G. Chitarin

Machine modification for active MHD control in RFX

Abstract Recent studies on RFP and Tokamak devices call for an active control of the MHD and resistive wall modes to induce plasma mode rotation and to prevent mode phase locking. The results obtained on RFX, where slow rotation of phase locked modes has been induced, support the possibility of extending active MHD mode control through a substantial modification of the device. A new first wall with an integrated system of electric and magnetic transducers has been realised. A close fitting 3 mm thick Cu shell replaces the 65 mm Al shell. A toroidal support structure (TSS) made of stainless steel replaces the shell in supporting all the forces acting on the torus. A system of 192 saddle coils is provided to actively control the MHD modes. This system completely surrounds the toroidal surface and allows the generation of harmonic fields with m=0 and m=1 poloidal wave number and with a toroidal spectrum up to n=24.

Detailed design optimization of the MITICA negative ion accelerator in view of the ITER NBI

The ITER Neutral Beam Test Facility (PRIMA) is presently under construction at Consorzio RFX (Padova, Italy). PRIMA includes two experimental devices: an ITER-size ion source with low voltage extraction, called SPIDER, and the full prototype of the whole ITER Heating Neutral Beams (HNBs), called MITICA.The purpose of MITICA is to demonstrate that all operational parameters of the ITER HNB accelerator can be experimentally achieved, thus establishing a large step forward in the performances of neutral beam injectors in comparison with the present experimental devices.The design of the MITICA extractor and accelerator grids, here described in detail, was developed using an integrated approach, taking into consideration at the same time all the relevant physics and engineering aspects. Particular care was taken also to support and validate the design on the basis of the expertise and experimental data made available by the collaborating neutral beam laboratories of CEA, IPP, CCFE, NIFS and JAEA. Considering the operational requirements and the other physics constraints of the ITER HNBs, the whole design has been thoroughly optimized and improved. Furthermore, specific innovative concepts have been introduced.

Physics design of the HNB accelerator for ITER

The physics design of the accelerator for the heating neutral beamline on ITER is now finished and this paper describes the considerations and choices which constitute the basis of this design. Equal acceleration gaps of 88 mm have been chosen to improve the voltage holding capability while keeping the beam divergence low. Kerbs (metallic plates around groups of apertures, attached to the downstream surface of the grids) are used to compensate for the beamlet–beamlet interaction and to point the beamlets in the right direction. A novel magnetic configuration is employed to compensate for the beamlet deflection caused by the electron suppression magnets in the extraction grid. A combination of long-range and short-range magnetic fields is used to reduce electron leakage between the grids and limit the transmitted electron power to below 800 kW.

Overview of RFX-mod results

With the exploration of the MA plasma current regime in up to 0.5 s long discharges, RFX-mod has opened new and very promising perspectives for the reversed field pinch (RFP) magnetic configuration, and has made significant progress in understanding and improving confinement and in controlling plasma stability. A big leap with respect to previous knowledge and expectations on RFP physics and performance has been made by RFX-mod since the last 2006 IAEA Fusion Energy Conference. A new self-organized helical equilibrium has been experimentally achieved (the Single Helical Axis—SHAx—state), which is the preferred state at high current. Strong core electron transport barriers characterize this regime, with electron temperature gradients comparable to those achieved in tokamaks, and by a factor of 4 improvement in confinement time with respect to the standard RFP. RFX-mod is also providing leading edge results on real-time feedback control of MHD instabilities, of general interest for the fusion community.

The RFX magnet system

Abstract The reversed field pinch (RFP) magnetic confinement requires both toroidal and poloidal components for the magnetic field induction. As in tokamaks, the former is provided by the toroidal and poloidal components for the the ohmic heating (OH) winding and the equilibrium field (EF) winding. The two induction field components have similar amplitudes, so that the toroidal component required in a RFP is about one order of magnitude lower than that in a tokamak with equal plasma current and aspect ratio. Owing to the local stability properties, the TF coils have to be located as close as possible to the plasma but, at the same time, the magnetic field ripple from TF coils (as well as any kinds of stray fields) must be kept to a minimum. Another peculiarity of the TF winding is that it is required to operate with a time-varying current and at high voltage levels. From these points of view, the design of RFP TF windings presents much less technological problems than a tokamak of similar size. The RFP requires a high toroidal loop voltage during fast current rise and a relatively high toroidal loop voltage during the flat top. The main consequences for the RFX PF magnet systems are as follows: • - large flux and energy to be inductively stored; • - very high voltage across the OH and EF winding terminals, giving rise to substantial insulation problems; • - the very fast rate required for current rise may produce remarkable skin effects within the OH conductors, so that large cross-section conductors have to be avoided; • - the electrodynamic forces acting on the OH winding are large and comparable with those in tokamaks; • - in case of any fault, currents in the coils can rise beyond safety limits at a very high rate, leading to extremely critical conditions for the machine integrity . From a mechanical point of view, both the OH and the EF windings are subject to working conditions similar to those experienced in tokamaks. Thus, RFX windings were manufactured with a similar technology. To detect the TF and PF winding faults a very fast, hard-wired system has been developed, which is able to elaborate signals from specific probes and to decide on necessary protective actions. The present contribution deals with the whole magnet system of RFX, including all TF and PF magnets, mechanical structure, magnetic and electric measurement instrumentation, as well as the fast fault detection system. After a review of the basic concepts representing the theoretical background behind the main choices, all aspects and features concerning the magnet design are presented in detail and deeply discussed. The manufacturing technology is then presented together with the main problems met during manufacture, development and acceptance tests and the methods adopted in order to solve them are explained. On-site assembly procedures, testing and the first integrated RFX operation are finally described.

Magnetic self organization, MHD active control and confinement in RFX-mod

RFX-mod is a reversed field pinch (RFP) experiment equipped with a system that actively controls the magnetic boundary. In this paper we describe the results of a new control algorithm, the clean mode control (CMC), in which the aliasing of the sideband harmonics generated by the discrete saddle coils is corrected in real time. CMC operation leads to a smoother (i.e. more axisymmetric) boundary. Tearing modes rotate (up to 100 Hz) and partially unlock. Plasma–wall interaction diminishes due to a decrease of the non-axisymmetric shift of the plasma column. With the ameliorated boundary control, plasma current has been successfully increased to 1.5 MA, the highest for an RFP. In such regimes, the magnetic dynamics is dominated by the innermost resonant mode, the internal magnetic field gets close to a pure helix and confinement improves.

Automated optimal design techniques for inverse electromagnetic problems

Two methods are considered for the solution of automated digital optimal design problems. The computer procedures based on these methods have been implemented and tested. The first method, based on a deterministic approach, considers a quadratic approximation of the cost function. The second, based on a stochastic approach, is derived from the simulated annealing algorithm. Both methods, implemented as computer codes, have been applied to the solution of a test synthesis problem where the magnetic field is generated by discrete coils. The deterministic method is substantially faster, especially when the calculation of the cost function is time consuming. On the other hand, the stochastic method gives good approximation of the global minimum independently of the initial conditions: as far as CPU time is concerned, the method is more expensive, but can be profitably used when the cost function can be calculated quickly and the number of design variables is large. >

Physics design of the injector source for ITER neutral beam injector (invited).

Two Neutral Beam Injectors (NBI) are foreseen to provide a substantial fraction of the heating power necessary to ignite thermonuclear fusion reactions in ITER. The development of the NBI system at unprecedented parameters (40 A of negative ion current accelerated up to 1 MV) requires the realization of a full scale prototype, to be tested and optimized at the Test Facility under construction in Padova (Italy). The beam source is the key component of the system and the design of the multi-grid accelerator is the goal of a multi-national collaborative effort. In particular, beam steering is a challenging aspect, being a tradeoff between requirements of the optics and real grids with finite thickness and thermo-mechanical constraints due to the cooling needs and the presence of permanent magnets. In the paper, a review of the accelerator physics and an overview of the whole R&D physics program aimed to the development of the injector source are presented.

The magnetic diagnostic set for ITER

This paper presents the multiple set of requirements for the ITER magnetic diagnostic systems and the current status of the various R&D activities performed by the EU partners.

High current regimes in RFX-mod

Optimization of machine operation, including plasma position control, density control and especially feedback control on multiple magnetohydrodynamic modes, has led RFX-mod to operate reliably at 1.5?MA, the highest current ever achieved on a reversed field pinch (RFP). At high current and low density the magnetic topology spontaneously self-organizes in an Ohmical helical symmetry, with the new magnetic axis helically twisting around the geometrical axis of the torus. The separatrix of the island disappears leaving a wide and symmetric thermal structure with large gradients in the electron temperature profile. The new topology still displays an intermittent nature but its overall presence has reached 85% of the current flat-top period. The large gradients in the electron temperature profile appear to be marginal for the destabilization of ion temperature gradient modes on the assumption that ions and electrons have the same gradients. There are indications that higher currents could provide the conditions under which to prove the existence of a true helical equilibrium as the standard RFP configuration.