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We present a technique to efficiently solve 3D eddy current problems in fusion devices, whose structures exhibit a geometrical symmetry. The proposed approach is based on the exploitation of symmetry via harmonic analysis and it is suitable to treat also problems where the field sources (plasma or external coils) do not share the domain symmetry. A simple test case is presented to describe the methodology. Then, a more complicated geometry is considered, which represents a realistic vacuum vessel of an ITER like fusion device (double layer structure with several portholes). The results are discussed for both axisymmetric and non-axisymmetric excitations in frequency domain. [ABSTRACT FROM AUTHOR]

In magnetic confinement fusion devices, close-fitting passive conducting structures surrounding the plasma are an efficient way to slow down the growth rate of magnetohydrodynamic (MHD) instabilities, enabling the possibility of an active stabilization by means of external actuators (coils). Nevertheless, a fully quantitative investigation, under realistic 3D boundary conditions, is still a very hard challenge. Moreover, the inclusion of fine geometrical details of actual passive structures might easily lead to impractical memory and computational time requirements if the problem is not carefully addressed. In this work we present an effective technique to solve eddy current problems in thin conductors of arbitrary topology by a boundary element method (BEM) based on a stream function. The proposed technique is applied to study the dynamic response of the stabilizing shell of RFX-mod, both to axisymmetric and non-axisymmetric external fields. [ABSTRACT FROM AUTHOR]

A major refurbishment of the toroidal complex of the RFX-mod device is in progress and will include the removal of the Inconel vacuum vessel and a modification of the stainless steel supporting structure to be made vacuum tight. The plasma facing graphite tiles will be mounted onto the inner surface of the copper shell so as to increase the plasma proximity factor. New operation regimes are expected to provide a significant reduction of the amplitude of RFP tearing modes with magnetic chaos mitigation and confinement improvement. On the other hand, due to the shorter distance from the passive structures, the plasma is expected to be even more sensitive to magnetic field errors at the plasma boundary, produced by the induced currents near the cuts of the same structures. In preparation to the new RFX-mod2 experiments, a thorough revision of the conditions triggering the error fields due to the eddy currents was undertaken. An analysis of the static and dynamic equilibrium magnetic field has been carried out in RFX-mod and RFX-mod2 to estimate the magnetic field driving the shell eddy currents. A lower equilibrium magnetic field should be required in RFX-mod2. Analyses of different configurations of the poloidal gap were also carried out by a specialized computational tool. A solution with a spaced poloidal gap and a more extended overlapping was envisaged capable of maintaining the same error fields at the closer plasma boundary with the same forcing field and meeting the more stringent insulation and assembly requirements of RFX-mod2. [ABSTRACT FROM AUTHOR]

After several years of successful RFX-mod operations, both in RFP and Tokamak configurations, an upgrade of the machine has been conceived. In particular, the vacuum vessel might be removed with the aim of improving passive MHD control and enabling plasma rotation at higher currents. In this paper, a detailed analysis of the new magnetic front-end is presented, with particular emphasis on the computation of the magnetic field errors generated at the poloidal gaps during transient phases of the discharge. At this purpose, a non-linear MHD equilibrium code, MAXFEA, has been used to provide the input data for the 3D electromagnetic analyses in the time domain carried out with the CAFE code. [ABSTRACT FROM AUTHOR]

RFX-mod is a device equipped with a state-of-the-art system for active control of MHD instabilities, which consists of 192 saddle coils and 192 radial field sensors. In order to further extend its operational space an upgrade of its magnetic front-end is now being studied. In this paper we focus on the implementation of fast and efficient algorithms for the 3D electromagnetic analysis of the effect of the designed magnetic front-end on the MHD control system in RFX-mod2, in the presence of complex conducting structures surrounding the plasma. [ABSTRACT FROM AUTHOR]

Abstract: Disruptions represent one of the main concerns for Tokamak operation, especially in view of fusion reactors, or experimental test reactors, due to the electro-mechanical loads induced by halo and eddy currents. The development of a predictive tool which allows to estimate the magnitude and spatial distribution of the halo current forces is of paramount importance in order to ensure robust vessel and in-vessel component design. With this aim, two numerical codes (CARIDDI, CAFE) have been developed, which allow to calculate the halo current path (resistive distribution) in the passive structures surrounding the plasma. The former is based on an integral formulation for the eddy currents problem particularized to the static case; the latter implements a pair of 3D FEM complementary formulations for the solution of the steady-state current conduction problem. A simplified plasma model is adopted to provide the inputs (halo current injected into the first wall). Two representative test cases (ITER symmetric and asymmetric VDEs) have been selected to cross check the results of the proposed approaches. [Copyright &y& Elsevier]