[en] This paper deals with the study of the time evolution of a free boundary plasma in a tokamak device, using an axisymmetric code. The model includes the plasma, the poloidal field coils (arbitrarily connected to external circuits) and the presence of massive conducting structures. (author). 11 refs.; 8 figs

[en] An integral formulation for eddy-current problems in nonmagnetic structures is presented. The solenoidality of the current density is assured by introducing a current vector potential T. This potential possesses only two scalar components, as the gauge chosen to ensure its uniqueness is T·u = O, where u is a prescribed vector field. The discrete analogue of this gauge and the boundary conditions are directly imposed by the shape functions. In massive structures, the two degrees of freedom are to be compared to four of the usual integral methods which exploit the presence of a scalar potential to ensure solenoidality. On the other hand, the procedure naturally reduces to the stream function approach when applied to thin shells. Finally, an integration procedure which guarantees symmetry and positive-definiteness of the inductance matrix is proposed. (author)

[en] Operational capability and sensitivity studies for the ITER Outline Design are presented. The operational objectives for ITER are to (i) demonstrate controlled ignition and extended burn in DT plasmas with a burn flat-top duration of ≥ 1000 s; (ii) conduct prolonged campaigns of extended-burn operation for in-situ testing of in-vessel and test article high-heat-flux and nuclear components; and (iii) demonstrate steady-state operation using a non-inductive current drive. For the component testing, the average neutron wall loading will be about 1 MW/m² and the fluence will be at least 1 MW·a/m², an end-of-life fluence capability for the basic machine components of up to 3 MW·a/m² will be adopted. An Outline Design has been developed that fulfills the ITER Detailed Technical Objectives. The ITER Technical Advisory Committee finds that the present design has been successful in its attempt to maximize physics and engineering performance while minimizing cost and complexity. (author). 6 refs, 3 tabs

[en] This paper presents the results of some dedicated experiments performed on Joint European Torus (JET), and related simulations, that clearly demonstrate the existence in JET of a neutral point (NP) for density limit disruptions. It has been observed that a plasma, specially designed to be set at different vertical equilibrium position without altering the shape, moves upwards (downwards) when the disruption is triggered with the plasma below (above) the NP. The CREATE_L linearised plasma response model applied to such configurations is able to predict and explain the most significant qualitative features of the experiment

[en] The CREATE-L linearized plasma response model can simulate the plasma current, position and shape responses to external coil currents and internal physical variations, also in the presence of 3D conducting structures whose eddy currents cannot be neglected. It has already been validated on existing tokamaks (TCV, FTU) and applied to tokamak under design (ITER) in order to investigate their control system capabilities. On these devices it has also allowed the design of high-order controllers. This paper presents the extension of the CREATE-L model to reversed field pinch (RFP) plasmas. The validation of this new model has been carried out against experimental data from RFX, the largest RFP machine

[en] A description is given of the poloidal field coil design system used for NET. This system is used to carry out sensitivity studies of the overall poloidal field coil currents to the plasma parameters (elongation, triangularity) and to the relative flux contributions from ohmic heating and equilibrium coils. The results from these are used to produce various poloidal field coil configurations for NET, and to examine the effects of machine size and maintenance access requirements on these configurations. (author)

[en] This paper describes the development and experimental validation of a simulation model for the design of FTU plasma radial position and current controllers. These controllers have been designed to be tested experimentally on the FTU tokamak. The results predicted in simulation were actually delivered during FTU operation, in two different discharges

[en] In tokamak reactors with elongated cross-section the plasma is unstable against vertical displacements. A significant contribution to the passive stabilization is given by the non-axisymmetric blanket shield structure surrounding the plasma. The paper aims to present a circuital model for the passive stabilization of the plasma, where the blanket/shield is schematized with a finite element network technique. It is shown how the efficiency of the passive system, measured by the residual growth rate of the instability, can be easily determined by solving an eigenvalue problems. The model is currently used in the INTOR/NET project

[en] In parallel with the programme to optimize the operation with a conventional divertor based on detached conditions to be tested on the ITER device, a project has been launched to investigate alternative power exhaust solutions for DEMO, aimed at the definition and the design of a divertor tokamak test facility (DTT). The DTT project proposal refers to a set of parameters selected so as to have edge conditions as close as possible to DEMO, while remaining compatible with DEMO bulk plasma performance in terms of dimensionless parameters and given constraints. The paper illustrates the DTT project proposal, referring to a 6 MA plasma with a major radius of 2.15 m, an aspect ratio of about 3, an elongation of 1.6–1.8, and a toroidal field of 6 T. This selection will guarantee sufficient flexibility to test a wide set of divertor concepts and techniques to cope with large heat loads, including conventional tungsten divertors; liquid metal divertors; both conventional and advanced magnetic configurations (including single null, snow flake, quasi snow flake, X divertor, double null); internal coils for strike point sweeping and control of the width of the scrape-off layer in the divertor region; and radiation control. The Poloidal Field system is planned to provide a total flux swing of more than 35 Vs, compatible with a pulse length of more than 100 s. This is compatible with the mission of studying the power exhaust problem and is obtained using superconducting coils. Particular attention is dedicated to diagnostics and control issues, especially those relevant for plasma control in the divertor region, designed to be as compatible as possible with a DEMO-like environment. The construction is expected to last about seven years, and the selection of an Italian site would be compatible with a budget of 500 M€. (paper)

[en] Structure and features of three codes (NAPS, PROTEUS and CARIDDI) developed for design and analysis related to electromagnetic transients and plasma engineering are briefly reviewed. Some applications concerning first wall and blanket are presented, and general conclusions regarding the engineering and modelling of these components are drawn. (author)