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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.

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.

The EFDA Close Support Unit (CSU) has initiated a technological development task to manufacture three full scale models of the austenitic steel cases of the ITER Toroidal Field (TF) coils. The main goals are to verify the feasibility of such a thick construction, assess the achievable material properties, and develop the manufacturing processes and quality control procedures. A first model (called Model 1), reproducing the geometry of the inner curved region of the TF coil, has been manufactured with thick forgings. A second model (Model 2), representing the geometry of one outer intercoil structure, has been produced by casting. A third model of the inner straight leg is being manufactured. The work has included a large campaign of development and qualification of welding processes and inspection procedures for thick case sections. Numerical models for the prediction of the geometrical deformations and residual stresses due to the welding have been developed. The goal of this task is the assessment of the manufacturing and quality procedures for the production of the ITER TF coil cases.

The requirements for ITER EDA were to achieve ignition with a good safety margin, and controlled long inductive burn. These requirements lead to a big device, which requested a too ambitious step to be undertaken by the world fusion community. More realistic objectives for a next step device shall be to demonstrate the net production of energy with a high energy gain factor (Q) and a high boot strap current fraction (>60%) which is required for a Fusion Power Plant (FPP). The Next Step Device (NSD) shall also allow operation flexibility in order to explore a large range of plasma parameters to find out the optimum concept for the fusion power plant prototype. These requirements could be too demanding for one single device and could probably be better explored in a strongly integrated world programme. The cost of one or more devices is the decisive factor for the choice of the fusion power development programme strategy. The plasma elongation and triangularity have a strong impact in the cost of the device and are limited by the plasma vertical position control issue. The distance between plasma separatrix and the toroidal field conductor does not vary a lot between devices. It is determined by the summe of the distance between first wall-plasma sepratrix and the thickness of the nuclear shield required to protect the toroidal field coil insultation. The thickness of the TF coil is determined by the allowable stresses and superconducting characteristics. The outer radius of the central solenoid is the result of an optimisation to provide the magnetic flux to inductively drive the plasma. Therefore, in order to achieve the objectives for Q and boot-strap current fractions at the minimum cost, the plasma aspect ratio and magnetic field value shall be determined. The paper will present the critical issues for the next device and will make considerations on the optimal way to proceed towards the realisation of the fusion power plant.

The TF coils for ITER (ITER, Final Design Report, December 1997) will use the concept of a circular thin walled Nb3Sn cable in conduit superconductor completely enclosed in an insulated groove in steel plates to form the coil pancakes. These are then stacked together to form the winding pack supported by a steel case. The concept is being demonstrated by the fabrication of a TFMC (E. Salpietro et al., Construction of a Toroidal Field Model Coil (TFMC) for ITER, MT-15 Fifteenth International Conference on Magnet Technology, Oct 20–24, 1997, Beijing, China; R. Maix et al., Manufacture of the ITER TF model coil (TMFC), Poster P1-217) with a bore of 1.4×2.5 m, a peak field of 8.77 T and a total current of 7.8 MAT. The coil will be tested during 1999 in the TOSKA facility at FZK, Karlsruhe, using the EURATOM LCT coil to provide an external field system. The objectives of the TFMC are as follows: 1. to develop and verify the full scale TF coil manufacturing techniques, in particular the following features: ◦ plate manufacturing (forming the grooves); ◦ fitting the conductor in the groove after heat treatment and insulation (i.e. predictable geometry change); ◦ closing the groove with a cover plate and plate insulation; ◦ fitting the winding into the case, gap filling and case closure; 2. to establish realistic manufacturing tolerances; 3. to bench-mark methods for the ITER TF coil acceptance tests, including insulation and impregnation process; 4. to obtain information on the coil’s mechanical behaviour, operating margins and in-service monitoring techniques, particularly for the insulation quality over fatigue cycles.

Objective of the Safety and Environmental Assessment of Fusion Power (SEAFP) [1] is to `assess the safety and environmental features of fusion on the basis of the conceptual studies of tokamak power reactors which incorporate safety and environmental features likely to ensure public acceptance and take into account the operating requirements of utilities'. The fusion reactors for commercial use are expected to enter in operation in several decades, therefore the assessment shall be as relevant as possible with respect to further development in physics and technology. The reactor has been conceived for monitored and filtered release of radioactivity to the environment during normal operation equivalent to a small fraction of the natural exposure; to avoid radioactivity release to the environment for all in-plant accidents and conceivable ex-plant accidents; to avoid disruption of community life outside the plant perimeter even in the case of rupture of confinement barriers, in particular by non-anticipated plant initiators. Consideration of possible fault conditions is an integrated part of the design of the main components and their associated support systems. the overall philosophy proposed is that all faults should be contained within an individual system regardless of the initiation fault or external action.

The TF model coil (TFMC) is one of the seven large RD the tests of the W 7-X stellarator prototype coil in an arrangement similar to the ITER TF model coil testing.

The first wall (FW) is one of the most loaded plasma-facing components; the heat flux is such that the thermal stresses are the most important design concern. In addition, the FW must resist the eddy current induced during plasma disruption and the high pressure of the coolant and should maintain its properties under a fast neutron flux (dose up to 3 MW m−2). The tubular solution is the most suitable to cope with the thermal stresses; the use of a double wall reduces the risk of leaks inside the vacuum vessel by avoiding the growth of cracks through both walls: the soft brazing in between the walls stops the growth of cracks from one tube to the other. The eddy currents induced in the tubes are low and the halo current flowing poloidally in the tubes exerts a radial pressure which is supported by the blanket box via supporting points provided in between the FW and the blanket. The tubes are protected with a coating of beryllium or boron carbide against the radiation heat load during disruption, and with a coating of copper against runaway electrons. Fins attached to the tubes are provided to cope with the change in the toroidal width of the FW along the poloidal direction. The fins are also protected by coatings. The tubes can be made of steel to resist a heat flux of up to 1 MW m−2. For higher heat loads, copper or vanadium can be used. The tubular FW can be replaced independently of the blanket. The thermohydraulic, electromagnetic and dynamic analyses confirm the viability of the solution proposed.

In principle, for given required plasma physics performance, the inputs for the design of the basic device (magnetic system + vacuum vessel) are rather well established. Conversely, for the design of the plasma facing components (PFCs) (e.g. first wall protection and divertor plates), a confirmation under the real operating conditions is required in the first stage of the machine operation. Therefore, to ensure success, the whole machine design is based upon rather conservative assumptions for the plasma physics, and the PFCs are designed to be replaceable. In the following the main features of the NET/ITER concept are described and key RD needed before starting construction is outlined.

NET will produce fusion reactions, which will result in the structure and internal components of the machine becoming radioactive. Consequently, NET will use remote handling techniques to perform maintenance. The strategy being followed to ensure that NET is fully compatible with the remote handling requirements is described together with a preliminary list of the remote handling equipment being developed.

The Next European Torus (NET) is a tokamak machine intended as the next step in the European fusion programme beyond the JET machine. As such it must have the capability to produce ignited plasmas and drive them for periods of up to several hundred seconds. One of the most critical areas in the design of fusion machines of the next generation, with both physics and technological objectives, is the plasma engineering. In this report we will consider the following aspects of the NET plasma engineering: (i) plasma configuration during start-up, burn and shut down; (ii) plasma control (equilibrium and stability) during burn; (iii) poloidal field coil power supply requirements; (iv) effect of disruptions. The poloidal field coil circuit must be designed not only to provide plasma equilibrium for a reference plasma during the machine “burn” phase, but also to start-up and shut down the plasma in a controlled fashion, while keeping within the technological constraints (maximum current density, maximum field, rate of change of field, power and voltage) imposed by superconducting coils and power supplies.