[en] A recently completed design of a Compact Reversed-Field Pinch Reactor based on pumped-limiter impurity control is used to estimate for the first time the impact of magnetic divertors. A range of divertor options for the low-toroidal-field RFP is examined, and a design selection is made constrained by considerations of field ripple (magnetic islands), blanket displacement, recirculating power, cost, heat flux, and access. Design choices based on diversion of minority (toroidal) field lead to a preference for (poloidally) symmetric or bundle divertor geometries. Coupling of magnetics and scrapeoff models with the above-mentioned constraints yields design points for both divertor approaches. (author)

[en] The sign of the toroidal magnetic field has a major effect on the divertor in Alcator C-Mod, determining whether the higher recycling is on the inboard or outboard side and leading to inboard/outboard temperature and density differences of up to a factor of ten. A recently published paper (Hutchinson et al 1995 Plasma Phys. Control. Fusion 37 1389) reported in detail on these observations, which are an indication of the importance of plasma particle drifts for understanding and modelling the divertor. The conference presentation covered these published results and also some additional results, documented here, in which the plasma flows associated with the field-direction-dependent asymmetries are measured using a Mach probe and a localized impurity puffing plume analysis technique. (author)

[en] The emergence of magnetic divertors as an impurity control and ash removal mechanism for future tokamak reactors bring on the need for further experimental verification of the divertor merits and their ability to operate at reactor relevant conditions, such as with auxiliary heating. This paper presents preliminary designs of a bundle and a poloidal divertor for Versator II, which can operate in conjunction with the existing 150 kW of LHRF heating or LH current drive. The bundle divertor option also features a new divertor configuration which should improve the engineering and physics results of the DITE experiment. Further design optimization in both physics and engineering designs are currently under way

[en] The present object of the fusion research is to accomplish the scientific break even by the year of 1986. In view of current progress in the field of Fusion reactor development, we decided to carry out the conceptual design of Tokamak-type fusion reactor during the year of 82-86 in order to acquire the principles of the fusion devices, find the engineering problems and establish the basic capabilities to develop the key techniques with originality. In this year the methods for calculating the locations of the poloidal coils and distribution of the magnetic field, which is one of the most essential and complicated task in the fusion reactor design works, were established. Study on the optimization of the design method of toroidal field coil was also done. Through this work, we established the logic for the design of the toroidal field coil in tokamak and utilize this technique to the design of small compact tokamak. Apart from the development work as to the design technology of tokamak, accelerating column and high voltage power supply (200 KVDC, 100 mA) for intense D-T neutron generator were constructed and now beam transport systems are under construction. This device will be used to develop the materials and the components for the tokamak fusion reactor. (Author)

[en] The Tokamak Physics Experiment (TPX) will be the first Tokamak using superconducting magnets for both the poloidal and toroidal field. It is designed for advanced Tokamak physics experiments in steady-state and long-pulse operation. The TPX superconducting magnets use an advanced cable-in-conduit conductor (CICC) design similar to that developed in support of the International Thermonuclear Experimental Reactor (ITER). The toroidal field magnets provide 4.0 T at 2.25 m with a stored energy of 1.05 GJ. The poloidal field magnets provide 18.0 V-s to ohmically start and control long burns of a 2.0 MA plasma

[en] Remote replacement of the ITER divertor will be required several times during the life of ITER. To facilitate its regular exchange, the divertor is assembled in the ITER vacuum vessel from 60 cassettes. Radial movers transport each cassette along radial rails through the handling ports and into the vessel where a toroidal mover lifts and transports the cassette around a pair of toroidal rails. Once at its final position the cassette is locked to the toroidal rails and is accurately aligned in both poloidal and toroidal directions. A further requirement on the divertor is to minimise the amount of activated waste to be sent to a repository. To this end the cassettes have been designed to allow the remote replacement, in a hot cell, of their plasma facing components. The paper describes the two facilities built at ENEA Brasimone, Italy, whose aim is to demonstrate the reliable remote maintenance of the divertor cassettes. (author)

[en] In a spherical tokamak, power handling at the divertor region becomes more difficult than that of conventional tokamaks because of the narrow space at the divertor region. Here a new poloidal-bundle divertor is proposed for a spherical tokamak reactor. The bundle divertor is located at the bottom and/or top region of the torus, in addition to several poloidal guide coils. These coils are installed inside of the toroidal coil, because the copper conductor is adaptable in spherical tokamak devices. The magnetic field line outside the separatrix is guided to the lower region of the torus by the poloidal guide coils, and extracted outside of the torus through the adjacent toroidal coils by the bundle divertor coils. The bundle divertor configuration does not affect the shape of the core plasma as much, and the magnetic field ripple, due to these bundle coils, is negligible

[en] The Superconducting Tokamak has different superconducting magnet systems, such as The Toroidal Field (TF), Poloidal field (PF) and Center solenoid (CS) magnet system. Each magnet coil has ∼ 1-2 nΩ low DC resistance joints as per the construction criteria and mechanical constraints. The measurement of such a low resistances is critical at the operating condition of 5 K helium temperature and 10 kA DC transport current. The development of electronics and instrumentation are challenging due to the measured signal intensity, large temperature gradient, large DC as well as time varing magnetic field ∼ 3 T and tokamak harsh noisy environment. A signal-conditioning electronics with large signal gain of 125 × 10"3 was developed for the low-DC superconducting joint resistance measurements. The measurement techniques followed to carry out these measurements by taking into account the thermo-electric potentials, lead resistance, non ohmic contacts, device heating etc. This paper presents the electronics design, measurement precision and different measurement techniques to measure the low-DC superconducting joint resistance.The paper also identifies the role of standard instruments and results of superconducting joints resistance measurement in the laboratory scale experiment. (author)

No abstract available

[en] The first results of divertor modelling for the DEMO-FNS divertor are presented. The divertor geometry open towards the private flux region is found important for improving the pumping conditions and reducing the power loading on the targets. (paper)