptual design of a Tokamak fusion power reactor, UWMAK-II with a special emphasis on the superconducting magnet designs is described. The reactor is designed to generate 5000 MW(th) during the plasma burn and to deliver 1716 MWsub(e) continuously. The structural material is 316 stainless steel and the primary coolant is helium. UWMAK-II is a low aspect ratio, low magnetic field design and includes a double null, axisymmetric poloidal field divertor for impurity control. In addition, a carbon curtain, made of two-dimensional woven carbon fiber, is mounted on the first vacuum chamber wall to protect the plasma from high Z impurities and to protect the first wall from erosion by charged particle bombardment. The blanket, which is designed to minimize the inventory of both tritium and lithium, utilizes a solid breeding material (LiAlO₂) with beryllium as a neutron multiplier. The total energy per fusion is 21.56 MeV, which is fairly high. The UWMAK-II toroidal field (TF) magnets are a set of 24 'extended D' superconducting coils of TiNb in Cu with stainless steel structure. The shield can be opened and a blanket module removed between coils without removing the TF coils. The vertical field (VF) coils have been deliberately placed inside the TF coils to minimize the energy stored in the poloidal magnetic field. The design philosophy for the VF coils is crucial when they are placed inside the TF set and this is discussed in detail. Toroidal coil design is discussed as based on a simple analytic solution for a constant tension 'D' sector. Magnet protection and safety circuits are described

[en] Thermonuclear reactors impose unique vacuum pumping problems involving very high pumping speeds, the handling of hazardous materials (tritium), extreme cleanliness requirements, and quantitative recovery of pumped materials. Two principal pumping systems are required for a fusion reactor, a main vacuum system for evacuating the torus and a vacuum system for removing unaccelerated deuterium from neutral beam injectors. The first system must pump hydrogen isotopes and helium while the neutral beam system can operate by pumping only hydrogen isotopes (perhaps only deuterium). The most promising pumping techniques for both systems appear to be cryopumps, but different cryopumping techniques can be considered for each system. The main vacuum system will have to include cryosorption pumps cooled to 4.2⁰K to pump helium, but the unburned deuterium-tritium and other impurities could be pumped with cryocondensation panels (4.2⁰K) or cryosorption panels at higher temperatures. Since pumping speeds will be limited by conductance through the ducts and thermal shields, the pumping performance for both systems will be similar, and other factors such as refrigeration costs are likely to determine the choice. The vacuum pumping system for neutral beam injectors probably will not need to pump helium, and either condensation of higher-temperature sorption pumps can be used

[en] The axisymmetric divertor field of the ASDEX Tokamak is induced by two triplets of multipole coils situated within the vacuum vessel above and below the plasma region. One coil of a triplet is wound with eight turns and two coils with four turns. The installation of the multipole coils requires, that the coils must be divided into two halves connected by demountable joints. The windings isolated with structural reinforced resin and placed in vacuum-tight stainless steel bellows are fixed to the vacuum vessel by 16 holdings. The multipole coils are series-connected with the toroidal field coils and powered by a flywheel generator. In order not to disturb the plasma build up, the multipole field is largely compensated in the plasma region by compensating coils situated outside the vacuum vessel. At reduced power it is also possible to work with time-dependent divertor fields by means of a thyristor convector. The maximum current in one conductor of the coils is 45 kA with a flat-top time of 5 s. The maximum tensile stresses in the joints are nearly 27 kp/mm²

[en] Desorption spectra of the Mo-N₂ and Mo-H₂ systems show a single desorption peak with first-order desorption kinetics at low temperatures and a group of desorption peaks with second-order desorption kinetics at higher temperatures. Results of surface coverage and sticking probability measurements are presented and discussed

[en] Prior to the design of large GJ toroidal magnet systems it is appropriate to procure small scale models, which can simulate their pertinent properties and allow to investigate their relevant phenomena. The important feature of the model is to show under which circumstances the system performance can be extrapolated to large magnets. Based on parameters such as the maximum magnetic field and the current density, the maximum tolerable magneto-mechanical stresses, a simple method of designing model magnets is presented. It is shown how pertinent design parameters are changed when the toroidal dimensions are altered. In addition some conductor cost estimations are given based on reactor power output and wall loading

[en] Film cooling stability is described for a composite conductor of superconductor filaments in a normal metal matrix. It is shown by an electromagnetic and thermal transient solution for current density, temperature and power generated that the following occurs after a filament has gone normal: the temperature rise ΔT approximately 50 K for 0.04 cm radius filaments and is less for smaller filaments, the current spread and temperature rise is fast, (approximately milliseconds), and the recovery to T=Tsub(c) is relatively slow, (approximately seconds). Recovery is assured if the heat flux generated at any temperature is less than the boiling heat flux in helium, in particular less than 0.15 W/cm², the minimum film boiling heat flux at 9 K

[en] Some aspects of the circuit which produces the plasma (3.8-4.8MA) and keeps it in equilibrium are described. The flux of JET transformer is driven almost equally in both directions -18...+16Vs. The main power source for the 'basic' performance (3.8MA plasma) is a generator-rectifier set rated at 70kA, 4.5kV D.C. and 600MJ energy. For the extended performance (4.8MA plasma) an additional thyristor bridge (80kA, 1.5kV, 800MJ) is used fed by the network. Two circuit breakers each rated at 24kV, 70kA will produce at the beginning of the pulse a transformer voltage of 140V/turn which raises the plasma fast, 10-20MA/s. The circuit can also produce fast rising fields necessary to compress a plasma (1MA) from 3.43 to 2.34m major radius in a time of 25-50ms. The power needed is some 400-500MW and is delivered by two inductors with 2x12.5MJ stored energy, rated at 12.5-25kA and 2x20kV. All active coils are connected in parallel and in addition passive coils may be used to give a 'shell effect' i.e. produce forces which counteract to quick movements of the plasma. Four fully controlled thyristor bridges each 13kA, 1kV are used to control currents and voltages in the coils and provide stability of the plasma position

[en] The problems of optimizing the location of the coil systems, stabilising shell and liner and the field error produced at the insulated gaps for a Reverse Field Pinch Experiment are discussed, taking account of plasma equilibrium conditions and field diffusion into the shell. The main constraints on the design are the need for field line concentricity with the liner, minimal diffusion effects and minimum field error at the insulated gaps