A high resolution neutron spectrometer (HRNS) system has been designed as a neutron diagnostic tool for ITER. The HRNS is dedicated to measurements of time resolved neutron energy spectra for both deuterium and deuterium-tritium (DT) plasmas. The main function of the HRNS is to determine the fuel ion ratio n t/n d in the plasma core with 20% uncertainty and a time resolution of 100 ms for a range of ITER operating scenarios from 0.5 MW to 500 MW in fusion power. Moreover, neutron spectroscopy measurements should also be possible in the initial deuterium phase of ITER experiments. A supplementary function of the HRNS is to provide information on the fuel ion temperature. Furthermore, the HRNS can be used as an additional line-of-sight (LOS) for the radial neutron camera. To meet these requirements, a set of four spectrometers positioned after each other along a single LOS has been designed. The detector techniques employed include a thin foil proton recoil spectrometer (TPR), a neutron diamond detector (NDD), a back-scattering time-of-flight system (bToF) and a forward time-of-flight system (fToF). The TPR system, positioned closest to the plasma, provides data at high fusion powers. For plasma conditions producing intermediate fusion power two neutron spectrometers are installed: NDD and bToF. The NDD is installed as the second instrument along the HRNS LOS after the TPR. The fToF spectrometer is dedicated for low tritium densities and pure deuterium operation. The paper summarizes the current state of the art of neutron spectroscopy useful in plasma diagnostics and the possibility of installing a dedicated HRNS for ITER in the designated diagnostic port. We conclude that the proposed HRNS system can fulfil the ITER measurement requirements over a broad range of plasma operational scenarios, including full power DT, start-up, ramp-down and pure D operations.

The Reverse Field eXperiment (RFX) explores the physics of Reverse Field Pinch (RFP) plasmas as an alternative confinement option with respect to Tokamak and Stellarator concepts. The toroidal field power supply of RFX, modified and upgraded during a long shut-down in 1999, is based on IGCTs (Integrated Gate Commutated Thyristors) and combines an extremely complex circuit topology and a fast, reliable and flexible control system. It includes, among other devices, 12 dc/ac converters rated for 3kV, 6kA, pulse duty operation (0.5s/10') and 14 static breakers rated for 4kV, 16kA, 128 MA2s. This paper will describe in detail the system, from the power to the control and protection sections. Experimental waveforms will also be reported and interesting indication about the system reliability and availability will be given, derived from more than ten years of operation.

FAST is a new machine proposed to support ITER experimental exploitation as well as to anticipate DEMO relevant physics and technology. FAST is aimed at studying, under burning plasma relevant conditions, fast particle (FP) physics, plasma operations and plasma wall interaction in an integrated way. FAST has the capability to approach all the ITER scenarios significantly closer than the present day experiments using deuterium plasmas. The necessity of achieving ITER relevant performance with a moderate cost has led to conceiving a compact tokamak (R = 1.82 m, a = 0.64 m) with high toroidal field (BT up to 8.5 T) and plasma current (Ip up to 8 MA). In order to study FP behaviours under conditions similar to those of ITER, the project has been provided with a dominant ion cyclotron resonance heating system (ICRH; 30MW on the plasma). Moreover, the experiment foresees the use of 6MW of lower hybrid (LHCD), essentially for plasma control and for non-inductive current drive, and of electron cyclotron resonance heating (ECRH, 4MW) for localized electron heating and plasma control. The ports have been designed to accommodate up to 10MW of negative neutral beams (NNBI) in the energy range 0.5-1MeV. The total power input will be in the 30-40MW range under different plasma scenarios with a wall power load comparable to that of ITER (P/R ~ 22MWm-1). All the ITER scenarios will be studied: from the reference H mode, with plasma edge and ELMs characteristics similar to the ITER ones (Q up to ?1.5), to a full current drive scenario, lasting around 170 s. The first wall (FW) as well as the divertor plates will be of tungsten in order to ensure reactor relevant operation regimes. The divertor itself is designed to be completely removable by remote handling. This will allow us to study (in view of DEMO) the behaviour of innovative divertor concepts, such as those based on liquid lithium. FAST is capable of operating with very long pulses, up to 170 s, despite being a copper machine. The magnets initial operation temperature is 30 K, with cooling provided by helium gas. The in vessel components, namely FW and divertor, are actively cooled by pressurized water above 80 oC. The same water is also used to bake the vacuum vessel. FAST is equipped with ferromagnetic inserts to keep the toroidal field magnet ripple down to 0.3%.