Fusion Engineering and Design 56–57 (2001) 947–952 An adaptive method to identify the plasma magnetic contour from magnetic and polarimetric measurements P. Bettini a, *, A. Formisano b , F. Trevisan a a Dipartimento di Ingegneria Elettrica, Gestionale e Meccanica, Uniersita ` di Udine, Via delle Scienze 208, I -33100 Udine, Italy b Dipartimento di Ingegneria dell’Informazione, Seconda Uniersita ` di Napoli, Via Roma 29, I -81031 Aersa, Italy Abstract The identification of a plasma magnetic contour can be cast as an inverse problem in which a set of equivalent currents, representing (the effect of) the actual plasma current distribution, have to be determined in order to best fit a set of external measurements. The diagnostic systems are typically provided with arrays of magnetic measurements, including both flux and field probes, positioned outside the plasma chamber, and some non magnetic measurements, such as motion Stark effect or multichord FIR polarimetry. In the framework of the method discussed, the basic external (magnetic) measurements have been integrated with internal (polarimetric) measurements to improve the identification process and gain information about the internal plasma current profile. The inverse problem can be formulated as the pseudo-inversion of the transfer matrix linking the unknown equivalent currents to the magnetic and polarimetric measurements. In this paper, the equivalent currents method is improved by means of a procedure that adaptively tries to allocate the EC to obtain the best estimate of the plasma contour. The adaptive strategy is presented in detail and its performance evaluated against two ITER equilibrium plasma configurations. © 2001 Elsevier Science B.V. All rights reserved. Keywords: Plasma magnetic contour; FIR polarimetry; Magnetic measurements; Polarimetric measurements www.elsevier.com/locate/fusengdes 1. Introduction In recent years, the applied research in the magnetically controlled thermonuclear fusion area has focused on the design, set up and testing of large tokamak-class devices. The plasma shape and position control system is one of the most critical issues in such a design. As a matter of fact, in the ITER machine the plasma instability time constants are in the order of some 100 ms (eg vertical instability growth time is 200 ms for the ITER FEAT layout). This in turn implies that a full estimation of the controlled variables must be provided in some 10 ms, preventing the adoption of full MHD equilibrium codes for the identifica- tion of plasma boundary. Under adequate hypothesis, valid in most of the relevant parts of tokamak operation, the recon- struction of plasma flux maps (i.e. the identifica- tion of plasma boundary) can be considered a magnetostatic problem, which is ill posed. A pos- sible approach to its regularisation can be the * Corresponding author. Tel.: +39-04-32558291; fax: +39- 04-32558258. E-mail address: bettini@uniud.it (P. Bettini). 0920-3796/01/$ - see front matter © 2001 Elsevier Science B.V. All rights reserved. PII:S0920-3796(01)00426-4